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Sample records for aecl radiochemical slowpoke reactor

  1. Utilization of the SLOWPOKE-2 research reactor

    International Nuclear Information System (INIS)

    SLOWPOKEs are typically low power research reactors that have a limited number of applications. However, a significant range of NAA can be performed with such reactors. This paper describes a SLOWPOKE-based NAA program that is performing a valuable series of studies in Jamaica, including geological mapping and pollution assessment. (author)

  2. SLOWPOKE: heating reactors in the urban environment

    International Nuclear Information System (INIS)

    Since global energy requirements are expected to double over the next 40 years, nuclear heating could become as important as nuclear electricity generation. To fill that need, AECL has designed a 10 MW nuclear heating plant for large buildings. Producing hot water at temperatures below 100 degrees Celsius, it incorporates a small pool-type reactor based on the successful SLOWPOKE Research Reactor. A 2 MW prototype is now being tested at the Whiteshell Nuclear Research Establishment in Manitoba, and the design of a 10 MW commercial unit is well advanced. With capital costs in the range $5 million to $7 million, unit energy costs could be as low as $0.02 per kWh, for a unit operating at 50% load factor over a 25-year period. By keeping the reactor power low and the water temperature below 100 degrees Celsius, much of the complexity of the large nuclear power plants can be avoided, thus allowing these small, safe, nuclear heating systems to be economically viable

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

    International Nuclear Information System (INIS)

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

  4. Keeping research reactors relevant: A pro-active approach for SLOWPOKE-2

    International Nuclear Information System (INIS)

    The SLOWPOKE is a small, inherently safe, pool-type research reactor that was engineered and marketed by Atomic Energy of Canada Limited (AECL) in the 1970s and 80s. The original reactor, SLOWPOKE-1, was moved from Chalk River to the University of Toronto in 1970 and was operated until upgraded to the SLOWPOKE-2 reactor in 1973. In all, eight reactors in the two versions were produced and five are still in operation today, three having been decommissioned. All of the remaining reactors are designated as SLOWPOKE-2 reactors. These research reactors are prone to two major issues: aging components and lack of relevance to a younger audience. In order to combat these problems, one SLOWPOKE -2 facility has embraced a strategy that involves modernizing their reactor in order to keep the reactor up to date and relevant. In 2001, this facility replaced its aging analogue reactor control system with a digital control system. The system was successfully commissioned and has provided a renewed platform for student learning and research. The digital control system provides a better interface and allows flexibility in data storage and retrieval that was never possible with the analogue control system. This facility has started work on another upgrade to the digital control and instrumentation system that will be installed in 2010. The upgrade includes new computer hardware, updated software and a web-based simulation and training system that will allow licensed operators, students and researchers to use an online simulation tool for training, education and research. The tool consists of: 1) A dynamic simulation for reactor kinetics (e.g., core flux, power, core temperatures, etc). This tool is useful for operator training and student education; 2) Dynamic mapping of the reactor and pool container gamma and neutron fluxes as well as the vertical neutron beam tube flux. This research planning tool is used for various researchers who wish to do irradiations (e.g., neutron

  5. 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)

  6. Operation of the SLOWPOKE-2 reactor in Jamaica

    International Nuclear Information System (INIS)

    Over the past sixteen years lCENS has operated a SLOWPOKE 2 nuclear reactor almost exclusively for the purpose of neutron activation analysis. During this period we have adopted a strategy of minimum irradiation times while optimizing our output in an effort to increase the lifetime of the reactor core and to maintaining fuel integrity. An inter-comparison study with results obtained with a much larger reactor at IPEN has validated this approach. The parameters routinely monitored at ICENS are also discussed and the method used to predict the next shim adjustment. (author)

  7. Operation of the SLOWPOKE-2 reactor in Jamaica

    Energy Technology Data Exchange (ETDEWEB)

    Grant, C.N.; Lalor, G.C.; Vuchkov, M.K. [University of the West Indies, Kingston (Jamaica)

    2001-07-01

    Over the past sixteen years lCENS has operated a SLOWPOKE 2 nuclear reactor almost exclusively for the purpose of neutron activation analysis. During this period we have adopted a strategy of minimum irradiation times while optimizing our output in an effort to increase the lifetime of the reactor core and to maintaining fuel integrity. An inter-comparison study with results obtained with a much larger reactor at IPEN has validated this approach. The parameters routinely monitored at ICENS are also discussed and the method used to predict the next shim adjustment. (author)

  8. A new safety principle for the SLOWPOKE reactor

    International Nuclear Information System (INIS)

    Slowpoke-2 (LEU core) is a pool type nuclear reactor with a maximum thermal power of 20 kW. It uses a pelletized uranium oxide fuel (19.9% enrichment) and provides a useful high neutron flux in the order of 1012 n.cm-2s-1. The key safety features built into the reactor design are the strictly limited amount of excess reactivity and the negative reactivity feedback characteristics, which provides a demonstrably safe self-limiting power excursion response to large reactivity insertions. However, the limited amount of excess reactivity also limits the continuous prolong reactor operation at full power. With a 3.7 mk excess reactivity, the reactor can operate for about one day at the full power, 20 kW, before this excess activity is lost due to temperature effects and Xe poisoning. A new safety concept is proposed in this paper to extend the continuous operation time to months by increasing the excess reactivity from 4 mk to 6 mk. This new concept has been demonstrated using a Matlab/simulink model of Slowpoke-2. (author)

  9. 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

  10. Removal of Beryllium Material during Decommissioning of a Slowpoke Reactor, Toronto, Canada

    International Nuclear Information System (INIS)

    The Slowpoke (acronym for Safe LOW-POwer Kritical Experiment) is a low energy, tank-in-pool type nuclear research reactor designed by the Atomic Energy of Canada Limited in the late 1960s. The fuel cage is surrounded by a beryllium assembly at the bottom of a water pool about 6 m deep. The beryllium reflects neutrons back into the core. Basically, the reactor is a subcritical mass of fuel, which the surrounding beryllium makes critical. The rate of reaction is controlled by inserting a neutron absorbing cadmium rod. Slowpokes have a maximum power of 100 kW and normally operate at about 20 kW. The University of Toronto SLOWPOKE-2 Reactor research services ended in December 1998, and the reactor was finally defuelled in June 2000. On 10 November 2000, the Canadian Nuclear Safety Commission (CNSC) issued the decommissioning licence to the University of Toronto for its SLOWPOKE-2 Nuclear Reactor Facility. The reactor decommissioning was completed in January 2001. The beryllium material was to have been shipped under the operating licence, but actually it was shipped under the decommissioning licence. The CNSC revoked the decommissioning licence for the University of Toronto SLOWPOKE-2 Reactor Facility on 24 February 2012, and the site was returned to the university for unrestricted site use. The following is a description of the incident involving the beryllium material management

  11. 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)

  12. Training new operators for the SLOWPOKE reactor at Ecole Polytechnique: theory and practice

    International Nuclear Information System (INIS)

    In the last two years we trained two new operators for the SLOWPOKE reactor at Ecole Polytechnique. In this paper we describe how the training program for these operators was designed. We also discuss the shortcomings that were identified in the program and the modifications it required when being put in use.

  13. A novel approach to the production of medical radioisotopes: the homogeneous SLOWPOKE reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [retired, Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Royal Canadian Navy, Ottawa, Ontario (Canada)

    2015-03-15

    In 2009, the unexpected 15-month outage of the Canadian NRU nuclear reactor resulted in a sudden 30% world shortage, with higher shortages experienced in North America than in Europe. Commercial radioisotope production is from just eight nuclear reactors, most being aging systems near the end of their service life. This paper proposes a more efficient production and distribution model. Tc-99m unit doses would be distributed to regional hospitals from ten integrated 'industrial radiopharmacies', located at existing licensed nuclear reactor sites in North America. At each site, one or more 20 kW Homogeneous SLOWPOKE nuclear reactors would deliver 15 litres of irradiated aqueous uranyl sulfate fuel solution daily to industrial-scale hot cells, for extraction of Mo-99; and the low-enriched uranium would be recycled. Purified Mo-99 would be incorporated in large Mo-99/Tc-99m generators for extraction of Tc-99m five days a week; and each automated hot-cell facility would be designed to load up to 7,000 Tc-99m syringes daily for road delivery to all of the nuclear medicine hospitals within a 3-hour range. At the current price of $20 per unit dose, the annual gross income from 10 sites would be approximately $360 million. The Homogeneous SLOWPOKE reactor evolved from the inherently safe SLOWPOKE-2 research reactor, with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors at the end-of-core life, enabling them to continue their primary missions of research and education, together with full time commercial radioisotope production. The Homogeneous SLOWPOKE reactor was modelled using both deterministic and probabilistic reactor simulation codes. The homogeneous fuel mixture is a dilute aqueous solution of low-enriched uranyl sulfate containing approximately 1 kg of U-235. The reactor is controlled by mechanical absorber rods in the beryllium reflector. Safety analysis was carried out for both normal operation and transient conditions. The most severe

  14. SLOWPOKE-2 31 years and still glowing

    International Nuclear Information System (INIS)

    In 1981, a SLOWPOKE-2 nuclear research reactor was installed at SRC Environmental Analytical Laboratories facility in Saskatoon and it has been operated trouble-free ever since it's commissioning. The SLOWPOKE reactor was designed by Atomic Energy of Canada Ltd (AECL) in the 1970's and is a low-energy, pool type reactor. The design is such that the heat produced by the reactor limits its reactivity and hence the reactor cannot run into uncontrolled power excursion, thus providing a high degree of inherent safety. Too small to be used as a source of power generation, the SLOWPOKE is used as a neutron source for an analytical technique known as neutron activation analysis (NAA). NAA is a non-destructive technique that allows for the analysis of many elements of the periodic table by producing radioactive isotopes which can then by analyzed using gamma spectroscopy, or by inducing fission in U 235 for analysis of uranium by delayed neutron counting. The history and basic design features of the reactor will be discussed. Applications of the NAA technique used at the laboratory will also be presented.

  15. Keeping research reactors relevant: a pro-active approach for SLOWPOKE-2 at RMC

    International Nuclear Information System (INIS)

    In 2001, the Royal Military College of Canada replaced its aging analogue SLOWPOKE-2 reactor control system with a digital control system. The system was successfully commissioned and has provided a renewed platform for student learning and research. An upgrade to the digital control and instrumentation system has been completed and will be installed in October 2010. The upgrade includes new computer hardware, updated software and a simulation and training system that will enhance training, education and research by licensed operators, students and researchers.

  16. 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.

  17. Rationalization and future planning for AECL's research reactor capability

    International Nuclear Information System (INIS)

    AECL's research reactor capability has played a crucial role in the development of Canada's nuclear program. All essential concepts for the CANDU reactors were developed and tested in the NRX and NRU reactors, and in parallel, important contributions to basic physics were made. The technical feasibility of advanced fuel cycles and of the organic-cooled option for CANDU reactors were also demonstrated in the two reactors and the WR-1 reactor. In addition, an important and growing radio-isotope production industry was established and marketed on a world-wide basis. In 1984, however, it was recognized that a review and rationalization of the research reactor capability was required. The commercial success of the CANDU reactor system had reduced the scope and size of the required development program. Limited research and development funding and competition from other research facilities and programs, required that the scope be reduced to a support basis essential to maintain strategic capability. Currently, AECL, is part-way through this rationalization program and completion should be attained during 1992/93 when the MAPLE reactor is operational and decisions on NRX decommissioning will be made. A companion paper describes some of the unique operational and maintenance problems which have resulted from this program and the solutions which have been developed. Future planning must recognize the age of the NRU reactor (currently 32 years) and the need to plan for eventual replacement. Strategy is being developed and supporting studies include a full technical assessment of the NRU reactor and the required age-related upgrading program, evaluation of the performance characteristics and costs of potential future replacement reactors, particularly the advanced MAPLE concept, and opportunities for international co-operation in developing mutually supportive research programs

  18. Standardization of the SLOWPOKE-2 reactor in Jamaica for routine NAA

    International Nuclear Information System (INIS)

    The International Centre for Environmental and Nuclear Sciences (ICENS) has been involved in conducting multipurpose geochemical surveys, the results of which were published in 'A Geochemical Atlas of Jamaican Soils'. The primary analytical tool for these studies was neutron activation analysis (NAA) using the SLOWPOKE-2 reactor at the Centre. The neutron flux of the SLOWPOKE-2 reactor is extremely stable, thus allowing a semi-absolute method for quantitative NAA. This has several advantages, but requires preparation and measurement of the single- or multi-element standards for each gamma-spectroscopy system (GSS). The NAA laboratory at ICENS operates three GSSs. The primary ('master') GSS was standardized using single element standards for over 50 elements, naturally occurring in most geological and biological materials. The standardization of the secondary GSS's was achieved by transferring of the elemental sensitivities of the master GSS using an instrumentation-free standardization approach. Implementation of this methodology and its utilization in the routine analytical work is described. (author)

  19. Electronic data collection for AECL NRU Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Klein, S., E-mail: kleins@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    An Electronic Data Collection program has been implemented at AECL's National Research Universal (NRU) Reactor to assists in the collection, trending and monitoring of system parameters. Prior to the implementation, daily surveillance was performed by Operators using hardcopy reading sheets. This was not conducive to proactive monitoring of equipment health as there was a significant delay from when the data was recorded to when it was accessible for trending. The electronically collected data allows for close to real-time monitoring of 971 unique instruments and approximately 7,200 data points per day. There are currently over 1,000,000 readings available in the database. (author)

  20. Electronic data collection for AECL NRU Research Reactor

    International Nuclear Information System (INIS)

    An Electronic Data Collection program has been implemented at AECL's National Research Universal (NRU) Reactor to assists in the collection, trending and monitoring of system parameters. Prior to the implementation, daily surveillance was performed by Operators using hardcopy reading sheets. This was not conducive to proactive monitoring of equipment health as there was a significant delay from when the data was recorded to when it was accessible for trending. The electronically collected data allows for close to real-time monitoring of 971 unique instruments and approximately 7,200 data points per day. There are currently over 1,000,000 readings available in the database. (author)

  1. The SLOWPOKE-2 nuclear reactor at the Royal Military College of Canada: applications for the Canadian Armed Forces

    International Nuclear Information System (INIS)

    The Royal Military College of Canada (RMCC) has a 20 kW SLOWPOKE-2 nuclear research reactor which is used for teaching and research. Since its commissioning, the reactor facility and instruments have been continuously upgraded to develop and enhance nuclear capabilities for the Canadian Armed Forces (CAF). Specific applications of neutron activation analysis (NAA), delayed neutron counting (DNC) and neutron imaging relevant to the CAF are discussed. (author)

  2. The SLOWPOKE-2 nuclear reactor at the Royal Military College of Canada: applications for the Canadian Armed Forces

    International Nuclear Information System (INIS)

    The Royal Military College of Canada (RMCC) has a 20 kW SLOWPOKE-2 nuclear research reactor which is used for teaching and research.Since its commissioning, the reactor facility and instruments have been continuously upgraded to develop and enhance nuclear capabilities for the Canadian Armed Forces (CAF). Specific applications of neutron activation analysis (NAA), delayed neutron counting (DNC) and neutron imaging relevant to the CAF are discussed. (author)

  3. An experimental investigation of fission product release in SLOWPOKE-2 reactors

    International Nuclear Information System (INIS)

    Increasing radiation fields due to a release of fission products in the reactor container of several SLOWPOKE-2 reactors fuelled with a highly-enriched uranium (HEU) alloy core have been observed. It is believed that these increases are associated with the fuel fabrication where a small amount of uranium-bearing material is exposed to the coolant at the end-welds of the fuel element. To investigate this phenomenon samples of reactor water and gas from the headspace above the water have been obtained and examined by gamma spectrometry methods for reactors of various burnups at the University of Toronto, Ecole Polytechnique and Kanata Isotope Production Facility. An underwater visual examination of the fuel core at Ecole Polytechnique has also provided information on the condition of the core. This report (Volume 1) summarizes the equipment, analysis techniques and results of tests conducted at the various reactor sites. The data report is published as Volume 2. (author). 30 refs., 9 tabs., 20 figs

  4. A program for the a priori evaluation of detection limits in instrumental neutron activation analysis using a SLOWPOKE II reactor

    International Nuclear Information System (INIS)

    A program that permits the a priori calculation of detection limits in monoelemental matrices, adapted to instrumental neutron activation analysis using a SLOWPOKE II reactor, is described. A simplified model of the gamma spectra is proposed. Products of (n,p) and (n,α) reactions induced by the fast components of the neutron flux that accompanies the thermal flux at the level of internal irradiation sites in the reactor have been included in the list of interfering radionuclides. The program calculates in a systematic way the detection limits of 66 elements in an equal number of matrices using 153 intermediary radionuclides. Experimental checks carried out with silicon (for short lifetimes) and aluminum and magnesium (for intermediate lifetimes) show satisfactory agreement with the calculations. These results show in particular the importance of the contribution of the (n,p) and (n,α) reactions in the a priori evaluation of detection limits with a SLOWPOKE type reactor

  5. The Development of Neutron Radiography and Tomography on a SLOWPOKE-2 Reactor

    Science.gov (United States)

    Bennett, L. G. I.; Lewis, W. J.; Hungler, P. C.

    Development of neutron radiography at the Royal Military College of Canada (RMC) started by trying to interest the Royal Canadian Air Force (RCAF) in this new non-destructive testing (NDT) technique. A Californium-252 based device was ordered and then installed at RMC for development of applicable techniques for aircraft by the first author. A second and transportable device was then designed, modified and used in trials at RCAF Bases and other locations for one year. This activity was the only foreign loan of the U.S. Californium Loan Program. Around this time, SLOWPOKE-2 reactors were being installed at four Canadian universities, while a new science and engineering building was being built at RMC. A reactor pool was incorporated and efforts to procure a reactor succeeded a decade later with a SLOWPOKE-2 reactor being installed at RMC. The only modification by the vendor for RMC was a thermal column replacing an irradiation site inside the reactor container for a later installation of a neutron beam tube (NBT). Development of a working NBT took several years, starting with the second author. A demonstration of the actual worth of neutron radiography took place with a CF-18 Hornet aircraft being neutron and X-radiographed at McClellan Air Force Base, Sacramento, CA. This inspection was followed by one of the rudders that had indications of water ingress being radiographed successfully at RMC just after the NBT became functional. The next step was to develop a neutron radioscopy system (NRS), initially employing film and then digital imaging, and is in use today for all flight control surfaces (FCS). With the third author, a technique capable of removing water from affected FCS was developed at RMC. Heating equipment and a vacuum system were utilized to carefully remove the water. This technique was proven using a sequence of near real time neutron images obtained during the drying process. The results of the drying process were correlated with a relative humidity

  6. SLOWPOKE: neutron activation analysis

    International Nuclear Information System (INIS)

    Neutron activation analysis permits the non-destructive determination of trace elements in crude oil and its derivatives at high sensitivity (up to 10-9 g/g) and good precision. This article consists of a quick survey of the method followed by an illustration based on the results of recent work at the SLOWPOKE reactor laboratory at the Ecole Polytechnique

  7. The Jamaican Slowpoke HEU-LEU core conversion

    International Nuclear Information System (INIS)

    The HEU core of the Jamaican SLOWPOKE research reactor is scheduled for conversion to LEU. The actual conversion process will most likely be contracted to Atomic Energy of Canada Limited (AECL). Preliminary calculations have indicated that the total activity of used HEU core in Jamaica (∼8 TBq) should be about half that of the Montreal used HEU core. There is sufficient infrastructure both onsite and offsite to maneuver the loaded transportation flask to the shipping vessel. Appropriate licenses for the importation of the new fuel and exportation of the used fuel will be applied for once a provisional timetable has been established. (author)

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

    International Nuclear Information System (INIS)

    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)

  9. 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)

  10. Planning a new research reactor for AECL: The MAPLE-MTR concept

    International Nuclear Information System (INIS)

    AECL Research is assessing its needs and options for future irradiation research facilities. A planning team has been assembled to identify the irradiation requirements for AECL's research programs and compile options for satisfying the irradiation requirements. The planning team is formulating a set of criteria to evaluate the options and will recommend a plan for developing an appropriate research facility. Developing the MAPLE Materials Test Reactor (MAPLE-MTR) concept to satisfy AECL's irradiation requirements is one option under consideration by the planning team. AECL is undertaking this planning phase because the NRU reactor is 35 years old and many components are nearing the end of their design life. This reactor has been a versatile facility for proof testing CANDU components and fuel designs because the CANDU irradiation environment was simulated quite well. However, the CANDU design has matured and the irradiation requirements have changed. Future research programs will emphasize testing CANDU components near or beyond their design limits. To provide these irradiation conditions, the NRU reactor needs to be upgraded. Upgrading and refurbishing the NRU reactor is being considered, but the potentially large costs and regulatory uncertainties make this option very challenging. AECL is also developing the MAPLE-MTR concept as a potential replacement for the NRU reactor. The MAPLE-MTR concept starts from the recent MAPLE-X10 design and licensing experience and adapts this technology to satisfy the primary irradiation requirements of AECL's research programs. This approach should enable AECL to minimize the need for major advances in nuclear technology (e.g., fuel design, heat transfer). The preliminary considerations for developing the MAPLE-MTR concept are presented in this report. A summary of AECL's research programs is presented along with their irradiation requirements. This is followed by a description of safety criteria that need to be taken into

  11. Performance of small reactors at universities for teaching, research, training and service (TRTS): thirty five years' experience with the Dalhousie University SLOWPOKE-2 reactor

    International Nuclear Information System (INIS)

    The Dalhousie University SLOWPOKE-2 Reactor (DUSR) facility, operated during 1976-2011, was the only research reactor in Atlantic Canada as well as the only one associated with a chemistry department in a Canadian university. The most outstanding features of the facility included: a rapid (100 ms) cyclic pneumatic sample transfer system, a permanently installed Cd-site, and a Compton-suppression gamma-ray spectrometer. The usage encompassed fundamental as well as applied studies in various fields using neutron activation analysis (NAA). The facility was used for training undergraduate/graduate students, postdoctoral fellows, technicians, and visiting scientists, and for cooperative projects with other universities, research organizations and industries. (author)

  12. Safety re-assessment of AECL test and research reactors

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor Maple-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX, and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX AND NRU). ZED-2 and PTR are operated by the Reactor Physics Branch; all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis, and the plan for future analysis. In addition, during the last four years safety assessments have been carried out on the new isotope-, heat-, and electricity-producing reactors, as part of the safety design review, commissioning and licensing activities

  13. Use of the Slowpoke-2 nuclear reactor at the Royal Military College of Canada for book conservation

    International Nuclear Information System (INIS)

    The present project investigated the use of the mixed radiation field produced by the SLOWPOKE-2 reactor to prolong the life of biodeteriorated books. Research into past studies of radiation treatment indicated that the primary biodeteriorating agents, insects and moulds, can be reduced enough to return books to the 'natural' level of infestation with a dose of 2-3kGy where they will age in a manner consistent with a 'normal' book. Based on research of the potential negative effects of irradiation on paper, including depolymerization, loss of paper strength and durability, discoloration, and harm to ink, it was found that at doses below 8kGy, at a dose rate of 2.4kGy, there is no serious harm to the paper. Based on a desired dose range of 2 to 8kGy, and the dimensions and flux mapping of the radiation field in the reactor pool, a 60cm x 58cm x 43.5cm vacuum-sealed box, with a Cadmium foil neutron shield, is proposed. A preliminary feasibility study suggests that the capital and operating costs of this irradiation procedure would be approximately C$15000 and C$600, respectively. (author)

  14. 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)

  15. The Jamaican SLOWPOKE-2 research reactor: neutron activation analysis in environmental and health studies

    International Nuclear Information System (INIS)

    In its 24 years of existence the reactor has been utilized mainly for Neutron Activation Analysis (NAA) and has played an important role in the development of research programs in the areas of archaeology, biology, chemistry, forensics, geochemistry, and mining as well as for the production of short lived radioisotopes for experimental work in the physics department. However, over the last fifth teen years our main thrust has been environmental geochemistry, agriculture and health related studies, with interesting results that have implications for land use, farming practices, diabetic control and dietary intakes during pregnancy. (author)

  16. INAA of trace elements in biological materials using the SLOWPOKE-2 reactor in Jamaica

    International Nuclear Information System (INIS)

    The biological standard reference materials Orchard Leaves SRM 1571 and Oyster Tissue SRM 1566a was analyzed by instrumental neutron activation analysis (INAA) at the International Centre for Environmental and Nuclear Sciences, Jamaica at (ICEN) and at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), Brazil. The comparison of the results with those obtained with the more powerful reactor are used to evaluate the possibilities of INAA for the analysis of biological samples at ICENS. The detection limits, the precision and accuracy of the results obtained in both laboratories are compared. The advantages and disadvantages of the different irradiation facilities are discussed. Some results obtained for Jamaican biological samples are also presented. (author)

  17. Radiochemicals

    International Nuclear Information System (INIS)

    In this catalogue those radioactive chemicals for research are listed which are produced by the Radiochemical Centre Amersham and our laboratories at Brunswick. The dates given for each product can understandably only be limited within the framework of such a catalogue. Additional dates and references to application technique can be obtained from us any time. Our programme is continually updated by new products. If a compound not listed in the catalogue should be required we ask for inquiry. Our working team for special syntheses will try to produce it according to our possibilities and our requirements. (orig.)

  18. Experimental and computational determination of radiation dose rates in the SLOWPOKE-2 research reactor at the Royal Military College of Canada

    International Nuclear Information System (INIS)

    The first SLOWPOKE-2 research reactor designed to use Low Enriched Uranium (LEU) dioxide fuel was commissioned at the Royal Military College of Canada/College militaire royal du Canada in September 1985. The enrichment is 19.89% in 235U . The reactor is a pool-type design, moderated and cooled by light water with the core located at the bottom of a 5.87 m pool, ensuring a 4.42 m layer of water above the core to provide the necessary radiation shielding. Cooling is by means of natural convection, the reactor power being limited to 20 kWth. The design allows free access to the reactor pool: in addition to the pneumatic irradiation system permitting the positioning of small samples within the beryllium reflector and close to the core, larger samples can be positioned in the pool against the reactor vessel using an 'elevator'-type positioning device. Several research projects have taken and are taking advantage of this equipment, most notably investigations of intense radiation effects on advanced polymers such as epoxies and poly-ether-ether-ketones (PEEKs). These research activities however require sound knowledge of the dose rates at the various irradiation sites within the reactor vessel, as well as at incremental positions within the pool. Accurate measurements of the particle fluxes were attempted before, but unfortunately yielded limited information due to the complexity of the composite radiation field and the capabilities of the instrumentation at that time. The present work aims at yet another attempt to gather sound dose rate data, using improved radiation detectors on one hand, and better computational resources (both hardware and software) on the other hand. All research on particle dose rates was performed with the reactor at steady-state half power (10 kWth) and at the reactor core mid-height. (author)

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

    International Nuclear Information System (INIS)

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

  20. Radiochemical characterisation of graphite from Juelich experimental reactor (AVR)

    International Nuclear Information System (INIS)

    Graphite built-in nuclear reactors may receive a high neutron dose for a long period. Depending on its chemical composition a lot of activation products are produced. In addition, there is more or less fission product contamination depending on the location. The migration of fission products may be supported by high temperatures which occur in high temperature reactors. At the Juelich 15 MWe High Temperature Gas-cooled experimental Reactor AVR (Arbeitsgemeinschaft Versuchsreaktor) two different types of nuclear graphite had been in use. High-purity graphite was used as basic material for core structures of the AVR. Insulation layers from carbon bricks (graphite with larger amounts of impurities) surrounding the graphite reflector were used to protect the metallic structures from high temperatures. For many reasons it is important to know the amount of contamination of graphite and carbon bricks with activation products and fission products. The head end of nuclear graphite analytics must be the incineration. Volatile activities (14C, 3H, 36Cl ...) must be caught for determination. In case of handling dustlike samples the incineration furnace must be small enough to be operated in a glove box. The resulting ashes can be used for determining all non volatile nuclides with different radiochemical methods. In early 1999 some graphite and carbon brick samples from AVR-reactor had been taken by drilling. The samples had been analysed in our laboratories at Juelich research centre. For incineration we used a vertical quartz-tube which dips at the bottom into a small electric furnace. Tritium, 14C and 36Cl are caught in washing bottles. After further preparation, they are determined by LSC. After dissolving the ashes, the elements are separated by ion exchange, extraction methods and HPLC. The radionuclides are then determined by a-spectrometry, LSC, low level g-spectrometry and x-ray spectrometry. (author)

  1. Radiochemical characterization of graphite from Juelich experimental reactor (AVR)

    International Nuclear Information System (INIS)

    Nuclear reactors which have in-built graphite may receive a high neutron dose for a long period. Depending on the chemical composition of the graphite, numerous activation products may result. In addition, the amount of fission product contamination will depend on the location of the graphite. The migration of fission products may be supported by the high temperatures which occur in high-temperature reactors. At the Juelich 15 MWe high-temperature gas-cooled experimental AVR (Arbeitsgemeinschaft Versuchsreaktor) reactor, two different types of nuclear graphite had been in use. High-purity graphite was used as a basic material for core structures of the AVR. Insulation layers of carbon bricks (graphite with larger amounts of impurities) surrounding the graphite reflector were used to protect the metallic structures from high temperatures. For various reasons it is important to know the degree of contamination of graphite and carbon bricks from activation and fission products. The optimum method for nuclear graphite analysis in decommissioning is by incineration. Volatile activities (14C, 3H, 36Cl, ...) have to be captured for analysis. In cases where dust-like samples are handled, the incineration furnace has to be small enough to be operated in a glove-box. The resulting ashes can be used for determining all non-volatile nuclides by different radiochemical methods. In early 1999, some graphite and carbon brick samples from the AVR reactor were obtained by drilling. The samples were then analysed in the laboratories at the Juelich research centre. For incineration a vertical quartz tube was used which dips at the bottom into a small electric furnace. Tritium, 14C and 36Cl were captured in washing bottles. After further preparation, they were analysed by liquid scintillation counting (LSC). After dissolving the ashes, the elements were separated by ion exchange, extraction methods and HPLC. The radionuclides were then determined by alpha-spectrometry, LSC, low

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

    International Nuclear Information System (INIS)

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

  3. Various applications using the SLOWPOKE-2 facility at RMC

    International Nuclear Information System (INIS)

    History will record that the reactor pool at the SLOWPOKE-2 Facility at RMC was one of the first SLOWPOKE pools to be constructed (mid 1970s), even though the reactor itself was the last SLOWPOKE reactor to be installed and commissioned (1985). The unique and very useful feature of the reactor pool is that it is uncovered, allowing for applications in addition to the NAA and radioisotope production applications initially advertised. Because the installation of a tangential neutron beam tube (NBT) had been planned from the beginning, an outer irradiation site inside the reactor container was replaced by a thermal column. Next, a positioning system was added to accept large objects such as flight control surfaces from DND's CF-18 fighter aircraft. Imaging of these surfaces using film is being phased out with the introduction of digital imaging. Very recently a tomography stage was designed and built and is now integrated into the neutron imaging system. Also in the open pool are three pulley and rope 'elevators', two of which allow for large samples to be exposed to various kinds of radiation directly outside of the reactor container. The third elevator is located against the west pool wall, which allows for sample exposure to radiation without any neutron contribution. At the time of negotiating the purchase of the reactor, a teaching package consisting of an in-pool borated ion chamber and an outlet thermocouple was ordered. Automatic irradiation and counting systems in the form of cyclic, pseudo-cyclic, and long counting options were added to the original manual irradiation option. This past summer (2010), a delayed neutron counting system (DNCS) was built and installed in the SLOWPOKE-2 Facility at RMC. Examples will be given for the above-mentioned applications.

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

    Energy Technology Data Exchange (ETDEWEB)

    Zeller, B., E-mail: bzeller@eclipsescientific.com [Eclipse Scientific Ltd., Waterloo, Ontario (Canada); Lombardi, L., E-mail: llombardi@utex.com [Utex Scientific Instruments, Mississauga, Ontario (Canada); Cyr, P., E-mail: pcyr@eclipsescientific.com [Eclipse Scientific Ltd., Waterloo, Ontario (Canada); Mair, H.D., E-mail: dmair@utex.com [Utex Scientific Instruments, Mississauga, Ontario (Canada); Ginzel, R., E-mail: rginzel@eclipsescientific.com [Eclipse Scientific Ltd., Waterloo, Ontario (Canada)

    2013-01-15

    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)

  5. 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)

  6. Radiochemical guidelines and process specifications for reactor shutdown: the EDF strategy

    International Nuclear Information System (INIS)

    Changes to French nuclear regulations made in June 2006 [1.] have made it necessary for EDF to modify its ruling principles. These modifications required the restructuring of radiochemical guidelines to better reflect their impact on nuclear safety, the environment and radioprotection. In accordance with these aims, a new authoritative document has been produced. This ruling document identifies all parameters with a potential impact on nuclear safety, radiological releases to the environment and personnel dose rates. These diagnostic and control parameters have been identified for a reactor in production and for a reactor during shutdown. For parameters related to a reactor in production, some indicators are used to evaluate impacts on availability, radioprotection and the environment during shutdown and on outage and to anticipate mitigation ways. On the other side, several parameters related to the stages of shutdown were also directly evaluated in order to minimize the impacts. This paper describes the EDF methodology used to establish operational documents: radiochemical guidelines and process specifications, and includes the following: - description of monitored parameters and their associated areas of risk; - justification of target values, frequencies of inspection and the required actions for the monitored parameters. The sizing methodology is based on theoretical studies and on EDF operational experience analysis. By implementing in the operational and technical specifications requirements linked to nuclear safety, radioprotection and environment respect, EDF will benefit from an improved compromise between these areas as well as an increased focus. (authors)

  7. Royal Military College of Canada SLOWPOKE-2 facility. Integrated regulating and instrumentation system (SIRCIS) upgrade project

    International Nuclear Information System (INIS)

    The SLOWPOKE-2 Facility at the Royal Military College of Canada has operated the only digitally controlled SLOWPOKE reactor since 2001 (Version 1.0). The present work describes ongoing project development to provide a robust digital reactor control system that is consistent with Aging Management as summarized in the Facility's Life Cycle Management and Maintenance Plan. The project has transitioned from a post-graduate research activity to a comprehensively managed project supported by a team of RMCC professional and technical staff who have delivered an update of the V1.1 system software and hardware implementation that is consistent with best Canadian nuclear industry practice. The challenges associated with the implementation of Version 2.0 in February 2012, the lessons learned from this implementation, and the applications of these lessons to a redesign and rewrite of the RMCC SLOWPOKE-2 digital instrumentation and regulating system (Version 3) are discussed. (author)

  8. Characterization of filter cartridges from the IEA-R1 reactor by radiochemical method

    International Nuclear Information System (INIS)

    The filter cartridges used in water purification system of research nuclear reactor IEA-R1 are considered radioactive wastes after their useful life. The characterization of these wastes is one of the stages of management, which aims to identify and quantify the radionuclides present, including those known as 'difficult to measure' (DTM) radionuclides. Establish a radiochemical analysis methodology for this type of waste is a difficult job, not only by the application of these techniques, but also by the amount of radionuclides that should be analyzed. In the waste produced in a nuclear reactor, the most important radionuclides are fission products, activation products and transuranic elements. Since these radionuclides emit gamma radiation not measurable in its decay process and consequently are difficult to measure, their concentrations can be estimated by indirect methods such as scale factors. This method is used to evaluate the DTM concentration, which is represented by alpha and beta nuclides using the correlation between them and the radionuclide key, a gamma emitter. The objective of this work is to describe a radiochemical analysis methodology for gamma emitter nuclides, present in the filter cartridges, evaluating the activity and concentrations by destructive assays. At the same time, two studies have been performed by non-destructive assays, the first one based on dose rates and the point kernel method to correlate the results and the second one based on calibration efficiency with Monte Carlo method. These studies belong to the radioactive waste characterization program that has been conducted at the Waste Management Laboratory of Nuclear and Energy Research Institute, IPEN-CNEN/SP. (author)

  9. District heating with SLOWPOKE energy systems

    International Nuclear Information System (INIS)

    The SLOWPOKE Energy System, a benign nuclear heat source designed to supply 10 thermal megawatts in the form of hot water for local heating systems in buildings and institutions, is at the forefront of these developments. A demonstration unit has been constructed in Canada and is currently undergoing an extensive test program. Because the nuclear heat source is small, operates at atmospheric pressure, and produces hot water below 100 degrees Celcius, intrinsic safety features will permit minimum operator attention and allow the heat source to be located close to the load and hence to people. In this way, a SLOWPOKE Energy System can be considered much like the oil- or coal-fired furnace it is designed to replace. The low capital investment requirements, coupled with a high degree of localization, even for the first unit, are seen as attractive features for the implementation of SLOWPOKE Energy Systems in many countries

  10. N.S. Savannah Reactor Vessel Metal Extraction and Radiochemical Analysis

    International Nuclear Information System (INIS)

    In early 2006 a project was concluded to determine radioisotopic inventory and Curie content of the N.S. Savannah Reactor Pressure Vessel (RPV), Internals and Neutron Shield Tank (NST) by extracting metal samples and performing radiochemical analysis. The objective of this project was to determine if the RPV and internals could be removed, packaged, shipped and disposed as Class A radioactive waste without opening the RPV or conducting further sampling of the RPV/Internals. The N.S. Savannah is de-fueled and has been shut down for 37 years. The following conclusions can be drawn from this project: - Results are consistent with previous analyses and are based upon conservative methodology and assumptions. - Nuclide concentration for the N/S Savannah reactor pressure vessel and internals package are shown to be within Class A disposal limits when averaged over the entire volume of metal in the Reactor Pressure Vessel and internals. - Performance of N.S. Savannah's nuclear reactor was excellent. During normal operations, the reactor seldom operated above 80% of its rated power level, thereby minimizing thermal stresses on the fuel cladding. In addition, the fuel rods were not subjected to any accident or severe transient conditions that could result in cladding breeches with subsequent release of fission products and fuel particles to the primary coolant loop. The trace quantities of Cesium-137 observed in the primary loop water indicate that some pinhole penetrations of fuel rod cladding may have occurred during operations. Another source of Cesium-137 could be the presence of uranium fuel on the exterior of the fuel rod cladding (tramp uranium), a condition not uncommon in the N.S. Savannah fuel fabrication time frame. Fissioning of this 'tramp uranium' would cause the rapid release of chemically active Cesium-137 into the reactor coolant. However, the absence of other fission products (e.g., Strontium-90) as well as uranium and transuranic isotopes in the reactor

  11. 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

  12. 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

  13. AECL sees improved earnings potential

    International Nuclear Information System (INIS)

    AECL (formerly Atomic Energy of Canada Limited) says it has turned the corner in the nuclear generating plant market, and predicts rising profits in the years to come. In its latest five year plan, the Canadian Crown corporation predicts sales revenues over the next four years will triple, to $666 million in 1995/96 from $191 million in the current fiscal year while profit is expected to jump to $39 million from $5 million. Late in 1990, the company signed a $400 million contract to build a second nuclear power plant at Wolsong, in South Korea, and says the South Korean government has expressed an interest in adding two more units to the site, which already has one operating CANDU reactor

  14. 177Lu radiochemical separation from 176Yb irradiated in high-flux research reactor SM

    International Nuclear Information System (INIS)

    Ytterbium and lutetium behaviour has been studied during electrolysis of aqueous solutions containing their chlorides and alkali metal citrate (Li, Na, K). The conditions providing the efficient extraction of ytterbium macro amounts into a mercury-pool cathode have been determined. Laboratory-scale experiments were performed to elaborate chromatographic procedures for 177Lu purification from ytterbium macro amounts and accompanying impurities including hafnium (177Lu radioactive decay product). The conditions providing the efficient separation of 177Lu from the above-mentioned impurities using cation-exchange (in α-hydroxy isobutyric acid) and extraction-chromatographic (impregnated with di-2-ethylhexyl phosphoric acid teflon powder as stationary phase and nitric or hydrochloric acids as eluant) methods have been found. Isotopically enriched ytterbium preparation (176Yb - 95.15; 174Yb - 2.47 atomic %) was purified from lutetium impurity and samples of the purified starting material were irradiated in the central neutron trap and beryllium reflector channel of the SM reactor. 177Lu was extracted from the irradiated targets by electroreduction of ytterbium on the mercury-pool cathode from lithium citrate solution. Cation exchange and extraction chromatography methods were used for subsequent purification of 177Lu. The radiochemical processing took about 50 hours. The results of analysis obtained by the spectrometry of X-ray and gamma radiation, mass-spectrometry and emission spectroscopy are as follows: Chemical form: 177LuCl3, solution in hydrochloric acid; solvent (HCl) concentration: 0.01 - 0.1 mol/l; 177Lu specific activity: ≥ 20 Ci/mg; 177mLu to 177Lu activity ratio: ≤ 0.02 %; total gamma emitters (Co-58, Co-60, Zn-65, Mn-54, Fe-59, Cr-51) to 177Lu activity ratio: ≤ 0.01 %; total mass of non-radioactive impurities (Cu, Zn, Al, Fe, Pb) to 177Lu activity ratio: ≤ 500 μg/Ci; total alpha emitters to 177Lu activity ratio ≤ 1x10-5 %. (author)

  15. 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

  16. AECL's plant Information Technologies

    International Nuclear Information System (INIS)

    The competitiveness of the world-wide energy market is a continual driving force for improvements to CANDU performance and lower operating, maintenance, and administration costs. As in other industries, advanced Information Technologies (IT) are changing the way we work and conduct business. The nuclear industry is no different and there exists strong incentives to improve work processes and provide faster and more flexible access to the information needed to effectively manage and maintain nuclear plant assets. AECL has responded to these forces through the development of a vision of integrated IT systems addressing all phases of nuclear plant development and operations. This includes the initial engineering, design, and construction processes as well as support to the long-term operations and maintenance. Integral to the AECL vision is the need for cost-effective engineering and operational configuration management systems, proactive maintenance processes and systems, and advanced plant surveillance and diagnostics. This paper presents the vision and describes the integrated information systems needed to manage both the design basis and operating plant data systems to ensure the cost-effective, long-term viability of CANDU plants. (author)

  17. 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

  18. 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)

  19. 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)

  20. AECL annual review 1992 - 1993

    International Nuclear Information System (INIS)

    1992/93 was a pivotal year for AECL, with the redirection of its strategic plan, the refocussing of its corporate mission, a change in its structural organization to meet new challenges, the contract with South Korea for Wolsong Units 3 and 4 and the Memorandum of Understanding with Saskatchewan. AECL looks forward to the next 12 months as a time of opportunity, confident in the knowledge that they possess the means to succeed. ills

  1. 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)

  2. Radiochemical procedures

    International Nuclear Information System (INIS)

    The modern counting instrumentation has largely obviated the need for separation processes in the radiochemical analysis but problems in low-level radioactivity measurement, environmental-type analyses, and special situations caused in the last years a renaissance of the need for separation techniques. Most of the radiochemical procedures, based on the classic works of the Manhattan Project chemists of the 1940's, were published in the National Nuclear Energy Series (NNES). Improvements such as new solvent extraction and ion exchange separations have been added to these methods throughout the years. Recently the Los Alamos Group have reissued their collected Radiochemical Procedures containing a short summary and review of basic inorganic chemistry - 'Chemistry of the Elements on the Basis of Electronic Configuration'. (A.L.)

  3. Annual report 1995-1996. AECL research No. AECL-11577

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-01

    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.

  4. 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

  5. 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.

  6. Development of advanced ceramics at AECL

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has a long history of developing ceramics for nuclear fission and fusion applications. AECL is now applying its multidisciplinary materials R and D capabilities, including unique capabilities in ceramic processing and nondestructive evaluation, to develop advanced ceramic materials for commercial and industrial applications. This report provides an overview of the facilities and programs associated with the development of advanced ceramics at AECL

  7. AECL's Experimental Fuel and Materials Test Loops in NRU

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Ltd (AECL) maintains two experimental fuel and materials test loops, U1 and U2, within the National Research Universal (NRU) reactor at Chalk River Laboratories (CRL). These loops operate at conditions typical of CANDU reactors. Each vertical test section (one in U1, two in U2) has the capacity to irradiate a test assembly 3 m in length and 10.16 cm in diameter; equivalent to six CANDU fuel bundles. The assembly is made up of six interchangeable bundles containing experimental fuels or materials test specimens. The fuel bundles can be 'fixed', with elements welded together into a rigid bundle, or 'demountable', where a frame with some fixed elements and element mounting mechanisms facilitate the placement of additional removable fuel elements. The materials test bundle has 30 fuel elements surrounding a 4.0 cm diameter tube. Currently, there are two specimen holder designs which fit within the tube: a ring of six 13.1 mm diameter specimen tubes, and a triangular assembly, 2.9 cm per side. In addition to standard fuel and materials irradiations, AECL has also performed instrumented test irradiations with modified test assemblies in the NRU Loops. The instrumented test irradiations were conducted in the Blowdown Test Facility (BTF; formerly part of U1) which simulated accident scenarios. AECL has recently qualified a new top closure plug for use with chemistry experiments in the loops. The plug provides electrical connections between instruments and the data acquisition hardware, through the pressure boundary, which will facilitate instrumented irradiations. In addition, an online gamma spectrometer is being added to the U2 loop to monitor loop coolant gamma activity and to facilitate fuel defect detection and characterization. The Canadian Supercritical Water Reactor (SCWR) fuel design will require irradiation testing. The reference fuel design, (Th, Pu)O2 fuel with high Pu content (13%), will require supporting fuel irradiations. AECL plans

  8. Follow-up of AECL employees involved in the decontamination of NRU in 1958

    International Nuclear Information System (INIS)

    In May 1958 the NRU reactor hall was badly contaminated by a damaged fuel rod that broke apart during its removal from the reactor. Radioactive fission products were spread around the reactor hall and into adjacent areas when a piece of the fuel rod fell into the maintenance pit and burned. AECL staff and others completed the decontamination in 2 1/2 months. This paper reports the results of a follow-up study of the AECL participants. No statistically significant increases in deaths from cancer or other diseases were found in this group

  9. Radiochemical methods

    International Nuclear Information System (INIS)

    This little volume is one of an extended series of basic textbooks on analytical chemistry produced by the Analytical Chemistry by Open Learning project in the UK. Prefatory sections explain its mission, and how to use the Open Learning format. Seventeen specific sections organized into five chaptrs begin with a general discussion of nuclear properties, types, and laws of nuclear decay and proceeds to specific discussions of three published papers (reproduced in their entirety) giving examples of radiochemical methods which were discussed in the previous chapter. Each section begins with an overview, contains one or more practical problems (called self-assessment questions or SAQ's), and concludes with a summary and a list of objectives for the student. Following the main body are answers to the SAQ's, and several tables of physical constants, SI prefixes, etc. A periodic table graces the inside back cover

  10. 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.

  11. 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.

  12. Radiochemical determination of the neutron capture cross sections of {sup 241}Am irradiated in the JMTR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shinohara, N.; Hatsukawa, Y.; Hata, K.; Kohno, N. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    The thermal neutron capture cross section {sigma}{sub 0} and Resonance integral I{sub 0} of {sup 241}Am leading to the production of {sup 242m}Am and {sup 242g}Am were measured by radiochemical method. The cross sections obtained in this study are {sigma}{sub 0}=60.9 {+-} 2.6 barn, I{sub 0}=213 {+-} 13 barn for {sup 241}Am(n,{gamma}){sup 242m}Am and {sigma}{sub 0}=736 {+-} 31 barn, I{sub 0}=1684 {+-} 92 barn for {sup 241}Am(n,{gamma}){sup 242g}Am. (author)

  13. 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

  14. A nuclear reactor for district heating

    International Nuclear Information System (INIS)

    Global energy requirements are expected to double over the next 40 years. In the northern hemisphere, many countries consume in excess of 25 percent of their primary energy supply for building heating. Satisfying this need, within the constraints now being acknowledged for sustainable global development, provides an important opportunity for district heating. Fuel-use flexibility, energy and resource conservation, and reduced atmospheric pollution from acid gases and greenhouse gases, are important features offered by district heating systems. Among the major fuel options, only hydro-electricity and nuclear heat completely avoid emissions of combustion gases. To fill the need for an economical nuclear heat source, Atomic Energy of Canada Limited has designed a 10 MW plant that is suitable as a heat source within a network or as the main supply to large individual users. Producing hot water at temperatures below 100 degrees C, it incorporates a small pool-type reactor based on AECL's successful SLOWPOKE Research Reactor. A 2 MW prototype for the commercial unit is now being tested at the Whiteshell Nuclear Research Establishment in Manitoba. With capital costs of $7 million (Canadian), unit energy costs are projected to be $0.02/kWh for a 10 MW unit operating in a heating grid over a 30-year period. By keeping the reactor power low and the water temperature below 100 degrees C, much of the complexity of the large nuclear power plants can be avoided, thus allowing these small, safe nuclear heating systems to be economically viable

  15. DEMONSTRATION SOLIDIFICATION TESTS CONDUCTED ON RADIOACTIVELY CONTAMINATED ORGANIC LIQUIDS AT THE AECL WHITESHELL LABORATORIES

    International Nuclear Information System (INIS)

    The AECL, Whiteshell Laboratory (WL) near Pinawa Manitoba, Canada, was established in the early 1960's to carry out AECL research and development activities for higher temperature versions of the CANDU(regsign) reactor. The initial focus of the research program was the Whiteshell Reactor-1 (WR-1) Organic Cooled Reactor (OCR) that began operation in 1965. The OCR program was discontinued in the early 1970's in favor of the successful heavy-water-cooled CANDU system. WR-1 continued to operate until 1985 in support of AECL nuclear research programs. A consequence of the Federal government's recent program review process was AECL's business decision to discontinue research programs and operations at the Whiteshell Laboratories and to consolidate its' activities at the Chalk River Laboratories. As a result, AECL received government concurrence in 1998 to proceed to plan actions to achieve closure of WL. The planning actions now in progress address the need to safely and effectively transition the WL site from an operational state, in support of AECL's business, to a shutdown and decommissioned state that meets the regulatory requirements for a licensed nuclear site. The decommissioning program that will be required at WL is unique within AECL and Canada since it will need to address the entire research site rather than individual facilities declared redundant. Accordingly, the site nuclear facilities are being systematically placed in a safe shutdown state and planning for the decommissioning work to place the facilities in a secure monitoring and surveillance state is in progress. One aspect of the shutdown activities is to deal with the legacy of radioactively contaminated organic liquid wastes. Use of a polymer powder to solidify these organic wastes was identified as one possibility for improved interim storage of this material pending final disposition

  16. 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)

  17. 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

  18. Radiochemical solar neutrino experiments

    International Nuclear Information System (INIS)

    Radiochemical experiments have been crucial to solar neutrino research. Even today, they provide the only direct measurement of the rate of the proton-proton fusion reaction, p+p→d+e++νe, which generates most of the Sun's energy. We first give a little history of radiochemical solar neutrino experiments with emphasis on the gallium experiment SAGE - the only currently operating detector of this type. The combined result of all data from the Ga experiments is a capture rate of 67.6±3.7 SNU. For comparison to theory, we use the calculated flux at the Sun from a standard solar model, take into account neutrino propagation from the Sun to the Earth and the results of neutrino source experiments with Ga, and obtain 67.3-3.5+3.9 SNU. Using the data from all solar neutrino experiments we calculate an electron neutrino pp flux of φpp♁=(3.41-0.77+0.76)×1010/(cm2-s), which agrees well with the prediction from a detailed solar model of φpp♁=(3.30-0.14+0.13)×1010/(cm2-s). Four tests of the Ga experiments have been carried out with very intense reactor-produced neutrino sources and the ratio of observed to calculated rates is 0.88±0.05. One explanation for this unexpectedly low result is that the cross section for neutrino capture by the two lowest-lying excited states in 71Ge has been overestimated. We end with consideration of possible time variation in the Ga experiments and an enumeration of other possible radiochemical experiments that might have been.

  19. Chemical and radiochemical constituents in water from wells in the vicinity of the Naval Reactors Facility, Idaho National Engineering Laboratory, Idaho, 1994--95

    International Nuclear Information System (INIS)

    The US Geological Survey, in response to a request from the US Department of Energy's Pittsburgh Naval Reactors Office, Idaho Branch Office, sampled water from 14 wells during 1994--95 as part of a long-term project to monitor water quality of the Snake River Plain aquifer in the vicinity of the Naval Reactors Facility, Idaho National Engineering Laboratory, Idaho. Water samples were analyzed for naturally occurring constituents and manmade contaminants. A total of 111 samples were collected from 10 monitoring wells and 4 production wells. Twelve quality-assurance samples also were collected and analyzed; 1 was a blank sample and 11 were replicate samples. The blank sample contained concentrations of one inorganic constituent, one organic constituent, and five radioactive constituents that were greater than the reporting levels. Concentrations of other constituents in the blank sample were less than their respective reporting levels. The 11 replicate samples and their respective primary samples generated 293 pairs of analytical results for a variety of chemical and radiochemical constituents. Of the 293 data pairs, 258 were statistically equivalent at the 95-percent confidence level; about 88 percent of the analytical results were in agreement

  20. Evaluation of ENDF/B-VI library with WIMS-AECL/RFSP Code system

    International Nuclear Information System (INIS)

    The object of this research is the evaluation of the cross-section charicteristics of ENDF/B-VI WIMS-AECL library against ENDF/B-V library previously used in the validation of WIMS-AECL code. validation of WIMS-AECL code had been carried out through the Phase-B post simulation of Wolsong Units 2, 3 and 4 before. Discrepancies between the calculated and measured values were thought to be mainly from observation errors and partly from the ENDF/B-V library. Till now, there had been various validation calculations for ENDF/B-VI library in the field of PWR but not in CANDU-PHWR. We herein, evaluated the ENDF/B-VI WIMS-AECL library for Wolsong Unit 4 by comparing the results with previous ones of ENDF/B-V for the same reactor unit with same WIMS/RFSP code system. It can be summarized that the Phase-B post simulation results of WIMS/RFSP with ENDF/B-VI are better than those of ENDF/B-V, because of less difference between calculated and measured values. There must be further study with different core conditions, however, for the exact evaluation of ENDF/B-VI WIMS-AECL library including calculations of many other physical parameters and the treatment of isotopes which is not in ENDF/B-VI but in ENDF/B-V

  1. Some highlights of research and development at AECL

    International Nuclear Information System (INIS)

    The research and development programs of AECL have as their goal the strengthening of the knowledge and ability necessary to achieve national objectives in the field of nuclear energy. These objectives include a nuclear reactor system appropriate to Canada's industrial capabilities, now realized, and the extension of that system, through scientific and technological development, to serve the nation's needs for the forseeable future. The Company's programs are carefully integrated and focused to use the available funding to maximum advantage. The research facilities on which the program depends are among the best in the world, and support a full spectrum of research from fundamental nuclear physics to full-scale power reactor component irradiation and testing. In this report it has only been possible to high-light some important facets of the programs in each of the principal areas currently employing our energies. (auth)

  2. 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.

  3. AECL: 60 years of contributing to Canada

    International Nuclear Information System (INIS)

    This paper traces the history of the Atomic Energy of Canada Limited. For 60 years AECL has contributed world class science and technology to Canada, while assisting Government on policy issues, enabling business innovation and technology transfer, and generating highly qualified workforce for Canadian industry.

  4. 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

  5. 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.

  6. 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)

  7. 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 analysis 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)

  8. Radiochemical analysis. Chapter 6

    International Nuclear Information System (INIS)

    A brief description is presented of sample selection and preparation, of radiochemical analysis principles, and of the determinations of tritium, phosphorus, strontium, zirconium, niobium, ruthenium, iodine, cesium, barium, cerium, polonium, radium, thorium, uranium, plutonium, and of rare earths. The occurrence of radionuclides in the environment and the prospects of radiochemical method applications are briefly described. (J.P.)

  9. Validation of MCNP and WIMS-AECL/DRAGON/RFSP for ACR-1000 applications

    Energy Technology Data Exchange (ETDEWEB)

    Bromley, Blair P.; Adams, Fred P.; Zeller, Michael B.; Watts, David G.; Shukhman, Boris V.; Pencer, Jeremy [AECL - Chalk River Laboratories, Chalk River (Canada)

    2008-07-01

    This paper gives a summary of the validation of the reactor physics codes WIMS-AECL, DRAGON, RFSP and MCNP5, which are being used in the design, operation, and safety analysis of the ACR-1000{sup R}. The standards and guidelines being followed for code validation of the suite are established in CSA Standard N286.7-99 and ANS Standard ANS-19.3-2005. These codes are being validated for the calculation of key output parameters associated with various reactor physics phenomena of importance during normal operations and postulated accident conditions in an ACR-1000 reactor. Experimental data from a variety of sources are being used for validation. The bulk of the validation data is from critical experiments in the ZED-2 research reactor with ACR-type lattices. To supplement and complement ZED-2 data, qualified and applicable data are being taken from other power and research reactors, such as existing CANDU{sup R} units, FUGEN, NRU and SPERT research reactors, and the DCA critical facility. MCNP simulations of the ACR-1000 are also being used for validating WIMS-AECL/ DRAGON/RFSP, which involves extending the validation results for MCNP through the assistance of TSUNAMI analyses. Code validation against commissioning data in the first-build ACR-1000 will be confirmatory. The code validation is establishing the biases and uncertainties in the calculations of the WIMS-AECL/DRAGON/RFSP suite for the evaluation of various key parameters of importance in the reactor physics analysis of the ACR-1000. (authors)

  10. 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)

  11. The LFR Radiochemical Facility

    International Nuclear Information System (INIS)

    The LFR Facility is a radiochemical laboratory designed and constructed with a hot-cells line, a glove-boxes and fume-hoods, all of them suited to work with radioactive materials. It is worth noticed the LFR capacity to carry on different research and development programs (R+D) in the Nuclear Fuel Cycle field, such as the burn up determination, absolute burn up measurement, the radiochemical analysis of different materials and solutions, the evaluation of radioactive waste immobilization processes, and researches on burnable poisons. (author)

  12. The keys to success in marketing small heating reactors

    International Nuclear Information System (INIS)

    The success of the SLOWPOKE Energy System requires acceptance of the SLOWPOKE reactor within the community where the reactor's energy is to be used. Public acceptance will be obtained once the public is convinced that this nuclear heat source is needed, safe and of economic benefit to the community. The need for a new application of nuclear energy is described and the ability of small reactors used for district heating to play that role is shown. The safety of the reactor is being demonstrated with the establishment of the SLOWPOKE Demonstration Reactor by Atomic Energy of Canada Limited and with open, candid discussion with the involved community. Economic arguments are reviewed and include discussion of quantitative and qualitative issues. (orig.)

  13. The keys to success in marketing small heating reactors

    International Nuclear Information System (INIS)

    The success of the SLOWPOKE Energy System requires acceptance of the SLOWPOKE reactor within the community where the reactor's energy is to be used. Public acceptance will be obtained once the public is convinced that this nuclear heat source is needed, safe and of economic benefit to the community. The need for a new application of nuclear energy is described and the ability of small reactors used for district heating to play that role is shown. The safety of the reactor is being demonstrated with the establishment of the SLOWPOKE Demonstration Reactor by Atomic Energy of Canada Limited and with open, candid discussion with the envolved community. Economic arguments are reviewed and include discussion of quantitative and qualitative issues. (author)

  14. AECL's experience in MOX fuel fabrication and irradiation

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited's mixed-oxide (MOX) fuel fabrication activities are conducted in the Recycle Fuel Fabrication Laboratories (RFFL) at the Chalk River Laboratories. The RFFl facility is designed to produce experimental quantities of CANDU MOX fuel for reactor physics tests or demonstration irradiations. From 1979 to 1987, several MOX fuel fabrication campaigns were run in the RFFl, producing various quantities of fuel with different compositions. About 150 bundles, containing over three tonnes of MOX, were fabricated in the RFFL before operations in the facility were suspended. In late 1987, the RFFL was placed in a state of active standby, a condition where no fuel fabrication activities are conducted, but the monitoring and ventilation systems in the facility are maintained. Currently, a project to rehabilitate the RFFL and resume MOX fuel fabrication is underway. The initial campaign will consist of the production of thirty-eight 37-element (U, Pu)O2 bundles containing 0.3 wt.% Pu in Heavy Element (H.E.) destined for physics tests in the zero-power ZED-2 reactor. An overview of AECL's MOX fuel irradiation program will be given. Post-irradiation examination results of (U,Pu)O2 bundles irradiated to burnups ranging from 18 to 49 GWd/te H.E. in the Nuclear Power Demonstration reactor are highlighted. The results demonstrated the excellent performance of CANDU MOX fuel to high burnup, at power ratings up to 45 kW/m. The paper also outlines the status of current MOX fuel irradiation tests, including the irradiation of various (U, Pu)O2 and (Th, Pu)O2 bundles. (author). 2 refs, 9 figs, 5 tabs

  15. Using deterministic methods for research reactor studies

    International Nuclear Information System (INIS)

    As an alternative to prohibitive Monte Carlo simulations, deterministic methods can be used to simulate research reactors. Using various microscopic cross section libraries currently available in Canada, flux distributions were obtained from DRAGON cell and supercell transport calculations. Then, homogenization/condensation is done to produce few-group nuclear properties, and diffusion calculations were performed using DONJON core models. In this paper, the multigroup modular environment of the code DONJON is presented, and the various steps required in the modelling of SLOWPOKE hexagonal cores are described. Numerical simulations are also compared with experimental data available for the EPM Slowpoke reactor. (author)

  16. Radiochemical separation methods

    International Nuclear Information System (INIS)

    Radiochemical separations make use of the traditional methods of separating elements: precipitation, ion exchange and solvent extraction, of which precipitation has been employed from the very beginning. A new method, accepted into wide use in the 1990s, is extraction chromatography, which combines solvent extraction as the separation method with column chromatography technology earlier used in ion exchange. These four methods are discussed in the paper. (author)

  17. The MNSR reactor

    International Nuclear Information System (INIS)

    This tank-in-pool reactor is based on the same design concept as the Canadian Slowpoke. The core is a right circular cylinder, 24 cm diameter by 25 cm long, containing 411 fuel pin positions. The pins are HEU-Aluminium alloy, 0.5 cm in diameter. Critical mass is about 900 g. The reactor has a single cadmium control rod. The back-up shutdown system is the insertion of a cadmium capsule in a core position. Excess reactivity is limited to 3.5mk. In both the MNSR and Slowpoke, the insertion of the maximum excess reactivity results in a power transient limited by the coolant/moderator temperature to safe values, independent of any operator action. This reactor is used primarily in training and neutron activation analysis. Up to 64 elements have been analyzed in a great variety of different disciplines. (author)

  18. Experience with the AECL digital console

    International Nuclear Information System (INIS)

    At the Pennsylvania State University (Penn State), it is believed that nuclear engineering students should have the opportunity to operate a nuclear reactor, perform experiments with reactor and radiation instrumentation, and perform a series of reactor physics experiments. These activities are done at the Penn State Breazeale Reactor (PSBR), a General Atomic Mark III TRIGA reactor. The PSBR, which is part of the Penn State Radiation Science and Engineering Center, is the longest operating university reactor in the US. Through periodic upgrades over its 34-yr lifetime, the facility has enabled the nuclear engineering department to familiarize students with the current instrumentation technology that they would expect to find in industry. The present reactor instrumentation and control system is >25 yr old and is no longer representative of current reactor instrumentation and control technology. The new reactor console was designed to provide the PSBR with safe and reliable operation in four different modes: manual, automatic, square wave, and pulse. It utilizes a reliable hardwired analog system for safety-related functions. All nonsafety-related functions are implemented with computer technology

  19. Radiochemical research in Switzerland

    International Nuclear Information System (INIS)

    The presentation summarizes ongoing research of the only Radiochemistry Laboratory in Switzerland, a joint unit of the Paul Scherrer Institute (PSI) and Bern University. It includes fundamental research (e.g. first ever chemical studies of heaviest elements) and applied studies (e.g. behaviour of radionuclides in liquid metal targets at spallation sources). In addition, the use of radiochemical techniques to environmental sciences is another topic of the research portfolio. Examples are nuclear dating of ice archives as well as the application of short-lived positron emitters to surface chemical investigations on aerosol particles. (author)

  20. Radiochemical synthesis of etomoxir

    International Nuclear Information System (INIS)

    Sodium 2-{6-(4-chlorophenoxy)hexyl}oxirane-2-carboxylate (Etomoxir) inhibits transport of fatty acids via the carnitine shuttle into mitochondria of muscle cells and prevents long chain fatty acids from providing energy through β-oxidation especially for muscle contraction. The objective of this synthesis is to develop a method for radioiodination of Etomoxir in order to explore its potential in diagnostic metabolic studies and molecular imaging. Thus, a method is described for the radiochemical synthesis and purification of ethyl 2-{6-(4-[131I]iodophenoxy)hexyl}oxirane-2-carboxylate (3) and 2-{6-(4-[131I]iodo-phenoxy)hexyl}oxirane-2-carboxylic acid (4). For the synthesis of these new agents, ethyl 2-{6-(4-bromophenoxy)hexyl}oxirane-2-carboxylate (1) and 2-{6-(4-bromophenoxy)hexyl}oxirane-2-carboxylic acid (2) were refluxed with [131I]NaI in the presence of anhydrous acetone at a temperature of 80 oC and 90 oC for a period of 3-4 hours, respectively. The method of radiolabeling, based on the nucleophilic exchange reaction, resulted in a radiochemical yield of 43% and 67% for compounds 3 and 4, respectively. This paper reports on the labeling of etomoxir with radioiodine as 124I labeled etomoxir may be of great importance in molecular imaging.

  1. 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)

  2. Verification of the AECL total system performance models

    International Nuclear Information System (INIS)

    An agreement between Atomic Energy of Canada, Limited (AECL) and the United States Department of Energy (USDOE) defines eight tasks in the study of topics relating to management of radioactive wastes. One task involves the verification of AECL's performance assessment code for their high level waste disposal program. In the agreement this task is given the title: Performance Assessment Technology Exchange. The quality assurance program established by AECL for this code requires that this task be performed. This paper presents an overview of the methods used for code verification and a progress report on the verification of the code. It describes the tools that have been developed to automatically examine code modules, check physical units, and prepare driver routines to exercise every line of code, and verify that it was executed correctly

  3. 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)

  4. The Maple reactor project

    International Nuclear Information System (INIS)

    MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)

  5. 11th radiochemical conference

    International Nuclear Information System (INIS)

    The conference met in four sesions which discussed: Separation methods, Radioanalytical methods, Labelled compounds and Miscellaneous. The first session discussed extraction methods, ion exchange and chromatographic separation of radioisotopes. The second session heard papers on the application of these methods, e.g., in geochemistry, on the use of radioactive tracers in radiochemical analysis and the use of activation analysis in the determination of trace elements. The third session heard papers on the preparation of labelled organic compounds with isotopes 3H, 14C, radioiodine and 32P, on the preparation of RIA kits and on the chemistry and radiopharmacology of technetium compounds. The other contributions which could not be heard in any of the three sessions discussed, e.g., the preparation of elements on the cyclotron and microtron, the production of a new 99mTc-generator, the participation of the IAEA in processing low- and medium-level radioactive wastes, etc. (E.S.)

  6. Radiochemical synthesis of etomoxir

    Energy Technology Data Exchange (ETDEWEB)

    Abbas, Hafiz G. [Institute of Nuclear Medicine and Oncology (INMOL), New Campus Road, Lahore (Pakistan); Yunus, M. [University of the Punjab, New Campus Road, Lahore (Pakistan); Feinendegen, Ludwig E., E-mail: feinendegen@gmx.ne [Department of Nuclear Medicine, Heinrich-Heine University Duesseldorf, Wannental 45, 88131 Lindau (Germany)

    2011-02-15

    Sodium 2-{l_brace}6-(4-chlorophenoxy)hexyl{r_brace}oxirane-2-carboxylate (Etomoxir) inhibits transport of fatty acids via the carnitine shuttle into mitochondria of muscle cells and prevents long chain fatty acids from providing energy through {beta}-oxidation especially for muscle contraction. The objective of this synthesis is to develop a method for radioiodination of Etomoxir in order to explore its potential in diagnostic metabolic studies and molecular imaging. Thus, a method is described for the radiochemical synthesis and purification of ethyl 2-{l_brace}6-(4-[{sup 131}I]iodophenoxy)hexyl{r_brace}oxirane-2-carboxylate (3) and 2-{l_brace}6-(4-[{sup 131}I]iodo-phenoxy)hexyl{r_brace}oxirane-2-carboxylic acid (4). For the synthesis of these new agents, ethyl 2-{l_brace}6-(4-bromophenoxy)hexyl{r_brace}oxirane-2-carboxylate (1) and 2-{l_brace}6-(4-bromophenoxy)hexyl{r_brace}oxirane-2-carboxylic acid (2) were refluxed with [{sup 131}I]NaI in the presence of anhydrous acetone at a temperature of 80 {sup o}C and 90 {sup o}C for a period of 3-4 hours, respectively. The method of radiolabeling, based on the nucleophilic exchange reaction, resulted in a radiochemical yield of 43% and 67% for compounds 3 and 4, respectively. This paper reports on the labeling of etomoxir with radioiodine as {sup 124}I labeled etomoxir may be of great importance in molecular imaging.

  7. 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

  8. AECL research programs in systems chemistry

    International Nuclear Information System (INIS)

    Research programs in Systems Chemistry are aimed at preserving the integrity of the many working systems in CANDU reactors and at minimizing chemistry-induced problems such as radiation field growth or fouling of surfaces. The topics of main concern are the chemistry and corrosion of steam generators, for it is in this general area that the potential for serious problems is very real

  9. The year 2000 (Y2k) Programme at AECL

    International Nuclear Information System (INIS)

    In the nuclear industry we make, in total, very extensive use of digital computers and equipment. While use of dates in our application may not be quite so extensive as in other businesses such as banking or insurance, dates are nonetheless employed, and are important in a variety of applications. Furthermore, date-related problems can sometimes propagate into overall system failures or computer crashes. Digital system or digital infrastructure failure can have serious potential consequences in a power plant, utility, or engineering design office. This in turn can have potential impact on public safety or the reliability of power production and delivery of electrical power to the public. A concerted effort is needed, and is underway by nuclear design organizations, and the nuclear utilities in order to identify and fix or avoid the problems in the short time that remains between now and the Year 2000. AECL have a substantial Year 2000 programme underway, addressing both the infrastructure systems at AECL, and AECL's products and services. High priority is placed, in the programme, on assisting AECL's customers with the Year 2000 issue. The programme, and some of the lesson learned to date, are described in this paper. The relationship to equipment vendors' and customers' Year 2000 programs is explained, and the importance of Year 2000 programmes conducted by the customers, to address systems and equipment which are under their control, is highlighted. (authors)

  10. Expert panel on hydrogeology; report to AECL Research (1992)

    International Nuclear Information System (INIS)

    In 1992 AECL Research convened a panel of external hydrogeological experts consisting of P.A. Domenico, G.E. Grisak, and F.W. Schwartz, to review AECL's proposed approach to siting a geological repository in the rocks of the Canadian Shield for the safe disposal of Canada's nuclear fuel wastes. In particular the panel was asked to provide its opinion on 1) the soundness of the technical approach developed to characterize the groundwater flow systems for the purpose of selecting a location for a disposal vault, 2) the validity and effectiveness of the geological case study used to demonstrate the performance assessment methodology based on the hydrogeological conditions observed at the Whiteshell Research Area, and 3) the adequacy of the hydrogeological information that AECL proposes to use in its Environmental Impact Statement (EIS) of the disposal concept. This report presents the findings, conclusions and recommendations of the hydrogeology review panel. The report was submitted to AECL Research in 1992 December. (author). 24 refs., 2 tabs., 4 figs

  11. Chemical and radiochemical constituents in water from wells in the vicinity of the naval reactors facility, Idaho National Engineering and Environmental Laboratory, Idaho, 1997-98

    Science.gov (United States)

    Bartholomay, Roy C.; Knobel, LeRoy L.; Tucker, Betty J.; Twining, Brian V.

    2000-01-01

    The U.S. Geological Survey, in response to a request from the U.S. Department of Energy?s Phtsburgh Naval Reactors Ofilce, Idaho Branch Office, sampled water from 13 wells during 1997?98 as part of a long-term project to monitor water quality of the Snake River Plain aquifer in the vicinity of the Naval Reactors Facility, Idaho National Engineering and Environmental Laboratory, Idaho. Water samples were analyzed for naturally occurring constituents and man-made contaminants. A totalof91 samples were collected from the 13 monitoring wells. The routine samples contained detectable concentrations of total cations and dissolved anions, and nitrite plus nitrate as nitrogen. Most of the samples also had detectable concentrations of gross alpha- and gross beta-particle radioactivity and tritium. Fourteen qualityassurance samples also were collected and analyze~ seven were field-blank samples, and seven were replicate samples. Most of the field blank samples contained less than detectable concentrations of target constituents; however, some blank samples did contain detectable concentrations of calcium, magnesium, barium, copper, manganese, nickel, zinc, nitrite plus nitrate, total organic halogens, tritium, and selected volatile organic compounds.

  12. Radiochemical examination on irradiated PHWR and FBTR fuels

    International Nuclear Information System (INIS)

    Radiochemical examination of irradiated fuels from Madras Atomic Power Station (MAPS) and Fast Breeder Test Reactor (FBTR) has been carried out in hot cells at Fuel Chemistry Division in Chemistry Group. Fuel was dissolved in concentrated nitric acid and small quantities of the fuel solution were transferred out of hot cells. Uranium and plutonium analysis were carried out using electro analytical techniques

  13. Decommissioning information management in decommissioning planning and operations at AECL (Ref 5054)

    International Nuclear Information System (INIS)

    As the AECL Decommissioning program has grown over the past few years, particularly with regard to long-term planning, so has its need to manage the records and information required to support the program. The program encompasses a diverse variety of facilities, including prototype and research reactors, fuel processing facilities, research laboratories, waste processing facilities, buildings, structures, lands and waste storage areas, many of which have changed over time. The decommissioning program involves planning, assessing, monitoring and executing projects to decommission the facilities. The efficient and effective decommissioning planning, assessment, monitoring and execution for the facilities and projects are dependent on a sound information base, upon which decisions can be made. A vital part of this Information Base is the ongoing management of historical facility records, including decommissioning records, throughout the full life cycle of the facilities. This paper describes AECL's and particularly DP and O's approach to: 1) Establishing a decommissioning records and information framework, which identifies what records and information are relevant to decommissioning, prioritizing the decommissioning facilities, identifying sources of relevant information and providing a user-friendly, electronic, search and retrieval tool for facility information accessible to staff. 2) Systematically, gathering, assessing, archiving and identifying important information and making that information available to staff to support their ongoing decommissioning work. 3) Continually managing and enhancing the records and information base and its support infrastructure to ensure its long-term availability. 4) Executing special information enhancement projects, which transform historic records into information for analysis. (author)

  14. Evaluation of radiochemical data usability

    International Nuclear Information System (INIS)

    This procedure provides a framework for implementation of radiochemical data verification and validation for environmental remediation activities. It has been developed through participation of many individuals currently involved in analytical radiochemistry, radiochemical validation, and validation program development throughout the DOE complex. It should be regarded as a guidance to use in developing an implementable radiochemical validation strategy. This procedure provides specifications for developing and implementing a radiochemical validation methodology flexible enough to allow evaluation of data useability for project-specific Data Quality Objectives (DQO). Data produced by analytical methods for which this procedure provides limited guidance are classified as open-quotes non-routineclose quotes radionuclides and methods, and analyses by these methods may necessitate adoption of modified criteria from this procedure

  15. 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)

  16. 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)

  17. Distributed fuel-management computation using RFSP, WIMS-AECL and PVM

    International Nuclear Information System (INIS)

    The Parallel Virtual Machine (PVM) software package was used to build an interface between RFSP and WIMS-AECL to enable history-based, local-parameter reactor fuel-management simulations in which batches of lattice-cell transport and burnup calculations can be made in parallel. The interface is based on the master/slave crowd-computation model. For slave computers numbering from one to twenty, the overhead spent by the one master preparing input for the slaves and processing their outputs was observed to be small in comparison with the computing time spent by the slaves themselves. Anticipating the availability of a much larger network of slaves in the future, two potential computational bottlenecks that might arise are described, and possible remedies for them are outlined. (author)

  18. Development and qualification of AECL computer codes for ACR safety analysis applications

    International Nuclear Information System (INIS)

    Over the past decade, AECL has developed and rigorously implemented a software Quality Assurance Program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. This paper provides an overview of the computer programs used in Advanced CANDU Reactor (ACR) safety analysis, including their purpose and linkages, and assessments of their applicability in the safety analyses of the ACR design. The paper also reviews the key elements of the Software Quality Assurance program applied in development and validation of these computer programs for ACR application including the computer code change control process and the documentation produced to support the qualification of the computer codes. (author)

  19. Radiochemical processing of activated targets

    International Nuclear Information System (INIS)

    The selection or development of an ''appropriate'' radiochemical processing scheme can be a complex and relatively difficult task. A variety of illustrative and alternative approaches to the various steps involved have been outlined. A few representative processing schemes have been examined in some detail. While these examples may serve to highlight the skills and versatility of the radiochemist, the most difficult aspect of radiochemical processing is frequently found in reconciling trade-offs. If the proper facilities are available, and the intended application and the limitations which that application impose on product quality are understood, a safe and reliable ''appropriate'' radiochemical processing scheme can almost always be developed. In the absence of such an understanding, even the best of facilities and technical skills are no guarantee of success

  20. Radiochemical analysis of chlorine-36

    International Nuclear Information System (INIS)

    The aim of this paper is to propose a radiochemical separation method of chlorine-36 from other beta-gamma emitters based on an oxidation technique where chlorine is trapped by NaOH. Chlorine-36 beta emissions are measured by liquid scintillation counting by the dual label technique in order to avoid the contamination produced by carbon-14 which is also trapped by NaOH and it is the main contaminant present in graphite samples. The sensitivity of this radiochemical method is high enough to achieve the needed thresholds for the radiological characterization of the radioactive materials in which this method can be applied

  1. 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

  2. Radiochemistry Lab Decommissioning and Dismantlement. AECL, Chalk River Labs, Ontario, Canada

    International Nuclear Information System (INIS)

    Atomic Energy of Canada (AECL) was originally founded in the mid 1940's to perform research in radiation and nuclear areas under the Canadian Defense Department. In the mid 50's The Canadian government embarked on several research and development programs for the development of the Candu Reactor. AECL was initially built as a temporary site and is now faced with many redundant buildings. Prior to 2004 small amounts of Decommissioning work was in progress. Many reasons for deferring decommissioning activities were used with the predominant ones being: 1. Reduction in radiation doses to workers during the final dismantlement, 2. Development of a long-term solution for the management of radioactive wastes in Canada, 3. Financial constraints presented by the number of facilities shutdown that would require decommissioning funds and the absence of an approved funding strategy. This has led to the development of a comprehensive decommissioning plan that is all inclusive of AECL's current and legacy liabilities. Canada does not have a long-term disposal site; therefore waste minimization becomes the driving factor behind decontamination for decommissioning before and during dismantlement. This decommissioning job was a great learning experience for decommissioning and the associated contractors who worked on this project. Throughout the life of the project there was a constant focus on waste minimization. This focus was constantly in conflict with regulatory compliance primarily with respect to fire regulations and protecting the facility along with adjacent facilities during the decommissioning activities. Discrepancies in historical documents forced the project to treat every space as a contaminated space until proven differently. Decommissioning and dismantlement within an operating site adds to the complexity of the tasks especially when it is being conducted in the heart of the plant. This project was very successful with no lost time accidents in over one hundred

  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. SLOB, a SLOWPOKE channel binding protein, regulates insulin pathway signaling and metabolism in Drosophila.

    Directory of Open Access Journals (Sweden)

    Amanda L Sheldon

    Full Text Available There is ample evidence that ion channel modulation by accessory proteins within a macromolecular complex can regulate channel activity and thereby impact neuronal excitability. However, the downstream consequences of ion channel modulation remain largely undetermined. The Drosophila melanogaster large conductance calcium-activated potassium channel SLOWPOKE (SLO undergoes modulation via its binding partner SLO-binding protein (SLOB. Regulation of SLO by SLOB influences the voltage dependence of SLO activation and modulates synaptic transmission. SLO and SLOB are expressed especially prominently in median neurosecretory cells (mNSCs in the pars intercerebralis (PI region of the brain; these cells also express and secrete Drosophila insulin like peptides (dILPs. Previously, we found that flies lacking SLOB exhibit increased resistance to starvation, and we reasoned that SLOB may regulate aspects of insulin signaling and metabolism. Here we investigate the role of SLOB in metabolism and find that slob null flies exhibit changes in energy storage and insulin pathway signaling. In addition, slob null flies have decreased levels of dilp3 and increased levels of takeout, a gene known to be involved in feeding and metabolism. Targeted expression of SLOB to mNSCs rescues these alterations in gene expression, as well as the metabolic phenotypes. Analysis of fly lines mutant for both slob and slo indicate that the effect of SLOB on metabolism and gene expression is via SLO. We propose that modulation of SLO by SLOB regulates neurotransmission in mNSCs, influencing downstream insulin pathway signaling and metabolism.

  5. Radiochemical studies of fast neutron induced reactions at KFA JUELICH

    International Nuclear Information System (INIS)

    The results of radiochemical study of neutron induced threshold reactions are reported. Integral cross-sections were measured by the activation technique. Tritium and 3He formation cross-sections were determined at 14.6 MeV neutron energy and on neutron spectrum from d-Be break-up reaction. Some preliminary systematic trends observed in the cross-section data described. Some general results of nuclear data measurement for reactor techology are discussed

  6. Radiochemical stability of radiopharmaceutical preparations

    Energy Technology Data Exchange (ETDEWEB)

    Martins, Patricia de A.; Silva, Jose L. da; Ramos, Marcelo P.S.; Oliveira, Ideli M. de; Felgueiras, Carlos F.; Herrerias, Rosana; Zapparoli Junior, Carlos L.; Mengatti, Jair; Fukumori, Neuza T.O.; Matsuda, Margareth M.N. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The 'in vitro' stability studies of the radiopharmaceutical preparations are an essential requirement for routine practice in nuclear medicine and are an important parameter for evaluating the quality, safety and efficacy required for the sanitary registration of pharmaceutical products. Several countries have published guidelines for the evaluation of pharmaceutical stability. In Brazil, the stability studies should be conducted according to the Guide for Conducting Stability Studies published in the Resolution-RE n. 1, of 29th July 2005. There are also for radiopharmaceutical products, two specific resolutions: RDC-63 regulates the Good Manufacturing Practices for Radiopharmaceuticals and RDC-64 provides the Registration of Radiopharmaceuticals, both published on the 18th December 2009. The radiopharmaceutical stability is defined as the time during which the radioisotope can be safely used for the intended purpose. The radiochemical stability can be affected by a variety of factors, including storage temperature, amount of radioactivity, radioactive concentration, presence or absence of antioxidants or other stabilizing agents. The radiochemical stability studies must be established under controlled conditions determined by the effective use of the product. The aim of this work was to evaluate the radiochemical stability of labeled molecules with {sup 131}I, {sup 123}I, {sup 153}Sm, {sup 18}F, {sup 51}Cr, {sup 177}Lu and {sup 111}In as well as {sup 67}Ga and {sup 201}Tl radiopharmaceuticals. Radiochemical purity was evaluated after production and in the validity period, with the maximum activity and in the recommended storage conditions. The analyses were carried out by thin-layer silica gel plate, paper chromatography and gel chromatography. The experimental results showed to be in accordance with the specified limits for all the analysed products. (author)

  7. Nuclear fuel cycle in Russia flows and parameters of nuclear materials reprocessing and produced at radiochemical plants

    International Nuclear Information System (INIS)

    The structure of nuclear cycle in Russia and nuclear material (NM) flows between radiochemical plants and reactors, as well as nuclear facilities using plutonium and regenerated uranium as input materials are reported. The properties and parameters of NM received by radiochemical nuclear facilities and shipped therefrom are especially considered. Research, power, and commercial reactors, spent fuel subassemblies, the irradiated uranium elements being reprocessing at radiochemical plants; major properties of reprocessed material important for NM accounting and control are listed. The flows of NM reprocessed and produced at radiochemical plants are shown schematically. Flows and major parameters of products, NM shipped and received by/at radiochemical plants as well as some parameters of containers essential from the standpoint of NM accounting and control are shown

  8. AECL's research and development program in environmental science and technology

    International Nuclear Information System (INIS)

    AECL's radiological research and development (R and D) program encompasses work on sources of radiation exposure, radionuclide transport through the environment and potential impacts on biota and on human health. The application of the radiation protection knowledge and technology developed in this program provides cradle-to-grave management for CANDU and related nuclear technologies. This document provides an overview of the Environmental Science and Technology (ES and T) program which is one of the technical areas of R and D within the radiological R and D program. The ES and T program uses science from three main areas: radiochemistry, mathematical modelling and environmental assessment. In addition to providing an overview of the program, this summary also gives specific examples of recent technical work in each of the three areas. These technical examples illustrate the applied nature of the ES and T program and the close coupling of the program to CANDU customer requirements. (author)

  9. 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

  10. 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.)

  11. Proceedings of the Tripartite Seminar on Nuclear Material Accounting and Control at Radiochemical Plants

    International Nuclear Information System (INIS)

    The problems of creation and operation of nuclear materials (NM) control and accounting systems and their components at radiochemical plants were discussed in seminar during November 2-6 of 1998. There were 63 Russian and 25 foreign participants in seminar. The seminar programme includes following sessions and articles: the aspects of State NM control and accountancy; NM control and accounting in radiochemical plants and at separate stages of reprocessing of spent nuclear fuel and irradiated fuel elements of commercial reactors; NM control and accountancy in storage facilities of radiochemical plants; NM control and accounting computerization, material balance assessment, preparation of reports; qualitative and quantitative measurements in NM control and accounting at radiochemical plants destructive analysis techniques

  12. Survey of research reactors

    International Nuclear Information System (INIS)

    A survey of reasearch reactors based on the IAEA Nuclear Research Reactor Data Base (RRDB) was done. This database includes information on 273 operating research reactors ranging in power from zero to several hundred MW. From these 273 operating research reactors 205 reactors have a power level below 5 MW, the remaining 68 reactors range from 5 MW up to several 100 MW thermal power. The major reactor types with common design are: Siemens Unterrichtsreaktors, 1.2 Argonaut reactors, Slowpoke reactors, the miniature neutron source reactors, TRIGA reactors, material testing reactors and high flux reactors. Technical data such as: power, fuel material, fuel type, enrichment, maximum neutron flux density and experimental facilities for each reactor type as well as a description of their utilization in physics and chemistry, medicine and biology, academic research and teaching, training purposes (students and physicists, operating personnel), industrial application (neutron radiography, silicon neutron transmutation doping facilities) are provided. The geographically distribution of these reactors is also shown. As conclusions the author discussed the advantages (low capital cost, low operating cost, low burn up, simple to operate, safe, less restrictive containment and sitting requirements, versatility) and disadvantages (lower sensitivity for NAA, limited radioisotope production, limited use of neutron beams, limited access to the core, licensing) of low power research reactors. 24 figs., refs. 15, Tab. 1 (nevyjel)

  13. The contribution of AECL CommercialProducts to nuclear medicine and radiation processing

    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 CommercialProducts 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 CommercialProducts dominates the world market. Another such example is the marketing of Mo-99, which is produced in the 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 'sterile male' technique and neutron activation and delayed neutron counting in uranium assay. (auth)

  14. Retrofit of AECL CAN6 seals into the Pickering shutdown cooling pumps

    International Nuclear Information System (INIS)

    The existing mechanical seals in the shutdown cooling (SDC) pumps at the eight-unit Pickering Nuclear Generating Station have caused at least seven forced outages in the last fifteen years. The SDC pumps were originally intended to run only during shutdowns, mostly at low pressure, except for short periods during routine testing of SDC isolation valves while the plant is operating at full pressure to verify that the emergency core injection system is available. Unfortunately, in practice, some SDC pumps must be run much more frequently than this to prevent overheating or freezing of components in the system while the plant is at power. This more severe service has decreased seal lifetime from about 8000 running hours to about 3000 running hours. Rather than tackling the difficult task of eliminating on-power running of the pumps, Pickering decided to install a more robust seal design that could withstand this. Through the process of competitive tender, AECL's CAN6 seal was chosen. This seal has a successful history in similarly demanding conditions in boiling water reactors in the USA. To supplement this and demonstrate there would be no 'surprises,' a 2000-hour test program was conducted. Testing consisted of simulating all the expected conditions, plus some special tests under abnormal conditions. This has given assurance that the seal will operate reliably in the Pickering shutdown cooling pumps. (author)

  15. Laboratory radiochemical and instrumental methods

    International Nuclear Information System (INIS)

    During the past 10 yr, the basic radiological laboratory support requirements had to be responsive to a number of industrial and regulatory concerns. In the latter part of the 1970s, the US Nuclear Regulatory Commission (NRC) issued generic radiological effluent technical specifications (10CFR50 Appendix I), which provided guidance on the minimum acceptable environmental radiation surveillance program requirements for the nuclear power industry. In the early 1980s, the Environmental Protection Agency (EPA) issued Radioactivity in Drinking Water Standards, and instituted a certification program for state and commercial laboratory processors, providing radioanalytical services in support of the EPA standard. In late December 1982, the NRC issued specific requirements (10CFR61) for licensing land disposal of low-level radioactive wastes containing licensed material. The presentation provides an overview of the typical radiochemical procedures and radiation detection instrumentation employed to process the various media and quantitate the radionuclides associated with the aforementioned programs

  16. Quick methods for radiochemical analysis

    International Nuclear Information System (INIS)

    Quick methods for radiochemical analysis, of adequate precision for the assay of a limited number of biologically important radionuclides, are important in the development of effective monitoring programs, particularly those that would be applied in emergency situations following an accidental release of radioactive substances. Methods of this type have been developed in a number of laboratories and are being altered and improved from time to time. The Agency invited several Member States to provide detailed information on any such procedures that might have been developed in their laboratories. From the information thus obtained a number of methods have been selected that appear to meet the criteria of speed and economy in the use of materials and equipment, and seem to be eminently suitable for those radionuclides that might be of major interest from the point of view of assessing the potential dose to persons following a serious dispersal of contamination. Refs, figs and tabs

  17. AECL's underground research laboratory: technical achievements and lessons learned

    International Nuclear Information System (INIS)

    During the development of the research program for the Canadian Nuclear Fuel Waste Management Program in the 1970's, the need for an underground facility was recognized. AECL constructed an Underground Research Laboratory (URL) for large-scale testing and in situ engineering and performance-assessment-related experiments on key aspects of deep geological disposal in a representative geological environment. Ale URL is a unique geotechnical research and development facility because it was constructed in a previously undisturbed portion of a granitic pluton that was well characterized before construction began, and because most of the shaft and experimental areas are below the water table. The specific areas of research, development and demonstration include surface and underground characterization; groundwater and solute transport; in situ rock stress conditions; temperature and time-dependent deformation and failure characteristics of rock; excavation techniques to minimize damage to surrounding rock and to ensure safe working conditions; and the performance of seals and backfills. This report traces the evolution of the URL and summarizes the technical achievements and lessons learned during its siting, design and construction, and operating phases over the last 18 years. (author)

  18. Fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    This paper includes some statements and remarks concerning the uranium silicide fuels for which there is significant fabrication in AECL, irradiation and defect performance experience; description of two Canadian high flux research reactors which use high enrichment uranium (HEU) and the fuels currently used in these reactors; limited fabrication work done on Al-U alloys to uranium contents as high as 40 wt%. The latter concerns work aimed at AECL fast neutron program. This experience in general terms is applied to the NRX and NRU designs of fuel

  19. 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)

  20. Activation analysis utilizing fast radiochemical separations and portable neutron generators

    International Nuclear Information System (INIS)

    Experience at the University of Michigan in adapting radiochemical procedures to short-half-lived radioisotopes and in working with neutron generators indicates that activation analysis can become a rapid routine industrial method for certain types of analysis. Portable neutron generators with thermal fluxes approaching 108 to 109 n/cm2 s, which have recently become available, offer the large analytical laboratory its own source of neutrons for activation analysis at a cost of about $ 20 000. In addition, rapid radiochemical procedures permit separation of activated products in a matter of minutes and completion of analyses for many elements in less than an hour. Rapid radiochemical separations have been devised for several elements. The sensitivities obtained from a 6-min irradiation of these elements are at least of the same order of magnitude and often better than those obtained with lengthy irradiations. Typical separation times are for vanadium 4 min and for silver 5 min. Thus it has been possible to make a complete analysis of samples ranging from meteorites to rat liver tissue, from marine biological ashes to crude petroleum, all within periods of 1/2 h to 1 h. These fast radiochemical techniques are also being applied to an evaluation of a Texas Nuclear Corporation neutron generator for use in activation analysis. This Cockcroft-Walton machine is small and compact; portable concrete shielding blocks suffice for protection. Although the present flux obtainable is less than 0.1% of the flux available in the pneumatic tube positions of the Michigan reactor, it is still sufficient to determine microgram amounts of an element such as vanadium. (author)

  1. MAPLE reactors for the secure supply of medical isotopes

    International Nuclear Information System (INIS)

    MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)

  2. Collected radiochemical and geochemical procedures

    Energy Technology Data Exchange (ETDEWEB)

    Kleinberg, J [comp.

    1990-05-01

    This revision of LA-1721, 4th Ed., Collected Radiochemical Procedures, reflects the activities of two groups in the Isotope and Nuclear Chemistry Division of the Los Alamos National Laboratory: INC-11, Nuclear and radiochemistry; and INC-7, Isotope Geochemistry. The procedures fall into five categories: I. Separation of Radionuclides from Uranium, Fission-Product Solutions, and Nuclear Debris; II. Separation of Products from Irradiated Targets; III. Preparation of Samples for Mass Spectrometric Analysis; IV. Dissolution Procedures; and V. Geochemical Procedures. With one exception, the first category of procedures is ordered by the positions of the elements in the Periodic Table, with separate parts on the Representative Elements (the A groups); the d-Transition Elements (the B groups and the Transition Triads); and the Lanthanides (Rare Earths) and Actinides (the 4f- and 5f-Transition Elements). The members of Group IIIB-- scandium, yttrium, and lanthanum--are included with the lanthanides, elements they resemble closely in chemistry and with which they occur in nature. The procedures dealing with the isolation of products from irradiated targets are arranged by target element.

  3. Canadian Experience in Application of Graded Approach for Safety Assessment of Research Reactors

    International Nuclear Information System (INIS)

    Full text: Research reactors are typically used for basic and applied research, education and training, production of isotopes, material testing, neutron activation analysis and other purposes. Most research reactors have a small potential for hazard to the public compared with power reactors. Safety assessment for the research reactors needs to be undertaken to evaluate compliance with safety requirements and to determine the measures to ensure reactor safety. Considering the different types of research reactors and their associated utilization, safety assessment should be commensurate with the potential hazard, ensuring that the design and operation of each reactor lead to adequate safety and defence in depth. The scope of presentation will cover the following topics: - Canadian regulatory framework for licensing research reactors; - Graded approach applied to safety assessment of the research reactors; - Use of graded approach to safety assessment of SLOWPOKE and NRU reactors. Canadian Nuclear Safety Commission (CNSC) has developed a regulatory framework for licensing small reactor facilities (including research reactors) that sets out requirements for the safety analysis and reactor design. CNSC staff considers each application individually in determining how much rigour and stringency are required for the safety assessment. All important factors affecting the overall reactor safety, such as safety system design, inherent safety features, the amount of fissile and fissionable materials, and the source terms are considered. The graded approach introduced, allows safety requirements to be implemented in such way that the level of safety assessment is proportional to the potential hazards posed by the research reactor. Licensing requirements vary with the type of facility and they may be applied in a graded fashion based on overall risk. Graded approach can be applied to all components of safety assessment including radiation risk, safety functions, defence in

  4. AECL's progress in developing the DUPIC fuel fabrication process

    International Nuclear Information System (INIS)

    Spent Pressurized Water Reactor (PWR) fuel can be used directly in CANDU reactors without the need for wet chemical reprocessing or reenrichment. Considerable experimental progress has been made in verifying the practicality of this fuel cycle, including hot-cell experiments using spent PWR fuels and out-cell trials using surrogate fuels. This paper describes the current status of these experiments. (author)

  5. 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)

  6. Self-sustainability of a research reactor facility with neutron activation analysis

    International Nuclear Information System (INIS)

    Long-term self-sustainability of a small reactor facility is possible because there is a large demand for non-destructive chemical analysis of bulk materials that can only be achieved with neutron activation analysis (NAA). The Ecole Polytechnique Montreal SLOWPOKE Reactor Facility has achieved self-sustainability for over twenty years, benefiting from the extreme reliability, ease of use and stable neutron flux of the SLOWPOKE reactor. The industrial clientele developed slowly over the years, mainly because of research users of the facility. A reliable NAA service with flexibility, high accuracy and fast turn-around time was achieved by developing an efficient NAA system, using a combination of the relative and k0 standardisation methods. The techniques were optimized to meet the specific needs of the client, such as low detection limit or high accuracy at high concentration. New marketing strategies are presented, which aim at a more rapid expansion. (author)

  7. 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

  8. 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

  9. 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

  10. The development, qualification and availability of AECL analytical, scientific and design codes

    International Nuclear Information System (INIS)

    Over the past several years, AECL has embarked on a comprehensive program to develop, qualify and support its key safety and licensing codes, and to make executable versions of these codes available to the international nuclear community. To this end, we have instituted a company-wide Software Quality Assurance (SQA) Program for Analytical, Scientific and Design Computer Programs to ensure that the design, development, maintenance, modification, procurement and use of computer codes within AECL is consistent with today's quality assurance standards. In addition, we have established a comprehensive Code Validation Project (CVP) with the goal of qualifying AECL's 'front-line' safety and licensing codes by 2001 December. The outcome of this initiative will be qualified codes, which are properly verified and validated for the expected range of applications, with associated statements of accuracy and uncertainty for each application. The code qualification program, based on the CSA N286.7 standard, is intended to ensure (1) that errors are not introduced into safety analyses because of deficiencies in the software, (2) that an auditable documentation base is assembled that demonstrates to the regulator that the codes are of acceptable quality, and (3) that these codes are formally qualified for their intended applications. Because AECL and the Canadian nuclear utilities (i.e., Ontario Power Generation, Bruce Power, Hydro Quebec and New Brunswick Power) generally use the same safety and licensing codes, the nuclear industry in Canada has agreed to work cooperatively together towards the development, qualification and maintenance of a common set of analysis tools, referred to as the Industry Standard Toolset (IST). This paper provides an overview of the AECL Software Quality Assurance Program and the Code Validation Project, and their associated linkages to the Canadian nuclear community's Industry Standard Toolset initiative to cooperatively qualify and support commonly

  11. Safety assessment for TA-48 radiochemical operations

    International Nuclear Information System (INIS)

    The purpose of this report is to document an assessment performed to evaluate the safety of the radiochemical operations conducted at the Los Alamos National Laboratory operations area designated as TA-48. This Safety Assessment for the TA-48 radiochemical operations was prepared to fulfill the requirements of US Department of Energy (DOE) Order 5481.1B, ''Safety Analysis and Review System.'' The area designated as TA-48 is operated by the Chemical Science and Technology (CST) Division and is involved with radiochemical operations associated with nuclear weapons testing, evaluation of samples collected from a variety of environmental sources, and nuclear medicine activities. This report documents a systematic evaluation of the hazards associated with the radiochemical operations that are conducted at TA-48. The accident analyses are limited to evaluation of the expected consequences associated with a few bounding accident scenarios that are selected as part of the hazard analysis. Section 2 of this report presents an executive summary and conclusions, Section 3 presents pertinent information concerning the TA-48 site and surrounding area, Section 4 presents a description of the TA-48 radiochemical operations, and Section 5 presents a description of the individual facilities. Section 6 of the report presents an evaluation of the hazards that are associated with the TA-48 operations and Section 7 presents a detailed analysis of selected accident scenarios

  12. Radiochemical studies on nuclear fission at Trombay

    Indian Academy of Sciences (India)

    Asok Goswami

    2015-08-01

    Since the discovery of nuclear fission in the year 1939, both physical and radiochemical techniques have been adopted for the study of various aspects of the phenomenon. Due to the ability to separate individual elements from a complex reaction mixture with a high degree of sensitivity and selectivity, a chemist plays a significant role in the measurements of mass, charge, kinetic energy, angular momentum and angular distribution of fission products in various fissioning systems. At Trombay, a small group of radiochemists initiated the work on radiochemical studies of mass distribution in the early sixties. Since then, radiochemical investigations on various fission observables have been carried out at Trombay in , , and heavy-ion-induced fissions. An attempt has been made to highlight the important findings of such studies in this paper, with an emphasis on medium energy and heavy-ion-induced fission.

  13. Automated Radiochemical Separation, Analysis, and Sensing

    International Nuclear Information System (INIS)

    Chapter 14 for the 2nd edition of the Handbook of Radioactivity Analysis. The techniques and examples described in this chapter demonstrate that modern fluidic techniques and instrumentation can be used to develop automated radiochemical separation workstations. In many applications, these can be mechanically simple and key parameters can be controlled from software. If desired, many of the fluidic components and solution can be located remotely from the radioactive samples and other hot sample processing zones. There are many issues to address in developing automated radiochemical separation that perform reliably time after time in unattended operation. These are associated primarily with the separation and analytical chemistry aspects of the process. The relevant issues include the selectivity of the separation, decontamination factors, matrix effects, and recoveries from the separation column. In addition, flow rate effects, column lifetimes, carryover from one sample to another, and sample throughput must be considered. Nevertheless, successful approaches for addressing these issues have been developed. Radiochemical analysis is required not only for processing nuclear waste samples in the laboratory, but also for at-site or in situ applications. Monitors for nuclear waste processing operations represent an at-site application where continuous unattended monitoring is required to assure effective process radiochemical separations that produce waste streams that qualify for conversion to stable waste forms. Radionuclide sensors for water monitoring and long term stewardship represent an application where at-site or in situ measurements will be most effective. Automated radiochemical analyzers and sensors have been developed that demonstrate that radiochemical analysis beyond the analytical laboratory is both possible and practical

  14. 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

  15. Rapid radiochemical separations in neutron activation analysis

    International Nuclear Information System (INIS)

    Rapid radiochemical separation procedures based on the removal of metal ions by columns of C18-bonded silica gel after selective complexation are examined and the simplicity of the method demonstrated by its application to the determination of Mn, Cu and Zn in neutron-activated biological material. The method is rapid and reliable and readily adaptable in all radiochemical laboratories. An alternative separation procedure for selenium in blood plasma involving desalination and concentration of the selenium protein complex by gel filtration or ultrafiltration is briefly discussed. (author)

  16. Effects of manipulating slowpoke calcium-dependent potassium channel expression on rhythmic locomotor activity in Drosophila larvae

    Directory of Open Access Journals (Sweden)

    Erin C. McKiernan

    2013-03-01

    Full Text Available Rhythmic motor behaviors are generated by networks of neurons. The sequence and timing of muscle contractions depends on both synaptic connections between neurons and the neurons’ intrinsic properties. In particular, motor neuron ion currents may contribute significantly to motor output. Large conductance Ca2+-dependent K+ (BK currents play a role in action potential repolarization, interspike interval, repetitive and burst firing, burst termination and interburst interval in neurons. Mutations in slowpoke (slo genes encoding BK channels result in motor disturbances. This study examined the effects of manipulating slo channel expression on rhythmic motor activity using Drosophila larva as a model system. Dual intracellular recordings from adjacent body wall muscles were made during spontaneous crawling-related activity in larvae expressing a slo mutation or a slo RNA interference construct. The incidence and duration of rhythmic activity in slo mutants were similar to wild-type control animals, while the timing of the motor pattern was altered. slo mutants showed decreased burst durations, cycle durations, and quiescence intervals, and increased duty cycles, relative to wild-type. Expressing slo RNAi in identified motor neurons phenocopied many of the effects observed in the mutant, including decreases in quiescence interval and cycle duration. Overall, these results show that altering slo expression in the whole larva, and specifically in motor neurons, changes the frequency of crawling activity. These results suggest an important role for motor neuron intrinsic properties in shaping the timing of motor output.

  17. Radiochemical measurement of neutron-spectrum averaged cross sections for the formation of {sup 64}Cu and {sup 67}Cu via the (n,p) reaction at a TRIGA Mark-II reactor. Feasibility of simultaneous production of the theragnostic pair {sup 64}Cu/{sup 67}Cu

    Energy Technology Data Exchange (ETDEWEB)

    Uddin, M. Shuza; Hossain, Syed Mohammod [Atomic Energy Research Establishment, Dhaka (Bangladesh). Inst. of Nuclear Science and Technology; Rumman-uz-Zaman, M. [Atomic Energy Research Establishment, Dhaka (Bangladesh). Inst. of Nuclear Science and Technology; Dhaka Univ. (Bangladesh). Dept. of Applied Chemistry and Chemical Engineering; Qaim, Syed M. [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Neurowissenschaften und Medizin (INM-5) - Nuklearchemie

    2014-09-01

    Integral cross sections of the {sup 64}Zn(n,p){sup 64}Cu and {sup 67}Zn(n,p){sup 67}Cu reactions were measured for the fast neutron spectrum of TRIGA Mark-II reactor at Savar, Dhaka, Bangladesh. A clean radiochemical separation was performed to isolate the copper radionuclides from the target element zinc. The radioactivities produced in the irradiation were measured by HPGe γ-ray spectroscopy. The neutron flux over the energy range 0.5-20 MeV was determined using the {sup 58}Ni(n,p){sup 58}Co monitor reaction. The measured results amount to 28.9 ± 2.0 mb and 0.84 ± 0.07 mb for the formation of {sup 64}Cu and {sup 67}Cu, respectively. These values are slightly lower than the respective values for a pure fission spectrum. The present results were compared with data calculated using the neutron spectral distribution and the recently critically analysed excitation function of each reaction given in the literature. The good agreement validates the reliability of those excitation functions. The feasibility of simultaneous production of {sup 64}Cu and {sup 67}Cu with fast neutrons is discussed. (orig.)

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

    International Nuclear Information System (INIS)

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

  19. The analysis and attribution of the time-dependent neutron background resultant from sample irradiation in a SLOWPOKE-2 reactor

    International Nuclear Information System (INIS)

    The Royal Military College of Canada (RMCC) has commissioned a Delayed Neutron Counting (DNC) system for the analysis of special nuclear materials. A significant, time-dependent neutron background with an initial maximum count rate, more than 50 times that of the time-independent background, was characterised during the validation of this system. This time-dependent background was found to be dependent on the presence of the polyethylene (PE) vials used to transport the fissile samples, yet was not an activation product of vial impurities. The magnitude of the time-dependent background was found to be irradiation site specific and independent of the mass of PE. The capability of RMCC's DNC system to analyze the neutron count rates in time intervals 235U contamination was present on each irradiated vial. However, Inductively Coupled Plasma-Mass Spectroscopy measurements of material leached from the outer vial surfaces after their irradiations found only trace amounts of uranium, 0.118 ± 0.048 ng of 235U derived from natural uranium. These quantities are insufficient to account for the time-independent background, and in fact could not be discriminated from the noise associated with time-independent background. It is suggested that delayed neutron emitters are deposited in the vial surface following fission recoil, leaving the main body of uranium within the irradiation site. This hypothesis is supported by the physical cleaning of the site with materials soaked in distilled water and HNO3, which lowered the background from a nominal 235U mass equivalent of 120 to 50 ng per vial. (author)

  20. Developing the MAPLE materials test reactor concept

    International Nuclear Information System (INIS)

    MAPLE-MTR is a new multipurpose research facility being planned by AECL Research as a possible replacement for the 35-year-old NRU reactor. In developing the MAPLE-MTR concept, AECL is starting from the recent design and licensing experience with the MAPLE-X10 reactor. By starting from technology developed to support the MAPLE-X10 design and adapting it to produce a concept that satisfies the requirements of fuel channel materials testing and fuel irradiation programs, AECL expects to minimize the need for major advances in nuclear technology (e.g., fuel, heat transfer). Formulation of the MAPLE-MTR concept is at an early stage. This report describes the irradiation requirements of the research areas, how these needs are translated into design criteria for the project and elements of the preliminary design concept

  1. Thermal-hydraulic interfacing code modules for CANDU reactors

    International Nuclear Information System (INIS)

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

  2. 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.

  3. Safety decommissioning regulations of radiochemical objects - the problem, requires urgent decision

    International Nuclear Information System (INIS)

    The necessity of planning and pursuance of the measures on decommissioning of radiochemical industry is discussed. Technological processes were stopped more than in 30 buildings and constructions of the PO Mayak. The characteristics of the technological buildings to be decommissioned were treated in the context of building peculiarities, function, character and level of contamination. An acceptable variant for reactor decommissioning invites development of the standard-legal aspects

  4. Improved CANDU fuel performance. A summary of previous AECL publications

    International Nuclear Information System (INIS)

    The fuel defect rate in CANDU power reactors has been very low (0.06%) since 1972. Most defects were caused by power ramping. The two measures taken to reduce the defect rate, by about an order of magnitude, were changes in the fuelling schemes and the introduction of thin coatings of graphite on the inside surface of the Zircaloy fuel cladding. Power ramping tests have demonstrated that graphite layers, and also baked poly-dimethyl-siloxane layers, between the UO2 pellets and Zircaloy cladding, increase the tolerance of fuel to power ramps. These designs are termed graphite CANLUB and siloxane CANLUB; fuel performance depends on coating parameters such as thickness, wear resistance and on environmental and thermal conditions during the curing of coatings. (author)

  5. An approach to neutronics analysis of candu reactors

    International Nuclear Information System (INIS)

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

  6. Automated radiochemical processing for clinical PET

    International Nuclear Information System (INIS)

    The Siemens RDS 112, an automated radiochemical production and delivery system designed to support a clinical PET program, consists of an 11 MeV, proton only, negative ion cyclotron, a shield, a computer, and targetry and chemical processing modules to produce radiochemicals used in PET imaging. The principal clinical PET tracers are [18F]FDG, [13N]ammonia and [15O]water. Automated synthesis of [18F]FDG is achieved using the Chemistry Process Control Unit (CPCU), a general purpose valve-and-tubing device that emulates manual processes while allowing for competent operator intervention. Using function-based command file software, this pressure-driven synthesis system carries out chemical processing procedures by timing only, without process-based feedback. To date, nine CPCUs have installed at seven institutions resulting in 1,200+ syntheses of [18F]FDG, with an average yield of 55% (EOB)

  7. Methods for training radiochemical technicians at ORNL

    International Nuclear Information System (INIS)

    The training of personnel to carry out radiochemical operations at ORNL is a formidable and recurrent task since repetitive, production-type operations are not involved, and programs are constantly shifting. It is essential that provisions be made for the routine retraining of personnel if they are to make effective contributions on a continuing basis. The present training methods have emerged as a result of thirty years experience in a variety of radiochemical pilot-plant programs. These programs have included operations performed in glove boxes, hot-cell manipulator work handling high-neutron-emitting isotopes, and the entire spectrum of remote solvent extraction operations. Present methods of training and the results obtained are summarized

  8. 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)

  9. 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

  10. Radiochemical analysis of military nuclear facilities

    International Nuclear Information System (INIS)

    Full text : Radiochemical Analysis is a branch of analytical chemistry comprising an aggregate of methods for qualitatively determining the composition and content of radioisotopes in the products of transformations. Safety and minimization of radiation impact on human and environment are important demand of operation of Military Nuclear Facilities (MNF). In accordance of recommendations of International Commission on Radiological Protection there are next objects of radiochemical analysis: 1) potential sources of radiochemical pollution; 2) environment (objects of environment, human environment including buildings, agricultural production, water, air et al.); 3) human himself (determination of dose from external and internal radiation, chemical poisoning). The chemical analysis can be carried out using, for example, the Gas Chromatography instrument whish separates chemical mixtures and identifies the components at a molecular level. It is one of the most accurate tools for analyzing environmental samples. The Gas Chromatography works on the principle that a mixture will separate into individual substances when heated. The heated gases are carried through a column with an inert gas (such as helium). As the separated substances emerge from the column opening, they flow into the Mass Spectrometry. Mass spectrometry identifies compounds by the mass of the analyte molecule. Newly developed portable Gas Chromatography and Mass Spectrometry are techniques that can be used to separate volatile organic compounds and pesticides. Other uses of Gas Chromatography, combined with other separation and analytical techniques, have been developed for radionuclides, explosive compounds such as royal demolition explosive and trinitrotoluene, and metals. So, based on the many years experience of operation of dangerous MNF, in concordance with norms of radiation and chemical safety it was considered that the tasks of the radiochemical analysis of Military Nuclear Facilities include

  11. Radiochemical separation of gold by amalgam exchange

    Science.gov (United States)

    Ruch, R.R.

    1970-01-01

    A rapid and simple method for the radiochemical separation of gold after neutron activation. The technique is based on treatment with a dilute indium-gold amalgam, both chemical reduction and isotopic exchange being involved. The counting efficiency for 198Au in small volumes of the amalgam is good. Few interferences occur and the method is applicable to clays, rocks, salts and metals. The possibility of determining silver, platinum and palladium by a similar method is mentioned. ?? 1970.

  12. 14th radiochemical conference. Booklet of abstracts

    International Nuclear Information System (INIS)

    The contributions dealt with the following topics: Radionuclides in the environment, radioecology; Nuclear analytical methods; Chemistry of actinide and trans-actinide elements; Ionizing radiation in science, technology, and arts and cultural heritage preservation; Production and application of radionuclides; Separation methods, speciation; Chemistry of nuclear fuel cycle, radiochemical problems in nuclear waste management; and Nuclear methods in medicine, radiopharmaceuticals, and radiodiagnostics, labelled compounds. Of the verbal and poster presentation, 192 have been input to INIS. (P.A.)

  13. Radiochemical aging of an epoxy network

    International Nuclear Information System (INIS)

    This thesis is to give a better understanding of the radiochemical aging of a thermoset resin under gamma irradiation. The conditions of aging are gamma irradiation under air with a dose rate of 2 kGy/h at 120 C. The requested lifetime is four years, it means a dose of 70 MGy. The first step of this work was the choice of a resistive epoxy resin. This choice was made thanks to the literature data. The high glass transition temperature and the high amount of aromatic groups were the main criteria of the final choice. After this choice, thermal and mechanical properties were followed under thermal and radiochemical aging: i) under thermal aging, after 600 hours at 220 C, the glass transition temperature remained unchanged. But, from a mechanical point of view, properties at break dramatically decreased. This embrittlement was assigned to a critical oxidized layer. The thickness of this layer was estimated about 30 μm. ii) the same kind of embrittlement was observed under radiochemical aging. Moreover, it appeared a decrease of the glass transition temperature when increasing the dose of irradiation. This indicates that the main degradation mechanism is chain scission under anaerobic atmosphere. We, then, proposed a mechanistic model associated with a kinetic model to predict the evolution of the glass transition temperature depending on the irradiation conditions. Parameters of the kinetic model were determined by solid NMR and ESR experiments. Comparison between experimental and calculated values at 120 C is satisfactory, a global good agreement was found. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Whitaker, S.H.; Brown, A.; Davison, C.C.; Gascoyne, M.; Lodha, G.S.; Stevenson, D.R.; Thorne, G.A.; Tomsons, D. [AECL Research, Whiteshell Labs., Pinawa, MB (Canada)

    1994-05-01

    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.

  15. Destructive radiochemical analysis of uraniumsilicide fuel for burnup determination

    Energy Technology Data Exchange (ETDEWEB)

    Gysemans, M.; Bocxstaele, M. van; Bree, P. van; Vandevelde, L.; Koonen, E.; Sannen, L. [SCK-CEN, Boeretang, Mol (Belgium); Guigon, B. [CEA, Centre de Cadarache, Saint Paul lez Durance (France)

    2004-07-01

    During the design phase of the French research reactor Jules Horowitz (RJH) several types of low enriched uranium fuels (LEU), i.e. <20% {sup 235}U enrichment, are studied as possible candidate fuel elements for the reactor core. One of the LEU fuels that is taken into consideration is an uraniumsilicide based fuel with U{sub 3}Si{sub 2} dispersed in an aluminium matrix. The development and evaluation of such a new fuel for a research reactor requires an extensive testing and qualification program, which includes destructive radiochemical analysis to determine the burnup of irradiated fuel with a high accuracy. In radiochemistry burnup is expressed as atom percent burnup and is a measure for the number of fissions that have occurred per initial 100 heavy element atoms (%FIMA). It is determined by measuring the number of heavy element atoms in the fuel and the number of atoms of selected key fission products that are proportional to the number of fissions that occurred during irradiation. From the few fission products that are suitable as fission product monitor, the stable Nd-isotopes {sup 143}Nd, {sup 144}Nd, {sup 145}Nd, {sup 146}Nd, {sup 148Nd}, {sup 150}Nd and the gamma-emitters {sup 137}Cs and {sup 144}Ce are selected for analysis. Samples form two curved U{sub 3}Si{sub 2} plates, with a fuel core density of 5.1 and 6.1 g U/cm{sup 3} (35% {sup 235}U) and being irradiated in the BR2 reactor of SCK x CEN{sup [1]}, were analyzed. (orig.)

  16. Trace analysis measurements in high-purity aluminium by means of radiochemical neutron and proton activation analysis

    International Nuclear Information System (INIS)

    The aim of the study consisted in the development of efficient radiochemical composite processes and activation methods for the multi-element determination of traces within the lower ng range in high-purity aluminium. More than 50 elements were determined with the help of activation with reactor neutrons; the selective separation of matrix activity (adsorption with hydrated antimony pentoxide) led to a noticeable improvement of detectability, as compared with instrumental neutron activation analysis. Further improvements were achieved with the help of radiochemical group separations in ion exchangers or with the help of the selective separation of the pure beta-emitting elements. Over 20 elements up to high atomic numbers were determined by means of activating 13 MeV protons and 23 Me protons. In this connection, improvements of the detection limit by as a factor of 10 were achieved with radiochemical separation techniques, as compared with pure instrumental proton activation analysis. (RB)

  17. Radiochemical synthesis of 14C-labelled pesticides

    International Nuclear Information System (INIS)

    Radioisotopic derivatives of pesticides labelled with either 14C or tritium are indispensable experimental tools for toxicology or metabolism studies required for registration of new compounds. The radiochemical synthetic pathways leading to the preparation of 14C-labelled pesticides of high specific activity, good chemical/radiochemical yield, and high radiochemical purity are presented for three groups of pesticides; triazines, aryl-haloids, and organometallic compounds. (N.T.). 10 refs., 1 tab

  18. Rapid automated batchwize radiochemical separation techniques

    International Nuclear Information System (INIS)

    The basic principles and specific techniques of rapid, automated radiochemical separation techniques that use batchwize separation methods are reviewed. The basic chemical technics include many standard methods used in analytical chemistry: precipitation, solvent extraction, ion exchange, distillation, volatilization, electrolysis and electrophoresis. Isotopic exchange, absorption and thermochromatography are examples of other techniques specially used in fast separation procedures. Auobatch techniques were used for the automatic process: silver isotope separation, technetium, palladium separation by solvent extraction, arsenic and antimony separation by volatile hydriole production, separation of individual rare earth fission products. (R.P.) 40 refs

  19. Radiochemical coupling of acrylic acid to polyvinylchloride

    International Nuclear Information System (INIS)

    Acrylic acid was coupled radiochemically to the surface of polyvinylchloride (PVC) foils. A 500 keV electron generator served as radiation source. After neutralization with ammonia, the surface of the PVC foils got hydrophilic properties. Their capacity of water uptake increased from 0,04 mg/cm2 to about 0,5 mg/cm2 and the condensation of water takes place in form of a clear transparent film and not in form of light scattering droplets. 6 refs., 20 figs., 8 tabs

  20. 13th Radiochemical Conference. Booklet of Abstracts

    International Nuclear Information System (INIS)

    The Conference included the following sessions: (i) Opening plenary presentations (6 contributions); (ii) Chemistry of natural radionuclides, discovery of radium and polonium (6 verbal presentations + 5 poster presentations); (iii) Radionuclides in the environment, radioecology (29 + 48); (iv) Activation analysis and other radioanalytical methods (36 + 49); (v) Ionizing radiation in science and technology (12 + 12); (vi) Chemistry of actinide and trans-actinide elements (11 + 14); (vii) Separation methods, speciation (18 + 41); (viii) Production and application of radionuclides (14 + 29); and (ix) Radiochemical problems in nuclear waste management (12 + 22). The majority of verbal presentations has been input to INIS, mostly in the form of the full authors' abstracts. (P.A.)

  1. Leakage evaluation in the PCV (Primary Containment Vessel) using chemical and radiochemical data

    International Nuclear Information System (INIS)

    Keeping the reliability of nuclear power plant operation, the primary coolant leakage in the PCV is strictly restricted by the Technical Specifications. It is very important to detect an indication of leakage and estimate the source of leakage to provide countermeasures. Usually the indication of leakage will be detected by increase of drain flow in the PCV sump. There are some possibilities of leakage sources in the PCV, such as reactor water, main steam, condensate, feedwater and closed cooling water. The leakage source contain different chemical and radiochemical species. This means that the leakage source can be presumed and detected by using chemical information from the PCV atmosphere and sump water. To detect the leakage indication and the source quickly and exactly, the PCV Leakage Detection Expert System has been developed. This paper describes how to evaluate the leakage indication and source in the PCV by using chemical and radiochemical data. (author)

  2. Determination of iodine in foodstuffs consumed in Libya using instrumental and radiochemical neutron activation analysis

    International Nuclear Information System (INIS)

    Iodine was determined in foodstuffs consumed in Libya employing two modes of NAA. The first mode was instrumental using short-time irradiation with epithermal neutrons behind a Cd shield (EINAA). The other mode utilized short-time irradiation with the reactor-pile neutrons followed by radiochemical separation (RNAA). The radiochemical separation procedure was based on the alkaline-oxidative fusion of samples and extraction of elemental iodine into chloroform. Separation yield determined using the radiotracer 131I was within the range of 90 to 95 %. For quality control purposes, standard reference materials were analyzed in both modes employed. Using RNAA, a detection limit of about 1 ng g-1 could be obtained indicating superiority of the method in measuring ultra-trace levels of iodine. On the other hand, more than one order of magnitude higher detection limit did not allow sufficiently accurate determination of iodine in Libyan foodstuffs using EINAA. (author)

  3. Radiochemical analysis of nuclear fuel burn-up and spent fuel key nuclides

    International Nuclear Information System (INIS)

    Destructive radiochemical analysis of spent nuclear fuels is an important tool to determine burn-up with high accuracy and to better understand the process of depletion and formation of actinides and fission products during irradiation as a result of fission and successive neutron capture. The resulting isotope inventories and nuclear databases that are created, are of high importance to evaluate the performance of nuclear fuels in a reactor, to evaluate computer codes applied for a safe transport, storage and disposal/reprocessing of spent fuels and to safeguard fissile material. The objective is to provide chemical and radiochemical analyses procedures for an accurate determination of isotopic compositions and concentrations of actinides and fission products in different types of industrial (UO2, MOX) and experimental nuclear fuels (UAlx, U3Si2, UMo, ...). For a burn-up determination program typically 21 actinides and fission products are analyzed. For an extended characterization program this can increase to up to approximately 50 isotopes

  4. Radiochemical compatibility of EPDM/PP blends

    International Nuclear Information System (INIS)

    The capacity of ethylene-propylene-diene terpolymer to be cross-linked by its exposure to high energy radiation was considered in this work. γ-irradiation of EPDM/PP (ethylene-propylene-diene terpolymer/polypropylene) blends proved that free radicals provided by PP can be grafted on the EPDM backbone. Gel content measurements over all blending concentrations revealed a maximum level cross-linking at around 200 kGy. Simultaneity of cross-linking and oxidative degradation in irradiated polymers requires the choice of proper conditions to obtain high durability. IR spectroscopic measurements of carbonyl and hydroxyl indexes emphasize that low concentrations in radiolytic products is attained at short exposure (less than 200 kGy). This work presents the difference between raw and degraded polypropylene used for mixture preparation. The radiochemical behaviour of similar compositions of blends is proved. Thermal stability of non-irradiated EPDM/PP blend was checked by oxygen uptake method that confirms the higher probability of polypropylene to provide free radicals in a large extent. The high temperature used in the oxidation testing of present blends requires radiochemical compatibility of polymers. (authors)

  5. Monitoring and control of Urex radiochemical processes

    International Nuclear Information System (INIS)

    There is urgent need for methods to provide on-line monitoring and control of the radiochemical processes that are currently being developed and demonstrated under the Global Nuclear Energy Partnership (GNEP) initiative. The methods used to monitor these processes must be robust (require little or no maintenance) and must be able to withstand harsh environments (e.g., high radiation fields and aggressive chemical matrices). The ability for continuous online monitoring allows the following benefits: - Accountability of the fissile materials; - Control of the process flowsheet; - Information on flow parameters, solution composition, and chemical speciation; - Enhanced performance by eliminating the need for traditional analytical 'grab samples'; - Improvement of operational and criticality safety; - Elimination of human error. The objective of our project is to use a system of flow, chemical composition, and physical property measurement techniques for developing on-line real-time monitoring systems for the UREX process streams. We will use our past experience in adapting and deploying Raman spectrometer combined with Coriolis meters and conductivity probes in developing a deployable prototype monitor for the UREX radiochemical streams. This system will be augmented with UV-vis-NIR spectrophotometer. Flow, temperature, density, and chemical composition and concentration measurements will be combined for real-time data analysis during processing. Currently emphasis of our research is placed on evaluation of the commercial instrumentation for the UREX flowsheet. (authors)

  6. Reactor physics studies for a pressure tube supercritical water reactor (PT-SCWR)

    International Nuclear Information System (INIS)

    Preliminary lattice physics and full core neutronic analysis have been performed for the pressure-tube supercritical water reactor (PT-SCWR). Current CANDU reactor physics codes (WIMS-AECL and RFSP) were used for modeling this reactor. A key challenge in the physics design of this reactor is the optimization of lattice parameters to achieve the appropriate balance between coolant void reactivity (CVR) and fuel utilization. A vertically-oriented, batch-fuelled reactor is considered, with an insulated pressure tube to accommodate the high coolant temperatures and pressures. The analysis shows the reactor physics conceptual feasibility of the design, although further optimization is required. (author)

  7. 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.

  8. Upgrading the NRU research reactor

    International Nuclear Information System (INIS)

    After a nearly two-year long detailed review, AECL Research decided that its NRU research reactor will complete its mission around the turn of the century. The company's original intentions for major refurbishment have been revised and upgrading work will now mainly comprise add-ons to existing systems - so that research projects and isotope production schedules can be met - and procedure modifications to ensure continued safe operation. (Author)

  9. Comments on nuclear reactor safety in Ontario

    International Nuclear Information System (INIS)

    The Chalk River Technicians and Technologists Union representing 500 technical employees at the Chalk River Nuclear Laboratories of AECL submit comments on nuclear reactor safety to the Ontario Nuclear Safety Review. Issues identified by the Review Commissioner are addressed from the perspective of both a labour organization and experience in the nuclear R and D field. In general, Local 1568 believes Ontario's CANDU nuclear reactors are not only safe but also essential to the continued economic prosperity of the province

  10. Reactor coolant pump seals: improving their performance

    International Nuclear Information System (INIS)

    Large CANDU plants are benefitting from transient-resistant four-year reliable reactor coolant pump seal lifetimes, a direct result of AECL's 20-year comprehensive seal improvement program involving R and D staff, manufacturers, and plant designers and operators. An overview of this program is presented, which covers seal modification design, testing, post-service examination, specialized maintenance and quality control. The relevancy of this technology to Light Water Reactor Coolant Pump Seals is also discussed

  11. Technology transfer programs using a low power nuclear reactor

    International Nuclear Information System (INIS)

    The SLOWPOKE II nuclear reactor developed by Atomic Energy of Canada Limited is well suited for neutron activation analysis and the production of small quantities of radionuclides. Emphasis has been placed on local research groups to transfer appropriate technology developed in their laboratories into the community. The development of several research protocols and associated technology is reviewed and their successful implementation into local industry is outlined. These include for example, the monitoring of environmental chlorinated compounds, the irradiation of gem stones, placer gold-mining efficiency measurements and measuring industrial flow-processes. (author) 6 refs.; 1 tab

  12. External hazards assessment of heating reactor installations in urban areas

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited is developing the 10-MW Slowpoke Energy System, a small, economically competitive reactor, for conventional district heating systems used at large industrial complexes, hospitals, and universities. This type of reactor would be situated within urban centers and likely adjacent to existing powerhouse complexes (desirable from an operation point of view). These existing complexes pose some interesting challenges in the area of hazards to the reactor facility from sources external to the facility, i.e., external hazards. Because of proximity of the reactor facility to the general public (the exclusion boundary is the reactor building itself), a detailed assessment of the external hazards is important for demonstrating feasibility and acquiring licensing approval. This paper briefly describes the methodology of an external hazards study

  13. Implementing the AECL decommissioning quality assurance program at the Chalk River and Whiteshell Laboratories

    International Nuclear Information System (INIS)

    This paper describes the approach and progress in developing, implementing and maintaining a quality assurance (QA) program for decommissioning at the nuclear facilities managed by Atomic Energy of Canada Limited (AECL). Decommissioning activities conducted by AECL are varied in nature, so the QA program must provide adequate flexibility, while maintaining consistency with accepted quality standards. Well-written documentation adhering to the applicable decommissioning standards is a key factor. Manager commitment and input during the writing of the documentation are also important to ensure relevance of the QA program and effectiveness of implementation. Training in the use of the quality assurance plan and procedures is vital to the understanding of the QA program. Beyond the training aspect there is a need for the quality assurance program to be supported by a QA subject expert who is able to advise the group in implementing the Quality Program with consistency over the range of decommissioning work activities and to provide continual assessment of the quality assurance program for efficiency and effectiveness, with a concomitant continuous improvement process. (author)

  14. Experimental and analysis methods in radiochemical experiments

    Science.gov (United States)

    Cattadori, C. M.; Pandola, L.

    2016-04-01

    Radiochemical experiments made the history of neutrino physics by achieving the first observation of solar neutrinos (Cl experiment) and the first detection of the fundamental pp solar neutrinos component (Ga experiments). They measured along decades the integral νe charged current interaction rate in the exposed target. The basic operation principle is the chemical separation of the few atoms of the new chemical species produced by the neutrino interactions from the rest of the target, and their individual counting in a low-background counter. The smallness of the expected interaction rate (1 event per day in a ˜ 100 ton target) poses severe experimental challenges on the chemical and on the counting procedures. The main aspects related to the analysis techniques employed in solar neutrino experiments are reviewed and described, with a special focus given to the event selection and the statistical data treatment.

  15. Mixing and sampling tests for Radiochemical Plant

    International Nuclear Information System (INIS)

    The paper describes results and test procedures used to evaluate uncertainly and basis effects introduced by the sampler systems of a radiochemical plant, and similar parameters associated with mixing. This report will concentrate on experiences at the Barnwell Nuclear Fuels Plant. Mixing and sampling tests can be conducted to establish the statistical parameters for those activities related to overall measurement uncertainties. Density measurements by state-of-the art, commercially availability equipment is the key to conducting those tests. Experience in the U.S. suggests the statistical contribution of mixing and sampling can be controlled to less than 0.01 % and with new equipment and new tests in operating facilities might be controlled to better accuracy

  16. Stop and regulating valves for radiochemical plants

    International Nuclear Information System (INIS)

    The papers on developing stop and regulating valves and selecting material for these valves are reviewed. The main technical requirements of the radiochemical production to the valves are: the fluid stream velocity of 0.2-10000 l/s, remote exchange, one year lifetime or 15O00 cycles, the pressure and temperature of the fluid up to 0.6 MPa and up to 120 deg C. It is presupposed to use 03Cr18Ni12 and 03Cr18Ni12MO2-3 type steels for these valves. The CrWCO type lining materials for the working medium of nitrogen acid are acknowledged unfit. Significantly better anticorrosion properties are found for the electrodes with the 1Cr18Ni Mo2-5 Mn2-4 type lining. The bellow valve design meets the suggested technical requirements most fully. Two perspective projects of the valve of this type are presented

  17. Radiochemical Analysis Methodology for uranium Depletion Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Scatena-Wachel DE

    2007-01-09

    This report provides sufficient material for a test sponsor with little or no radiochemistry background to understand and follow physics irradiation test program execution. Most irradiation test programs employ similar techniques and the general details provided here can be applied to the analysis of other irradiated sample types. Aspects of program management directly affecting analysis quality are also provided. This report is not an in-depth treatise on the vast field of radiochemical analysis techniques and related topics such as quality control. Instrumental technology is a very fast growing field and dramatic improvements are made each year, thus the instrumentation described in this report is no longer cutting edge technology. Much of the background material is still applicable and useful for the analysis of older experiments and also for subcontractors who still retain the older instrumentation.

  18. Ignition Failure Mode Radiochemical Diagnostics Initial Assessment

    International Nuclear Information System (INIS)

    Radiochemical diagnostic signatures are well known to be effective indicators of nuclear ignition and burn reaction conditions. Nuclear activation is already a reliable technique to measure yield. More comprehensively, though, important quantities such as fuel areal density and ion temperature might be separately and more precisely monitored by a judicious choice of select nuclear reactions. This report details an initial assessment of this approach to diagnosing ignition failures on point-design cryogenic National Ignition Campaign targets. Using newly generated nuclear reaction cross section data for Scandium and Iridium, modest uniform doping of the innermost ablator region provides clearly observable reaction product differences between robust burn and failure for either element. Both equatorial and polar tracer loading yield observable, but indistinguishable, signatures for either choice of element for the preliminary cases studied

  19. OPERATIONAL EXPERIENCE: UPGRADED MPC AND A SYSTEMS FOR THE RADIOCHEMICAL PLANT OF THE SIBERIAN CHEMICAL COMBINE

    International Nuclear Information System (INIS)

    The success of reducing the risk of nuclear proliferation through physical protection and material control/accounting systems depends upon the development of an effective design that includes consideration of the objectives of the systems and the resources available to implement the design. Included among the objectives of the design are facility characterization, definition of threat, and identification of targets. When considering resources, the designer must consider funds available, rapid low-cost elements, technology elements, human resources, and the availability of resources to sustain operation of the end system. The Siberian Chemical Combine (SCC) is a multi-function nuclear facility located in the Tomsk region of Siberia, Russia. Beginning in 1996, SCC joined with the United States Department of Energy (US/DOE) Material Protection, Control, and Accounting (MPC and A) Program to develop and implement MPC and A upgrades for the Radiochemical, Chemical Metallurgical, Conversion, Uranium Enrichment, and Reactor Plants of the SCC. At the Radiochemical Plant the MPC and A design and implementation process has been largely completed for the Plutonium Storage Facility and related areas of the Radiochemical Plant. Design and implementation of upgrades for the Radiochemical Plant include rapid physical protection upgrades such as bricking up of doors and windows, and installation of security-hardened doors. Rapid material control and accounting upgrades include installation of modern balances and bar code equipment. Comprehensive MPC and A upgrades include the installation of access controls to sensitive areas of the Plant, alarm communication and display (AC and D) systems to detect and annunciate alarm conditions, closed circuit (CCTV) systems to assess alarm conditions, central and secondary alarm station upgrades that enable security forces to assess and respond to alarm conditions, material control and accounting upgrades that include upgraded physical

  20. Development of Commercial Neutron Activation Analysis Service with a Small Reactor

    International Nuclear Information System (INIS)

    It has been shown that with sufficient motivation the staff of a SLOWPOKE type reactor facility can develop a commercial NAA service generating enough revenues to pay the salaries of all those involved as well as reactor maintenance costs. The NAA service should be fast and continuously available; industry often requires a turn-around time of one day. At the École Polytechnique NAA Laboratory, years of work have led to the successful development of a hybrid NAA method combining the k0 method and the improved relative method. It offers large savings in time as well as improved flexibility and accuracy. (author)

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

    International Nuclear Information System (INIS)

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

  2. Radiochemical reprocessing of V-Cr-Ti alloy and its feasibility study

    Science.gov (United States)

    Bartenev, S. A.; Kvasnitskij, I. B.; Kolbasov, B. N.; Romanov, P. V.; Romanovskij, V. N.

    2004-08-01

    An extraction scheme for radiochemical reprocessing of an activated vanadium-chromium-titanium alloy after a fusion reactor decommissioning was developed and checked experimentally. It is based on extraction of V, Cr and Ti freed of activation products from the alloy dissolved in nitric acid. The solution of di-2-ethyl-hexyl-phosphoric acid (D2EHPA) in a hydrocarbon solvent (dodecane) serves as an extractant. It takes 50 extraction steps to recover V, Cr and Ti down to an effective dose rate Technical and economic analysis suggests that the reprocessing alternative is more attractive economically than the burial of spent V-Cr-Ti alloy components.

  3. Methods for nuclear material control used in the basic production of a typical radiochemical plant

    International Nuclear Information System (INIS)

    Techniques for destructive and non-destructive assay of the component and isotopic composition of nuclear materials are described, namely gravimetric, titrimetric, coulometric, mass spectrometry, as well as those based on registration of neutron and γ radiations. Their metrologic characteristics are described. The techniques described are suggested to be used for nuclear material (NM) control and accounting purposes at the model radiochemical plant for processing irradiated fuel subassemblies from power reactors. The measurement control program is also described. This program is intended for the measurement quality assurance in the framework of NM control and accountancy system

  4. Safety analysis for non-power reactors

    International Nuclear Information System (INIS)

    Non-power reactors have been operating in Canada since 1945, with NRU (National Research Universal, 1957) being the oldest operating non-power reactor. Presently, there are five generic 'types' of non-power reactors: NRU, ZED-2, SLOWPOKE, MNR and MAPLE, the latter undergoing commissioning as the MDS Medical Isotope Reactor. These reactors range in thermal power from 200 Watts to more than 100 MW. Other non-power reactors are likely to be built for new applications and to replace older reactors. The uniqueness of each reactor, the wide range of power levels and the evolution of safety philosophy over time have lead to non-uniform practices for safety analysis. This non-uniformity may be a problem for the preparation by the licensee and review by the regulator of the safety analysis report required for licensing of the reactor facility. Clearly, there is no universally applicable practice, while at the same time, expectations for safety analyses have evolved in order to demonstrate higher levels of overall safety. This paper examines a new 'graded approach' to preparing the safety analysis report for reactors of diverse features but with a common standard of safety. It discusses necessary content, methods and the training and qualification of the safety analyst. (author)

  5. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

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

  6. CANDU nuclear reactor technology

    International Nuclear Information System (INIS)

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

  7. 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.)

  8. 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)

  9. 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

  10. 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

  11. AECL review of CANDU 6 design in light of the Ontario Hydro nuclear IIPA technical findings

    International Nuclear Information System (INIS)

    In the spring of 1997, Ontario Hydro (OH) conducted an Independent, Integrated Performance Assessment (IIPA) to address long-standing management, process and equipment issues within the Ontario Hydro Nuclear (OHN) organization and its multi-unit CANDU stations. This review included six Safety System Functional Inspections (SSFIs) on: Bruce A Emergency Coolant Injection System; Bruce B Service Water Systems; Darlington Compressed Air Systems; Pickering Electrical Distribution Systems; Fire Protection (Programmatic); In-Service Environmental Qualification Program (Programmatic). Overall, the OHN inspections found that 'the design of the CANDU plant is robust and plant hardware (including equipment and materials), for the most part, is adequately reliable.' However, the SSFIs also identified a number of deficiencies in the areas of management, control of design/engineering, operations, training, maintenance, testing and quality assurance. Atomic Energy of Canada Limited (AECL) has undertaken an in-depth review of all design-related issues to assess their applicability and impact on the current CANDU 6 design. The AECL review has determined that equipment/design and programmatic deficiencies identified at the OLIN plants have been addressed in the current CANDU 6 design through an effective design feedback process and the application of modem codes and standards that were not in place during the design of the early OHN stations. Many of the design-related SSFI findings can be attributed to inadequate configuration management and the impact of unauthorized design modifications. Problems in these areas can arise at any nuclear station and prevention requires adherence to quality engineering procedures and documentation processes. (author)

  12. Multielement and automated radiochemical separation procedures for activation analysis

    International Nuclear Information System (INIS)

    In recent years the demand for information about the distribution of elements at trace concentration levels in high purity materials and in biological, environmental and geological specimens has increased greatly. Neutron activation analysis can play an important role in obtaining the required information. Radiochemical separations are required in many of the applications mentioned. A critical review of the progress made over the last 15 years in the development and application of radiochemical separation schemes for multielement activation analysis and in their automation is presented. About 80 radiochemical separation schemes are reviewed. Advantages and disadvantages of the automation of radiochemical separations are critically analysed. The various machines developed are illustrated and technical suggestions for the development of automated machines are given. (author)

  13. A radiochemical assay for biotin in biological materials

    International Nuclear Information System (INIS)

    A radiochemical assay for biotin is described. The assay was sensitive to one nanogram and simple enough for routine biotin analyses. The assay yielded results which were comparable to those obtained from a microbiological assay using Lactobacillus plantarum. (author)

  14. The use of robots for automation in the radiochemical laboratory

    International Nuclear Information System (INIS)

    The use of robotic systems for automated processes such as overnight operations, procedures involving radiation hazards in radiochemical laboratories is discussed. Particular reference is made to their use in analytical problems. Their flexibility is emphasised. (U.K.)

  15. CANDU: The fuel conserving reactor

    International Nuclear Information System (INIS)

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

  16. Radiochemical studies on primary coolant circuit of Kori 1 nuclear power plant

    International Nuclear Information System (INIS)

    I. The Radiochemical studies on the reactor coolant of Kori 1 nuclear power plant. α and γ spectrometry were tried to analyze the uranium content in the reactor coolant crud sample. The results from the analysis show that both methods were not applicable because of the great influences of other nuclides but the uranium. The investigation of the fission products release from the nuclear fuel rod to the reactor coolant shows that the releasing type is the diffusion. It was found out that the radioactivity levels of gamma emitting nuclide in the Kori 1 reactor coolant were rather higher than the standard. By the measurements of gamma emitting nuclides, it was found out that there were crud elements in the reactor coolant as the states of both the ions and the particles. And the ionic states are much more abundant than the particles. The cesium ratio in the reactor coolant indicates that the fuel burnt up is 7,000-8,O00 MWd/t. II. The analysis of the accumulated crud on the spent fuel rod of Kori 1 nuclear power plant. The crud sampling devices and the similar fuel assembly for the mock up test were constructed. The experiment was performed at the Kori 1 spent fuel storage pit and we have prepared ''The Technical Procedure for the Crud Sampling (5-3-8-5)''. (Author)

  17. Reactors

    International Nuclear Information System (INIS)

    Purpose: To provide a spray cooling structure wherein the steam phase in a bwr reactor vessel can sufficiently be cooled and the upper cap and flanges in the vessel can be cooled rapidly which kept from direct contaction with cold water. Constitution: An apertured shielding is provided in parallel spaced apart from the inner wall surface at the upper portion of a reactor vessel equipped with a spray nozzle, and the lower end of the shielding and the inner wall of the vessel are closed to each other so as to store the cooling water. Upon spray cooling, cooling water jetting out from the nozzle cools the vapor phase in the vessel and then hits against the shielding. Then the cooling water mostly falls as it is, while partially enters through the apertures to the back of the shielding plate, abuts against stoppers and falls down. The stoppers are formed in an inverted L shape so that the spray water may not in direct contaction with the inner wall of the vessel. (Horiuchi, T.)

  18. Thorium fuel-cycle studies for CANDU reactors

    International Nuclear Information System (INIS)

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

  19. 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)

  20. Radiochemical verification and validation in the environmental data collection process

    International Nuclear Information System (INIS)

    A credible and cost effective environmental data collection process should produce analytical data which meets regulatory and program specific requirements. Analytical data, which support the sampling and analysis activities at hazardous waste sites, undergo verification and independent validation before the data are submitted to regulators. Understanding the difference between verification and validation and their respective roles in the sampling and analysis process is critical to the effectiveness of a program. Verification is deciding whether the measurement data obtained are what was requested. The verification process determines whether all the requirements were met. Validation is more complicated than verification. It attempts to assess the impacts on data use, especially when requirements are not met. Validation becomes part of the decision-making process. Radiochemical data consists of a sample result with an associated error. Therefore, radiochemical validation is different and more quantitative than is currently possible for the validation of hazardous chemical data. Radiochemical data include both results and uncertainty that can be statistically compared to identify significance of differences in a more technically defensible manner. Radiochemical validation makes decisions about analyte identification, detection, and uncertainty for a batch of data. The process focuses on the variability of the data in the context of the decision to be made. The objectives of this paper are to present radiochemical verification and validation for environmental data and to distinguish the differences between the two operations

  1. RECONSTRUCTION OF 131I RELEASES FROM STACKS OF THE RADIOCHEMICAL PLANT OF THE MAYAK PRODUCTION ASSOCIATION FOR THE PERIOD FROM 1948 TO 1967

    Energy Technology Data Exchange (ETDEWEB)

    Glagolenko, Y. V.; Drozhko, Evgeniy G.; Mokrov, Y.; Pyatin, N. P.; Rovny, Sergey I.; Anspaugh, L. R.; Napier, Bruce A.

    2008-06-01

    The method of reconstruction of 131I releases from the Mayak PA Radiochemical Plant stacks for the period from 1948 to 1967 is proposed and the results of reconstruction are given. During this period of time, no continuous routine experimental monitoring of release was performed. As a result, reconstruction was carried out on the basis of earlier obtained data on deliveries of 131I to the radiochemical plants with irradiated uranium from the Mayak PA graphite - uranium reactors. The reconstruction also used calculation - experimental data on the iodine distribution in process solutions and in ventilation exhaust gases from the radiochemical plants, as well as in archive information on the efficiency of iodine trapping with the help of gas purification facilities. Available experimental data on 131I releases from the stacks of the radiochemical plants are given. The reconstruction results are presented as average monthly and annual releases of 131I from the stacks of radiochemical plants B and DB. The results are intended to be used for estimating doses to the population living in the vicinity of the enterprise in the 1950s-1960s.

  2. Reconstruction Of 131I Releases From Stacks Of The Radiochemical Plant Of The Mayak Production Association For The Period From 1948 To 1967

    International Nuclear Information System (INIS)

    The method of reconstruction of 131I releases from the Mayak PA Radiochemical Plant stacks for the period from 1948 to 1967 is proposed and the results of reconstruction are given. During this period of time, no continuous routine experimental monitoring of release was performed. As a result, reconstruction was carried out on the basis of earlier obtained data on deliveries of 131I to the radiochemical plants with irradiated uranium from the Mayak PA graphite - uranium reactors. The reconstruction also used calculation - experimental data on the iodine distribution in process solutions and in ventilation exhaust gases from the radiochemical plants, as well as in archive information on the efficiency of iodine trapping with the help of gas purification facilities. Available experimental data on 131I releases from the stacks of the radiochemical plants are given. The reconstruction results are presented as average monthly and annual releases of 131I from the stacks of radiochemical plants B and DB. The results are intended to be used for estimating doses to the population living in the vicinity of the enterprise in the 1950s-1960s.

  3. Studies on groundwater flow and radionuclide migration at underground environments. Final report of collaboration research between JAERI and AECL

    Energy Technology Data Exchange (ETDEWEB)

    Ogawa, Hiromichi; Nagao, Seiya; Yamaguchi, Tetsuji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2001-01-01

    The Japan Atomic Energy Research Institute (JAERI) conducted a collaboration program Phase II with the Atomic Energy of Canada Limited (AECL) from 1994 to 1998. The program was started to contribute the establishment of safety assessment methodology for the geological disposal of high-level radioactive wastes on the basis of the results from the Phase I program (1987-1993). The Phase II program consisted of following experimental items: (1) radionuclide migration experiments for quarried blocks (1m x 1m x 1m) of granite with natural fracture under in-situ geochemical conditions at 240 m level of Underground Research Laboratory of AECL; (2) study on the effects of dissolved organic materials extracted from natural groundwaters on radionuclide migration; (3) study on groundwater flow using environmental isotopes at two different geologic environments; (4) development of groundwater flow and radionuclide transport model for heterogeneous geological media. The mobility of radionuclides was retarded in the fracture by the deep geological conditions and the fracture paths. The groundwater humic substances with high molecular size were enhanced for the mobility of radionuclides in the sand and granitic media due to the complexation. The application of {sup 36}Cl and {sup 129}I for the analysis on the long-term groundwater flow can be validated on the basis of investigation at the URL site. Moreover, the geostatistical model for the analysis on groundwater flow and radionuclide migration was developed, and was able to describe the groundwater flow and the migration of environmental tracers at AECL sites. This report summaries the results of the Phase II program between JAERI and AECL. (author)

  4. Chemical and radiochemical specifications - PWR power plants

    International Nuclear Information System (INIS)

    Published by EDF this document gives the chemical specifications of the PWR (Pressurized Water Reactor) nuclear power plants. Among the chemical parameters, some have to be respected for the safety. These parameters are listed in the STE (Technical Specifications of Exploitation). The values to respect, the analysis frequencies and the time states of possible drops are noticed in this document with the motion STE under the concerned parameter. (A.L.B.)

  5. Radiochemical studies on environmental radioactivity in Sudan

    International Nuclear Information System (INIS)

    Measurements of uranium and thorium isotopes, 226 Ra, 210 Po, 228 Ra, 40 K and fallout radionuclide 137 Cs in soil samples collected from different districts in Sudan, rock phosphate samples collected from the uro and kurun rock phosphate deposits in the eastern part of the Nuba mountains in Western Sudan, and surface marine sediments and marine organisms collected from the sudanese coastal waters of the Red Sea have been made using a high resolution gamma-spectrometry, radiochemical separation and α spectrometry. The external exposure due to γ radiation from the ground has been calculated. The average exposure was found to be 45.4 ± 21.3 nGy/h, corresponding to the annual dose equivalent of 278 μSv/y. With the exception of some areas, the calculated exposure falls within the global wide range of outdoor radiation exposure given in the UNSCEAR publications. The nation-wide average concentrations of 226 Ra, 238 U, 232 Th, 40 K and 137 Cs determined were 31.6 ± 27, 20.1 ± 16.4, 19.1 ± 8.1, 280.3 ± 137.6 and 4.1 ± 4.3 Bq/Kg, respectively. This shows that there is little contamination due to fallout radioactivity at survey sites. The exchangeable radium fraction constitutes 19-24% of the total radium content. The data show that 238 U and its decay products are the principal contributors of radioactivity in both phosphate deposits at Uro and Kurun. The equivalent mass concentrations of uranium in the Uro rock phosphate fall within the range that could be economically recovered as the by-product of fertilizer industry. The mean activity concentrations weighted by average agricultural consumption of 300 kg/ha of untreated ground rock fertilizer resulted in an annual distribution of 120.63 Bq Ra/m2 with Uro rock and 12.97, 0.21 and 4.24 Bq/m2 respectively, with Kurun rock fertilizer. The external radiation exposure over agricultural areas was estimated 23.41 x 10 -9 Gy/h and 2.59 x 10 -9 Gy/h at 1 m above ground level for Uro and Kurun rock phosphate fertilizers

  6. Radiochemical analysis for nuclear waste management in decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    Hou, X. (Technical Univ. of Denmark, Risoe National Lab. for Sustainable Energy. Radiation Research Div., Roskilde (Denmark))

    2010-07-15

    The NKS-B RadWaste project was launched from June 2009. The on-going decommissioning activities in Nordic countries and current requirements and problems on the radiochemical analysis of decommissioning waste were discussed and overviewed. The radiochemical analytical methods used for determination of various radionuclides in nuclear waste are reviewed, a book was written by the project partners Jukka Lehto and Xiaolin Hou on the chemistry and analysis of radionuclide to be published in 2010. A summary of the methods developed in Nordic laboratories is described in this report. The progresses on the development and optimization of analytical method in the Nordic labs under this project are presented. (author)

  7. Production and radiochemical separation of rhodium-105 for radiopharmaceutical applications

    International Nuclear Information System (INIS)

    Production of no carrier added 105Rh by thermal neutron irradiation of natural Ru target and its radiochemical separation to obtain radionuclidically and radiochemically pure species is described. Irradiation planning, post irradiation chemical separation of 105Rh from ruthenium and iridium radionuclidic contaminants are discussed in detail. 100 mg of natural Ru was irradiated in a thermal neutron flux of 3 x 1013 neutrons/cm2/s for 7 days. Irradiated target was dissolved in a mixture of KIO4 and KOH in presence of water. Solvent extraction with CCl4 was used for removal of 97Ru and 103Ru, whereas solvent extraction with TBP was used for the removal of traces of 192Ir impurity. Different radiochemical separation techniques were tested for the recovery of 105Rh and its final purification from the accompanying radionuclidic impurities. The salts associated with the product were subsequently removed by fractional precipitation and by cation exchange chromatography. The radionuclidic purity of the 105Rh separated was estimated by gamma ray spectrometry; no radionuclidic contamination was observed. The radiochemical purity of the 105RhCl3 was ascertained by paper chromatography and paper electrophoresis. (orig.)

  8. Trace Analysis of Ancient Gold Objects Using Radiochemical Neutron Activation

    CERN Document Server

    Olariu, A; Constantinescu, O; Badica, T; Popescu, I V; Besliu, C; Leahu, D; Olariu, Agata; Constantinescu, Mioara; Leahu, Doina

    1999-01-01

    Radiochemical neutron activation analysis has been applied to investigate the microelements in gold samples with archaeological importance. Chemical separation has allowed the determination of traces of Ir, Os, Sb, Zn, Co, Fe, Ni. Instrumental neutron activation analysis has been used for the determination of Cu.

  9. One-step high-radiochemical-yield synthesis of [18F]FP-CIT using a protic solvent system

    International Nuclear Information System (INIS)

    Although [18F] fluoropropylcarbomethoxyiodophenylnortropane (FP-CIT) is a promising radiopharmaceutical for dopamine transporter imaging, it has not been used for clinical studies because of low radiochemical yield. The purpose of our study was to develop a new radiochemistry method using a protic solvent system to obtain a high radiochemical yield of [18F]FP-CIT in single-step manual and automatic preparation procedures. [18F]F- was trapped on a QMA Sep-Pak cartridge or PS-HCO3 cartridge and eluted with Cs2CO3/K222 buffer or TBAHCO3, respectively, or 8 μl of TBAOH was added directly to [18F]F-/H218O solution in a reactor without using a cartridge. After drying, [18F] fluorination was performed with 2-6 mg of mesylate precursor, 100 μl of CH3CN and 500 μl of t-BuOH at 50-120oC for 5-30 min, followed by high-performance liquid chromatography (HPLC) purification to obtain the final product. For comparison, the same procedure was performed with a tosylate precursor. Manual synthesis gave a decay-corrected radiochemical yield of 52.2±4.5%, and optimal synthesis conditions were as follows: TBAOH addition, 4 mg of precursor, 100oC and 20 min of [18F] fluorination (n=3). We obtained low radiochemical yields of [18F]FP-CIT with carbonate elution systems such as Cs2CO3 or TBAHCO3. We also developed an automatic synthesis method based on manual synthesis results. In automatic production, we obtained a decay-corrected radiochemical yield of 35.8±5.2% after HPLC purification, and we did not have any synthesis failures (n=14). Here, we describe our new method for the synthesis of [18F]FP-CIT using a protic solvent system. This method gave a high radiochemical yield with high reproducibility and might enable [18F]FP-CIT to be used clinically and commercially

  10. Build your own Candu reactor

    International Nuclear Information System (INIS)

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

  11. Radiochemical analysis of radionuclides difficult to measure for waste characterization in decommissioning of nuclear facilities

    International Nuclear Information System (INIS)

    Some radiochemical analytical methods for the determination of important beta-radionuclides for decommissioning are presented. An analytical method is briefly described, which is used for the determination of 3H and 14C in steel and aluminum by combustion using commercial oxidizer. A leaching method was developed for the determination of 3H in the contaminated silica gel. A simple distillation method is presented for the determination of 14C in heavy water and wastewater sample. A method developed for the simultaneous determination of 3H, 14C, 36Cl, 55Fe, 63Ni, 41Ca and 129I in concrete, graphite, aluminium, lead, and steel is presented. The developed methods have been successfully used to analyse various materials for characterization of the waste during the decommissioning of Danish nuclear reactors. (author)

  12. Radiochemical reprocessing of V-Cr-Ti alloy and its feasibility study

    International Nuclear Information System (INIS)

    An extraction scheme for radiochemical reprocessing of an activated vanadium-chromium-titanium alloy after a fusion reactor decommissioning was developed and checked experimentally. It is based on extraction of V, Cr and Ti freed of activation products from the alloy dissolved in nitric acid. The solution of di-2-ethyl-hexyl-phosphoric acid (D2EHPA) in a hydrocarbon solvent (dodecane) serves as an extractant. It takes 50 extraction steps to recover V, Cr and Ti down to an effective dose rate <12.5 μSv/h, permitting the refabrication of these metals without biological shielding from ionizing radiation. Technical and economic analysis suggests that the reprocessing alternative is more attractive economically than the burial of spent V-Cr-Ti alloy components

  13. Radiochemical studies in the development of deep geological repository in the Czech Republic

    International Nuclear Information System (INIS)

    In this contribution the main achievements of radiochemical studies performed within the framework of the Czech DGR development programme are summarized and further plans outlined. The results of selection of the most dangerous radionuclides in spent fuel assemblies from VVER 440 reactors, based on spent fuel inventory calculations and analyses of migration rate of radionuclides to the environment, are presented in the first part of the contribution. It is shown that 14C, 129I, 126Sn, 135Cs, 36Cl, 79Se, 226Ra, 237Np, 229Th, and 242Pu belong among the most dangerous radionuclides in the Czech disposal concept. Problems with the determination of migration parameters of radionuclides are described in the second part of this contribution. (author)

  14. Thorium base fuels reprocessing at the L.P.R. (Radiochemical Processes Laboratory) experimental plant

    International Nuclear Information System (INIS)

    The availability of the LPR (Radiochemical Processes Laboratory) plant offers the possibility to demonstrate and create the necessary technological basis for thorium fuels reprocessing. To this purpose, the solvents extraction technique is used, employing TBP (at 30%) as solvent. The process is named THOREX, a one-cycle acid, which permits an adequate separation of Th232 and U233 components and fission products. For thorium oxide elements dissolution, the 'chopp-leach' process (installed at LPR) is used, employing a NO3 H 13N, 0.05M FH and 0.1M Al (NO3)3, as solvent. To adapt the pilot plant to the flow-sheet requirements proposed, minor modifications must be carried out in the interconnection of the existing decanting mixers. The input of the plant has been calculated by Origin Code modified for irradiations in reactors of the HWR type. (Author)

  15. The high-temperature gas-cooled reactor VGR-50 for power-technological use

    International Nuclear Information System (INIS)

    The VGR-50 reactor is the first experimental commercial prototype of a reactor with continuous fuel circulation in the reactor core. The development of the reactor design is described together with is use for electricity production and as a source of radiations for radio-chemical and energo-technological processes. (U.K.)

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

    International Nuclear Information System (INIS)

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

  17. Safety upgrades to the NRU research reactor

    International Nuclear Information System (INIS)

    The NRU (National Research Universal) Reactor is a 135 MW thermal research facility located at Chalk River Laboratories, and is owned and operated by Atomic Energy of Canada Limited. One of the largest and most versatile research reactors in the world, it serves as the R and D workhorse for Canada's CANDU business while at the same time filling the role as one of the world's major producers of medical radioisotopes. AECL plans to extend operation of the NRU reactor to approximately the year 2005 when a new replacement, the Irradiation Research Facility (IRF) will be available. To achieve this, AECL has undertaken a program of safety reassessment and upgrades to enhance the level of safety consistent with modem requirements. An engineering assessment/inspection of critical systems, equipment and components was completed and seven major safety upgrades are being designed and installed. These upgrades will significantly reduce the reactor's vulnerability to common mode failures and external hazards, with particular emphasis on seismic protection. The scheduled completion date for the project is 1999 December at a cost approximately twice the annual operating cost. All work on the NRU upgrade project is planned and integrated into the regular operating cycles of the reactor; no major outages are anticipated. This paper describes the safety upgrades and discusses the technical and managerial challenges involved in extending the operating life of the NRU reactor. (author)

  18. Innovative reactor technologies - Enabling success

    International Nuclear Information System (INIS)

    Many innovative reactors are being discussed, offering advantages in economics, sustainability, environmental impact, versatility and efficiency. To be successful, however, innovative reactors must meet the requirements for a successful build project. This requires achieving the mixture of innovation and proveness required to meet the first-of-a-kind hurdle. Based on the successful CANDU 6 reactor, a design still being built today, the ACR adds specific innovations in key areas chosen to achieve a balanced design. Capital cost has been significantly reduced by optimising the reactor-core design and simplifying systems. Key changes in this area include a move from a heavy water coolant to a light water coolant, and the adoption of SEU fuel. Construction times have also been reduced by using a modular design that takes advantage of modern construction techniques. Operating performance has been enhanced through improvements in system materials, equipment layout and component specifications. In parallel with these priorities, design adaptations have been applied so as to increase safety margins and defence-in-depth, again adding to the confidence in ACR licensability. The ACR development plan includes early review by regulators to reduce licensing risk, with international regulatory review having commenced. Overall, this places the ACR in a good position to meet the first-of-a-kind challenge, a necessary condition to enabling the success of an innovative reactor. AECL sees a logical evolution from the ACR, via increasing temperature and pressure capability, to the SCWR (Supercritical Water Reactor). AECL's CANDU-X program is already looking at designs for this concept. Inherent features of both ACR and the fuel channel SCWR lend themselves to different fuel cycles for the future. One of the prominent characteristics of the heavy-water moderated fuel channel reactor approach is the high potential for innovation. The evolutionary path allows innovation in practical steps

  19. 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

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

    International Nuclear Information System (INIS)

    Microbial populations in groundwaters from AECL's Underground Research Laboratory (URL) range from 103 to 105 cells/mL. Based on the total dissolved organic carbon (DOC), nitrate and phosphate content of these waters, populations of about 105 to 107 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 105 to 4 x 106 cells/mL, regardless of the initial size of the population (∼101 to 104 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 conditions. Similar effects were observed in R2A. This confirms a more efficient use of

  1. First experimental results of the thermal behaviour of AECL's CANSTOR spent fuel dry storage module

    International Nuclear Information System (INIS)

    This paper presents the first experimental results of the thermal behavior of AECL's CANSTOR spent fuel dry storage module. The CANSTOR module is an air-cooled concrete vault about 22 m long, 8 m wide and 7 m high. It can store 12000 CANDU spent fuel bundles inside 200 baskets which are stacked into two rows of 10 storage cylinders. The first module was built on the site of Hydro-Quebec's Gentilly-2 station during the summer of 1995. Dissipation of the residual heat generated by the spent fuel is a major factor in spent fuel dry storage design and one of the key elements for its licensing. The fuel temperature must be kept below 160 deg C to avoid oxidation. Experiments on a mock-up and calculations showed that the air cooling circuit provides at least 15 deg C margin for the fuel with 6-year cooled fuel subject to the ambient design temperature of 40 deg C. Nevertheless, the Atomic Energy Control Board of Canada (AECB) requested Hydro-Quebec to monitor the temperatures and limit the age of the fuel to more than 8-year cooled. During the construction, fourteen temperature sensors were installed to measure the temperature of the air, concrete and top of storage cylinders. A computer based data acquisition system has been used to collect the data, starting before the first fuel was loaded. The first loading campaign occurred during the fall of 1995, mainly during the months of October and November. The module was half filled with 6000 bundles that had been cooled in the spent fuel bay for more than 8 years, in accordance with the AECB license. No loading was done during the 1995-1996 winter. This provided a few months of data with quasi-constant power dissipation. This paper presents this data and compares it with the calculations used in support of the licensing submission. It is shown that fuel of much less than 8-year cooled could be loaded into the CANSTOR module. (author)

  2. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Palleck, S.J.; Sim, K.S. [Atomic Energy of Canada Ltd., Mississauga, Ontario (Canada); Gheorghiu, C. [Institute of Nuclear Research (Romania)

    2001-07-01

    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)

  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. 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

  6. Radiochemical analyses for radionuclide estimation in environmental samples

    International Nuclear Information System (INIS)

    Radioactivity is not only a residuary product of nuclear energy. It is also a normal constituent of the earth's crust. The stellar material from which the earth was formed about 4.5 billion years ago contained many unstable nuclides. The majority of these unstable nuclides have long time since decayed into stable elements. However, some of the original (primordial) nuclides, whose half-lives, are about as long as the earth's age, are still present. In recent decades the activity levels have been enhanced by the addition of man-made radionuclides, mainly from fallout due to the atmospheric testing of nuclear weapons during the 1950' s and 1960' s and from controlled and accidental discharges of radioactive effluents from nuclear installations. Variety of radiochemical techniques are employed in the determination of natural radionuclides, transuranes, and fission products present in water, soil, ores, tailings, vegetation, biological tissue, filters, resins, etc. at low levels. Sample breakdown and radionuclide solubilization is accomplished by wet-ashing or dry-ashing in a muffle furnace, depending on the volatility of the radionuclide(s ) of interest. Because of the low-level nature of these samples, subsequent radiochemical separators are generally done sequentially from a single sample. Radiochemical operations include coprecipitation, ion exchange, and solvent extraction. Each sample is spiked with the appropriate carriers and yield tracers prior to radiochemical analysis. Following separation, each radionuclide fraction is converted to a suitable form for counting using precipitation or electrodeposition. This source is counted for alpha, beta, or gamma using alpha spectrometry, gas proportional counting, liquid scintillation counting, or gamma spectroscopy. (authors)

  7. Decontamination and decommission of a radiochemical laboratory building complex

    International Nuclear Information System (INIS)

    Full text: Handling of unsealed radioactive substances for research and development purposes in chemical or pharmaceutical industries or research centres as well as production of radioactive substances (e.g. for applications in nuclear medicine or industry) requires operation of special radiochemical laboratories. In general, operation of radiochemical laboratories is strongly regulated by the government and national authorities. The operator needs a permit related to radiological protection. In general, technical requirements for such facilities are very high. To ensure high safety standards with respect to the employees and the environment, several radiological protection measures have to be taken. These measures (for example special shielding or ventilation and waste water systems) depend on various factors, e.g. activity in use, kind of nuclides, chemical properties and volatility of substances. In order to close-down such radiochemical laboratories some radiological protection measures have to be maintained to ensure protection of both humans and the environment induced by possible residual contaminations within the facility including technical inventory. However, a later reuse of the facility as a non-radioactive facility requires removal of all radioactive contamination with respect to national regulation. Resulting radioactive wastes have to be disposed of under control of competent authorities. Based on the experience of a decontamination and decommission project for a former radiochemical laboratory complex, the main steps necessary to release such a facility are discussed. Analytical aspects of initial conditions, necessary organisational structures within the project, resources needed estimation and exploration of the radiological situation in the laboratory, elaboration of a measuring strategy and decontamination methods as well as different waste disposal routes in relation to different waste types are reported. (author)

  8. Radiochemical regularities of migration mobility of Chernobyl' discharge radionuclides

    International Nuclear Information System (INIS)

    Data on the radionuclude (RN) migration in environment later the Chernobyl' accident are generalized. Introduction of fallout of the radioactive discharge into environment causes necessity to account and to study different factors of geochemical and physicochemical character determining further RN behaviour in the medium. For a well-founded forecast of the behaviour it is necessity to use a complex of radiochemical and physicochemical research, lying in the base of radiation monitoring of environment. 1 refs

  9. Rapid radiochemical separation of zirconium-95 and niobium-95

    International Nuclear Information System (INIS)

    A rapid method for the quantitative separation of 95Zr and 95Nb has been developed. The method is based on the ion flotation of cationic zirconium complex ions with sodium lauryl sulfate (NaLS) from niobium which is masked with hydrogen peroxide. The separation was applied to mixtures of 95Zr and 95Nb initially in oxalic acid solution and quantitative recoveries of the radiochemically pure radioisotopes were obtained. (orig.)

  10. Study on the Radiochemical Separation of 142La

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In order to increase the diagnostic sensitivity of nuclear material fine fission, short half-life fission-product nuclides are used. The precision of many nuclear data of short half-life fission-products is not well, so they must be measured by more precise method. At first separating and preparing radiochemically pure radionuclide is needed.As the existence of more than one isotopes of an element, it is impossible that individual

  11. Analysis of radiochemical impurities in radiopharmaceuticals. Pt. 1

    International Nuclear Information System (INIS)

    In a study of the relevant literatur analytical methods for the determination of radiochemical purity of pharmaceuticals were compiled in form of a dictionary. It contains data on studies of 71 radiopharmaceuticals and a total of 123 analyses. About half of the substances were labelled with I 131. The most frequently used investigation method was thin-layer chromatography (75 analyses), followed by paper chromatography (36 analyses), electrophoresis (10 analyses) and high-pressure liquid chromatography (2 analyses). (orig.)

  12. A radiochemical assay for glycolytic activity in dental plaque

    International Nuclear Information System (INIS)

    A radiochemical technique for the rapid and precise measurement of glucose utilization of fresh samples of dental plaque is described. The method appears to be a sensitive indicator of the actual glycolytic ability of an organized microbial plaque including activity in the Embden-Meyerhof pathway as well as heterolactic fermentation and the Entner-Doudoroff pathway. It was found that the glycolytic rate of plaque associated with periodontal pockets was significantly higher than that for plaque not so associated. (author)

  13. Radiochemical Mix Diagnostic in the Presence of Burn

    Energy Technology Data Exchange (ETDEWEB)

    Hayes, Anna C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-01-28

    There is a general interest in radiochemical probes of hydrodamicalmix in burning regions of NIF capsule. Here we provide estimates for the production of 13N from mixing of 10B ablator burning hotspot of a capsule. By comparing the 13N signal with x-ray measurements of the ablator mix into the hotspot it should be possible to estimate the chunkiness of this mix.

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

    International Nuclear Information System (INIS)

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

  15. Radiochemical studies in development of deep geological repository in the Czech Republic

    International Nuclear Information System (INIS)

    The coordinated development of a deep geological repository (DGR) started in the Czech Republic as early as 1993. The first stage of DGR development programme was focused mainly on gathering information from foreign programmes. The next stage, which is now in progress, consists in preparation of methodologies and laboratory research focused on acquiring data needed for proving safety of the Czech concept of the deep geological repository. Radiochemical studies play a crucial role in the safety assessment of a deep geological repository. Solubility of radionuclides in pore water of repository materials and sorption of radionuclides on these materials present primary parameters needed for the repository safety evaluation. In this contribution the main achievements of radiochemical studies performed in the framework of the Czech DGR development programme are summarized and further plans outlined. The results of selection of critical radionuclides of spent fuel from WWER 440 and 1000 reactors, based on spent fuel inventory calculations and review of solubility of radionuclides in groundwaters are presented in the first part of the contribution. Sorption experiments have been focused primarily on the development of methodology and understanding the problems connected with description of transport of radionuclides from spent fuel through bentonite buffer to geosphere. 137Cs was used in most of sorption experiments devoted to studying methodology problems mainly due to the simplicity of its species and easy detection. It is shown that even evaluation of sorption of this radionuclide, occurring only in single valence state and not forming complexes, is not straightforward and cannot be described by simple Kd model. Some of the results of modelling of sorption of Cs using a surface complexation model are shown. Sorption of technetium, as a representative of redox sensitive radionuclides, has been studied under various conditions. It is shown that the presence of some

  16. OE Management at Research Technology Operations, Chalk River Laboratories, AECL, Canada

    International Nuclear Information System (INIS)

    Brief description of nuclear facility. A nuclear installation consisting of a 130 MW research reactor and 13 licensed nuclear facilities, staffed by ∼2600 employees, on three distinct sites. Main activities include: (1) Reactor development; (2) CRL nuclear operations; (3) Research and development; (4) Isotope production; (5) Waste management and decommissioning. Overview of OE arrangements. A centralized OE group that is permanently resourced and trained to support the organization. The group is spread over two time zones and supported by a cadre of permanently dedicated OE coordinators and action tracking coordinators throughout the organization

  17. MAPLE: a Canadian multipurpose reactor concept for national nuclear development

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited, following an investigation of Canadian and international needs and world-market prospects for research reactors, has developed a new multipurpose concept, called MAPLE (Multipurpose Applied Physics Lattice Experimental). The MAPLE concept combines H2O- and D2O-moderated lattices within a D2O calandria tank in order to achieve the flux advantages of a basic H2O-cooled and moderated core along with the flexibility and space of a D2O-moderated core. The SUGAR (Slowpoke Uprated for General Applied Research) MAPLE version of the conept provides a range of utilization that is well suited to the needs of countries with nuclear programs at an early stage. The higher power MAPLE version furnishes high neutron flux levels and the variety of irradiation facilities that are appropriate for more advanced nuclear programs

  18. Rapid Radiochemical Analyses in Support of Fukushima Nuclear Accident - 13196

    Energy Technology Data Exchange (ETDEWEB)

    Maxwell, Sherrod L.; Culligan, Brian K.; Hutchison, Jay B. [Savannah River National Laboratory, Building 735-B, Aiken, SC 29808 (United States)

    2013-07-01

    discussed. Air filter samples were reported within twenty-four (24) hours of receipt using rapid techniques published previously. [11] The rapid reporting of high quality analytical data arranged through the U.S. Department of Energy Consequence Management Home Team was critical to allow the government of Japan to readily evaluate radiological impacts from the nuclear reactor incident to both personnel and the environment. SRNL employed unique rapid methods capability for radionuclides to support Japan that can also be applied to environmental, bioassay and waste management samples. New rapid radiochemical techniques for radionuclides in soil and other environmental matrices as well as some of the unique challenges associated with this work will be presented that can be used for application to environmental monitoring, environmental remediation, decommissioning and decontamination activities. (authors)

  19. Rapid Radiochemical Analyses in Support of Fukushima Nuclear Accident - 13196

    International Nuclear Information System (INIS)

    within twenty-four (24) hours of receipt using rapid techniques published previously. [11] The rapid reporting of high quality analytical data arranged through the U.S. Department of Energy Consequence Management Home Team was critical to allow the government of Japan to readily evaluate radiological impacts from the nuclear reactor incident to both personnel and the environment. SRNL employed unique rapid methods capability for radionuclides to support Japan that can also be applied to environmental, bioassay and waste management samples. New rapid radiochemical techniques for radionuclides in soil and other environmental matrices as well as some of the unique challenges associated with this work will be presented that can be used for application to environmental monitoring, environmental remediation, decommissioning and decontamination activities. (authors)

  20. Rapid radiochemical analyses in support of Fukushima nuclear accident

    International Nuclear Information System (INIS)

    ) hours of receipt using rapid techniques published previously. The rapid reporting of high quality analytical data arranged through the U.S. Department of Energy Consequence Management Home Team was critical to allow the government of Japan to readily evaluate radiological impacts from the nuclear reactor incident to both personnel and the environment. SRNL employed unique rapid methods capability for radionuclides to support Japan that can also be applied to environmental, bioassay and waste management samples. New rapid radiochemical techniques for radionuclides in soil and other environmental matrices as well as some of the unique challenges associated with this work will be presented that can be used for application to environmental monitoring, environmental remediation, decommissioning and decontamination activities.

  1. RAPID RADIOCHEMICAL ANALYSES IN SUPPORT OF FUKUSHIMA NUCLEAR ACCIDENT

    Energy Technology Data Exchange (ETDEWEB)

    Maxwell, S.

    2012-11-07

    reported within twenty-four (24) hours of receipt using rapid techniques published previously. The rapid reporting of high quality analytical data arranged through the U.S. Department of Energy Consequence Management Home Team was critical to allow the government of Japan to readily evaluate radiological impacts from the nuclear reactor incident to both personnel and the environment. SRNL employed unique rapid methods capability for radionuclides to support Japan that can also be applied to environmental, bioassay and waste management samples. New rapid radiochemical techniques for radionuclides in soil and other environmental matrices as well as some of the unique challenges associated with this work will be presented that can be used for application to environmental monitoring, environmental remediation, decommissioning and decontamination activities.

  2. 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

  3. Radiochemical estimation of neutron fluence of Hiroshima and Nagasaki atomic bombs

    International Nuclear Information System (INIS)

    Purpose: To estimate neutron fluence of Hiroshima and Nagasaki atomic bombs by radiochemical methods. Methods: Thermal neutron fluence at the time of explosion was estimated from the results of radiochemical analysis of residual 60Co in iron materials or iron products. Results: Materials were obtained through the kindness of Dr. Masanori Nakaidzum. The distribution of neutron fluence in Hiroshima and Nagasaki can be determined by measuring the residual radioactivity of many pieces of material by radiochemical methods presented in the paper

  4. IFPE/AECL-BUNDLE, Fission Gas Release and Burnup Analysis, PHWR Fuel

    International Nuclear Information System (INIS)

    Description: Prototype Candu Fuel bundles for the CANDU6 (bundle NR) and Bruce (bundle JC) reactors were irradiated in the NRU experimental reactor at Chalk River Laboratories in experimental loop facilities under typical Candu reactor conditions, except that they were cooled using light water. NEA-1596/01 - Description: Bundle JC was a prototype 37-element fuel bundle for the Bruce-A Ontario Hydro reactors. This pressurized heavy water reactor (PHWR) design utilizes a heavy water moderator and pressurize heavy water coolant. For irradiation in the NRU reactor, the centre fuel element was removed and replaced by a central tie rod for irradiation purposes in the vertical test section. Coolant for the test was pressurized light water under typical PHWR conditions of 9 to 10.5 MPa and 300 deg. C. The fuel elements used 1.55 wt% U-235 in U uranium dioxide fuel and were clad with Zircaloy-4 material. The bundles' elements were coated with a graphite coating. The fuel is somewhat atypical of 37 element-type fuel since the length to diameter ratio (l/d) is large (1.73) due to the pellets being ground down from a OD of 14.3 mm to 12.12 mm. The outer element burnup averaged approximately 640 MWh/kgU on discharge. Outer element powers varied between 57 kW/m near the beginning of life and 23 kW/m at discharge. Due to the long irradiation, the bundle experienced 153 short shutdowns, and 129 longer duration shutdowns. No element instrumentation was used during the irradiation. However, the bundle was subjected to extensive post-irradiation examination (PIE) that included dimensional changes, fission gas release, fuel burnup analysis, and metallography that included grain size measurement. NEA-1596/02 - Description: Bundle NR was a prototype 37-element fuel bundle for the Candu 600 reactor. This pressurized heavy water reactor (PHWR) design utilizes a heavy water moderator and pressurized heavy water coolant. For irradiation in the NRU reactor, the centre fuel element was

  5. 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)

  6. 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)

  7. 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)

  8. Radiochemical separation methods for preparation of biomedical cyclotron radionuclides

    International Nuclear Information System (INIS)

    A short review of the radiochemical methods for preparation of widely used or promising cyclotron-produced radionuclides for nuclear medicine and biomedical or environmental studies is given. The presented data include the current status of the production of some gamma-emitters (97Ru, 111In, 123I, 201Tl), generator-pairs (68Ge/68Ga, 82 Sr/82Rb, 128Ba/128Cs, 178W/178Ta), radioisotopes for metabolism studies (26Al, 67Cu, 237Pu) and actinides tracers for environmental researches (235Np, 236Np, 236Pu). The conditions for preparation of high-purity isotopes have been investigated and procedures including target chemistry design were developed. (author)

  9. Fast radiochemical separations with an automated rapid chemistry apparatus

    International Nuclear Information System (INIS)

    The microcomputer controlled Automated Rapid Chemistry Apparatus, ARCA, is described together with the He(KCl) gas-jet and the target and recoil chamber as it was developed and used in experiments at the heavy ion accelerator UNILAC. This set-up allows in a fast and reproducible way to carry out automated high performance liquid chromatographic separations in a chemically inert apparatus. Its modular design makes a large variety of different types of radiochemical separations easily possible. As examples a group separation from our search for superheavy elements and a separation of the elements Md, No and Lr is discussed. (orig.)

  10. Establishment of a radiochemical procedure for the obtainment strontium titanate

    International Nuclear Information System (INIS)

    A research that aims to develop radiochemical procedures for the separation and solidification of 137Cs and 90Sr from fission products solutions has been carried out at the Radiochemistry Division of IPEN-CNEN/SP. In a previous paper(1), a schematic outline of the process steps for the separation of 90Sr from fission products mixture was shown. In the present paper, the experimental conditions for solidification of strontium as strontium titanate have been studied. According to literature data, this compound offers the most suitable chemical form to use 90Sr as beta rays source. (author)

  11. Studies on some Indian paints for radiochemical plants

    International Nuclear Information System (INIS)

    The choice of paints in areas subjected to contamination and radiation in nuclear installation need special attention. The types of generic coatings are examined with reference to these requirements. Among those examined, certain types of epoxy paints are found to be attractive for these applications. Samples of epoxy paints obtained from some Indian manufacturers are tested for their suitability. Decontaminability and radiation resistance properties are also evaluated with special reference to radiochemical plants. Important specifications for such applications are listed. This report summarizes the results of these studies. (author)

  12. Guiding Principles for Sustainable Existing Buildings: Radiochemical Processing Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Pope, Jason E.

    2013-11-11

    In 2006, the United States (U.S.) Department of Energy (DOE) signed the Federal Leadership in High Performance and Sustainable Buildings Memorandum of Understanding (MOU), along with 21 other agencies. Pacific Northwest National Laboratory (PNNL) is exceeding this requirement and, currently, about 25 percent of its buildings are High Performance and Sustainable Buildings. The pages that follow document the Guiding Principles conformance effort for the Radiochemical Processing Laboratory (RPL) at PNNL. The RPL effort is part of continued progress toward a building inventory that is 100 percent compliant with the Guiding Principles.

  13. Correction and verification of AECL Bonner Sphere response matrix based on mono-energetic neutron calibration performed at NPL

    International Nuclear Information System (INIS)

    The AECL Bonner Sphere Spectrometer (BSS) was taken to National Physical Laboratory (NPL) for calibration in mono-energetic neutron fields and bare 252Cf neutron fields. The mono-energetic radiations were performed using ISO-8529 prescribed neutron energies: 0.071, 0.144, 0.565, 1.2, 5 and 17 MeV. A central SP9 proportional counter was also evaluated at the NPL thermal neutron calibration facility in order to assess an effective pressure of 3He inside the counter, i.e. number density of 3He atoms. Based on these measurements and methods outlined by Thomas and Soochak, a new BSS response matrix was generated. The response matrix is then verified by unfolding spectra corresponding to various neutron fields. Those are NPL bare 252Cf source, National Institute of Standards and Technology bare and heavy water moderated 252Cf source and 241AmBe calibration source located at National Research Council. A good agreement was observed with expected neutron fluence rates, as well as derived dosimetric quantities, such as International Commission on Radiological Protection-74 ambient dose equivalent. The AECL BSS response matrix was created based on methods proposed by Wiegel et al., Thomas and Thomas and Soochak. The response matrix was further corrected for the mono-energetic neutron measurements taken and NPL. In order to experimentally verify the response matrix, four neutron measurements were taken at three laboratories: NPL, NIST and NRC. Good agreement with expected values both for integrated neutron fluence and derived dosimetric quantities was observed in all four cases. (authors)

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

    International Nuclear Information System (INIS)

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

  15. METHODS FOR RECONSTRUCTION OF RADIONUCLIDE COMPOSITION AND ACTIVITY OF FISSION PRODUCTS ACCUMULATED IN THE IRRADIATED URANIUM AT THE MOMENT OF ITS RADIOCHEMICAL REPROCESSING AT PLANT “B”, “MAYAK” PA IN THE EARLY 1950s

    Energy Technology Data Exchange (ETDEWEB)

    Glagolenko, Y. V.; Drozhko, Evgeniy G.; Mokrov, Y.; Rovny, Sergey I.; Lyzhkov, A. V.; Anspaugh, L. R.; Napier, Bruce A.

    2008-06-01

    The article describes calculation procedure for reconstruction of radionuclide composition and activity of fission fragments accumulated in the irridated uranium from “Mayak” PA graphite-uranium reactors at the moment, when irradiation is completed, and at the moment, when the uranium is transferred to radiochemical processing (plant B) in the early 1950s. The procedure includes a reactor model and a cooling pool model. It is based on archive data on monthly uranium unloading and loading in the reactor and in the cooling pool of each reactor. The objects of reconstruction include: order of reloading of uranium versus its location radius in the reactor core; duration of irradiation and radionuclide composition of fission fragments for each radius; order of uranium removal from the cooling pool; effective time of uranium storage in the pool; radionuclide composition and activity of fission fragments in the irradiated uranium delivered to radiochemical reprocessing daily and on average for each month. The model is intended for use in reconstruction of parameters of radionuclide release source into the atmosphere and the source of liquid radioactive waste generation at the “Mayak” PA radiochemical plant.

  16. Methods For Reconstruction Of Radionuclide Composition And Activity Of Fission Products Accumulated In The Irradiated Uranium At The Moment Of Its Radiochemical Reprocessing At Plant 'B', 'Mayak' PA In The Early 1950s

    International Nuclear Information System (INIS)

    The article describes calculation procedure for reconstruction of radionuclide composition and activity of fission fragments accumulated in the irradiated uranium from 'Mayak' PA graphite-uranium reactors at the moment, when irradiation is completed, and at the moment, when the uranium is transferred to radiochemical processing (plant B) in the early 1950s. The procedure includes a reactor model and a cooling pool model. It is based on archive data on monthly uranium unloading and loading in the reactor and in the cooling pool of each reactor. The objects of reconstruction include: order of reloading of uranium versus its location radius in the reactor core; duration of irradiation and radionuclide composition of fission fragments for each radius; order of uranium removal from the cooling pool; effective time of uranium storage in the pool; radionuclide composition and activity of fission fragments in the irradiated uranium delivered to radiochemical reprocessing daily and on average for each month. The model is intended for use in reconstruction of parameters of radionuclide release source into the atmosphere and the source of liquid radioactive waste generation at the 'Mayak' PA radiochemical plant.

  17. Development of a neutron tomography system using a low flux reactor

    Science.gov (United States)

    Hungler, P. C.; Bennett, L. G. I.; Lewis, W. J.; Bevan, G. A.; Gabov, A.

    2011-09-01

    A neutron tomography instrument was designed and developed at the Royal Military College (RMC) of Canada with Queen's University to enhance these institutions' non-destructive evaluation capabilities. The neutron imaging system was built around a Safe Low-Power C(K)ritical Experiment (SLOWPOKE-2) nuclear research reactor. The low power and physical geometry of the reactor required that a novel design be developed to facilitate tomography. A unique rotisserie style rotary stage and clamping apparatus was developed. Furthermore, the low flux at the image plane (3×10 4 n cm -2 s -1), necessitated that the image acquisition and reconstruction processes be optimized. Tomographs of numerous samples were obtained using the new tomography instrument at RMC.

  18. Development of a neutron tomography system using a low flux reactor

    International Nuclear Information System (INIS)

    A neutron tomography instrument was designed and developed at the Royal Military College (RMC) of Canada with Queen's University to enhance these institutions' non-destructive evaluation capabilities. The neutron imaging system was built around a Safe Low-Power C(K)ritical Experiment (SLOWPOKE-2) nuclear research reactor. The low power and physical geometry of the reactor required that a novel design be developed to facilitate tomography. A unique rotisserie style rotary stage and clamping apparatus was developed. Furthermore, the low flux at the image plane (3x104 n cm-2 s-1), necessitated that the image acquisition and reconstruction processes be optimized. Tomographs of numerous samples were obtained using the new tomography instrument at RMC.

  19. Radiochemical analysis of waters and mud of Euganean spas (Padua

    Directory of Open Access Journals (Sweden)

    Cianchi A.

    2012-04-01

    Full Text Available The area around the Euganean Hills (North-East Italy is concerned with thermal phenomena known and used for therapeutic purposes since ancient times. The thermal waters collected in this area have taken up a natural radionuclides content due to the leaching of hot and permeable deep rocks, with which they come into contact, before their rising to the surface. During the "maturation" process of the mud used for treatment purposes, the thermal waters make happen a complex series of biochemical changes and release a series of chemical species to the mud, resulting, in particular, in an enrichment phenomenon for some radionuclides. In this work, the first radiochemical analysis extended to all the Euganean Thermal District is reported. In particular, chemical analyses of mud, as well as radiochemical analyses of both mud and waters were performed; the enrichment of the radioisotopes in mud used for treatments was also documented. The results show that the 226Ra content in mud, during the "maturation" process, presents an enrichment even of one order of magnitude with respect to the value found in the unprocessed mud. Furthermore, in the same thermal waters, high concentrations of "unsupported" 222Rn have been found, which have shown to be not completely negligible both for people under treatment and particularly for spa workers.

  20. Facility Effluent Monitoring Plan for the 325 Radiochemical Processing Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Shields, K.D.; Ballinger, M.Y.

    1999-04-02

    This Facility Effluent Monitoring Plan (FEMP) has been prepared for the 325 Building Radiochemical Processing Laboratory (RPL) at the Pacific Northwest National Laboratory (PNNL) to meet the requirements in DOE Order 5400.1, ''General Environmental Protection Programs.'' This FEMP has been prepared for the RPL primarily because it has a ''major'' (potential to emit >0.1 mrem/yr) emission point for radionuclide air emissions according to the annual National Emission Standards for Hazardous Air Pollutants (NESHAP) assessment performed. This section summarizes the airborne and liquid effluents and the inventory based NESHAP assessment for the facility. The complete monitoring plan includes characterization of effluent streams, monitoring/sampling design criteria, a description of the monitoring systems and sample analysis, and quality assurance requirements. The RPL at PNNL houses radiochemistry research, radioanalytical service, radiochemical process development, and hazardous and radioactive mixed waste treatment activities. The laboratories and specialized facilities enable work ranging from that with nonradioactive materials to work with picogram to kilogram quantities of fissionable materials and up to megacurie quantities of other radionuclides. The special facilities within the building include two shielded hot-cell areas that provide for process development or analytical chemistry work with highly radioactive materials and a waste treatment facility for processing hazardous, mixed radioactive, low-level radioactive, and transuranic wastes generated by PNNL activities.

  1. Facility Effluent Monitoring Plan for the 325 Radiochemical Processing Laboratory

    International Nuclear Information System (INIS)

    This Facility Effluent Monitoring Plan (FEMP) has been prepared for the 325 Building Radiochemical Processing Laboratory (RPL) at the Pacific Northwest National Laboratory (PNNL) to meet the requirements in DOE Order 5400.1, ''General Environmental Protection Programs.'' This FEMP has been prepared for the RPL primarily because it has a ''major'' (potential to emit >0.1 mrem/yr) emission point for radionuclide air emissions according to the annual National Emission Standards for Hazardous Air Pollutants (NESHAP) assessment performed. This section summarizes the airborne and liquid effluents and the inventory based NESHAP assessment for the facility. The complete monitoring plan includes characterization of effluent streams, monitoring/sampling design criteria, a description of the monitoring systems and sample analysis, and quality assurance requirements. The RPL at PNNL houses radiochemistry research, radioanalytical service, radiochemical process development, and hazardous and radioactive mixed waste treatment activities. The laboratories and specialized facilities enable work ranging from that with nonradioactive materials to work with picogram to kilogram quantities of fissionable materials and up to megacurie quantities of other radionuclides. The special facilities within the building include two shielded hot-cell areas that provide for process development or analytical chemistry work with highly radioactive materials and a waste treatment facility for processing hazardous, mixed radioactive, low-level radioactive, and transuranic wastes generated by PNNL activities

  2. The OMEGA Gas Sampling System and Radiochemical Diagnostic Development

    Science.gov (United States)

    Stoyer, Mark; Hudson, Bryant; Sangster, Craig; Freeman, Charlie; Schwartz, B.; Olsen, M.

    2001-10-01

    Radiochemical diagnostics for the National Ignition Facility (NIF) will address important issues such as shell rho-R, mix and charged particle production in ignition and near-ignition capsules. Many reaction products from charged particle reactions are noble gases. A gas sampling system for obtaining radiochemical samples following OMEGA shots has been assembled at LLNL and is being installed on the target chamber at OMEGA. Results of benchtop tests and possibly target chamber background collections with such a system will be discussed. A primary goal is to demonstrate reproducible collection efficiencies for this new technical capability of near 100include measuring collection efficiencies for certain reaction processes and to test the collection scheme for other low energy reaction products. Should high collection efficiencies be demonstrated, and the background be low and well-characterized, test reactions of 18O(alpha,n)21Ne or 80Kr(n,2n)79Kr and 38Ar(n,2n)37Ar will be investigated at OMEGA. In addition, other collection schemes are being considered for reactions that do not result in a noble gas isotope. Some simulations of expected activations from several capsule designs will be discussed. This work was performed under the auspices of the U.S. Department of Energy by the University of California, Lawrence Livermore National Laboratory under Contract No. W-7405-Eng-48.

  3. Volume reduction of radiochemical effluents by thermosyphon evaporators

    International Nuclear Information System (INIS)

    The concentration and containment of radiochemical effluents enhances safety during transportation, storage and disposal. The evaporation technique is very effective, in reducing the volume of waste, and in separating the low-activity distillate. Quantitative evaluation of technical, economic and safety aspects of various type of evaporators for volume reduction of radiochemical effluents, and existing experience and research in the field are reviewed to point out potential areas of development. The important conclusions emerging from the current research in many diverse areas; especially boiling accompanied with two-phase thermal hydraulics in thermosyphon evaporators, investigations in predicting accurate thermophysical properties, new equations for estimation of progressive increase in fouling resistance, recent improvements in process control and instrumentation, as well as design analysis and development are summarized. The experience of a wide range of operation and design parameters are described to evolve more reliable design methods and rational operation procedures. The new recommendations are useful to review, examine and optimize the process and design parameters of existing and forthcoming liquid waste concentration systems

  4. A radiochemical analyses of metastudtite and leachates from spent fuel

    International Nuclear Information System (INIS)

    Immersion of commercial spent nuclear fuel (CSNF) in deionized water produced two novel corrosion products after a two-year contact period. Another unexpected result was that suspensions of aggregates were observed to form at the air-water interface for each of five samples. These solids were characterized, by SEM and XRD to be nearly pure metastudtite (UO4-2H2O); while the corrosion present on the surface of the fuel itself was determined to be studtite (UO4-2H2O). The occurrence of the floating phase prompted a radiochemical analysis of these solids. This chemical analysis was a unique opportunity to study the relatively pure corrosion phase for incorporation of radionuclides. The analysis indicated that high concentration of 90Sr, 137Cs, 99Tc, and that lower concentrations 237Np, 238, 239Pu and 243, 244Cm had partitioned with the air-water interface aggregates. The concentrations of 241Am were two orders of magnitude lower than the expected inventory in the suspended solids. The radiochemical analyses of the several leachate samples provide preliminary solubility data for the hydrogen peroxide leaching of CSNF and these data are compared to leaching of the same fuel in J-13 and deionized waters. The extent of fuel dissolution in these media are discussed

  5. Radiochemical studies on thyroid function in Sudanese newborn

    International Nuclear Information System (INIS)

    In this study two thyroid related hormones of 200 neonates were investigated, in order to study the prevalence of congenital hypothyroidism in Khartoum state. Radiochemical iodinated anti-thyroid stimulating hormone (TSH) monoclonal antibody and and thyroxine labeled antigen were used as a radiotracer. Radiochemical chlora min T method of radioiodination, which based on chlomin T as strong oxidizing agent and sodium metabisulphite as strong reducing agent was used to prepare the iodine 125 labeled radio tracer of TSH and T4 hormones. There for two radiotracer were used in sensitive and specific radioimmunoassay (RIA) for T4 and TSH. Cord blood was used at the moment of delivery to obtain serum samples for the investigations. The mean of most neonates was normal in T4 (117 nmol/1) and the TSH (1.8 mu/1), while the normal levels ranging from (50-150 nmol/ 1) and (0.4-4 mu/1) respectively. There was strong correlation between T4 and TSH, p>0.01. Two neonates showed high TSH level and low T4 level. In this study the prevalence of congenital hypothyroidism was found to be 1% which correlates with the international incidence. (Author)

  6. Analytical performance of radiochemical method for americium determination in urine

    International Nuclear Information System (INIS)

    This paper presents an analytical method developed and adapted for separation and analysis of Plutonium (Pu) isotopes and Americium (Am) in urine samples. The proposed method will attend the demand of internal exposure monitoring program for workers involved mainly with dismantling rods and radioactive smoke detectors. In this experimental procedure four steps are involved as preparation of samples, sequential radiochemical separation, preparation of the source for electroplating and quantification by alpha spectrometry. In the first stage of radiochemical separation, plutonium is conventionally isolated employing the anion exchange technique. Americium isolation is achieved sequentially by chromatographic extraction (Tru.spec column) from the load and rinse solutions coming from the anion exchange column. The 243Am tracer is added into the sample as chemical yield monitors and to correct the results improving the precision and accuracy. The mean recovery obtained is 60%, and the detection limit for 24h urine sample is 1.0 mBq L-1 in accordance with the literature. Based in the preliminary results, the method is appropriate to be used in monitoring programme of workers with a potential risk of internal contamination. (author)

  7. Design features of the laboratory-scale radiochemical immobilization system

    International Nuclear Information System (INIS)

    Under the High-Level Waste Immobilization Program, the Pacific Northwest Laboratory (PNL) is studying various ways to solidify high-level nuclear wastes. A variety of waste forms and processes are being investigated, with the most highly developed process being spray calcination coupled with in-can melting. This report describes a remote laboratory-scale system that was designed for the purpose of investigating the effects of different operating conditions and waste compositions on the product and on the effluents generated. It is termed laboratory-scale because of its nominal 1 L/h feed rate as compared to well over 300 L/h for full-scale equipment at PNL. The equipment currently consists of a feed system, a spray calciner, an in-can melter, and an effluent control system. It is operated in a shielded radiochemical hot cell using radioactive high-level liquid waste (HLLW) to answer questions on the deposition of radiochemicals during actual waste processing. The effluent control system can be modified in order to test different effluent systems, one of which has been proposed by the Savannah River Laboratories (SRL) for use in the Savannah River Plant vitrification system. The laboratory-scale system can also be used to test alternative immobilization processes, since spray calcination is a common processing step in many alternative waste form flowsheets. Thus, only the addition of a specific forming step such as pelletizing or sintering is necessary

  8. Radiochemical changes in polymers used for packaging drugs

    International Nuclear Information System (INIS)

    Polymers are now widely used for packaging drugs sterilized with ionizing radiation. Sterilizing doses of radiation can cause changes in the structure and characteristics of polymers. Gas chromatography methods were used to study the composition of gaseous products formed during the radiation sterilization of low- and high-pressure polythene, polymethyl methacrylate (PMMA), PVC, and divinylstyrene and isoprenestyrene rubbers (DST-30 and IST-30), and it is shown that radiation sterilization causes complex radiochemical changes. Infra-red spectroscopy was used to investigate the effect of radiation sterilization on the structure of poly-4-methylpentene-1, low- and high-pressure polythene, PMMA, DST-30, IST-30 and PVC, and it is shown that gamma radiation causes radiochemical changes in polymers leading to the formation of ketones, aldehydes, acids, spirits, esters, unsaturated compounds, etc. Effective constants for the rate of formation of ketones and acids are determined and it is found that they are practically identical and independent of the thickness of the specimen within the range 30 to 150 μm. It is shown that radiation sterilization causes insignificant changes to the ageing mechanism of DST-30 in air and water. The investigation showed that the greatest resistance to the effects of radiation sterilization and aqueous solutions of insulin is offered by dimthyl siloxane rubber, amongst the rubbers, and polythene amongst the polyolefines. (author)

  9. The influence of non-aqueous radiochemical processes on radiation parameters of spent fuel and radioactive wastes

    International Nuclear Information System (INIS)

    The influence of the technology applied for separation of radioactive elements on radiation parameters of fuel and wastes when using non-aqueous radiochemical processing of spent fuels are studied. The results of calculational modelling the fuel recycle in the BREST-1200 reactor closed fuel cycle are considered. The data characterizing contribution of separate elements in potential biological danger (dose) and the dependence of the potential biological danger of the wastes on regenerated fuel cooling time are discussed. It is shown that plutonium and americium give the main contributions into the fuel potential biological danger in time period of 40-1000 years. For monitored cooling of 120-150 years the balance between natural uranium potential biological danger and that of wastes at different waste compositions is achievable. The fission product contributions into potential biological danger differ slightly for different variants of the processing technology. The 99Tc contribution is noticeable only in the case of metallurgical processing. The conclusion is made that differences in radiochemical technologies applied for waste fracturing and fuel purification degree do not influence in principle on capabilities for radiation balance achieving. For a long-time perspective the radiation balance is determined by plutonium, americium and their decay products. The technology peculiarities may change radiation characteristics of wastes only at separate stages of cooling and do not affect greatly the radiation balance as a whole

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

    International Nuclear Information System (INIS)

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

  11. Nuclear research centres in the 21st century: An AECL perspective

    International Nuclear Information System (INIS)

    The nuclear energy programme of Canada started at Chalk River Laboratories with the setting up of Zero Energy Experimental Site in 1945. One of the early research reactors of Canada, the National Research Universal (NRU) continues to provide 70% of the world requirement of isotopes for medical and industrial applications. A CANDU prototype (208 MW(e)) came on line in 1967 and based on this concept, Canada has a large nuclear power programme. The role of nuclear research centres has evolved with time starting with strategic research in the initial phases through to implementation of technology, building and supporting industry, and carrying out advanced technology development. Most of these centres have important assets in terms of licensed sites, trained personnel, research reactors, shielded facilities and expertise for handling large quantities of radioactivity and high tech laboratories for advanced R and D. These centres would, therefore, continue to play an important role in emission free and economic energy generation, nuclear medicine, food irradiation and industrial applications. Nuclear research centres in different countries are at various stages of development and have many unique features. However, there are generic issues and much will be gained by developing a shared vision for the future and implementing programmes in a collaborative manner. (author)

  12. Rapid radiochemical methods for the determination of alpha and beta radionuclides (a review)

    International Nuclear Information System (INIS)

    Radiochemical laboratories are important actors in emergency preparedness. The review summarizes the development of rapid radioanalytical methods with respect to detection and separation techniques. The review is intended as a guide for laboratories wishing to select suitable rapid radiochemical methods, viable in emergency situations. (orig.)

  13. PLM and the single reactor utility - or how a single reactor utility can face the PLM issues

    International Nuclear Information System (INIS)

    Although Gentilly-2 reactor was planned to last for 30 years, its life could be significantly shorter if nothing were done, whereas retubing and refurbishment after, say, 25 years should result in an extension of service life to 45-50 years. In the long run, dimensional changes rather than hydriding may prove to be the pressure tubes' life limiting factor. Hydro Quebec, New Brunswick Power and AECL have an agreement to cooperate in developing a life management program for CANDU-6 reactors. The author expresses the opinion that cost-benefit criteria should be introduced in regulatory decision making. 6 refs., 9 figs

  14. Mobile air filtration units for NPPs and radiochemical facilities

    International Nuclear Information System (INIS)

    Alongside the decommissioning of a range of nuclear installations and ongoing construction of new NPPs, it has become apparent that mobile filtration and ventilation devices (MFVD) are now in demand. The units are to carry out local purification of air, remove various aerosols from process operations (welding, cutting, metal works, etc.), as well as remove toxic gases that accompany them in the production rooms of nuclear stations and radiochemical facilities. The MFVD includes filtration modules, which are placed within a single device mounted on a trolley and consists of filters of three stages of purification and a body of the filters; as well as a radial fan with electrical motor, a frequency converter and a motor starter, exhaust devices or flexible hoses. The technical characteristics of the MFVD combined with acceptable overall dimensions make the unit a very good solution capable of providing the required air purification that significantly improves the working conditions, personnel and environmental safety

  15. 15th radiochemical conference: Booklet of abstracts and conference programme

    International Nuclear Information System (INIS)

    The conference was structured as follows: Opening plenary lectures (6 lectures); Topic 1 - radionuclides in the environment, radioecology (22 verbal presentations (VPs), 23 poster presentations (PPs)); Topic 2 - nuclear analytical methods (22 VPs, 32 PPs); Topic 3 - chemistry of actinide and transactinide elements (8 VPs, 10 PPs); Topic 4 - radiation chemistry (9 VPs, 5 PPs); Topic 5 - production and application of radionuclides (17 VPs, 6 PPs); Topic 6 - separation methods, speciation (21 VPs, 23 PPs); Topic 7 - chemistry of nuclear fuel cycle, radiochemical problems in nuclear waste management (20 VPs, 16 PPs); Topic 8 - nuclear methods in medicine, radiopharmaceuticals, and radiodiagnostics, labelled compounds (8 VPs, 7 PPs); and Panels (2 introductions). (P.A.)

  16. Handling of Ammonium Nitrate Mother-Liquid Radiochemical Production - 13089

    International Nuclear Information System (INIS)

    The aim of the work is to develop a basic technology of decomposition of ammonium nitrate stock solutions produced in radiochemical enterprises engaged in the reprocessing of irradiated nuclear fuel and fabrication of fresh fuel. It was necessary to work out how to conduct a one-step thermal decomposition of ammonium nitrate, select and test the catalysts for this process and to prepare proposals for recycling condensation. Necessary accessories were added to a laboratory equipment installation decomposition of ammonium nitrate. It is tested several types of reducing agents and two types of catalyst to neutralize the nitrogen oxides. It is conducted testing of modes of the process to produce condensation, suitable for use in the conversion of a new technological scheme of production. It is studied the structure of the catalysts before and after their use in a laboratory setting. It is tested the selected catalyst in the optimal range for 48 hours of continuous operation. (authors)

  17. Improvement in the degradation resistance of LDPE for radiochemical processing

    Science.gov (United States)

    Zaharescu, Traian; Pleşa, Ilona; Jipa, Silviu

    2014-01-01

    The effect of rosemary extract on radiochemical stability of low density polyethylene was studied by chemiluminescence, FT-IR spectroscopy and differential scanning calorimetry after γ(137Cs)-irradiation at processing low doses (10 and 20 kGy) in respect of pristine material. The additive concentrations (1, 2 and 5 wt%) induced a significant improvement in radiation stability, especially at high temperatures, for example 200 °C, which is proved chiefly by lower values of chemiluminescence intensities. The comparison of neat and rosemary-modified LDPE samples has revealed the protection action of this natural extract, which delays efficiently the propagation of oxidative degradation in γ-exposed polyethylene. The most evident proof for antioxidative protection efficiency promoted by rosemary is the smooth changes in hydroxyl and carbonyl indexes calculated on LDPE/5 wt% rosemary samples at all exposure doses.

  18. Paleomagnetism and radiochemical age estimates for Late Brunhes polarity episodes

    International Nuclear Information System (INIS)

    Several reversed polarity magnetozones occur within deep-sea sediment core CH57-8 from the Greater Antilles Outer Ridge, within sediment of latest Pleistocene/Late Brunhes age. The uppermost reversed interval spanning 31 data points coincides with the X faunal zone of the Last Interglacial Period. Radiochemical dating of cores CH57-8 and KN25-4 has shown that all the reversed polarity magnetozones are significantly younger than the Brunhes/Matuyama boundary at 0.7 m.y.B.P. A variation of the excess 230Th method was used, in which 210Po and 238U were the actual radionuclides measured. In a third core from the Mid-Atlantic Ridge, the 210Po results were similar to those which others obtained earlier by direct 230Th measurements. (Auth.)

  19. Fast analysis procedure of radiochemical coordinat uptake for methotrexate

    International Nuclear Information System (INIS)

    Under this invention, a radio-chemical analysis is submitted to determine the concentration of methotrexate or its equivalents in analysis in a biological medium. The amounts taken up of the labelled compound and the known concentrations of the unlabelled compound to be determined are radio-isotopically related to a first system containing a pre-determined amount of the labelled compound and a pre-determined amount of the unlabelled compound. In a second system, identical to the first, save that the sample of the biological medium to be analyzed takes the place of the unlabelled compound, the amount of labelled compound taken up is determined radio-isotopically. The concentration of the compound in the sample is then determined by correlation of the labelled compound uptake determined in the second system with the relation determined in the first system. The radio-isotopic relations and determinations may be made by direct and sequential analytical techniques

  20. Development of robotic plasma radiochemical assays for positron emission tomography

    Energy Technology Data Exchange (ETDEWEB)

    Alexoff, D.L.; Shea, C.; Fowler, J.S.; Gatley, S.J.; Schlyer, D.J. [Brookhaven National Lab., Upton, NY (United States). Dept. of Chemistry

    1995-12-01

    A commercial laboratory robot system (Zymate PyTechnology II Laboratory Automation System; Zymark Corporation, Hopkinton, MA) was interfaced to standard and custom laboratory equipment and programmed to perform rapid radiochemical analyses for quantitative PET studies. A Zymark XP robot arm was used to carry out the determination of unchanged (parent) radiotracer in plasma using only solid phase extraction methods. Robotic throughput for the assay of parent radiotracer in plasma is 4--6 samples/hour depending on the radiotracer. Robotic assays of parent compound in plasma were validated for the radiotracers [{sup 11}C]Benztropine, [{sup 11}C]cocaine, [{sup 11}C]clorgyline, [{sup 11}C]deprenyl, [{sup 11}C]methadone, [{sup 11}C]methylphenidate, [{sup 11}C]raclorpride, and [{sup 11}C]SR46349B. A simple robot-assisted methods development strategy has been implemented to facilitate the automation of plasma assays of new radiotracers.

  1. Development of robotic plasma radiochemical assays for positron emission tomography

    International Nuclear Information System (INIS)

    A commercial laboratory robot system (Zymate PyTechnology II Laboratory Automation System; Zymark Corporation, Hopkinton, MA) was interfaced to standard and custom laboratory equipment and programmed to perform rapid radiochemical analyses for quantitative PET studies. A Zymark XP robot arm was used to carry out the determination of unchanged (parent) radiotracer in plasma using only solid phase extraction methods. Robotic throughput for the assay of parent radiotracer in plasma is 4--6 samples/hour depending on the radiotracer. Robotic assays of parent compound in plasma were validated for the radiotracers [11C]Benztropine, [11C]cocaine, [11C]clorgyline, [11C]deprenyl, [11C]methadone, [11C]methylphenidate, [11C]raclorpride, and [11C]SR46349B. A simple robot-assisted methods development strategy has been implemented to facilitate the automation of plasma assays of new radiotracers

  2. Personal monitoring in the radiochemical production of 125I

    International Nuclear Information System (INIS)

    To estimate occupational exposure during the radiochemical production of 125I, measurements, of 125I deposited in the thyroids of personnel were carried out, beginning in 1986. Each measurement was designed to control later values, which were based on: the thyroid dose accumulated during the year prior to the measurement and during the monitoring period, the committed annual thyroid dose assessed during 240 days before the measurement, and the dose that would be received after complete decay of the measured activity of 125I in the thyroid. Due to the improvement in safety in the production process, quarterly thyroid doses to personnel decreased approximately three-fold during the first year of monitoring. Data on individual quarterly 125I thyroid content and doses over 1.5 years of control for 20 workers are presented. Average annual thyroid doses of personnel varied from 15 to 30 mSv.y-1 in 1987-1992. (Author)

  3. Radiochemical methods for studying lipase-catalyzed interesterification of lipids

    International Nuclear Information System (INIS)

    Reactions involving lipase-catalyzed interesterification of lipids, which are of commendable interest in biotechnology, have been monitored and assayed by radiochemical methods using 14C-labeled substrates. Medium chain (C12 plus C14) triacylglycerols were reacted in the presence of an immobilized lipase from Mucor miehei and hexane at 450C with methyl [1-14C]oleate, [1-14C]oleic acid, [carboxyl-14C]trioleoylglycerol, [1-14C]octadecenyl alcohol, and [U-14C]glycerol, each of known specific activity. The reactions were monitored and the rate of interesterification determined by radio thin layer chromatography from the incorporation of radioactivity into acyl moieties of triacylglycerols (from methyl oleate, oleic acid, and trioleoylglycerol), alkyl moieties of wax esters (from octadecenyl alcohol), and into glycerol backbone of monoacylglycerols and diacylglycerols (from glycerol). (orig.)

  4. Polytrimethylsylylpropyne gas separation membranes modified by radiochemical grafting of divinylbenzene

    International Nuclear Information System (INIS)

    A radiochemical method was employed to obtain poly(1-trimethylsilyl-1-propyne)(PTMSP)-divinylbenzene (DVB) grafted films. DVB monomer vapors were absorbed by the PTMSP, and the grafting reaction was thereafter accomplished by 60Co γ-irradiation in a nitrogen atmosphere. The films so obtained were tested for nitrogen-oxygen separation. The performances of the membranes were studied as functions of time and percent of grafting. The DVB-grafted membranes show an increased selectivity factor and stability with time. The experimental data and some SEM observations confirm the presence of large voids in the PTMSP matrix. These voids are responsible for permeability changes during operation and disappear after the grafting procedure. 8 refs., 5 figs

  5. The OMEGA Gas Sampling System and Radiochemical Diagnostics for NIF

    Science.gov (United States)

    Stoyer, Mark; Sangster, Craig; Hudson, Bryant; Lougheed, Ron; Freeman, Charlie; Schwartz, Brook-Eden; Olsen, Michele

    2000-10-01

    Radiochemical diagnostics for the National Ignition Facility (NIF) will address important issues such as shell rho-R, mix and charged particle production in ignition and near-ignition capsules. Development of key tools for these diagnostics has been progressing on NOVA and OMEGA laser systems. A limitation of the sample collection techniques currently being used is the solid angle for collection of post-shot debris, which is maximally about 1collections at NOVA. Because of the large standoff for NIF (5 m), the solid angle subtended would be expected to be much less without development of expandable foil collection schemes. Many reaction products from charged particle reactions are noble gases. A gas sampling system for obtaining radiochemical samples following OMEGA shots is currently being assembled at LLNL. Results of benchtop tests with such a system will be discussed. A primary goal is to demonstrate reproducible collection efficiencies for this new technical capability of near 100Secondary goals include measuring collection efficiencies for certain reaction processes and to test the collection scheme for other low energy reaction products. Should high collection efficiencies be demonstrated, test reactions of 18O(alpha,n)21Ne and 79Br(p,n)79Kr will be investigated at OMEGA as mix-diagnostics for NIF. Note that it may be possible to use the 18O or Br already in most capsules, circumventing some target development issues. The gas sampling system will be designed in such a way as to not preclude the addition of carrier gas to the target chamber following an experiment for "flushing" of other, perhaps non-gaseous, reaction products.

  6. 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)

  7. Radiochemical neutron activation analysis of high pure palladium and platinum by ion exchange chromatography

    International Nuclear Information System (INIS)

    Full text: The palladium and platinum are widely used for jewel manufacture because of their beautiful white color. However the most part of these metals are widely adopted in the world as catalysts. Many works on analytical chemistry of platinum group elements published during last years are devoted to determination of platinum and palladium in other materials. There are no articles on analysis technique of the palladium and platinum purity published during last 20 years. Available publications are very old and are published till 70th of the last century, and implement chemical and spectral methods. At the same time, the palladium and platinum are very suitable for NAA. Therefore the purpose of our research was development of high-sensitivity and multielement techniques of radiochemical neutron activation analysis of a high pure palladium and platinum. Research of nuclear characteristics of palladium and platinum has shown that radioactive nuclides with different yields are formed under the reactor neutrons. 109, 111, 111mPd, 109m, 111Ag, 191,197, 199Pt, 199Au are the most important among them. 109Pd separation factor is equal to 1*105 at palladium analysis, whereas 197Pt and 199Au separation factor is equal to 1*104 at the platinum analysis every other day after irradiation. Palladium and platinum can be separated by precipitation, extraction and ion exchange methods. For separation of radioactive nuclide of the matrix elements from the impurity elements we used ion exchange chromatography system Dowex-1x8 - 1 M HNO3 for palladium and Dowex-1x8 - 0.1 M HNO3 for platinum. At the HNO3 acid concentrations variation from 0,1 M to 1 M more then 25 elements have distribution factors less than 1 and 10 elements have distribution factors 5 while matrix elements have distribution factors higher than 100. It allows an effective separation of these elements from palladium and platinum. Optimum sizes of the chromatographic column and the column effluent volume was obtained

  8. Radiochemical assessment of core water as an indicator of fuel reliability behaviour: A case study

    International Nuclear Information System (INIS)

    Evaluation of coolant activity during nuclear reactor operation is the first step in assessing fuel reliability. Monitoring the activity of selected fission product isotopes in the off-gas system or in the primary coolant provides useful information on fuel status. Non-destructive and destructive methods were used to determine fission products as well as activation products in liquid samples taken from the ET-RR-1 research reactor core water, and examined for indications of fuel leakage during operation.The ET-RR1 is a 2 MW light water reactor at Inshass, Egypt, operated with aluminium cladded 10% enriched uranium oxide. Gamma spectrometry, based on a hyperpure germanium detector, was used to evaluate specific activities of present radionuclides. The release rate of 137Cs was used to characterize the fuel assemblies while investigating the presence of corrosion pits along the external fuel clad. Radiochemical analysis of core water before and 1 h after full power operation indicated high activity of 137Cs.About 70% of the 137Cs was already present in the core water before operation. Caesium-137 to 60Co ratios, before and after operation, were similar at 274 and 210, respectively.As 239Np was detected in core water after operation (58 ± 1.4 kBq·L-1), the total uranium content in the water before operation was determined using laser fluorimetry to evaluate U traceability, if any. The total uranium content of the core water before and after operation was 0.9 and 1.6 ppm, while water samples from the new and old spent fuel pools showed 1.2 and 2.1 ppm, respectively. Neptunium-239 can be effectively employed as an indicator of degraded fuel conditions and of significant fuel loss.Among others, 131I, 133I, 135Xe, 134Cs, 137Cs, 91Sr, 90Sr, 140La and 140Ba were determined. The specific activities of 137Cs and 90Sr measured in the reactor core water before and after operation were about 18 and 12.7 kBq·L-1, and 26 and 15.3 kBq·L-1, respectively. The specific activities

  9. US DOE-AECL cooperative program for development of high-level radioactive waste container fabrication, closure, and inspection techniques

    International Nuclear Information System (INIS)

    The US Department of Energy (DOE) and Atomic Energy of Canada Limited (AECL) plan to initiate a cooperative research program on development of manufacturing processes for high-level radioactive waste containers. This joint program will benefit both countries in the development of processes for the fabrication, final closure in a hot-cell, and certification of the containers. Program activity objectives can be summarized as follows: to support the selection of suitable container fabrication, final closure, and inspection techniques for the candidate materials and container designs that are under development or are being considered in the US and Canadian repository programs; and to investigate these techniques for alternate materials and/or container designs, to be determined in future optimization studies relating to long-term performance of the waste packages. The program participants will carry out this work in a conditional phased approach, and the scope of work for subsequent years will evolve subject to developments in earlier years. The overall term of this cooperative program is planned to run roughly three years. 5 refs., 2 tabs

  10. AECL international standard problem ISP-41 FU/1 follow-up exercise (Phase 1): Containment Iodine Computer Code Exercise: Parametric Studies

    International Nuclear Information System (INIS)

    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- 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 from 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. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Mistry, Sunjay [AECL, Ontario (Canada)

    2013-07-01

    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

  12. 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

  13. Advanced CANDU reactor, evolution and innovation

    International Nuclear Information System (INIS)

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

  14. Validation of computer codes used in the safety analysis of Canadian research reactors

    International Nuclear Information System (INIS)

    AECL has embarked on a validation program for the suite of computer codes that it uses in performing the safety analyses for its research reactors. Current focus is on codes used for the analysis of the two MAPLE reactors under construction at Chalk River but the program will be extended to include additional codes that will be used for the Irradiation Research Facility. The program structure is similar to that used for the validation of codes used in the safety analyses for CANDU power reactors. (author)

  15. Canada's reactor exports

    International Nuclear Information System (INIS)

    A brief sketch of the development of Canada's nuclear exports is presented and some of the factors which influence the ability to export reactors have been identified. The potential market for CANDUs is small and will develop slowly. The competition will be tough. There are few good prospects for immediate export orders in the next two or three years. Nonetheless there are reasonable opportunities for CANDU exports, especially in the mid-to-late 1980s. Such sales could be of great benefit to Canada and could do much to sustain the domestic nuclear industry. Apart from its excellent economic and technical performance, the main attraction of the CANDU seems to be the autonomy it confers on purchasing countries, the effectiveness with which the associated technology can be transferred, and the diversification it offers to countries which wish to reduce their dependence on the major industrial suppliers. Each sales opportunity is unique, and marketing strategy will have to be tailored to the customer's needs. Over the next decade, the factors susceptible to Canadian government action which are most likely to influence CANDU exports will be the political commitment of the government to those reactor exports, the performance established by the four 600 MWe CANDUs now nearing completion, the continuing successful operation of the nuclear program in Ontario, and the co-ordination of the different components of Canada's nuclear program (AECL, nuclear industry, utilities, and government) in putting forth a coherent marketing effort and following through with effective project management

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

    International Nuclear Information System (INIS)

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

  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. Livermore pool-type reactor

    International Nuclear Information System (INIS)

    The Livermore Pool-Type Reactor (LPTR) has served a dual purpose since 1958--as an instrument for fundamental research and as a tool for measurement and calibration. Our early efforts centered on neutron-diffraction, fission, and capture gamma-ray studies. During the 1960's it was used for extensive calibration work associated with radiochemical and physical measurements on nuclear-explosive tests. Since 1970 the principal applications have been for trace-element measurements and radiation-damage studies. Today's research program is dominated by radiochemical studies of the shorter-lived fission products and by research on the mechanisms of radiation damage. Trace-element measurement for the National Uranium Resource Evaluation (NURE) program is the major measurement application today

  19. Safety upgrades to the NRU research reactor

    International Nuclear Information System (INIS)

    The NRU (National Research Universal) Reactor is a 135 MW thermal research facility located at Chalk River Laboratories. AECL owns and operates the multi-purpose research reactor that serves as the primary R and D facility for supporting the CANDU business. The reactor is also a major producer of the world's medical radioisotopes. Since NRU was started up in 1957, it has operated in a consistent and safe manner with an overall annual capacity factor of approximately 80 %. The demands on the operation to perform experiments and produce radioisotopes were increased significantly when the NRX (National Research Experimental) shut down in 1992. Radioisotope customers demand an uninterrupted supply of short-lived radioisotopes e g Molybdenum-99, while experimental researchers require frequent shutdowns to accommodate fuel and materials programs. A two year systematic review and assessment of NRU to determine the condition and state of the facility was completed in 1991. This engineering assessment was complemented by safety analyses which focused on systems and components critical to safety. Reactor aging, obsolescence, current codes, and hazards vulnerability (especially, seismic) were emphasized during the analyses. This initial assessment concluded that the overall condition of NRU was good and there was no undue risk to the public or environment with the present operation. In addition, seven major upgrades were identified to enhance reactor safety to satisfy modern standards. In 1992, the AECL executive approved the Upgrades Project. Implementation of the seven upgrades were then included in the Facility Authorization document that defines the limiting conditions for safe operation with the Chalk River site license. The Atomic Energy Control Board would approve and license the upgrades under the change control provisions of the FA. Each upgrade and/or assessment recommendation (minor modification) had to be implemented without adversely affecting the current

  20. Shielding design for research and education reactor

    International Nuclear Information System (INIS)

    For the purpose of education and research at the University, 20-KW powered SLOWPOKE-2 research reactor has been chosen as a prototype reactor. In order to study the safety characteristics of the reactor, exposure rate has been estimated at the pool boundary. Reactor core as a radiation source is assumed to be cylindrical volume source. Thus point kernel integration method can be applied to determine the exposure rate. For the sake of simplicity, calculation was done only for the prompt fission gamma rays and fission product gamma rays. As a result, the maximum exposure rate at the pool boundary was estimated to be 18R/min at the same height of the center of the core. In order to examine the accuracy for the point kernel integration method, two shielding experiments were carried out: one for the water tank only and the other for with concrete blocks outside the water tank. Water tank was made of wood pieces which is 13.4cm wide, 1.5cm thick and 2.15m long. Thus the water tank has the total dimension of 1 m radius and 2.1 m height. The experiment was carried out for the radiation source of 0.968 mCi Co-60 at the center of the water tank and the penetrated gamma rays were measured at 5 different detector positions. For the measurement and analysis of the responses, NaI(T1) 3''x3'' detector and 256 channel multichannel analyzer was utilized. To convert pulse height distribution to the exposure rate, Moriuchi conversion factor was adopted. Data from the calculations by point kernel method were well agreed within 10% band with the data from the the experiments. (Author)

  1. Waste treatment at the Radiochemical Engineering Development Center

    International Nuclear Information System (INIS)

    At the Radiochemical Engineering Development Center (REDC) irradiated targets are processed for the recovery of valuable radioisotopes, principally transuranium nuclides. A system was recently installed for treating the various liquid alkaline waste streams for removal of excess radioactive contaminants at the REDC. Radionuclides that are removed will be stored as solids and thus the future discharge of radionuclides to liquid low level waste tank storage will be greatly reduced. The treatment system is of modular design and is installed in a hot cell (Cubicle 7) in Building 7920 at the REDC where preliminary testing is in progress. The module incorporates the following: (1) a resorcinol-formaldehyde resin column for Cs removal, (2) a cross flow filtration unit for removal of rare earths and actinides as hydroxide, and (3) a waste solidification unit. Process flowsheets for operation of the module, key features of the module design, and its computer-assisted control system are presented. Good operability of the cross flow filter system is mandatory to the successful treatment of REDC wastes. Results of tests to date on the operation of the filter in its slurry collection mode and its slurry washing mode are presented. These tests include the effects of entrained organic solvent in the waste stream feed to the filter

  2. Radiochemical evaluation of a new brain receptor imaging agent

    International Nuclear Information System (INIS)

    We report about the radiochemical evaluation of a new serotonin-1A (5-HT1A) receptor imaging agent. The new derivative of WAY 100635, viz. C1-(2 methoxyphenyl)-(4- mercaptoethyl)-piperazine, was labelled with technetium-99m using thiocresol through 99mTc(V)-glucoheptonate precursor. The labelling was carried out at room temperature within 10 minutes using 370-740 MBq of 99mTc-pertechnetate. The specific activity of the '2+1+1' mixed ligand complex was about 40 GBq/ml. The labelling efficiency and the stability of the labelled compound were monitored by ITLC-SG, solvent extraction and reverse-phase HPLC. The labelling efficiency exceeded 95% and remained high about 4 hours if stored at room temperature or in a refrigerator at 4 deg C. The results give evidence of a high labelling efficiency and stability of the ligand used. The labelled ligand seems to hold promise within the family of existing radiopharmaceuticals

  3. Human erythrocyte thiol methyltransferase: radiochemical microassay and biochemical properties

    International Nuclear Information System (INIS)

    A radiochemical microassay for the measurement of thiol methyltransferase (TMT) activity in human red blood cell (RBC) membranes has been developed. Both 2-mercaptoethanol and dithiothreitol were used as substrates for the enzyme. The pH optimum of the reaction was approximately 9.0 when glycine-NaOH was used as a buffer. The apparent Michaelis-Menten (Ksub(M)) value for the methyl donor for the reaction, S-adenosyl-L-methionine, was 43 μmol/l. Human RBC TMT activity was neither activated nor inhibited by Ca2+, Mg2+, or tropolone, but the enzyme was inhibited by SKF 525A and by reagents that react with sulfhydryl grcups. The mean TMT activity in blood from 289 randomly selected adult white subjects was 10.93 +- 3.22 units per mg protein (mean +- S.D.). The activity was the same in samples from men and women. The results of experiments in which TMT activity was measured in mixtures of RBC membranes with relatively ''low'' and relatively ''high'' activities provided no evidence that individual variations in the enzyme activity were due to variations in endogenous TMT activators or inhibitors. (Auth.)

  4. Development of a radiochemical sensor. Part I: Feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Tarancon, A. [Departament de Quimica Analitica, Facultat de Quimica, Universitat de Barcelona, C/ Marti Franques 1, 08028 Barcelona (Spain); Garcia, J.F. [Departament de Pintura, Facultat de Belles Arts, Universitat de Barcelona, C/ Pau Gargallo 4, 08028 Barcelona (Spain)]. E-mail: jfgarcia@apolo.qui.ub.es; Rauret, G. [Departament de Quimica Analitica, Facultat de Quimica, Universitat de Barcelona, C/ Marti Franques 1, 08028 Barcelona (Spain)

    2005-05-04

    The evolution of nuclear activities and criteria for radiation protection have led to a continuous increase in measures to monitor and control the environment and therefore in the number of determinations required for such purposes. Classical analytical procedures are time-consuming, labor-intense and generate a large amount of waste. The alternative use of sensors for such determinations has seen very limited development. The present study focuses on the evaluation of the behavior of a prototype radiochemical sensor for liquid effluents. The sensor is based on a receptor made of a plastic scintillator and is capable of continuous, on-time and accurate remote quantification of the activity of alpha, beta and beta-gamma emitters. Low-level active solutions of {sup 90}Sr/{sup 90}Y, {sup 238}Pu, {sup 134}Cs and {sup 60}Co in matrices of groundwater, seawater and drinking water were quantified with prediction errors lower than 10% in most cases. The study also yields information about light generation and transmission and transductor configuration that will be useful in the design of future versions of this sensor.

  5. Development of a radiochemical sensor. Part I: Feasibility study

    International Nuclear Information System (INIS)

    The evolution of nuclear activities and criteria for radiation protection have led to a continuous increase in measures to monitor and control the environment and therefore in the number of determinations required for such purposes. Classical analytical procedures are time-consuming, labor-intense and generate a large amount of waste. The alternative use of sensors for such determinations has seen very limited development. The present study focuses on the evaluation of the behavior of a prototype radiochemical sensor for liquid effluents. The sensor is based on a receptor made of a plastic scintillator and is capable of continuous, on-time and accurate remote quantification of the activity of alpha, beta and beta-gamma emitters. Low-level active solutions of 90Sr/90Y, 238Pu, 134Cs and 60Co in matrices of groundwater, seawater and drinking water were quantified with prediction errors lower than 10% in most cases. The study also yields information about light generation and transmission and transductor configuration that will be useful in the design of future versions of this sensor

  6. Kinetic study of the radiochemical ageing of polyethylene

    International Nuclear Information System (INIS)

    Various bulk or multilayer low density polyethylene samples were irradiated by cobalt gamma rays (60Co) in air and at ambient temperature. The thickness distribution of carbonyl groups concentration displays a sharp inflexion zone at an invariant depth of about 180 μm for the dose rate under investigation. The oxidation rate is practically zero in the core zone of thicker samples (≥500 μm). An investigation by DSC showed that chemicrystallisation process takes places in the oxidized zones, whereas the well known destruction of crystallites predominates in the center of the sample. Experiments at various oxygen pressures were made in order to determine the rate constants for the unperturbed oxidation process. Then, various kinetic models based on Fick's law, for the diffusion controlled oxidation were compared and discussed, showing that any explanation involving a variation of the permeability cannot be invoked. On this basis, explanations of the shape of the oxidation profiles are proposed, leading to a mathematical prediction of the thickness of the oxidized layer. Large difference is found near the boundaries between the experimental and theoretical results. Hypotheses are made and discussed to explain this discrepancy. The evolution of mechanical properties, for hdPE during radiochemical ageing is also presented; the influence of the oxidized layer on mechanical properties is shown

  7. Radiochemical separation of 90Sr from high level waste: scaled-up studies

    International Nuclear Information System (INIS)

    Radiochemical separation of 90Sr was carried out from High Level Waste (HLW) using a combination of different chemical techniques. This paper describes various steps for separation as well as the modifications incorporated during the scaling-up. (author)

  8. Automated radiochemical synthesis and biodistribution of [11C]l-α-acetylmethadol ([11C]LAAM)

    International Nuclear Information System (INIS)

    Long-acting opioid agonists methadone and l-α-acetylmethadol (LAAM) prevent withdrawal in opioid-dependent persons. Attempts to synthesize [11C]-methadone for PET evaluation of brain disposition were unsuccessful. Owing, however, to structural and pharmacologic similarities, we aimed to develop [11C]LAAM as a PET ligand to probe the brain exposure of long-lasting opioids in humans. This manuscript describes [11C]LAAM synthesis and its biodistribution in mice. The radiochemical synthetic strategy afforded high radiochemical yield, purity and specific activity, thereby making the synthesis adaptable to automated modules. - Highlights: • Radiochemical synthesis of opioid [11C]l-α-acetylmethadol (LAAM) described for the first time. • High radiochemical yield, purity and specific activity. • Easily reproducible and adaptable synthesis to any C-11 automated modules. • [11C]LAAM utility as a PET radiopharmaceutical for assessing brain penetration

  9. Radiochemical and biological control for metaiodobenzylguanidine (MIBG) labelled with 131I

    International Nuclear Information System (INIS)

    This study shows the standardization of the radiochemical control of MIBG-131I in eletrophoretic system and also the biological control in Wistar rat for a period of time, not longer than 60 minutes after the tracer administration. (author)

  10. Radiochemical and biological control of metaiodobenzyl-guanidine (MIBG) labeled with 131I

    International Nuclear Information System (INIS)

    This study shows the standardization of the radiochemical control of MIBG - 131I in eletrophoretic system and also the biological control in Wistar rat for a period of time, not longer than 60 minutes after tracer administration. (author)

  11. Enhanced CANDU 6 Reactor

    International Nuclear Information System (INIS)

    Full text: The CANDU 6 power reactor is visionary in its approach, remarkable for its on-power refuelling capability and proven over years of safe, economical and reliable power production. Developed by Atomic Energy of Canada Ltd, the CANDU 6 design offers excellent performance utilizing state-of-the-art technology. The first CANDU 6 plants went into service in the early 1980's as leading edge technology and the design has been continuously advanced to maintain superior performance with an outstanding safety record. The first CANDU 6 plants- Gentilly 2 and Point Lepreau in Canada, Embalse in Argentina and Wolsong- Unit 1 in Korea have been in service for more than 21 years and are still producing electricity at peak performance and to the end of 2004, their average lifetime Capacity Factor was 83.2%. The newer CANDU 6 units in Romania (Cernavoda 1), Korea (Wolsong-Units 2, 3 and 4) and Qinshan (Phase III- Units 1 and 2) have also been performing at outstanding levels. The average lifetime Capacity Factor of the 10 CANDU 6 operating units around the world has been 87% to the end of 2004. Building on these successes, AECL is committed to the further development of this highly successful design, now focussing on meeting customer's needs for reduced costs, further improvements to plant operation and performance, enhanced safety and incorporating up-to-date technology as warranted. This has resulted in AECL embarking on improving the CANDU 6 design through an upgraded product termed as the 'Enhanced CANDU 6' (EC6)- which incorporates several attractive but proven features that will make the CANDU 6 reactor even more economical, safer and easier to operate. Some of the key features that will be incorporated in the EC6 include increasing the plant's power output, shortening the overall project schedule, decreasing the capital cost, dealing with obsolescence issues, optimizing maintenance outages and incorporating lessons learnt through feedback obtained from the

  12. Quality assurance in radiochemical laboratories. Proceedings of the workshop contributions ZAKVARAL'09

    International Nuclear Information System (INIS)

    The aim of this collective work is an increase of knowledge level, expertness and qualification of personnel in the frame of general principles and practical radiation protection and environment protection with consecutive application at completion radiochemical analyses, evaluation of conversion parameters of riskiness of contaminants in practical application for chemical contaminants, ionizing radiation and radionuclides. The published collective work consists of 13 contributions from regular seminar meeting 'Quality assurance in radiochemical laboratories - ZAKVARAL 09', extended with 6 rigorous works in this area.

  13. GC/MS characterization of KSM-17, a phosphonic acid extractant of importance in radiochemical separation

    International Nuclear Information System (INIS)

    Phosphonic acids and their derivatives are finding increasing applications in the extractions of many useful radiochemical separations. Indigenously synthesized material is found to have many synthetic impurities; or else, a pure compound under harshness of radiochemical separation conditions would accumulate such undesirable species. GC/MS is a potential tool for confirming the identity of such compounds while the study of mass fragmentation pattern would also provide probable mechanism of its radiolytic degradation

  14. Biological and radiochemical quality control of indigenous 99mTc-radiopharmaceutical kits

    International Nuclear Information System (INIS)

    Biological and radiochemical quality control of indigenous (Pinscan) diagnostic cold kits of Methylene Diphosphonate (MDP), Tin-colloid and Diethylene Triamine Pentaacetic Acid (DTPA) was performed in parallel with imported Amersham's kits (Amerscan). The results of radiochemical purity, sterility, apyrogenicity and biodistribution of indigenous (Pinscan) kits were good and quantitatively and qualitatively comparable to those obtained with Amersham's (Amerscan) imported kits. (author) 21 refs.; 8 tabs

  15. Current studies of biological materials using instrumental and radiochemical neutron activation analysis

    International Nuclear Information System (INIS)

    Instrumental neutron activation analysis still remains the preferred option when analysing the trace element distribution in a wide rage of materials by neutron activation analysis. However, when lower limits of detection are required or major interferences reduce the effectiveness of this technique, radiochemical neutron activation analysis is applied. This paper examines the current use of both methods and the development of rapid radiochemical techniques for analysis of the biological materials, hair, cow's milk, human's milk, milk powder, blood and blood serum

  16. A selective separation method for 93Zr in radiochemical analysis of low and intermediate level wastes from nuclear power plants

    International Nuclear Information System (INIS)

    The zirconium isotope 93Zr is a long-lived pure β-particle-emitting radionuclide produced from 235U fission and from neutron activation of the stable isotope 92Zr and thus occurring as one of the radionuclides found in nuclear reactors. Due to its long half life, 93Zr is one of the radionuclides of interest for the performance of assessment studies of waste storage or disposal. Measurement of 93Zr is difficult owing to its trace level concentration and its low activity in nuclear wastes and further because its certified standards are not frequently available. A radiochemical procedure based on liquid-liquid extraction with 1-(2-thenoyl)-3,3,3-trifluoroacetone in xylene, ion exchange with Dowex resin and selective extraction using TRU resin has to be carried out in order to separate zirconium from the matrix and to analyze it by liquid scintillation spectrometry technique (LSC). To set up the radiochemical separation procedure for 93Zr, a tracer solution of 95Zr was used in order to follow the behavior of zirconium during the process by γ-ray spectrometry through measurement of the 95Zr. Then, the protocol was applied to low level waste (LLW) and intermediate level waste (ILW) from nuclear power plants. The efficiency detection for 63Ni was used to determination of 93Zr activity in the matrices analyzed. The limit of detection of the 0.05 Bq l-1 was obtained for 63Ni standard solutions by using a sample:cocktail ratio of 3:17 mL for OptiPhase HiSafe 3 cocktail. (author)

  17. Design of a multipurpose research reactor

    International Nuclear Information System (INIS)

    The availability of a research reactor is essential in any endeavor to improve the execution of a nuclear programme, since it is a very versatile tool which can make a decisive contribution to a country's scientific and technological development. Because of their design, however, many existing research reactors are poorly adapted to certain uses. In some nuclear research centres, especially in the advanced countries, changes have been made in the original designs or new research prototypes have been designed for specific purposes. These modifications have proven very costly and therefore beyond the reach of developing countries. For this reason, what the research institutes in such countries need is a single sufficiently versatile nuclear plant capable of meeting the requirements of a nuclear research programme at a reasonable cost. This is precisely what a multipurpose reactor does. The Mexican National Nuclear Research Institute (ININ) plans to design and build a multipurpose research reactor capable at the same time of being used for the development of reactor design skills and for testing nuclear materials and fuels, for radioisotopes production, for nuclear power studies and basic scientific research, for specialized training, and so on. For this design work on the ININ Multipurpose Research Reactor, collaborative relations have been established with various international organizations possessing experience in nuclear reactor design: Atomehnergoeksport of the USSR: Atomic Energy of Canada Limited (AECL); General Atomics (GA) of the USA; and Japan Atomic Energy Research Institute

  18. Phenolsulphotransferase in human tissue: radiochemical enzymatic assay and biochemical properties

    International Nuclear Information System (INIS)

    Phenolsulphotransferase (EC 2.8.2.1) (PST) is an important catecholamine and drug metabolizing enzyme. Optimal conditions have been determined for the accurate measurement of PST activity in the human platelet, human renal cortex, and human jejunum with a radiochemical microassay. 3-Methoxy-4-hydroxyphenylglycol (MHPG) and 35S-3'-phosphoadenosine-5'-phosphosulfate (35S-PAPS) were the substrates for the reaction. The apparent Michaelis-Menten (Ksub(m)) values for MHPG with platelet, renal cortex, and jejunum were 1.09, 0.46 and 1.16 mmol/l, respectively. Apparent Ksub(m) values for PAPS in the same tissues were 0.14, 0.13 and 0.21 μmol/l. The pH optimum of the reacton in all three tissues was approximately 6.2-6.8 with three different buffer systems. The coefficients of variation for the assay of platelet, renal cortex, and jejunal activities were 6.2%, 3.4% and 4.4%, respectively. Mean platelet PST activity in blood samples from 75 randomly selected adult subjects was 5.0 +- 1.72 mmol of MHPG sulfate formed per hour per mg of platelet protein (8.3 X 10-5 +- 2.9 X 10-5 μmol min-1 mg-1, mean +- S.D.). There was a 5-fold intersubject variation in platelet PST activity within two standard deviations of the mean value. Experiments in which partially purified human erythrocyte PST was added to platelet, kidney and gut homogenates under these assay conditions provided evidence that endogenous PST inhibitors did not affect the observed enzyme activity. (Auth.)

  19. Improved radiochemical assay analyses using TRITON depletion sequences in SCALE

    International Nuclear Information System (INIS)

    With the release of TRITON in SCALE 5.0, Oak Ridge National Laboratory has made available a rigorous two-dimensional (2D) depletion sequence based on the arbitrary-geometry 2D discrete ordinates transport solver NEWT. TRITON has recently been further enhanced by the addition of depletion sequences that use KENO V.a and KENO-VI for three-dimensional (3D) transport solutions. The Monte Carlo-based depletion sequences add stochastic uncertainty issues to the solution, but also provide a means to perform direct 3D depletion that can capture the effect of leakage near the ends of fuel assemblies. Additionally, improved resonance processing capabilities are available to TRITON using CENTRM. CENTRM provides lattice-weighted cross sections using a continuous energy solution that directly treats the resonance overlap effects that become more important in high-burnup fuel. And beginning with the release of SCALE 5.1 in the summer of 2006, point data and fine-structure multigroup libraries derived from ENDF/B-VI evaluations will be available. The combination of rigorous 2D and 3D capabilities with improved cross section processing capabilities and data will provide a powerful and accurate means for the characterization of spent fuel, making it possible to analyze a broad range of assembly designs and assay data. This in turn will reduce biases and uncertainties associated with the preduction of spent fuel isotopic compositions. This paper describes advanced capabilities of the TRITON sequence for depletion calculations and the results of analyses performed to date for radiochemical assay data. (author)

  20. Radiochemical assay for ACh: modifications for sub-picomole measurements

    International Nuclear Information System (INIS)

    Details are given of modifications which have been made to a radiochemical technique for the determination of picomole levels of ACH (Goldberg, A., and McCaman, R., 1973, J. Neurochem., vol. 20, 1; Goldberg, A., and McCaman, R., 1974, in Choline and Acetylcholine: Handbook of Chemical Assay Methods, ed. Hanin, I., Raven Press, New York). The procedure involved the conversion of ACh to choline-32PO4 in the presence of AT32P and the enzymes, acetylcholinesterase and choline kinase. A substantial increase in sensitivity (from 2.5 to 0.04 pmol) has been achieved by increasing the specific activity of the [γ-32P]ATP and decreasing the reagent blank. The number of manipulative steps has been reduced. The ion exchange columns, prepared in Pasteur pipettes, could be reused indefinitely. Assays could be carried out with 8 to 500 μg wet weight of tissue. Values for ACh determined by the technique were proportional to tissue over a 500-fold range, from approximately 0.1 to 50 μg protein. The increased sensitivity was demonstrated by comparison of the results for the ACh content of individual soma determined by both procedures. The modified assay procedure gave 125,000 cpm in an assay for ACh in single Aplysia neurons, showing that considerably smaller cells than these could be assayed for ACh. Only a fraction of the cell extract is needed to obtain reliable measurements, and the remainder of the extract can then be used in other assays. (U.K.)

  1. Solar neutrino measurement with radiochemical gallium detector (GALLEX)

    Science.gov (United States)

    von Ammon, Reinhard

    1994-04-01

    The GALLEX experiment for the detection of solar neutrinos by means of a radiochemical gallium detector is operated by groups from Italy, France, Germany, Israel and the USA in the Gran Sasso underground laboratory (LNGS) near L'Aquila (Italy). It consists of (1) the technical scale tank made of glass fiber reinforced polyester fabric containing 101 metric tons (54 cu m) of a highly concentrated (8 moles per liter) GaCl3 solution; (2) a gas sparging system for desorption of GeCl4 which has been formed by interaction of the neutrinos with gallium according to Ga-71 + nue yields Ge-71 + e(-) and by addition of ca. 1 mg of a stable Ge isotope; (3) the absorption columns for concentration of GeCl4 into a volume of 1 l of water; (4) the laboratory scale apparatus for conversion of GeCl4 to GeH4 and mixing with the counting gas Xe; (5) the counter filling station, and (6) the low level proportional counters. Contributions of possible side reactions which have to be corrected for, e.g. by cosmic muons, fast neutrons and alpha-emitters are discussed, as well as the purification of the target solution from long-lived ( t1/2 = 271 d) cosmogenic Ge-68. A first preliminary result after one year of solar neutrino measurement is presented. This constitutes the first direct measurement of the basic proton-proton fusion reaction in the core of the sun. This result, appreciably below the predictions of the standard solar model (SSM) (132 Solar Neutrino Units (SNU)) can be interpreted, together with the results of the chlori ne and KAMIOKANDE experiments either by astrophysics or by neutrino oscillations (Mikheyev-Smirnov-Wolfenstein (MSW) effect). The solar neutrino measurements are continuing and a calibration experiment with a Cr-51 source is in preparation.

  2. 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)

  3. Radiochemical analysis of 3H, 14C, 55Fe, 63Ni in waste samples

    International Nuclear Information System (INIS)

    Full text: In the nuclear waste from decommissioning of a nuclear facility, most of the beta activity is contributed from 14C, 3H, 55Fe, and 63Ni, especially in the first 10 years after the close of reactor. The determination of these radionuclides in radioactive waste is important for decommissioning and disposal of the waste. 14C, 3H, and 63Ni I are pure beta emitter; and 55Fe decay by electron capture. It is therefore necessary to decompose sample and separate individual radionuclide from matrix element and other interfering nuclides before measurement of their radioactivity. In our laboratory, analytical methods have been developed for the determination of these radionuclides in various nuclear wastes for the decommissioning of nuclear facilities. An oxidizing combustion method has been developed to decompose graphite, concrete, metals, and other solid samples. 14C and 3H released from the sample are separated trapped and measured by liquid scintillation counter (LSC). By this method the sample preparation time can be shortened to only 2-3 minutes. The detection limit of this method for 14C and tritium are 0.96 and 0.58 Bq/g graphite and 0.11 and 0.06 Bq/g concrete respectively. The interference of other radionuclides in samples is insignificant. A radiochemical separation procedure based on precipitation, ion exchange and extraction chromatography have been developed for the determination of 55Fe and 63Ni. The decontamination factors of the developed method are higher than 104 for the interfering radionuclides. The chemical recoveries for both Fe and Ni are higher than 80%. The detection limits of this method are 0.018 Bq for 55Fe and 0.014 Bq for 63Ni. The developed method has been successfully applied for the analysis of heavy concrete, graphite, steel, aluminium, and lead from nuclear reactor and the estimation of inventories of these radionuclides in nuclear waste have been carried out. In addition, the methods were also applied for the analysis of

  4. Fuel burnup characteristics for the NRU research reactor

    International Nuclear Information System (INIS)

    The driver fuel of the NRU research reactor at AECL, Chalk River is a low enriched uranium (LEU) fuel alloy of Al-61 wt% U3Si, consisting of particles of U3Si dispersed in a continuous aluminum matrix, with 19.8% U235 in uranium. This paper describes the burnup characteristics for this type of fuel in NRU, including the determination of fuel depletion using the neutronic simulation code TRIAD, comparisons between simulated and measured burnup values, and the regulatory licensing operational average fuel burnup limit. (author)

  5. Fuel burnup characteristics for the NRU research reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

    The driver fuel of the NRU research reactor at AECL, Chalk River is a low enriched uranium (LEU) fuel alloy of Al-61 wt% U{sub 3}Si, consisting of particles of U{sub 3}Si dispersed in a continuous aluminum matrix, with 19.8% U235 in uranium. This paper describes the burnup characteristics for this type of fuel in NRU, including the determination of fuel depletion using the neutronic simulation code TRIAD, comparisons between simulated and measured burnup values, and the regulatory licensing operational average fuel burnup limit. (author)

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

    International Nuclear Information System (INIS)

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

  7. Reliable reactor coolant pump seal performance - the station's role

    International Nuclear Information System (INIS)

    During the early days of the Canada deuterium uranium (CANDU) power reactor program, operators and designers learned that close attention to reactor coolant pump (RCP) seals was imperative for achieving high-capacity factors. This lesson was driven home by unpredictable and frequent seal failures in the following early CANDU plants. Those seal failures caused forced outages, maintenance/dose burdens, and heavy-water losses. Because then-available industrial seal technology proved inadequate in providing satisfactory fixes, Atomic Energy of Canada Limited (AECL) began a major effort to understand seal performance, develop improved designs, and evolve the station technology needed to attain the RCP seal reliable lifetime requirement of 4 yr. The payback has been huge: Fixes have been successfully implemented and excellent performance is now being achieved with AECL improved RCP seals. In this paper, the CANDU RCP seal experience, the methodology (with emphasis on the station's role) for attaining reliable long RCP seal life, and the adaptability of this technology to US light water reactors (LWRs) are discussed

  8. The low power miniature neutron source reactors: Design, safety and applications

    International Nuclear Information System (INIS)

    The Chinese Miniature Neutron Source Reactor (MNSR) is a low power research reactor with maximum thermal neutron flux of 1 x 1012 n.cm-2.s-1 in one of its inner irradiation channels and thermal power of approximately 30kW. The MNSR is designed based on the Canadian SLOWPOKE reactor and is one of the smallest commercial research reactors presently available in the world. Its commercial versions currently in operation in China, Ghana, Iran, Nigeria, Pakistan and Syria, is considered as an excellent tool for Neutron Activation Analysis (NAA), training of Scientist, and Engineers in nuclear science and technology and small scale radioisotope production. The paper highlights the basic design and theory of the commercial MNSR, its safety features, applications and advantages over the Chinese Prototype. The experimental flux characteristics determined in this work and in similar studies by other authors reveal that the commercial MNSR has more flux stability, longer life span, higher negative temperature coefficient of reactivity and low under-moderation compared to its prototype in China. The result shows that the facility is safe for reactor physics experiments, teaching and training of students and also ideal for application of NAA for the determination of elemental composition of biological and environmental samples. It can also be a useful tool for geochemical and soil fertility mapping. (author)

  9. Mo-99 production on a LEU solution reactor

    International Nuclear Information System (INIS)

    A pilot homogenous reactor utilizing LEU has been developed by the Kurchatov Institute in Moscow along with their commercial partner TCI Medical. This solution reactor operates at levels up to 50 kilowatts and has successfully produced high quality Mo-99 and Sr-89. Radiochemical extraction of medical radionuclides from the reactor solution is performed by passing the solution across a series of inorganic sorbents. This reactor has commercial potential for medical radionuclide production using LEU UO2SO4 fuel. Additional development work is needed to optimize multiple 50 kilowatt cores while at the same time, optimizing production efficiency and capital expenditure. (author)

  10. Luncheon address: Development of the CANDU reactor

    International Nuclear Information System (INIS)

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

  11. Substoichiometric radiochemical separation and their application to neutron activation analysis technique

    International Nuclear Information System (INIS)

    Radiochemistry is the chemistry of substances which are detected by their nuclear radiations. The first radiochemical separation was used by Marie and Pierre Curie. The present paper discusses the concept of purity and the importance of various radiochemical separation methods such as precipitation, volatilization, high vacuum distillation, ring oven technique, ion exchange, solvent extraction, etc. NAA can be carried out by comparator or absolute method. Comparator method can be performed by radiochemical or instrumental technique. Radiochemical separation for NAA requires that the final product must be in well-defined chemical form, so as to obtain correct chemical yield. The chemical yield determination is time-consuming and important, especially when the product of activation has a short half-life. Ruzika and Star/proposed a substoichiometric procedure which uses subequivalent amount of the reagent corresponding to the amount of carrier added. If the carrier and reagent added to the sample and standard are the same, and the same fraction is isolated free from other activation products, the chemical yield determination is not required. The process becomes quantitative and time saving. The present paper discusses some new, rapid and selective method developed for the substoichiometric radiochemical separation and estimation of some elements

  12. Innovative reactor technologies - Enabling success

    International Nuclear Information System (INIS)

    Full text: 1. Introduction. AECL has been pursuing innovation in reactor design using an evolutionary approach. The Advanced CANDU Reactor, or ACR design is the logical next step in the CANDU fuel-channel reactor design process, and achieves major improvements in economics while expanding safety margins. ACR technology is also the start of the long-term development path for CANDU fuel channel reactors. This path fits in with the long-term development directions identified in initiatives such as the Generation IV International Forum (GIF) and INPRO. AECL sees the product evolution from the ACR to the Supercritical Water-costed Reactor (SCWR), one of the concepts identified by GIF for further development. One of the prominent characteristics of the heavy-water moderated fuel channel reactor approach is the high potential for innovation. 2. Background to Development Strategy. As part of the pressurized water reactor family, CANDU's share many characteristics with other light water reactors, while retaining a set of distinctive features: High pressure water coolant in individual fuel channels, with low-pressure, low temperature moderator; Horizontal fuel channel design with on-power refuelling; Simple, easily-fabricated fuel bundle design; Use of heavy water to improve neutron efficiency. Since the original CANDU reactors were first built, the advent of deregulated, competitive energy markets has strongly emphasized the importance of low capital cost and short construction time. The ACR design innovations are chosen to respond to this evolution of energy markets, while retaining the proven features of the CANDU line. 3. Innovative CANDU Product Designs. The ACR-700 design is an evolutionary development of familiar CANDU technology, adding a carefully chosen set of innovations to the major improvements in economics, operations and safety margins: Slightly enriched uranium fuel contained in CANFLEX bundles; Light water replacing heavy water as the reactor coolant; More

  13. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

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

  14. Understanding temperatures and pressures during short radiochemical reactions.

    Science.gov (United States)

    Lazari, Mark; Irribarren, Jonathan; Zhang, Shuang; van Dam, R Michael

    2016-02-01

    Automated radiosynthesizers are critical for the reliable, routine production of PET tracers. To perform reactions in these systems, the temperature of the reactor heater is controlled, and the liquid temperature within the reaction vessel is presumed to closely follow. In reality, the liquid temperature can lag by several minutes and generally does not reach the heater temperature. Furthermore, because different synthesizers have different heating mechanisms and geometries, discrepancies are certain to exist between the actual temperatures experienced by the reaction mixture on different synthesizers. For dissimilar reactors, this can necessitate re-optimization of conditions when adapting a synthesis from one system to another, especially for the short-duration reactions common in radiochemistry. Herein, we study the relationship between the temperatures of the reactor heater and reaction liquid for various solvents using the ELIXYS radiosynthesizer as a representative example of a vial-based system. Our aims are to quantitatively illustrate this discrepancy to the community and provide data necessary to enable efficient translation of protocols between other radiosynthesizers and the ELIXYS. PMID:26706993

  15. Water chemistry management in cooling system of research reactor in JAERI

    International Nuclear Information System (INIS)

    The department of research reactor presently operates three research reactors (JRR-2, JRR-3M and JRR-4). For controlling and management of water and gas in each research reactor are performed by the staffs of the research reactor technology development division. Water chemistry management of each research reactor is one of the important subject. The main objects are to prevent the corrosion of water cooling system and fuel elements, to suppress the plant radiation build-up and to minimize the radioactive waste. In this report describe a outline of each research reactor facilities, radiochemical analytical methods and chemical analytical methods for water chemistry management. (author)

  16. Determination of the tritium content in the reactor heavy water, Phase II

    International Nuclear Information System (INIS)

    Measurement results of the 3H activity in non-irradiated water and after reactor operation are presented. Methods were developed for sampling and radiochemical water purification by ion exchange and multiple distillation. Methods for absolute measurement of soft beta radiation of tritium were established. Migration of tritium through the heavy water RA reactor system was monitored. Results were compared with other measured reactor parameters

  17. Microwave assisted rapid and improved radiochemical method for the estimation of uranium in leaf samples

    International Nuclear Information System (INIS)

    In this paper, we report the development of a rapid and improved radiochemical method assisted by microwave technique for the determination of uranium in leafy samples for use in radiological emergency situations, where quick assessment of radioactivity is required. About 200 g of fresh leaf sample was ashed in a microwave muffle furnace, followed by digestion in a microwave digester and radiochemically separated using UTEVA resin. Counting was performed in a passivated ion implanted planar silicon alpha spectrometer after electroplating the separated sample for uranium. By this method, the time of sample processing and analysis was reduced to few hours from about 2 days and also the radiochemical recovery of uranium has considerably enhanced as compared to conventional methods. (author)

  18. Microwave assisted rapid and improved radiochemical method for the estimation of uranium in leaf samples

    International Nuclear Information System (INIS)

    In this paper, we report the development of a rapid and improved radiochemical method assisted by microwave technique for the determination of uranium in leafy samples for use radiological emergency situations, where in the quick assessment of radioactivity is required. About 200 g of fresh garden leaf sample was ashed in a microwave muffle furnace, followed by digestion in a microwave digester and radiochemically separated using UTEVA resin. Counting was performed in a PIPS detector alpha spectrometer after electroplating the separated sample for uranium. By this method, the time of sample processing and analysis was reduced to few hours from about two days and also the radiochemical recovery of uranium has considerably enhanced as compared to conventional methods. (author)

  19. Radiolabeling, quality control and radiochemical purity assessment of 99mTc-HYNIC-TOC

    International Nuclear Information System (INIS)

    Somatostatine receptors are widely expressed by several tumors, especially of the neuroendocrine origin. In vivo images of these tumors using radiolabeled somatostatine analogues became a useful clinical tool in oncology. The aim of this work was the radiolabeling of the somatostatine analogue HYNIC-TOC with 99mTc as well as the evaluation of the radiochemical stability and quality control of labeled complex. 99mTc-HYNIC-TOC was produced by labeling conditions using 20 μg of peptide, 20 mg of tricine and 10 mg of EDDA as coligands, 1110 MBq of 99mTc (99Mo-99mTc IPEN-TEC generator) and 15 μg of SnCl2.2H2O. The reaction proceeds for 10 minutes at boiling water bath. Radiochemical purity of labeled preparation was evaluated by different chromatographic systems: ITLC-SG in methanol:ammonium acetate (1:1); TLC-SG in sodium citrate buffer 0.1 N pH 5.0 and methylethylketone, and HPLC employing column C-18, 5 μm, 4.6 mm x 250 mm, UV (220 nm), radioactivity detectors, 1 mL/minute flow of acetonitrile and trifluoroacetic acid solution 0.1 %. Labeled compound has been found radiochemically stable for 5 hours and radiochemical purity was higher than 90 %. The thin layer chromatographic systems enabled the separation of radiochemical species presented in the labeled mixture as well as HPLC system. The labeling procedure studied resulted in high radiochemical yield and easy preparation. Future works include the preparation of a lyophilized reagent to make feasible the preparation of 99mTc-HYNIC-TOC at nuclear medicine services in order to study the clinical potential of the radiopharmaceutical in diagnostic and staging of neuroendocrine tumors. (author)

  20. Recycling of vanadium alloys in fusion reactors

    International Nuclear Information System (INIS)

    The feasibility of reprocessing a vanadium alloy after its use as a structural material in a fusion reactor, in order to enable its subsequent hands-on recycling within the nuclear industry, has been determined. For less neutron-exposed components, clearance of materials has also been considered. A conceptual model for the radiochemical processing of the alloy has been developed and tested experimentally. Using di-2-ethyl-hexyl-phosphoric acid it is possible to purify the components of the V-Cr-Ti alloy after its exposure in a fusion reactor down to the required level of activation product concentrations

  1. Kit preparation of 153Sm-EDTMP and factors affecting radiochemical purity and stability

    International Nuclear Information System (INIS)

    A fast kit method was developed for the production of 153Sm-EDTMP in two steps avoiding the use of nitric acid, evaporation and sterilization of the final solution by autoclave. Methods of analysis for the determination of chemical and radiochemical purity in the radiopharmaceutical solution were established. Factors affecting radiochemical purity and stability of the complex as the molar ratio of EDTMP/Sm, concentration of phosphate buffer and neutralization of EDTMP prior kit preparation were also analyzed. The use of this radiopharmaceutical in rabbits and patients showed selective skeletal uptake. (author). 5 refs., 4 figs., 3 tabs

  2. Automation of Extraction Chromatograhic and Ion Exchange Separations for Radiochemical Analysis and Monitoring

    International Nuclear Information System (INIS)

    Radiochemical analysis, complete with the separation of radionuclides of interest from the sample matrix and from other interfering radionuclides, is often an essential step in the determination of the radiochemical composition of a nuclear sample or process stream. Although some radionuclides can be determined nondestructively by gamma spectroscopy, where the gamma rays penetrate significant distances in condensed media and the gamma ray energies are diagnostic for specific radionuclides, other radionuclides that may be of interest emit only alpha or beta particles. For these, samples must be taken for destructive analysis and radiochemical separations are required. For process monitoring purposes, the radiochemical separation and detection methods must be rapid so that the results will be timely. These results could be obtained by laboratory analysis or by radiochemical process analyzers operating on-line or at-site. In either case, there is a need for automated radiochemical analysis methods to provide speed, throughput, safety, and consistent analytical protocols. Classical methods of separation used during the development of nuclear technologies, namely manual precipitations, solvent extractions, and ion exchange, are slow and labor intensive. Fortunately, the convergence of digital instrumentation for preprogrammed fluid manipulation and the development of new separation materials for column-based isolation of radionuclides has enabled the development of automated radiochemical analysis methodology. The primary means for separating radionuclides in solution are liquid-liquid extraction and ion exchange. These processes are well known and have been reviewed in the past.1 Ion exchange is readily employed in column formats. Liquid-liquid extraction can also be implemented on column formats using solvent-impregnated resins as extraction chromatographic materials. The organic liquid extractant is immobilized in the pores of a microporous polymer material. Under

  3. The role of high performance liquid chromatography in radiochemical/radiopharmaceutical synthesis and quality assurance

    International Nuclear Information System (INIS)

    The usefulness of HPLC in all areas of radiopharmaceutics has been demonstrated in numerous laboratories, particularly in the development of in-house radiopharmaceuticals for SPECT and PET. HPLC continues to be a powerful tool in preparation and quality assurance (QA) as illustrated in such areas as chemical and radiochemical identification; product separation and isolation; preparative scale purification; and specific activity determination. A review of established HPLC techniques in radiopharmaceutics will be presented. Examples from the literature as well as newer applications will be used in an attempt to assess and define the present-day role of HPLC in the preparation of radiochemicals and radiopharmaceuticals with emphasis on QA

  4. Radiochemical procedures for determination of selected members of the uranium and thorium series

    International Nuclear Information System (INIS)

    The radiochemical procedures contained in this manual are adaptations of those developed and published by many radiochemists. In many cases the identity of the originator is not clear and usually modifications in the original procedure have been made by subsequent workers. Nearly all of the basic radiochemical techniques and separations in use today were developed during the Manhattan Project and can be found in U.S.A.E.C. reports published from 1945 to 1953. This manual contains methods for the determination of Pb-210, Po-210; Ra-226, Ra-228, Th-228, Th-230 and Th-232. (auth)

  5. Inorganic and Radiochemical Analysis of AW-101 and AN-107 Tank Waste

    International Nuclear Information System (INIS)

    This report presents the inorganic and radiochemical analytical results for AW-101 and AN-107 as received materials. The analyses were conducted in support of the BNFL Proposal No. 30406/29274 Task 5.0. The inorganic and radiochemical analysis results obtained from the as received materials are used to provide initial characterization information for subsequent process testing and to provide data to support permit application activities. Quality Assurance (QA) Plan MCS-033 provides the operational and quality control protocols for the analytical activities, and whenever possible, analyses were performed to SW-846 equivalent methods and protocols

  6. Cadmium determination in biological samples using neutron activation analysis with radiochemical separations

    International Nuclear Information System (INIS)

    Chile has 7500 km of coastline on the Southern Pacific ocean,with about 4500 km of continental coastline that contains a variety of different geographical zones.This variety means that there is a great diversity of marine resources such as fish, shellfish and seaweeds. The utilization of these resources has been increasing in recent years making this sector an economically important one. The catch as of May 2002 came to 1.9 million tons and exports of the different species amounted to US$611.5 million as of April.But this important economic resource is being threatened by the technical demands imposed by importing countries, mainly the specific requirements for sanitary certification for fishery export products, depending on the markets of destination. The chemical element cadmium is one of the most strictly controlled elements due some shellfish accumulate a large amount of this element and to its high toxicity. The Chilean standard's analytical procedures for cadmium determination in hydro biological products, which must be met by laboratories that certify and control these products for export, are now being evaluated. Through its Chemical Metrology Unit, the Chilean Nuclear Energy Commission is strongly supporting this sector by preparing the secondary reference or control materials, and it has developed and implemented nuclear analytical methods for the certification of these materials, which will be used mostly in collaborative studies. This work describes the methodology developed for the determination of cadmium in biological samples, particularly in shellfish and fish. The method is based on neutron activation analysis with radiochemical separations, using the radioisotopes 115Cd and 115mIn, generated in the samples by bombarding with neutrons in a nuclear reactor. The samples were digested at 110oC with H2SO4 and H2O2 and then the radioactive cadmium element was separated from the other elements present in the samples using a Bio Rad AG 2-X8 resin

  7. 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)

  8. 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)

  9. Rapid radiochemical analysis of 131I in environmental samples using a well-type Ge-detector

    International Nuclear Information System (INIS)

    Large-scale production of 131I in a nuclear reactor, the gaseous nature of 131I, and its selective uptake by the human thyroid gland, make this radioisotope a health hazard in the event of a nuclear accident. The maximum concentration of 131I in drinking water has been set at 1 pCi/l. Human ingestion of 131I through the grass-cow-milk pathway makes milk an environmentally significant matrix to be monitored for. A simple and a rapid radiochemical procedure for the analysis of 131I in water and milk samples is reported. A quick single-step separation on anion-exchange resin concentrates radioiodine from large sample volumes. The resin is then directly counted in the cavity of a low-background well-type HPGe detector that has high counting efficiency for X-rays and low-energy γ-radiation. Chemical recovery is evaluated from the intensity of the 29.6 keV X-rays of the 129I spike, and 131I is assayed through the intensity of its 364.5 keV γ-peak. The method's minimum detection limit is 0.5 pCi 131I based on a 1 liter sample and a 200-minute count. (author)

  10. Recycled uranium: An advanced fuel for CANDU reactors

    International Nuclear Information System (INIS)

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

  11. The next generation of CANDU reactor: evolutionary economics

    International Nuclear Information System (INIS)

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

  12. Chances dim for Sask. reactor

    International Nuclear Information System (INIS)

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

  13. 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

  14. Fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    For a period of about 10 years AECL had a significant program looking into the possibility of developing U3Si as a high density replacement for the UO2 pellet fuel in use in CANDU power reactors. The element design consisted of a Zircaloy-clad U3Si rod containing suitable voidage to accommodate swelling. We found that the binary U3Si could not meet the defect criterion for our power reactors, i.e., one month in 300 degree C water with a defect in the sheath and no significant damage to the element. Since U3Si could not do the job, a new corrosion resistant ternary U-Si-Al alloy was developed and patented. Fuel elements containing this alloy came close to meeting the defect criterion and showed slightly better irradiation stability than U3Si. Shortly after this, the program was terminated for other reasons. We have made much of this experience available to the Low Enrichment Fuel Development Program and will be glad to supply further data to assist this program

  15. Radiochemical studies on corrosion products of oral biomaterials

    International Nuclear Information System (INIS)

    The work given in this thesis deals with a radioactive tracer study of the sorption of the corrosion products of dental amalgams and antimony on human teeth, porcelain and acrylic materials, used as dental restorative material. Sorption was investigated in presence of water and liquids commonly intaken by man; namely tea with or without sugar, soluble coffee ( Nescaffee) with or without sugar and/or milk, red tea (karkadeh or hibiscus) with or without sugar and chicken soup. The radioactive isotopes of Ag, Sn, Zn (amalgam components) and antimony were prepared by their irradiation in the nuclear reactor; 110m Ag, 113Sn, 65Zn and 124 Sb were thereby produced. The percent uptake of each studied element was evaluated from the depletion of radioactivity of the corresponding radioactive tracer in the given medium containing a tooth (human or artificial)

  16. Radiochemical measurement of fast neutrons using a Ca(NO3)2 aqueous solution

    International Nuclear Information System (INIS)

    A radiochemical detector has been built to measure fast neutrons (E>2.5 MeV), using the reaction n + 40 Ca → 37 Ar + 4 He. The target consists in 470 litres of an aqueous solution of calcium nitrate. This detector has been used in the GALLEX environment at the Gran Sasso Underground Laboratory. (authors). 19 refs., 10 figs., 4 tabs

  17. Radiochemical measurement of mass distribution in 16O+238U reaction at sub-barrier energy

    International Nuclear Information System (INIS)

    In the present, radiochemical study of the mass distribution in 16O+238U has been carried out at sub-barrier energy to investigate the nature of mass distribution in CFF and TF channels. In addition, cross sections of evaporation residues formed in one nucleon transfer/pick-up reactions have also been measured

  18. Computer aided piping layout design in radiochemical plants- an improved software package

    International Nuclear Information System (INIS)

    A software package was developed and it was successfully implemented for the piping layout design of the four process cells of the Kalpakkam Reprocessing Project. This paper discusses in detail all the improvements and modifications that are being carried out in the package so that it becomes more meaningful and useful for implementation for the forthcoming radiochemical plants

  19. On the form formation during flotation processing of the waste waters of radiochemical plant

    International Nuclear Information System (INIS)

    Froth formation in the process of flotational treatment of imitated and real sewage from radiochemical processes with a high content of anionic and nonionogenic surfactants was investigated. It is suggested that water-soluble polyelectrolyte VPK-402 should be used to reduce carryover of the waters treated to the froth

  20. Foam formation during flotation treatment of wastewaters from the radiochemical industry

    International Nuclear Information System (INIS)

    Foam formation is studied during flotation treatment of simulated and actual wastewaters from the radiochemical industry that contain large amounts of anionic and nonionic surface-active substances. Use of the water-soluble polyelectrolyte VPK-402 is recommended in order to reduce losses of treated waters to the foam. 5 refs., 5 tabs

  1. Features of radiochemical polymerization of the PEh-265 polyester lacquer on thermal insulating substrates

    International Nuclear Information System (INIS)

    The peculiarities of radiation polymerization of polyester lacquer on thermal insulating substrates have been investigated. The same features of polymerization on both pearlite and lignin substrates were studied. Physical, mechanical and thermal protective properties of the created materials were detected. It is shown that radiochemical modification of the surface layer on perlite or lignin substrates gives advanced heat-insulating materials

  2. Safety requirements in the design of research reactors: A Canadian perspective

    International Nuclear Information System (INIS)

    In Canada, the formal development of safety requirements for the design of research reactors in general began under an inter-organizational Small Reactor Criteria Committee. This committee developed safety and licensing criteria for use by several small reactor projects in their licensing discussions with the Atomic Energy Control Board. The small reactor projects or facilities represented included the MAPLE-X10 reactor, the proposed SES-10 heating reactor and its prototype, the SDR reactor at the Whiteshell Laboratories, the Korea Multipurpose Research Reactor (a.k.a., HANARO) in Korea, the SCORE project, and the McMaster University Nuclear Reactor. The top level set of criteria which form a safety philosophy and serve as a framework for more detailed developments was presented at an IAEA Conference in 1989. AECL continued this work to develop safety principles and design criteria for new small reactors. The first major application of this work has been to the design, safety analysis and licensing of the MAPLE 1 and 2 reactors for the MDS Nordion Medical Isotope Reactor Project. This paper provides an overview of the safety principles and design criteria. Examples of an implementation of these safety principles and design criteria are drawn from the work to design the MAPLE 1 and 2 reactors. (author)

  3. Synthesis of 195mPt radiolabelled cis-diamminedichloroplatinum(II) of high chemical and radiochemical purity using high performance liquid chromatography

    International Nuclear Information System (INIS)

    An improved method is described for the synthesis of 195mPt-radiolabelled cis-diamminedichloroplatinum(II). An amount of 10 mg of 95% enriched 194Pt was irradiated for 75 h in the hydraulic conveyer of the Kyoto University Reactor at a thermal neutron flux of approximately 8.15x1013 n.cm-2.sec-1 and the 195mPt-radiolabelled CDDP was purified using HPLC. The chemical yield is 61%, chemical purity is greater than 99.74%, the radiochemical purity is nearly 100%, and the specific activity is 7.4x106 Bq mg-1 CDDP (200 μCi mg-1 CDDP). (author) 9 refs.; 5 figs.; 1 tab

  4. The Northern Marshall Islands radiological survey: A quality control program for radiochemical and gamma spectroscopy analysis

    International Nuclear Information System (INIS)

    From 1979 to 1989, approximately 25,000 Post Northern Marshall Islands Radiological Survey (PNMIRS) samples were collected, and over 71,400 radiochemical and gamma spectroscopy analyses were performed to establish the concentration of 90Sr, 137Cs, 241Am, and plutonium isotopes in soil, vegetation, fish, and animals in the Northern Marshall Islands. While the Low Level Gamma Counting Facility (B379) in the Health and Ecological Assessment (HEA) division accounted for over 80% of all gamma spectroscopy analyses, approximately 4889 radiochemical and 5437 gamma spectroscopy analyses were performed on 4784 samples of soil, vegetation, terrestrial animal, and marine organisms by outside laboratories. Four laboratories were used by Lawrence Livermore National Laboratory (LLNL) to perform the radiochemical analyses: Thermo Analytical Norcal, Richmond, California (TMA); Nuclear Energy Services, North Carolina State University (NCSU); Laboratory of Radiation Ecology, University of Washington (LRE); and Health and Ecological Assessment (HEA) division, LLNL, Livermore, California. Additionally, LRE and NCSU were used to perform gamma spectroscopy analyses. The analytical precision and accuracy were monitored by including blind duplicates and natural matrix standards in each group of samples analyzed. On the basis of reported analytical values for duplicates and standards, 88% of the gamma and 87% of the radiochemical analyses in this survey were accepted. By laboratory, 93% of the radiochemical analyses by TMA; 88% of the gamma-ray spectrometry and 100% of the radiochemistry analyses by NCSU; 89% of the gamma spectroscopy and 87% of the radiochemistry analyses by LRE; and 90% of the radiochemistry analyses performed by HEA's radiochemistry department were accepted

  5. 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

  6. 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.

  7. Calculation and comparisons with measurement of fast neutron fluxes in the material testing facilities of the NRU research reactor

    International Nuclear Information System (INIS)

    The NRU reactor at Chalk River provides three irradiation facilities to study the effects of fast neutrons (E> 1 MeV) on reactor materials for assessing material damage and deformation. The facilities comprise two types of fast neutron rods (Mark 4 and Mark 7), and a Material Test Bundle (MTB) irradiated in a loop site. This paper describes the neutronic simulation of these testing facilities using the WIMS-AECL and TRIAD codes, and comparisons with the fast neutron flux measurements using iron-wire activation techniques. It also provides comparisons of flux levels, neutron spectra, and size limitations of the experimental cavities between these test facilities. (author)

  8. Calculation and comparisons with measurement of fast neutron fluxes in the material testing facilities of the NRU research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    The NRU reactor at Chalk River provides three irradiation facilities to study the effects of fast neutrons (E> 1 MeV) on reactor materials for assessing material damage and deformation. The facilities comprise two types of fast neutron rods (Mark 4 and Mark 7), and a Material Test Bundle (MTB) irradiated in a loop site. This paper describes the neutronic simulation of these testing facilities using the WIMS-AECL and TRIAD codes, and comparisons with the fast neutron flux measurements using iron-wire activation techniques. It also provides comparisons of flux levels, neutron spectra, and size limitations of the experimental cavities between these test facilities. (author)

  9. Radiochemical methods used by the IAEA's laboratories at Siebersdorf for the determination of 90Sr, 144Ce and Pu radionuclides in environmental samples collected for the International Chernobyl Project

    International Nuclear Information System (INIS)

    During the IAEA's International Chernobyl Project to assess the radiological consequences of the nuclear reactor accident, the Agency's Laboratories at Seibersdorf participated in the collection and analyses of environmental samples from the Soviet Union. Under Task 2 of this effort, the determination of the activity concentrations of 90Sr and the alpha-emitting Pu radionuclides was important for the corroboration of the official USSR environmental contamination maps. The present paper describes in detail the sampling methods and radiochemical procedures used for the 90Sr, 144Ce, 238Pu and 239,240Pu analyses in these samples with emphasis on the grass and soil treatments. (Author)

  10. Pre-licensing of the Advanced CANDU Reactor

    International Nuclear Information System (INIS)

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

  11. Dry Storage of Spent Fuel Discharged from Research Reactors in Canada

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) manages the spent fuel discharged from most of Canada's research reactors. These research reactors have been operated to support Canada's nuclear R and D programmes and medical isotope production. The spent fuel inventory consists of dozens of types and configurations, including intact and non-intact fuels and totalling approximately 95 MTHM. The fuels are a mixture of both high and low enrichments, and are typically aluminium-clad with various fuel core materials, including uranium metal, uranium dioxide, uranium-aluminium alloy, and uranium silicide-aluminium dispersion fuels. The discharged spent fuels are initially wet stored at reactor fuel bays for approximately two years before being transferred to dry storage facilities, at AECL's Chalk River Laboratories site and Whiteshell Laboratories site. After 50-100 years of interim dry storage, the spent fuel will be sent to permanent geological disposal as part of Canada's long term management programme for spent nuclear fuel. This paper presents the strategy and practices with regard to the management and storage of spent research reactor fuel in Canada. (author)

  12. Implementation of advanced electrochemical oxidation for radiochemical concentrate treatment

    International Nuclear Information System (INIS)

    Water treatments in Nuclear Power Plants include ion exchange, evaporation and mechanical filtration techniques. These technologies are used to control the chemical release and to treat coolant in light water reactor types from chemicals and most importantly, from radioactive nuclides. Most of the conventional methods are efficient, but at the same time producing aqueous concentrates with high organic load. Before final storage, the level of organic content of those concentrates must be reduced. Advanced electrochemical oxidation with Boron Doped Diamond (BDD) electrodes are being investigated in laboratory- and pilot scale for treatment of dilute and concentrated aqueous waste streams at Vattenfall-Ringhals NPP. BDD anodes and cathodes are having high over potential against water electrolysis, and therefore well suitable for oxidation of organics. Dilute wastewater, such as laundry water, which has an initial COD level of around 500 mg/l, was reduced to a level of < 20 mg/l in the laboratory. Evaporator concentrates, with a TS content of 3% and pH of 7-8, were treated in pilot scale of 800 liters, working in batch operation mode, at temperatures between 25-50 deg. C. Initial COD levels between 2500 and 8000 mg/l in concentrate was reduced to < 100 mg/l at the first tests and later to < 300 mg/l. The advanced electrochemical oxidation is proven to be a promising technique for radioactive concentrate treatment. Long-term operation is still ongoing to evaluate the performance of the electrodes, cell components and overall process efficiency. (authors)

  13. Scientific and technical conference. Problems and horizons of development of chemical and radiochemical control in nuclear energetics. Collection of summaries of reports

    International Nuclear Information System (INIS)

    During scientific and technical conference on problems of development of chemical and radiochemical control in nuclear energetics following themes were considered: the problems of methodological and instrumental assurance of chemical and radiochemical control at working nuclear power plants and nuclear energetic units; modern conceptions of automation systems construction of chemical and radiochemical control on the basis of intellectual measuring channels; the ways of decision of generally system problems of organization and management of chemical and radiochemical control using computed technologies; the problems of certification of chemical and radiochemical methods of measuring in nuclear energetics

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

    International Nuclear Information System (INIS)

    To replace the aging NRU reactor, AECL has developed the concept for a dual-purpose national Irradiation Research Facility (IRF) that tests fuel and materials for CANDU (CANada Deuterium Uranium) reactors and performs materials research using extracted neutron beams. The IRF includes a MAPLE reactor in a containment building, experimental facilities, and support facilities. At a nominal reactor power of 40 MWt, the IRF will generate powers up to 1 MW in natural-uranium CANDU bundles, fast-neutron fluxes up to 1.4 x 1018 n·m-2·s-1 in Zr-alloy specimens, and thermal-neutron fluxes matching those available to the NRU beam tubes. (author). 9 refs., 5 tabs., 2 figs

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

    International Nuclear Information System (INIS)

    To replace the aging NRU reactor, AECL has developed the concept for a dual-purpose national Irradiation Research Facility (IRF) that tests fuel and materials for CANDU (CANada Deuterium Uranium) reactors and performs materials research using extracted neutron beams. The IRF includes a MAPLE reactor in a containment building, experimental facilities, and support facilities. At a nominal reactor power of 40 MWt, the IRF will generate powers up to 1 MW in natural-uranium CANDU bundles, fast-neutron fluxes up to 1.4 x 1018 N·m-2·s-1 in Zr-alloy specimens, and thermal-neutron fluxes matching those available to the NRU beam tubes. (author). 9 refs., 5 tabs., 2 figs

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

    International Nuclear Information System (INIS)

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

  17. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    International Nuclear Information System (INIS)

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  18. Case study of inventory difference (ID) computation and analysis based on radiochemical plant model

    International Nuclear Information System (INIS)

    Inventory Difference (ID) computation and analysis is an urgent task of high priority in the field of nuclear material control and accountancy. In this paper this task is considered from the point of view of studying different practical cases (case study) in order to upgrade qualification of nuclear material control and accounting specialists. Training courses which are regularly held in Russian Methodological and Training Centre, SCI, Obninsk training center and discussions with specialists during those courses confirm how urgent this task is. In this paper the model of radiochemical plant is considered, for this case the practical tasks and solutions have been developed. The case study given in the paper is the first version of ID calculation and analysis for a radiochemical plant

  19. Some problems concenrning the use of automated radiochemical separation systems in destructive neutron activation analysis

    International Nuclear Information System (INIS)

    The present state of a long term program is reviewed. It was started to elaborate a remote controlled automated radiochemical processing system for the neutron activation analysis of biological materials. The system is based on wet ashing of the sample followed by reactive desorption of some volatile components. The distillation residue is passed through a series of columns filled with selective ion screening materials to remove the matrix activity. The solution is thus ''stripped'' from the interfering radioions, and it is processed to single-elements through group separations using ion-exchange chromatographic techniques. Some special problems concerning this system are treated. (a) General aspects of the construction of a (semi)automated radiochemical processing system are discussed. (b) Comparison is made between various technical realizations of the same basic concept. (c) Some problems concerning the ''reconstruction'' of an already published processing system are outlined. (T.G.)

  20. Trace determination of uranium and thorium in biological samples by radiochemical neutron activation analysis

    International Nuclear Information System (INIS)

    Radiochemical neutron activation analysis (RNAA) is an excellent method for determining uranium and thorium; it offers unique possibilities for their ultratrace analysis using selective radiochemical separations. Regarding the favourably sensitive nuclear characteristics of uranium and of thorium with respect to RNAA, but the different half-lives of their induced nuclides, two different approaches were used. In the first approach uranium and thorium were determined separately via 239U, 239Np and 233Pa. In the second approach these elements were 239239233 determined simultaneously in a single sample using U and/or Np and Pa. Isolation of induced nuclides was based on separation by extraction and/or anion exchange chromatography. Chemical yields were measured in each sample aliquot using added 235U, 238Np and 231Pa radioisotopic tracers. (author)

  1. Radiochemical neutron activation analysis of trace lanthanoids in geological and cosmochemical samples

    International Nuclear Information System (INIS)

    In order to determine trace lanthanoids in geological and cosmochemical samples, an analytical procedure for radiochemical neutron activation analysis (RNAA) was developed, where lanthanoids are radiochemically purified through precipitations of hydroxides and fluorides, and cation exchange using HBr as an eluent. Chemical yields are determined by reactivation. The procedure was applied to the Allende meteorite and geological standard rocks. Our data for Allende are in excellent agreement with literature values, and those values for standard rocks, JP-1 (peridotite) and JF-2 (feldspar), in which lanthanoids are depleted by orders of 1 to 2 compared with those in Allende, seem to be reasonable, although not properly evaluated because of large scatterings in their literature data. This suggests that the present RNAA procedure is highly effective in determining trace lanthanoids (less than 1 ng) in rock samples. (orig.)

  2. RADCHEM - Radiochemical procedures for the determination of Sr, U, Pu, Am and Cm

    International Nuclear Information System (INIS)

    An accurate determination of radionuclides from various sources in the environment is essential for assessment of the potential hazards and suitable countermeasures both in case of accidents, authorised release and routine surveillance. Reliable radiochemical separation and detection techniques are needed for accurate determination of alpha and beta emitters. Rapid analytical methods are needed in case of an accident for early decision-making. The objective of this project has been to compare and evaluate radiochemical procedures used at Nordic laboratories for the determination of strontium, uranium, plutonium, americium and curium. To gather detailed information on the procedures in use, a questionnaire regarding various aspects of radionuclide determination was developed and distributed to all (sixteen) relevant laboratories in the Nordic countries. The response and the procedures used by each laboratory were then discussed between those who answered the questionnaire. This report summaries the findings and gives recommendation on suitable practice. (au)

  3. RADCHEM - Radiochemical procedures for the determination of Sr, U, Pu, Am and Cm

    Energy Technology Data Exchange (ETDEWEB)

    Sidhu, R. [Inst. for Energy Technology (Norway)

    2006-04-15

    An accurate determination of radionuclides from various sources in the environment is essential for assessment of the potential hazards and suitable countermeasures both in case of accidents, authorised release and routine surveillance. Reliable radiochemical separation and detection techniques are needed for accurate determination of alpha and beta emitters. Rapid analytical methods are needed in case of an accident for early decision-making. The objective of this project has been to compare and evaluate radiochemical procedures used at Nordic laboratories for the determination of strontium, uranium, plutonium, americium and curium. To gather detailed information on the procedures in use, a questionnaire regarding various aspects of radionuclide determination was developed and distributed to all (sixteen) relevant laboratories in the Nordic countries. The response and the procedures used by each laboratory were then discussed between those who answered the questionnaire. This report summaries the findings and gives recommendation on suitable practice. (au)

  4. Radiochemical separation for determining of some trace elements in standard biological materials

    International Nuclear Information System (INIS)

    A radiochemical separation method has been developed to determine the elements W, Cd, Cr, U, Th e Co in three biological materials of botanic origin used as SRM's: Peach Leaves, Apples Leaves and the new proposed material Spinach. The aim was to obtain more information for these elements whose values are not yet determined or are given only as suggested values. The radiochemical procedure was based on chromatographic separation using resin Chelex 100 in H Ac 0.1 M-N H4 Ac 0.1 M at pH 4.8. All the experimental data e results obtained are described and compared with the literature values. (author). 10 refs, 4 tabs

  5. Exploring rapid radiochemical separations at the University of Tennessee Radiochemistry Center of Excellence

    International Nuclear Information System (INIS)

    The University of Tennessee formed its Radiochemistry Center of Excellence (RCoE) in 2013 with support from the U.S. National Nuclear Security Administration. One of the major thrusts of the RCoE is to develop deeper understanding of rapid methods for radiochemical separations that are relevant to both general radiochemical analyses as well as post-detonation nuclear forensics. Early work has included the development and demonstration of rapid separations of lanthanide elements in the gas phase, development of a gas-phase separation front-end for ICP-TOF-MS analysis, and the development of realistic analytical surrogates for post-detonation debris to support methods development. (author)

  6. Rapid radiochemical separation of short-lived radionuclides in neutron-activated samples

    International Nuclear Information System (INIS)

    Radiochemical separation procedures based on the removal of metal ions by columns of C18-bonded silica gel after selective complexation are examined and the simplicity of the method demonstrated by its application to determination of Mn, Cu and Zn in neutron-activated biological material from the following solutions (pH 0-10, sulphate concentration 0,18M and 1,44M SO4): 8-hydroxyquinoline (oxine), ammonium pyrrolidinedithiocarbamate (APDC), cupferron (CUP), 1-(2-pyridylazo)-2-naphthol (PAN), 1-(2'-thiazolylazo)-2-naphthol (TAN), 4-(2-pyridylazo) resorcinol (PAR), diethylammonium diethyldithiocarbamate (DDC), potassium ethyl xanthate (PEX), acetylacetone (AcAc) or thenoyltrifluoracetone (TTA). The method is rapid and reliable and readily adaptable in all radiochemical laboratories

  7. Standardization of rapid radiochemical separation method for 90Sr analysis in sediment

    International Nuclear Information System (INIS)

    In case of major nuclear accident leading to radioactivity release, 90Sr and 137Cs are the important long lived radionuclides of significance to the environment. In the present study, a rapid and reliable method using Solid Extraction Chromatography (SEC) was standardized for separation of Sr in soil samples for quick response to such emergency situations. For the purpose of standardization, the soil samples were spiked with known amount of stable Sr carrier and 90Sr activity. The present paper discuss the studies carried out on the column capacity of SEC (Sr-spec resin), radiochemical as well as gravimetric yield of the procedure standardized and the effect of repetitive use of Sr-spec resin on its exchange capacity. The overall radiochemical recovery of the procedure (> 60%) was comparable to the gravimetric recovery and was found to reduce with the repetitive use. (author)

  8. Radiochemical solar neutrino experiment using 81Br(nu, e-)81Kr

    International Nuclear Information System (INIS)

    Both geochemical and radiochemical experiments based on the interaction 81Br(nu,e-)81Kr to detect 7Be solar neutrinos have been suggested as a logical extension of the 37Cl experiment of Davis et al. The 81Br experiment, however, requires the development of a direct counter for the slowly decaying 81Kr. Progress toward such a detector based on Resonance Ionization Spectroscopy (RIS) is discussed

  9. Study of performance characteristics of a radiochemical method to determine uranium in biological samples

    International Nuclear Information System (INIS)

    In this paper is described a methodology to calculate detection limit (Ld), quantification level (Lq) and minimum detectable activity (MDA) in a radiochemical method for determination of uranium in urine samples. The concentration is measured by fluorimetry and alpha gross activity using liquid scintillation counting (LSC). The calculation of total propagated uncertainty on a spike sample is presented. Furthermore, the major sources of uncertainty and percentage contribution in both measurements are assessed. (author)

  10. Radiochemical separation of actinides for their determination in environmental samples and waste products

    Energy Technology Data Exchange (ETDEWEB)

    Gleisberg, B. [Nuclear Engineering and Analytics Rossendorf, Inc. (VKTA), Dresden (Germany)

    1997-03-01

    The determination of low level activities of actinides in environmental samples and waste products makes high demands on radiochemical separation methods. Artificial and natural actinides were analyzed in samples form the surrounding areas of NPP and of uranium mines, incorporation samples, solutions containing radioactive fuel, solutions and solids resutling from the process, and in wastes. The activities are measured by {alpha}-spectrometry and {gamma}-spectrometry. (DG)

  11. Radiochemical separation of carrier free 204,206Bi from α-irradiated thallium oxide target

    International Nuclear Information System (INIS)

    Carrier-free 204,206Bi produced by the α-activation of Tl2O3 target, was separated from the target matrix using two different radiochemical methods, (i) liquid-liquid extraction (LLX) with methyl isobutyl ketone (MIBK)-HCl system and (ii) solid-liquid extraction with inorganic ion exchanger, zirconium vanadate from HCl medium. The separation was found to be maximum around pH 2. (author)

  12. Radiochemical neutron activation analysis: the continuous need of this analysis mode

    Czech Academy of Sciences Publication Activity Database

    Kučera, Jan

    Como: Indico, 2012. s. 107-107. [NRC-8, EuCheMS International Conference on Nuclear and Radiochemistry. 16.09.2012-21.09.2012, Como] R&D Projects: GA ČR(CZ) GBP108/12/G108; GA MŠk LM2011019 Institutional support: RVO:61389005 Keywords : low-level element determination * radiochemical separation * reference matarials Subject RIV: CB - Analytical Chemistry, Separation http://indico.cern.ch/conferenceDisplay.py/abstractBook?confId=183405

  13. New radiochemical methods for determination of 237Np a 241Pu using extraction chromatography (Presentation)

    International Nuclear Information System (INIS)

    Thesis was focused on the development of a new methodology for the separation of anthropogenic transuranium radionuclides 237Np a 241Pu from different kinds of matrices. The analytical methods used in this study were based on extraction chromatography and were optimized according to the sample type. The proposed radiochemical procedure is a combination of two algorithms, which represent the separation of radionuclides by using extraction chromatographic sorbents TEVA resin and TRU resin supplied by Eichrom Technologies LLC. 239Np a 237Np were selectively captured on sorbent TEVA resin in oxidation state 4+. TRU resin was used for purification of plutonium fraction from interfering americium radionuclide. 242Pu and 239Np radionuclides as tracers have been used to monitor the radiochemical yields of separation. Before every radiochemical separation tracer radionuclide 239Np was obtained by separation from the parent radionuclide 243Am, which is in radioactive equilibrium to 239Np. The average yield of chemical separation was 69,3% for 239Np at 277 keV energy line and 65,9% at 228 keV energy line. The NPL AH-B08069 (2008) samples which consist of the mixture of alpha-radionuclides were used for the modification and optimization of separation method used for separation of Np and Pu in model samples. This method provided high radiochemical yields of 239,240Pu (95,0 ± 3,5)% and 237Np (87,9 ± 3,0)%.. Reliability of the method was verified by applying our modified separation procedures on reference materials IAEA-375 and IAEA-414 supplied by International Atomic Energy Agency. A good agreement between the results is obtained by this procedure and the certified values were found. Samples of contaminated soils from the area of Nuclear power plant A-1 Jaslovske Bohunice which is stored temporarily before disposal were analyzed using developed separation procedure. Specific activity of investigated radionuclides was determined in these samples. (author)

  14. New radiochemical methods for determination of 237Np a 241Pu using extraction chromatography

    International Nuclear Information System (INIS)

    Thesis was focused on the development of a new methodology for the separation of anthropogenic transuranium radionuclides 237Np a 241Pu from different kinds of matrices. The analytical methods used in this study were based on extraction chromatography and were optimized according to the sample type. The proposed radiochemical procedure is a combination of two algorithms, which represent the separation of radionuclides by using extraction chromatographic sorbents TEVA resin and TRU resin supplied by Eichrom Technologies LLC. 239Np a 237Np were selectively captured on sorbent TEVA resin in oxidation state 4+. TRU resin was used for purification of plutonium fraction from interfering americium radionuclide. 242Pu and 239Np radionuclides as tracers have been used to monitor the radiochemical yields of separation. Before every radiochemical separation tracer radionuclide 239Np was obtained by separation from the parent radionuclide 243Am, which is in radioactive equilibrium to 239Np. The average yield of chemical separation was 69,3% for 239Np at 277 keV energy line and 65,9% at 228 keV energy line. The NPL AH-B08069 (2008) samples which consist of the mixture of alpha-radionuclides were used for the modification and optimization of separation method used for separation of Np and Pu in model samples. This method provided high radiochemical yields of 239,240Pu (95,0 ± 3,5)% and 237Np (87,9 ± 3,0)%.. Reliability of the method was verified by applying our modified separation procedures on reference materials IAEA-375 and IAEA-414 supplied by International Atomic Energy Agency. A good agreement between the results is obtained by this procedure and the certified values were found. Samples of contaminated soils from the area of Nuclear power plant A-1 Jaslovske Bohunice which is stored temporarily before disposal were analyzed using developed separation procedure. Specific activity of investigated radionuclides was determined in these samples. (author)

  15. Radiochemical detection of dihydrodiol dehydrogenase: distribution of the indomethacin sensitive enzyme in rat tissues

    International Nuclear Information System (INIS)

    Dihydrodiol dehydrogenase catalyzes the NADP+ dependent oxidation of trans-dihydrodiols of polycyclic aromatic hydrocarbons (PAH) which are potent proximate carcinogens. The authors have developed a highly sensitive radiochemical assay for this enzyme in which the oxidation of trans-1,2-dihydroxy-3,5-cyclohexadiene, a model substrate for trans-dihydrodiol proximate carcinogens, is coupled to O-methylation catalyzed by catechol O-methyl transferase. Using S-adenosyl-(3H-methyl)-methionine as methyl donor at a specific activity of 0.1 nCi/pmol and extracting the product, 3H-o-methoxyphenol, the assay provides a 5000 fold increase in sensitivity over the existing spectrophotometric method. The radiochemical assay was validated by comparing the K/sub m/ and V/sub max/ values for rat liver cytosol with those derived spectrophotometrically. In both instances there was close agreement between values (K/sub m/ = 0.77 +/- 0.11 mM and V/sub max/ = 2.14 +/- 0.13 nmoles/min/mg protein determined radiochemically; K/sub m/ = 0.96 +/- 0.10 mM and V/sub max/ = 6.31 +/- 0.50 nmoles/min/mg protein determined spectrophotometrically). Using the radiochemical method, dihydrodiol dehydrogenase activity was detected in the following rat tissues: liver > lung > heart > small intestine > testis > seminal vesicle > bladder > prostate > spleen. Specific activities ranged between 0.944 and 0.016 nmoles/min/mg protein. In liver, lung, and testis, which are sites of PAH metabolism, the dehydrogenase is sensitive to inhibition by low μM concentrations of indomethacin, suggesting that this drug can prevent the detoxification of proximate carcinogens by this route

  16. Media effects on radiochemical corrosion at high-output gamma irradiation facilities

    International Nuclear Information System (INIS)

    Gamma irradiation of metals at high dose rate conditions may induce or accelerate a wide variety of electrochemical corrosion processes. Examination of failures encountered in irradiation facilities due to corrosion indicated that, above a threshold value for atmospheric humidity, the electrode reactions are chiefly controlled by the action of radiolytic products arising from the electrolyte during gamma irradiation. Thus, the nature of the corrosive medium provides the decisive variable factor influencing the overall effect of radiochemical corrosion. (author)

  17. Radiochemical determination and separation or total radium, 226Ra and 224Ra

    International Nuclear Information System (INIS)

    Radiochemical purification and separation of radium has been carried out and the determination of total radium solubilized in aqueous samples has been studied assuming that all the alpha emitters of the sample have their origin in the 226Ra and elements of its desintegration chain. Also, the activities of 22Ra and 226 Ra have been evaluated separately doing a measurement after the chemical separation of the radium and another one 10 days after. (Author) 9 refs

  18. Radiochemical separation of thorium from 18O induced reaction with natural uranium

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    A radiochemical procedure used to separate and purify trace concentra tion thorium produced in heavy ion reaction with uranium targets is presented. The procedure can rapidly yield thorium fraction suitable for gamma-ray spectroscopy studies. The resultant gamma-ray spectra showed that Th was isolated from a large number of elements produced in the reaction, and there were only a few contaminat ing activities of isotopes of Sc, Cd, In, etc. The decontamination factors for the main reaction products are given.

  19. Recent trends in the concept of specific activity: Impact on radiochemical and radiopharmaceutical producers

    International Nuclear Information System (INIS)

    In the radiochemical and radiopharmaceutical industry, the concepts and subsequent specification used for determining the purity of the radiopharmaceutical product are of concern to both the regulator and the producer. It is therefore of profound importance that these concepts such as specific radioactivity are used correctly and their meaning fully understood. Recent changes in the pharmacopoeias are evaluated and the implications thereof discussed. On the basis thereof suggestions are made for definitions, specifications and tests

  20. Protein binding studies with radiolabeled compounds containing radiochemical impurities. Equilibrium dialysis versus dialysis rate determination

    DEFF Research Database (Denmark)

    Honoré, B

    1987-01-01

    The influence of radiochemical impurities in dialysis experiments with high-affinity ligands is investigated. Albumin binding of labeled decanoate (97% pure) is studied by two dialysis techniques. It is shown that equilibrium dialysis is very sensitive to the presence of impurities resulting...... in erroneously low estimates of the binding affinity and in inconsistent results at varying albumin concentrations. Dialysis rate determination (R. Brodersen et al. (1982) Anal. Biochem. 121, 395-408) is less sensitive to impurities. Udgivelsesdato: 1987-Apr...

  1. Statistical analysis of radiochemical measurements of TRU radionuclides in REDC waste

    Energy Technology Data Exchange (ETDEWEB)

    Beauchamp, J.; Downing, D.; Chapman, J.; Fedorov, V.; Nguyen, L.; Parks, C.; Schultz, F.; Yong, L.

    1996-10-01

    This report summarizes results of the study on the isotopic ratios of transuranium elements in waste from the Radiochemical Engineering Development Center actinide-processing streams. The knowledge of the isotopic ratios when combined with results of nondestructive assays, in particular with results of Active-Passive Neutron Examination Assay and Gamma Active Segmented Passive Assay, may lead to significant increase in precision of the determination of TRU elements contained in ORNL generated waste streams.

  2. Radiochemical aspects of production and processing of radiometals for preparation of metalloradiopharmaceuticals

    OpenAIRE

    Zhernosekov, Konstantin P.

    2006-01-01

    Radiometals play an important role in nuclear medicine as involved in diagnostic or therapeutic agents. In the present work the radiochemical aspects of production and processing of very promising radiometals of the third group of the periodic table, namely radiogallium and radiolanthanides are investigated. The 68Ge/68Ga generator (68Ge, T½ = 270.8 d) provides a cyclotron-independent source of positron-emitting 68Ga (T½ = 68 min), which can be used for coordinative labelling. However, f...

  3. Production of 48V in a nuclear reactor via secondary tritons

    International Nuclear Information System (INIS)

    The production of 48V in a nuclear reactor, induced on titanium by tritons generated from the 6Li(n, t)4 He reaction, and eventually 7Li(n, n't)4He, is described. Samples of lithium titanate were irradiated for an irradiation cycle (120 h) in the RA-3 reactor, belonging to Ezeiza Atomic Centre. After a radiochemical separation, the characteristic radiations from 48V were identified in the gamma ray spectra of the vanadium fractions. (orig.)

  4. Standard test methods for chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of nuclear-grade plutonium metal

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2004-01-01

    1.1 These test methods cover procedures for the chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of nuclear-grade plutonium metal to determine compliance with specifications.

  5. A comparison of different radiochemical methods applicable for the determination of plutonium isotopes in urine via alpha spectrometry

    International Nuclear Information System (INIS)

    The aim of the present work was to compare the performance of four of the most widely adopted radiochemical procedures making use of different extraction methods for the determination of plutonium in urine samples via alpha spectrometry

  6. Radiochemical studies related to the development of new production routes of some diagnostic and therapeutic radionuclides

    International Nuclear Information System (INIS)

    Nuclear reaction cross section measurements were done in connection with the development of new production routes of the therapeutic and diagnostic radionuclides 32P, 71As, 72As, 73As, 74As, 82Sr, 90Y, 153Sm and 169Yb. Investigations on the production of n.c.a. 73Se using novel targetry were also performed. Integral cross sections were measured for the natS(n,p)32P, natZr(n,p)90Y and natEu(n,p)153Sm reactions using a 14 MeV d(Be) neutron field. The neutron spectrum was characterised using multiple foil activation and the code SULSA. Existing cross section data were validated within 10 - 15 %, thereby substantiating earlier evaluated and recommended excitation functions of the investigated reactions. It is inferred that for production of radionuclides via the (n,p) reaction, a fast neutron spectral source (e.g. spallation or fusion) would be better suited than a fission reactor. Proton and α-particle induced reactions were investigated in the high-mass area for the production of 153Sm and 169Yb via alternative routes. Measurements were done for the first time on the natNd(a,n)153Sm process over the energy range of 10 to 26.5 MeV and the possible production yield of 153Sm amounts to 2 GBq. The excitation function of the 169Tm(p,n)169Yb reaction was determined over the energy range from threshold to 45 MeV and compared with the results of nuclear model calculation based on the ALICE-IPPE code. A good agreement was found. The calculated possible production yields are lower than those via the conventional (n,γ) production route, but the produced 153Sm and 169Yb are in no-carrier-added form. Cross sections were also measured with regard to the production of 71As, 72As, 73As and 74As via the natGe(p,xn) processes and the results were compared with those from the ALICE-IPPE calculations. Possible yields were calculated together with potential impurities. The various processes contributing to the formation of 71As in the irradiation of natGe were analysed by performing

  7. Validation of micro-depletion method for CANDU® reactors for the core-tracking simulations

    International Nuclear Information System (INIS)

    The WIMS-AECL / DRAGON / RFSP reactor physics code set was used to simulate a core tracking scenario, which constitutes more than 400 full-power days. 102 vanadium detectors were used to record the local fluxes. These cases were run by using the micro-depletion method, embedded in the RFSP code. The calculated diffusion flux at the locations of the vanadium detectors were compared with the site measurement data. The average difference between the calculated flux and the measurement was about 2 %. (author)

  8. Production equipment for radiochemical separation of 202Tl from metallic thallium proton irradiated target

    International Nuclear Information System (INIS)

    Thallium-201 finds widely application in nuclear medicine cardio diagnostics and myocardial imaging. The production device is proposed for production of thallium radiopharmaceutical - 201TlCl by quantitative separation of thallium-201 from thallium target and lead (carrier and the contaminants- 200Pb and 203Pb). The proposed equipment is created, designed and constructed in the Radiochemical laboratory of the Institute for Nuclear Research and Nuclear Energy for experimenting and production of TI pharmaceuticals with defined physical-chemical and biological characteristics. The production equipment is supplied with all necessary technical devices, reaction vessels and communications for the chemical processes, ensuring radiochemical and radioisotope purity of thallium pharmaceuticals at preliminary set technological conditions and parameters. The technological schemes allow the production to be done in inert atmosphere in gas flow (argon or nitrogen) and ensures hermetic work in a specially designed for the purpose radiation protected box. The equipment ensures high reliability, high radiochemical yield, high purity, non-pyrogenity and sterility of 201Tl radiopharmaceuticals for medical diagnostics

  9. Separation and collection of iodine, sulfur, and phosphorous anion complexes for subsequent radiochemical analysis

    International Nuclear Information System (INIS)

    We developed a method to separate anion complexes of sulfur, iodine, and phosphorus to enable determination by radiochemical techniques. This method involves ion chromatographic separation of the anion complexes from other highly emitting radioactive species such as cesium-137 and strontium-90 which interfere with radiochemical analysis. We essentially use the ion chromatograph as a sample pretreatment method. The samples are injected onto a cation exchange column which allows the anions to pass through while retaining the positively charged species. These anions are collected in the column effluent and measured by nuclear counting methods. The method was developed to enable measurement of trace radionuclides in radioactive waste and in environmental samples. Trace radionuclides which are present in concentrations of only a few hundred disintegrations per minute per milliliter can be separated and then analyzed using liquid scintillation counting analysis. This paper establishes the separation and collection protocol, collection efficiencies for sulfur, iodine, and phosphorus anion standards, and overall efficiencies and detection limits for the separation and subsequent radiochemical analysis of iodine-129 from both environmental level and high salt waste samples

  10. Radiochemical aging of an epoxy network; Vieillissement radiochimique d'un reseau epoxyde

    Energy Technology Data Exchange (ETDEWEB)

    Devanne, Th

    2003-05-01

    This thesis is to give a better understanding of the radiochemical aging of a thermoset resin under gamma irradiation. The conditions of aging are gamma irradiation under air with a dose rate of 2 kGy/h at 120 C. The requested lifetime is four years, it means a dose of 70 MGy. The first step of this work was the choice of a resistive epoxy resin. This choice was made thanks to the literature data. The high glass transition temperature and the high amount of aromatic groups were the main criteria of the final choice. After this choice, thermal and mechanical properties were followed under thermal and radiochemical aging: i) under thermal aging, after 600 hours at 220 C, the glass transition temperature remained unchanged. But, from a mechanical point of view, properties at break dramatically decreased. This embrittlement was assigned to a critical oxidized layer. The thickness of this layer was estimated about 30 {mu}m. ii) the same kind of embrittlement was observed under radiochemical aging. Moreover, it appeared a decrease of the glass transition temperature when increasing the dose of irradiation. This indicates that the main degradation mechanism is chain scission under anaerobic atmosphere. We, then, proposed a mechanistic model associated with a kinetic model to predict the evolution of the glass transition temperature depending on the irradiation conditions. Parameters of the kinetic model were determined by solid NMR and ESR experiments. Comparison between experimental and calculated values at 120 C is satisfactory, a global good agreement was found. (author)

  11. Radio-UHPLC: A tool for rapidly determining the radiochemical purity of technetium-99m radiopharmaceuticals?

    International Nuclear Information System (INIS)

    Determining the radiochemical purity (RCP) of technetium-99m (99mTc) radiopharmaceuticals using the method described in the package insert is a time-consuming process, requiring particular attention in order to achieve accurate RCP results. The purpose of this study was to evaluate whether radio-ultra high performance liquid chromatography (radio-UHPLC) may be an alternative method for RCP testing of 99mTc-tetrofosmin, 99mTc-MAG3 and 99mTc-sestamibi. Results obtained using radio-UHPLC were in excellent agreement with the standard method, with total analysis time being reduced to less than 3 min. - Highlights: • Radiochemical purity of 3 technetium-99m radiopharmaceuticals was evaluated using radio-UHPLC. • Results obtained were in agreement with those obtained using the standard method. • Analysis time was less than 3 min using radio-UHPLC. • Radio-UHPLC could be proposed as an alternative technique for radiochemical purity determination

  12. Papers presented by A.E.C.L. to the International Conference of the Canadian Nuclear Association

    International Nuclear Information System (INIS)

    The International Conference of the Canadian Nuclear Association was held in Toronto, Ontario, Canada on May 25-27, 1964. There were six papers presented by Atomic Energy of Canada Limited. The titles were: I. Canada - A Nuclear Power Plant Supplier, by J.L. Gray; II. Nuclear Power Development in Canada and Other Countries, by W.B. Lewis; III. The Development and Some Applications of Cobalt-60 Irradiators, by R.F. Errington; IV. The Definition and Achievement of Development Targets for the Canadian Power Reactor Program, by A.J. Mooradian; V. Recent Applications of Tracers in the Physical Sciences in Canada, by R.H. Betts and J.A. Davies; and, VI. Economic Comparison of Oyster Creek, Nine Mile Point and CANDU-type Stations under Canadian Conditions, by G.A. Pon and R.L. Beck.

  13. Radiochemical quality control of sup(99m)Tc-labelled immunoglobulin G by immobilised protein A from staphylococcus aureus

    International Nuclear Information System (INIS)

    Large fractions of immunoglobulin G (IgG) from mammalian species show high affinities for protein A isolated from staphylococcus aureus. The radiochemical purity of sup(99m)Tc-labelled IgG was determined by means of protein A, covalently bound to sepharosis, by a column radiochromatographic technique and by a simpler and more rapid technique in vials. Different labelling methods produced different radiochemical purities. Limitations and applications of the testing systems are discussed. (orig.)

  14. CANDU 6 - the highly successful medium sized reactor

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeun Seung; Eom, Young Sam; Lee, Suk Hee; Ryu, Seung Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    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. Hydrogen problems related to reactor accidents

    International Nuclear Information System (INIS)

    At reactor accidents, the combustion of hydrogen causes pressure and temperature transients which pose supplementary loads in containment. In certain conditions, they could reach hazardous levels and impair the integrity of the containment and the operability of the safety systems. The mechanisms of chemical reactions specific for the hydrogen-oxygen system are presented. Conditions in which combustion can occur and the various combustion modes, including the transition to detonation are also described. The related safety aspects and mitigation methods are discussed. Examples for particular applications and safety approaches for various types of reactors, included those promoted for the advanced reactors are also given. Presentation of the experimental research completed at AECL-Research, Whiteshell Laboratory is given, where the multi-point ignition effects for constant volume and for vented combustion of dry hydrogen-air mixtures in various geometries have been investigated. Various aspects of modelling and simulation of hydrogen combustion are discussed. The adaptations and the new models implemented in the codes VENT and CONTAIN, aimed to widen the simulation capabilities of hydrogen combustion models are described. The capabilities and limitations of the modelling assumptions of these two codes are also evaluated. (EG) (11 tabs., 39 ills., 82 refs.)

  18. 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

  19. Fast reactors and nonproliferation

    International Nuclear Information System (INIS)

    1.Three aspects of nonproliferation relevant to nuclear power are: Pu buildup in NPP spent fuel cooling ponds (∼ 104 t in case of consumption of ∼ 107 t cheap uranium). Danger of illegal radiochemical extraction of Pu for weapons production; Pu extraction from NPP fuel at the plants available in nuclear countries, its burning along with weapon-grade Pu in NPP reactors or in special-purpose burners; increased hazard of nuclear weapons sprawl with breeders and closed fuel cycle technology spreading all over the world. 2.The latter is one of major obstacles to creation of large-scale nuclear power. 3.Nuclear power of the first stage using 235 U will be able to meet the demands of certain fuel-deficient countries and regions, replacing ∼ 5-10% of conventional fuels in the global consumption for a number of decades. 4.Fast reactors of the first generation and the currently employed fuel technology are far from exhausting their potential for solving economic problems and meeting the challenges of safety, radioactive waste and nonproliferation. Development of large-scale nuclear power will become an option accepted by society for solving energy problems in the following century, provided a breeder technology is elaborated and demonstrated in the next 15-20 years, which would comply with the totality of the following requirement: full internal Pu breeding deterministic elimination of severe accidents involving fuel damage and high radioactivity releases: fast runaway, loss of coolant, fires, steam and hydrogen explosions, etc.; reaching a balance between radioactive wastes disposed of and uranium mined in terms of radiation hazard; technology of closed fuel cycle preventing its use for Pu extraction and permitting physical protection from fuel thefts;economic competitiveness of nuclear power for most of countries and regions, i.e. primarily the cost of NPPs with fat reactors is to be below the cost of modern LWR plants, etc

  20. N Reactor

    Data.gov (United States)

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

  1. Preparation, radiochemical purity control and stability of 99mTc-mertiatide (Mag-3)

    International Nuclear Information System (INIS)

    Scintigraphic image analysis of 99mTc-mertiatide (Mag-3, mercaptoacetyltriglycine) clearance provides the determination of the blood flow, the tubular transit time and the excretion as well from both kidneys. Radiopharmaceutical routine recommends a radiochemical purity control before administration of the product to a patient. The main objective of this study is to develop a Mag-3 labeling procedure that fits better than the previous one in our daily routine production of radiopharmaceuticals. Increasing proportions of 99mTc-Mag-3 were measured during the heating and cooling steps of the Mag-3 labeling procedure. High performance liquid chromatography (HPLC) analysis was used to confirm the results of a rapid radiochemical quality control assay on standard instant thin-layer chromatography-silica gel (ITLC-SG) paper. The reconstitution time takes 20-25 minutes from the harvest of pertechnetate to a ready-for-use calibrated patient syringe. The HPLC profile of 99mTc-Mag-3 including its minor impurities remains unchanged for 24-48 hours after reconstitution. The application of a programmable Peltier-directed device for heating/cooling provides a better control of the temperature course. The procedure proposed fully meets the labeling criteria recommended by the supplier and can be performed with a minimum of attention within a time-span that we formerly needed for solely the radiochemical purity control assay. Moreover, 99mTc-Mag-3 prepared in this way seems to be considerably more stable than mentioned in the manufacturer's instructions. (author)

  2. Radiochemical analysis of 90Sr, 41Ca, 129I and 36Cl in waste samples

    International Nuclear Information System (INIS)

    Full text: The decommissioning of a nuclear facility requires estimating the total inventory of radioactivity in various materials and its variation with time, which has to be carried out by the determination of the radioactivity of various radionuclides presented in the materials. Of all materials in the nuclear facilities, graphite, concrete, and steel are the main low-medium radioactive waste due to their large volume. Besides the neutron activation products of components and impurity in the materials including 36Cl and 41Ca, some fission products, such as 90Sr, 99Tc, 129I, and 137Cs also exist in the materials due to the contamination of the leaked nuclear fuel. Of these radionuclides, the determination of gamma emitters is easier and can be directly carried out by gamma spectroscopy without any radiochemical separation. But the beta and alpha emitters including 3H, 14C, 36Cl, 41Ca, 55Fe, 63Ni, 90Sr, 9 and 129I and some transuranics, have to be determined by radiochemical analysis including a completely separation of individual radionuclides from matrix and other radionuclides before measurement by beta counting, alpha spectrometry or mass spectrometry. This work presents radiochemical analytical methods developed in our laboratory in the recent years for the determination of 36Cl, 41Ca, 90Sr and 129I in bio-shielding concrete, graphite and metals. for the decommissioning of nuclear facilities. Besides individual procedure for the purification of various radionuclides, a combined procedure is also developed and presented for the simultaneous determination of all four radionuclides from one sample. (author)

  3. Radiochemical determination of 210 Pb and 226Ra in petroleum sludges and scales

    International Nuclear Information System (INIS)

    The oil extraction and production, both onshore and offshore, can generate different types of residues, such as sludge, that is deposited in the water/oil separators, valves and storage tanks and scales, which form i the inner surface of ducts and equipment. Analyses already carried out through gamma spectrometry indicated the existence of high radioisotope concentration. However, radionuclides emitting low-energy gamma-rays, such as 210 Pb, are hardly detected by that technique. Consequently, there is a need to test alternative techniques to determine this and other radionuclides from the 238 U series. This work, therefore, focuses on the radiochemical determination of the concentration of 210Pb, and 226 Ra in samples of sludge and scale from the oil processing stations of the UN-SEAL, a PETROBRAS unit responsible for the exploration and production of petroleum in Sergipe and Alagoas. The sludge and scale samples went through a preliminary process of extraction of oil, in order to separate the solid phase, where the largest fraction of the radioactivity is concentrated. After oil removal, the samples were digested using alkaline fusion as an option for dissolution. Finally, their activity concentration was determined for the samples of sludge and scales, using and alternative radiochemical method, which is based on ionic exchange. The activity concentration found for 210Pb varied from 1,14 to 507,3 kBq kg-1. The values for 226Ra were higher, varying from 4,36 to 3.445 kBq kg-1. The results for 226Ra were then compared with the ones found for the same samples of sludge and scales using gamma spectrometry. The results of the comparison confirm the efficiency of the methodology used int hi work, that is, radiochemical determination by means of ionic exchange. (author)

  4. System comparative analysis of the most advanced pressured water reactors (PWR, WWER) and boiling water reactors (BWR) projects with the aim to choose the reactors for NPP construction in Kazakhstan

    International Nuclear Information System (INIS)

    Full text: The official decision on construction of a Nuclear Power Plant (NPP) in Kazakhstan has been accepted by the Kazakhstan government. The results on the choice of the power reactors projects of the NPP are given in the report. The choice has been carried out with the aim to develop recommendation on reactors of the NPP for construction in Kazakhstan. The choice of the reactors was based on the system comparative analysis of the most advanced power reactors projects using 15 criteria system of the nuclear, radiating and ecological safety and economic competitiveness. Following Pressurized Water Reactor (PWR, WWR) projects have been subjected to the system comparative analysis: 1) Large Sized Reactors (700 MW(el) and up): such as EPR, developed by Germany Siemens and France Framatome companies; CANDU-9, heavy-water reactor, developed by Atomic Energy of Canada Ltd (AECL); System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation, developed by Korea Power Engineering Company, Inc.; APWR, Japanese advanced reactor, developed by Japan Atomic Power Company, Japan, Mitsubishi Heavy Industries, Japan and Westinghouse Electric Company, USA; WWER-1000 (V-392) - development by Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor, development by Westinghouse, USA/Genesi, Italy. 2) Medium Sized Reactors (300 MWe - 700 MWe): AP-600, passive PWR, developed by the Westinghouse company; CANDU-6, heavy-water reactor, developed by Atomic Energy of Canada Ltd (AECL); An-tilde-600, passive PWR, developed by Nuclear Power Institute of China; WWER-640, Russian passive reactor, developed by 0KB ''Gidropress'' Experimental and Design Office, Russian Federation; MS-600, developed by Mitsubishi Company; KSNP-600, developed by Korea Power Engineering Company, Inc., South Korea. 3) Small Sized Reactors (a few MWe- 300 MWe): IRIS, reactor of IV generation, developed by the International Corporation of 13

  5. The EC6 - an enhanced mid-sized reactor with fuel cycle applications

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has two CANDU reactor products matched to markets: the Enhanced CANDU 6 (EC6), a modern 700 MWe-class design, and the Advanced CANDU Reactor (ACR-1000), a 1200 MWe-class Gen III+ design. Both reactor types are designed to meet both market-, and customer-driven needs; the ACR-1000 design is 90% complete and market-ready. The EC6 incorporates the CANDU 6's well-proven features, and adds enhancements that make the reactor even safer and easier to operate. The EC6 is the only mid-sized reactor with a proven pedigree that meets modern reactor expectations and regulatory standards. It is sized for smaller grids and also has outstanding fuel-cycle capability. The EC6 has domestic and offshore market pull and is the current focus of AECL's development program; market interest in the ACR-1000 is anticipated in the longer term. Some of the key features incorporated into the EC6 include upgrading containment and seismic capability to meet modern standards, shortening the overall project schedule, addressing obsolescence issues, optimizing maintenance outages and incorporating lessons learnt through feedback obtained from the operating plants. The EC6 utilizes modern computers and a distributed control system housed in an advanced control room which, along with automated testing and on-line diagnostics, make the plant easier and safer to operate, with minimal operator intervention. The first deployment of the EC6 is anticipated in Canada; off-shore markets are also being pursued. The EC6 burns natural uranium as standard. But, high neutron economy, on-power refuelling, a simple fuel bundle, and the fundamental CANDU fuel channel design provide the EC6 with the flexibility to accommodate a range of advanced fuels. (author)

  6. Determination of trace elements in bottled water in Greece by instrumental and radiochemical neutron activation analyses

    International Nuclear Information System (INIS)

    Four different bottled water brands sold in Greece in the winter of 2001-2002 were analyzed for a wide range of chemical elements, using neutron activation analysis (NAA). The elements Na and Br were determined instrumentally (INAA), whereas the other metals and trace elements radiochemically (RNAA). The results indicated that the mean level of all the elements determined in the samples were well within the European Union (EU) directive on drinking water and accomplish the drinking water standards of the World Health Organisation (WHO) as well as of the Food and Drug Administration (FDA). (author)

  7. Radiochemical methods for the determination of subnanogram amounts of cadmium in environmental samples

    Energy Technology Data Exchange (ETDEWEB)

    Shamaev, V.I.

    1986-02-01

    A radiochemical method has been developed for the determination of cadmium, based on an interpolation method with the addition of an interfering element (zinc). Using extraction by Dithizone in chloroform form from alkaline media it is possible to determine cadmium with a detection limit of about 2x10/sup -10/ M and quite high selectivity. Combination of the method with a preliminary substoichiometric concentration allows the detection limit to be reduced to about 2x10/sup -11/ M and the selectivity to be increased significantly. The method was used to determine cadmium in environmental samples.

  8. Advanced liquid and solid extraction procedures for ultratrace determination of rhenium by radiochemical neutron activation analysis

    Science.gov (United States)

    Mizera, J.; Kučera, J.; Řanda, Z.; Lučaníková, M.

    2006-01-01

    Radiochemical neutron activation analysis (RNAA) procedures for determination of Re at the ultratrace level based on use of liquid-liquid extraction (LLE) and extraction chromatography (EXC) have been developed. Two different LLE procedures were used depending on the way of sample decomposition using either 2-butanone or tetraphenylarsonium chloride in CHCl3. EXC employed new solid extractant materials prepared by incorporation of the liquid trioctyl-methyl-ammonium chloride into an inert polyacrylonitrile matrix. The RNAA procedures presented have been compared and applied for Re determination in several biological and environmental reference materials.

  9. Energy and Water Conservation Assessment of the Radiochemical Processing Laboratory (RPL) at Pacific Northwest National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, Stephanie R.; Koehler, Theresa M.; Boyd, Brian K.

    2014-05-31

    This report summarizes the results of an energy and water conservation assessment of the Radiochemical Processing Laboratory (RPL) at Pacific Northwest National Laboratory (PNNL). The assessment was performed in October 2013 by engineers from the PNNL Building Performance Team with the support of the dedicated RPL staff and several Facilities and Operations (F&O) department engineers. The assessment was completed for the Facilities and Operations (F&O) department at PNNL in support of the requirements within Section 432 of the Energy Independence and Security Act (EISA) of 2007.

  10. An alkaline kit formulation to obtain [99mTc]MAG3 in high radiochemical yields

    International Nuclear Information System (INIS)

    An alkaline kit formulation (ph 9) to obtain [99mTc]MAG3 with radiochemical purities over 98% has been developed, avoiding the addition of filtered air to the vial, the use of large amounts of 99mTc activity (i.e., 3.7 GBq) or the reconstitution of large volumes. The use of this radiopharmaceutical in mice showed a minimal accumulation in the hepatobiliary system (0.37 ± 0.3 % I.D., 1 h postinjection). However, in rabbits we always obtained good image quality. (author) 7 refs.; 1 fig.; 2 tabs

  11. New procedures of radiochemical neutron activation analysis for ultratrace determination of rhenium

    International Nuclear Information System (INIS)

    Radiochemical neutron activation analysis (RNAA) procedures for determination of Re at the ultratrace level based on use of liquid-liquid extraction (LLE) and extraction chromatography (EXC) have been developed. Two different LLE procedures were used depending on the way of sample decomposition using either 2-butanone or tetraphenylarsonium chloride in CHCl3. EXC employed new solid extractant materials prepared by incorporation of the liquid trialkyl-methylammonium chloride into an inert polyacrylonitrile matrix. The RNAA procedures presented were compared and applied to Re determination in various biological and environmental reference materials. (author)

  12. Advanced liquid and solid extraction procedures for ultratrace determination of rhenium by radiochemical neutron activation analysis

    International Nuclear Information System (INIS)

    Radiochemical neutron activation analysis (RNAA) procedures for determination of Re at the ultratrace level based on use of liquid-liquid extraction (LLE) and extraction chromatography (EXC) have been developed. Two different LLE procedures were used depending on the way of sample decomposition using either 2-butanone or tetraphenylarsonium chloride in CHCl3. EXC employed new solid extractant materials prepared by incorporation of the liquid trialkyl-methylammonium chloride into an inert polyacrylonitrile matrix. The RNAA procedures presented have been compared and applied for Re determination in several biological and environmental reference materials. (author)

  13. Determination of mercury and selenium in consumed food items in Libya using instrumental and radiochemical NAA

    International Nuclear Information System (INIS)

    Instrumental and radiochemical neutron activation analysis (INAA and RNAA, respectively) were used to analyze several consumed food items in Libya for the detection of low level concentrations of mercury and selenium. Selenium was determined using both short- and long-term irradiation, while mercury was determined in the long-term irradiation mode. At RNAA, after wet-ashing of samples in a microwave digestion unit, mercury was extracted with Ni(DDC)3/CHCl3, and selenium was precipitated in elemental form with ascorbic acid. For quality control, NIST reference materials were analyzed using the same procedures as for the food samples. The results of the analytical modes used were compared. (author)

  14. Distribution of platinum in patients treated with cisplatin determined by radiochemical neutron activation analysis

    DEFF Research Database (Denmark)

    Heydorn, K.; Rietz, B.; Krarup-Hansen, A.

    1998-01-01

    Cisplatin is used in a successful treatment of testicular cancer and some related conditions, but several toxic effects have been observed. Knowledge about the distribution of platinum in the human body after treatment with massive doses of cisplatin might provide clues to the origin of side...... effects, and a study was initiated to provide such information by the analysis of postmortem samples by our method of radiochemical neutron activation analysis (RNAA). Autopsy samples of kidney, liver, lung, muscle, and pancreas were taken with stainless steel scalpels together with samples of nerve...

  15. Conceptual Design for the Pilot-Scale Plutonium Oxide Processing Unit in the Radiochemical Processing Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Lumetta, Gregg J.; Meier, David E.; Tingey, Joel M.; Casella, Amanda J.; Delegard, Calvin H.; Edwards, Matthew K.; Jones, Susan A.; Rapko, Brian M.

    2014-08-05

    This report describes a conceptual design for a pilot-scale capability to produce plutonium oxide for use as exercise and reference materials, and for use in identifying and validating nuclear forensics signatures associated with plutonium production. This capability is referred to as the Pilot-scale Plutonium oxide Processing Unit (P3U), and it will be located in the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. The key unit operations are described, including plutonium dioxide (PuO2) dissolution, purification of the Pu by ion exchange, precipitation, and conversion to oxide by calcination.

  16. Design of the Laboratory-Scale Plutonium Oxide Processing Unit in the Radiochemical Processing Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Lumetta, Gregg J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Meier, David E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tingey, Joel M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Amanda J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Delegard, Calvin H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Edwards, Matthew K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Orton, Robert D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rapko, Brian M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smart, John E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-05-01

    This report describes a design for a laboratory-scale capability to produce plutonium oxide (PuO2) for use in identifying and validating nuclear forensics signatures associated with plutonium production, as well as for use as exercise and reference materials. This capability will be located in the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. The key unit operations are described, including PuO2 dissolution, purification of the Pu by ion exchange, precipitation, and re-conversion to PuO2 by calcination.

  17. A quantitative radiochemical study of ionic and molecular transport in bovine dental enamel

    International Nuclear Information System (INIS)

    A radiochemical method was developed to determine quantitatively and simultaneously the transport of up to four different compounds through dental enamel. The compounds chosen were [3H]-sorbitol, [14C]-glycerol, 36Cl- and 86Rb+. Effective diffusion coefficients of these compounds determined at 40C were considerably different for different specimens of enamel. Thus values for [3H]-sorbitol varied for 0.04 x 10-8 to 2.5 x 10-8 cm2s-1. Most enamel membranes showed an ion selective behaviour by which the cations were more mobile than the anions. A molecular sieve effect was observed for glycerol and sorbitol. (author)

  18. Sequential radiochemical separations from Alpine Wetland soils (Boreon, France) with emphasis on 90Sr measurement

    International Nuclear Information System (INIS)

    A radiochemical procedure to extract plutonium, americium and strontium from soils is presented. Strontium was separated from americium and plutonium fraction at the beginning of the method to increase the Sr recovery. The studied soils coming from an Alpine wetland site contain a big amount of iron which was eliminated by an oxalate precipitation before the column step. The hydroxide precipitation should be made by adding iron of known quantity to avoid interference. The procedure was validated by reference soils from IAEA. Plutonium-238, 239, 240, 241Am, 90Sr and 137Cs activities are given and some isotopic ratios are calculated in order to know the origin of the radionuclides. (author)

  19. Trace element evaluation of different varieties of chewing gum by radiochemical neutron activation analysis

    International Nuclear Information System (INIS)

    Extensive use of chewing gums, by children in particular, entails the evaluation of trace element contents in them. Radiochemical neutron activation analysis (RNAA) was successfully employed to determine the concentration of 35 trace elements (essential, toxic and nonessential) in eight different brands of chewing gum generally consumed in Rawalpindi/Islamabad area. Comparison of trace element data of our work with literature has been presented. None of the elements detected in the brands of chewing gum examined was found to be present at a level representing a substantial contribution to the total dietary intake of the element. (author)

  20. Design of the Laboratory-Scale Plutonium Oxide Processing Unit in the Radiochemical Processing Laboratory

    International Nuclear Information System (INIS)

    This report describes a design for a laboratory-scale capability to produce plutonium oxide (PuO2) for use in identifying and validating nuclear forensics signatures associated with plutonium production, as well as for use as exercise and reference materials. This capability will be located in the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. The key unit operations are described, including PuO2 dissolution, purification of the Pu by ion exchange, precipitation, and re-conversion to PuO2 by calcination.

  1. Rapid and accurate determination of radiochemical purity of sup(99m)Tc compounds

    International Nuclear Information System (INIS)

    The wide spread use of sup(99m)Tc-labelled radiopharmaceuticals and limitation of the short half-life of the isotope, is associated with an urgent need for a rapid, simple but accurate method for determining the radiochemical purity of the compound. A short paper chromatographic (KK) or thin layer chromatographic (KLT) method using 95% methanol or 0.9% saline solution as solvents, has solved the problem. With these methods, the amount of free sup(99m)Tc pertechnetate in a compound, can be determined in only a few minutes. These methods compare satisfactorily with lengtheir procedures. (author)

  2. 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)

  3. Advancing the CANDU reactor: From generation to generation

    International Nuclear Information System (INIS)

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

  4. Computer code qualification program for the Advanced CANDU Reactor

    International Nuclear Information System (INIS)

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

  5. Multi-element characterization of silicon nitride powders by instrumental and radiochemical neutron activation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Franek, M.; Krivan, V. (Ulm Univ. (Germany). Sektion Analytik und Hoechstreinigung)

    1992-07-15

    An optimized instrumental neutron activation analysis method was applied to the comprehensive trace characterization of good- and high- purity silicon nitride powders of different origins. Experimental modes are given for 55 elements leading to limits of detection below 1 ng g[sup -] [sup 1] for 28 elements, between 1 and 100 ng g[sup -1] for 19 elements and higher than 100 ng g[sup -1] for 8 elements. For the removal of the radionuclides [sup 140]La, [sup 182]Ta and [sup 187]W, which cause the major activity in certain types of materials, radiochemical procedures based in cation exchange from 2 M HCl and anion exchange from 2 M HF were developed. [sup 64]Cu was selectively extracted with dithizone from 10 M HF for counting the 511-keV line. By radiochemical neutron activation analysis, the limits of detection were improved by up to three orders of magnitude. Comparison with results obtained by inductively coupled plasma (ICP) atomic emission spectrometry and ICP mass spectrometry shows satisfactory agreement and demonstrates the advantages of neutron activation analysis especially when low elements contents are to be determined. (author). 30 refs.; 2 figs.; 6 tabs.

  6. Spectroscopic Online Monitoring for Process Control and Safeguarding of Radiochemical Fuel Reprocessing Streams - 13553

    International Nuclear Information System (INIS)

    There is a renewed interest worldwide to promote the use of nuclear power and close the nuclear fuel cycle. The long term successful use of nuclear power is critically dependent upon adequate and safe processing and disposition of the used nuclear fuel. Liquid-liquid extraction is a separation technique commonly employed for the processing of the dissolved spent nuclear fuel. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. This paper discusses application of absorption and vibrational spectroscopic techniques supplemented by physicochemical measurements for radiochemical process monitoring. In this context, our team experimentally assessed the potential of Raman and spectrophotometric techniques for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid and Pu(IV)/Np(V)/Nd(III), respectively, in solutions relevant to spent fuel reprocessing. Both techniques demonstrated robust performance in the repetitive batch measurements of each analyte in a wide concentration range using simulant and commercial dissolved spent fuel solutions. Static spectroscopic measurements served as training sets for the multivariate data analysis to obtain partial least squares predictive models, which were validated using on-line centrifugal contactor extraction tests. Satisfactory prediction of the analytes concentrations in these preliminary experiments warrants further development of the spectroscopy-based methods for radiochemical safeguards and process control. (authors)

  7. Spectroscopic online monitoring for process control and safeguarding of radiochemical streams

    International Nuclear Information System (INIS)

    This paper summarizes application of the absorption and vibrational spectroscopic techniques supplemented by physicochemical measurements for radiochemical process monitoring. In this context, our team experimentally assessed the potential of Raman and spectrophotometric techniques for online real-time monitoring of the U(VI)/nitrate ion/nitric acid and Pu(IV)/Np(V)/Nd(III), respectively, in solutions relevant to spent fuel reprocessing. These techniques demonstrate robust performance in the repetitive batch measurements of each analyte in a wide concentration range using simulant and commercial dissolved spent fuel solutions. Spectroscopic measurements served as training sets for the multivariate data analysis to obtain partial least squares predictive models, which were validated using on-line centrifugal contactor extraction tests. Satisfactory prediction of the analytes concentrations in these preliminary experiments warrants further development of the spectroscopy-based methods for radiochemical process control and safeguarding. Additionally, the ability to identify material intentionally diverted from a liquid-liquid extraction contactor system was successfully tested using on-line process monitoring as a means to detect the amount of material diverted. (authors)

  8. A direct reading on-line flowrate meter for use in radiochemical plant

    International Nuclear Information System (INIS)

    A device for measurement and remote direct reading display of the flowrates of streams in a radiochemical plant is described. The device is interposed in the measured stream and consists of a syphon pot with a specially developed attachment on the discharge line. Differential pressure switches are used to trigger a timer device at set levels in the pot and the time required for filling the pot during each cycle is measured and is used to compute and display the flowrate. The device is accurate and reliable and is simple to fabricate and install. It is maintenance-free since it has no moving parts. It is also suggested that a manometer with conductive contacts could be used in place of the d.p. switches. The background and various stages of development of the device are described. The operating data is tabulated and parameters required for plant applications are indicated in detail. A simple method to detect and correct for errors due to drift in d.p. switch setting is also outlined. Sketches of typical syphon pot, the schematic of the apparatus and suggested layout for application in radiochemical plant are also included. (author). 11 figures, 6 tables

  9. Standardization of the radiochemical method to determine Ra-226 in the Chillon River

    International Nuclear Information System (INIS)

    The present work shows the development of five radiochemical methods in order to determine Ra-226 in river waters for human consumption, through gamma spectrometry. These methods are used in order to obtain this radionuclide as Ba(Ra)S04. The method tested was found to be the most appropriate due to its lower cost, high radiochemical yield obtained (83,53 %) on average and the obtained results show smaller variability. This standardized method is applied to water samples of the Chillon River, outcoming an activity of 0,1526 Bq/L. Assuming inhabitants of Huarangal Valley consume 2 liters of this river water daily. The annual ingest is about 111,398 Bq on average, which constitutes the 1,58 % of the allowed radioactivity limit given by the regulation on sanitary protection against ionizing radiation. Furthermore physical chemical analysis in waters of the Chillon River were performed. The results found are within the allowed ranges considered by the World Health Organization (WHO), the Technical Standards National Institute (ITINTEC - PERU) and Law of Waters. Chillon River waters can be destined for the human consumption according to the physical-chemical analysis performed. The results obtained about radioactivity levels in these samples do not show radio-sanitarian hazards for inhabitants of Huarangal Valley

  10. Determination of selenium in human diets by radiochemical neutron activation analysis

    International Nuclear Information System (INIS)

    A post-irradiation radiochemical separation technique was tested for the determination of selenium levels in diet samples, collected by using a duplicate portion technique, from both rural and urban population groups in Turkey. The technique involved sample irradiation, acid digestion, selective distillation, precipitation and filtration steps. During the separations it was possible to determine the yield of each sample using a stable selenium carrier. An average chemical yield of 71 ± 3% was obtained for the radiochemical neutron activation analysis. For samples from urban and rural regions, the average selenium concentrations obtained were 0.14 ± 0.04 and 0.07 ± 0.02 mg kg-1, respectively. It was also possible to determine daily dietary selenium intakes, which were found to be 81 ± 41 μg and 23 ± 11μg for the urban and rural groups, respectively. Although daily selenium intakes were found for a small number of subjects in this study, the separation technique developed can be used for determination of the selenium status in larger population groups. (author)

  11. Very accurate determination of trace amounts of selenium in biological materials by Radiochemical Neutron Activation Analysis

    International Nuclear Information System (INIS)

    Selenium is both a toxic and an essential trace element for humans and animals. The purpose of this work was to elaborate a very accurate (definitive) method for the determination of selenium traces in different types of biological materials. The method is based on a combination of neutron activation and quantitative and very selective radiochemical separation of selenium by ion-exchange and extraction chromatography, followed by gamma-spectrometric measurement of 75Se. Three amines: 2,3-diaminonaphtalene, 3,3'-diaminobenzidine and 4-nitro-phenyldiamine supported on Bio Beads SM-2 or Amberlite XAD-4 were chosen to batch experiments. Using 3,3'-diaminobenzidine tracer experiments were carried out with the unirradiated biological samples. They have proved that the whole radiochemical separation procedure is quantitative. Gamma-ray spectrum of the selenium fraction practically did not show any other activities except background peaks. The obtained results demonstrate good agreement of results obtained by our new '' definitive '' method for the determination of selenium with the certified values

  12. Anion effect on radiochemical stability of room-temperature ionic liquids under gamma irradiation

    International Nuclear Information System (INIS)

    Radiochemical stability of imidazolium-based ionic liquids constituted of the BuMeIm+ cation and associated with four commonly used anions (X-: Tf2N-, TfO-, PF6- and BF4-) has been investigated under gamma irradiation for high irradiation doses (up to 2.0 MGy). The anion effect has been examined by quantifying the radiolytic yields of disappearance for cation and anions and by identifying corresponding radiolysis products with several analytical techniques. On the one hand, a large number of radiolysis products are formed throughout the irradiation in ionic liquid solutions, resulting from reactions of primary generated species of cation and anion by indirect radiolysis. Primary generated species can react together throughout the irradiation by indirect radiolysis to form numerous radiolysis products in small quantities, indicating that several complex degradation pathways are involved for these radiation doses. This degradation pattern has been confirmed by identification of numerous gaseous radiolytic products. On the other hand, quantitative studies show that radiochemical stabilities of ionic liquids are in the same range of values as systems envisioned in nuclear fuel reprocessing with relatively low hydrogen yields. Indeed, this present work emphasizes the suitability of ionic liquids for applications in the nuclear fuel cycle. (authors)

  13. RAPID AUTOMATED RADIOCHEMICAL ANALYZER FOR DETERMINATION OF TARGETED RADIONUCLIDES IN NUCLEAR PROCESS STREAMS

    International Nuclear Information System (INIS)

    Some industrial process-scale plants require the monitoring of specific radionuclides as an indication of the composition of their feed streams or as indicators of plant performance. In this process environment, radiochemical measurements must be fast, accurate, and reliable. Manual sampling, sample preparation, and analysis of process fluids are highly precise and accurate, but tend to be expensive and slow. Scientists at Pacific Northwest National Laboratory (PNNL) have assembled and characterized a fully automated prototype Process Monitor instrument which was originally designed to rapidly measure Tc-99 in the effluent streams of the Waste Treatment Plant at Hanford, WA. The system is capable of a variety of tasks: extraction of a precise volume of sample, sample digestion/analyte redox adjustment, column-based chemical separations, flow-through radiochemical detection and data analysis/reporting. The system is compact, its components are fluidically inter-linked, and analytical results can be immediately calculated and electronically reported. It is capable of performing a complete analytical cycle in less than 15 minutes. The system is highly modular and can be adapted to a variety of sample types and analytical requirements. It exemplifies how automation could be integrated into reprocessing facilities to support international nuclear safeguards needs

  14. Trace characterization of high-purity nickel by instrumental and radiochemical neutron activation analysis

    International Nuclear Information System (INIS)

    The high activity of the radionuclides 65Ni (t1/2=2.52 h) and 58Co (t1/2=70.8 d) imposes severe limitations on the performance of direct instrumental neutron activation analysis of nickel. The extent of the interference of the 58Co depends on the ratio of the fluxes of the fast and thermal neutrons. A method of selective removal of cobalt, based on extraction with β-nitroso-α-naphthol has been developed for the purpose of radiochemical NAA. Separation yields have been determined for 36 elements. The detection limits obtainable with both the instrumental and the radiochemical method are -4 μg/g for the elements Au, Eu, Ir, La, Sc and Sm, between 10-4 and 10-3 μg/g for Cr, Cs, Hf, Hg, Lu, Re, Sb, Ta, Th, Tm and Yb, between 10-3 and 10-2 μg/g for As, Ag, Br, Ce, Ga, Na, Ru, Se, W and Zn, and in the range 0.01-1 μg/g for Ba, Cd, Co, Fe, In, K, Mo, Nd, Pd, Rb, Sn, U and Zr. (orig.)

  15. Conceptual design for comprehensive automation in radiochemical analysis of bioassay samples

    International Nuclear Information System (INIS)

    Bioassay Laboratory of Health Physics Division is entrusted with the task of carrying out the bioassay monitoring of occupational workers from various plants/divisions of BARC for various radionuclides like Pu, U, Th, 90Sr, 3H etc. On the average about 1400-1500 analyses are performed on 700-800 urine samples collected annually from radiation workers. The workload has increased by 1.5 to 2.0 times in recent past and is expected to increase further due to expanding nuclear programmes of the Department. Therefore, it was planned to carry out automation in various stages of bioassay sample handling, processing and analysis under the XI plan programme. Automation work in Bioassay Lab. is planned to be taken-up in three stages namely, automation in initial processing of i) urine samples, ii) fecal samples and iii) automation in radiochemical analysis of bioassay samples. In the initial phase, automation in radiochemical analysis of bioassay samples has been taken up

  16. Determination of chromium, cobalt and nickel in tissue samples by radiochemical activation analysis

    International Nuclear Information System (INIS)

    A radiochemical neutron activation analysis method for the determination of chromium, cobalt and nickel in tissue samples. A radiochemical neutron activation analysis method for the determination of chromium, cobalt and nickel has been developed. The destruction device used consisted of a combined wet-ashing-distillation and ion-exchange system. Six samples could be treated at the same time. The samples were wet-ashed with H*L2SO*L4-H*L2O*L2 mixture. Volatile elements were distilled as bromide compounds with HBr*H-. The distillation residue in 8M HCl was passed through hydrated antimony pentoxide (HAP) in order to remove disturbing *H2*H4Na-activity and through a Dowex 2 x 8 column so as to retain *H6*H0Co (formed from *H5*H8Ni). Chromium was elutriated from the column and precipitated as Cr(OH)*L3 for the removal of disturbing *H3*H2P-activity. The standards and samples were treated in a similar manner each so that the yield determination is not necessarily needed. The yields by tracer experiments were (43 +- 5) % for Cr, (93 +- 4) % for Co and (88 +- 14) % for Ni. The precision and accuracy of the method were studied by using reference materials of the National Bureau of Standards (NBS) and the International Atomic Energy Agency (IAEA)

  17. Radiochemical quality control of kits labelled with Tc-99m produced at IPEN-CNEN/SP

    International Nuclear Information System (INIS)

    The radiopharmaceuticals labelled with Tc-99m are routinely used in Nuclear Medicine Laboratories. A large number of these employ tin (II) reagents to reduce Tc (pertechnetate-VII) to a lower valence state thereby making it more able to complex forming reactions. The miniaturized chromatography system of Tc-99m labelled compounds using Whatman 3MM (8 x 1 cm) as a support and 30% NaC1: 0,9% 'NaC1: 85% MeOH and acetone as a solvent permits to assay the radiochemical purity in a few minutes after preparation. In addition this method introduced in routine work not only determines Tc-99m (pertechnetate) but also determines reduced Tc - 99m unbound to the radiopharmaceuticals (hydrolyzed reduced Tc-99m). The lyophilized kits for labelling with Tc-99m produced at IPEN-CNEN/SP are: MDP, DTPA, HSA, GHA, HIDA, Pyro, MAA, MIAA, Sulfur Colloid, Dextran-500, Sn.Cit. and Phytate. Radiochemical quality control of these kits were performed at the first day of preparation and during 12 months for determining' their validity for use. All preparation showed high yield of labelling (95-99%) during this period of time. (author)

  18. Radiochemical determination of 237NP in soil samples contaminated with weapon grade plutonium

    Science.gov (United States)

    Antón, M. P.; Espinosa, A.; Aragón, A.

    2006-01-01

    The Palomares terrestrial ecosystem (Spain) constitutes a natural laboratory to study transuranics. This scenario is partially contaminated with weapon-grade plutonium since the burnout and fragmentation of two thermonuclear bombs accidentally dropped in 1966. While performing radiometric measurements in the field, the possible presence of 237Np was observed through its 29 keV gamma emission. To accomplish a detailed characterization of the source term in the contaminated area using the isotopic ratios Pu-Am-Np, the radiochemical isolation and quantification by alpha spectrometry of 237Np was initiated. The selected radiochemical procedure involves separation of Np from Am, U and Pu with ionic resins, given that in soil samples from Palomares 239+240Pu levels are several orders of magnitude higher than 237Np. Then neptunium is isolated using TEVA organic resins. After electrodeposition, quantification is performed by alpha spectrometry. Different tests were done with blank solutions spiked with 236Pu and 237Np, solutions resulting from the total dissolution of radioactive particles and soil samples. Results indicate that the optimal sequential radionuclide separation order is Pu-Np, with decontamination percentages obtained with the ionic resins ranging from 98% to 100%. Also, the addition of NaNO2 has proved to be necessary, acting as a stabilizer of Pu-Np valences.

  19. Radiochemical separation and quality assessment for the 68Zn target based 64Cu radioisotope production

    International Nuclear Information System (INIS)

    The radiochemical separation of the different radionuclides (64Cu, 67Cu, 67Ga, 66Ga, 56Ni, 57Ni, 55Co, 56Co, 57Co, 65Zn, 196Au ) induced in the Ni supported Cu substrate - 68Zn target system, which was bombarded with the 29.0 MeV proton beam, was performed by ion-exchange chromatography using successive isocratic and/or concentration gradient elution techniques. The overlapped gamma-ray spectrum analysis method was developed to assess the 67Ga and 67Cu content in the 64Cu product and even in the post-67Ga production 68Zn target solution without the support of radiochemical separation. This method was used for the assessment of 64+67Cu radioisotope separation from 67Ga , the quality control of 64Cu product and the determination of the 68Zn (p,2p)67Cu reaction yield. The improvement in the targetry and the optimization of proton beam energy for the 68Zn target based 64Cu and 67Ga production were proposed based on the stopping power and range of the incident proton and on the excitation functions, reaction yields and different radionuclides induced in the target system. (author)

  20. Radiochemical procedures for analysis of Pu, Am, Cs and Sr in water, soil, sediments and biota samples

    International Nuclear Information System (INIS)

    The Environmental Radioactivity Analysis Laboratory (ERAL) was established as an analytical facility. The primary function of ERAL is to provide fast and accurate radiological data of environmental samples. Over the years, many radiochemical procedures have been developed by the staffs of ERAL. As result, we have found that our procedures exist in many different formats and in many different notebooks, documents and files. Therefore, in order to provide for more complete and orderly documentation of the radiochemical procedures that are being used by ERAL, we have decided to standardize the format and compile them into a series of reports. This first report covers procedures we have developed and are using for the radiochemical analysis of Pu, Am, Cs, and Sr in various matrices. Additional analytical procedures and/or revisions for other elements will be reported as they become available through continuation of these compilation efforts

  1. The use of cuprous iodide as a precipitation matrix in the radiochemical determination of 131I in milk

    International Nuclear Information System (INIS)

    As a result of the implementation of the As Low As is Reasonably Achievable philosophy to the nuclear power industry, recent U.S. Nuclear Regulatory Commission requirements have prompted high sensitivity radiochemical analysis for the measurement of 131I in milk. The most recognized and commonly employed technique incorporates costly palladium iodide as the final precipitate in the radiochemical purification of the iodine chemical species. The procedure presented in this paper outlines the many advantages of using cuprous iodide as the final precipitate. These include lower cost per analysis, consistent recoveries, better precipitate matrix and good self absorption characteristics. Typical lower limit of detection values and operating characteristics obtained for high sensitivity β-γ analysis as well as gas proportional counting and a comparison of radiochemical and Ge(Li) spectrometric results for environmental samples collected during a recent Chinese weapons fallout incident are presented. (author)

  2. Radiochemical studies on the neutron- and proton-induced 7Be emission at energies up to 100 MeV

    International Nuclear Information System (INIS)

    Cross sections for 7Be emission in neutron and proton induced reaction on medium and heavy mass nuclei were measured up to 100 MeV using the activation method, radiochemical separations and γ-ray spectroscopy. For this purpose chemical separations were developed and modified to optain 7Be in a radiochemically pure form. In this work the excitation functions of (p,7Be) reactions on the target nuclei V, Nb, Au and Bi could be measured radiochemically for the first time in the energy range of 35 to 100 MeV. The cross sections for gold amount to a few μb, for vanadium to several tens of μb and for niobium to several hundred μb. (orig./HSI)

  3. Adaptation of WWR-K reactor for irradiation work at low power

    International Nuclear Information System (INIS)

    According to INAA in Sept 2000 there were 288 research reactors in operation in the world. 67 % of these reactors were operating in power mode below 1.1 MW. The profile of an available scientific and technological program depends to a great extent on reactor power. High-power reactors are used mainly for isotope production and material testing, specially for needs of power generation programs. Low-power reactors are used mainly for education and for irradiation for purposes of activation analysis. TRIGA, Slowpoke and the Chinese Mini Reactor are typical instruments in this class. High neutron flux can also be utilized for activation analysis. In this case, samples for irradiation should be tightly sealed in aluminium or in quartz, and organic samples can be irradiated for a relatively short time only. Another important issue is that reactor operation on a high power level is connected with higher expenses for fuel, for electric power, for coolant and for general maintenance. Many reactor centers use to charge prices like USD 200 to 400 for 1 day irradiation of a single container 10-12 cm high and 2.5 to 5 cm in diameter. For many users this price is prohibitive. In a low-power reactor all the expenses are smaller and samples of organic origin can be much larger. The WWR-K is 6 MW light water research reactor with the maximum neutron flux 1.4·1014 n·cm-2·s-1 and, correspondingly, rather high operation expenses. However, it can be operated at lower power levels, with much less electric power consumption for cooling systems. To check a possibility of use of the WWR-K as a low-power reactor, to decrease expenses for activation analysis, three experimental runs were performed on 200 k W power. Samples were irradiated in a specially built aluminium container 6 cm in diameter and 60 cm high. Inside the aluminium container there were located 6·120 ml bottles (jars). The total volume available in one irradiation was 720 ml. Analysis of economical and technological

  4. Contribution of the TRIGA - INR Pitesti reactor to implementation of National Program for Nuclear Power

    International Nuclear Information System (INIS)

    The TRIGA reactor of INR Pitesti, designed for nuclear fuel and structural materials testing, was commissioned in 1979. It has two reactor cores completely independent that share the same pool, which are practically two distinct nuclear reactors, a 14 MW steady-state unit and a second pulse reactor, working at 20,000 MW/pulse. The last unit may also be operated in a steady state regime of 500 KW power. Being of pool type it allows easy handling of the irradiation devices. The two TRIGA reactors are described. These reactors are used mainly for irradiation testing, particularly, for Cernavoda NPP fuel elements and production of radioisotopes as, for instance, Ir-131, Mo-99 for non-destructive industrial analyses and Co-60 for cobalt therapy. Also, programs for experimental physical research were developed as for instance crystallographic studies by means of neutron diffraction, prompt gamma spectrometry for isotopic composition of Ga poison at Cernavoda NPP and neutron activation analyses. A program RERTR is now undergoing for converting the highly enriched-fuel research reactors into slightly enriched-fuel reactors. This project is developed in the frame of a cooperation with IAEA-Vienna and DOE - USA. Also a program of power cycling testing for the study of CANDU fuel power followup in collaboration with AECL - Canada is currently implemented. Several research programs were established aiming at testing slightly enriched fuel for Cernavoda NPP, testing of CANDU fuel in LOCA accident conditions, preparation of radiopharmaceuticals, etc

  5. Determination of the chemical and radiochemical purity and specific radioactivity of [18F]FDG by HPLC

    International Nuclear Information System (INIS)

    High performance liquid chromatography (HPLC) in combination with the radioactivity detection is the best control method for the radiochemical purity of [18F]FDG. An anion exchange separation mechanism allows isocratic separation of carbohydrates. Using a strong basic eluent, the weakly acid carbohydrates form anions and are therefore retained on the anion exchange resin. The chemical and radiochemical purity and specific radioactivity can be determined simultaneously by including in the chromatographic system a mass detector sensitive, enough for quantitative determination of the product species. (orig.)

  6. HPLC-ICP-MS compared with radiochemical detection for metabolite profiling of H-3-bromohexine in rat urine and faeces

    DEFF Research Database (Denmark)

    Jensen, B.P.; Gammelgaard, B.; Hansen, S.H.; Andersen, J.V.

    2005-01-01

    H-3-Bromohexine was dosed to rats as a model compound to allow comparison of HPLC-ICP-MS detection on bromine to radiochemical detection in an in vivo drug metabolism study. Metabolite profiles were obtained in urine and faeces extracts. No influence of the methanol gradient on the bromine response...... was observed in the range of 18 - 75% methanol. The sensitivity obtained by HPLC- ICP-MS was almost two orders of magnitude better than on-line H-3 radiochemical detection. For ICP- MS, the limit of detection was calculated to be 69 nM Br ( injection volume 100 mu l), corresponding to an absolute...

  7. Stability of Y-90 Zevalin: Radiochemical purity evaluation using instant thin layer and size exclusion high performance liquid chromatography

    International Nuclear Information System (INIS)

    Aim: The purpose of the study was to evaluate the stability of typical Y-90 Zevalin (IDEC Pharmaceuticals Corp) patient doses, either maintained at room temperature or refrigerated, using the manufacturer's recommended instant thin layer chromatography procedure and confirming the results using size exclusion high performance liquid chromatography (HPLC) to evaluate radiochemical purity. Material and Methods: Following radiolabeling of Y-90 Zevalin, two patient doses were withdrawn into a 10-ml syringe. One patient dose, consisting of 41.2 mCi Y-90 Zevalin in 10 ml, was refrigerated. The other patient dose, consisting of 31.2 mCi Y-90 Zevalin in 7.3 ml, was maintained at room temperature. At selected time intervals after formulation, ranging from 0.5 to 49 hrs, radiochemical purity evaluations were performed using instant thin layer chromatography (ITLC-SG) with normal saline and size exclusion HPLC using a TSKgel G3000SW molecular sizing column. For each time interval, five separate samples were analyzed and the data statistically summarized. Results: Following initial radiolabeling, the radiochemical purity of all preparations evaluated was greater than 95%, as demonstrated by both chromatographic methods. At 24 hours post radiolabeling, the mean radiochemical purity of Y-90 Zevalin, refrigerated or maintained at room temperature, was 95.4% ± 0.3% (s.d.) and 86.3% ± 1.2% (s.d), respectively using instant thin layer chromatography. At 48 hours post radiolabeling, the mean radiochemical purity of Y-90 Zevalin, refrigerated or maintained at room temperature, was 91.0% ±± 0.8% (s.d.) and 86.0% ± 2.0% (s.d.), respectively using instant thin layer chromatography. In general, size exclusion HPLC confirmed the chromatographic results. With increased time post radiolabeling, an increase in radiolabeled low molecular weight components (Y-90 DTPA) and an increase in radiolabeled high molecular weight components and/or aggregates was observed. These radiochemical

  8. Present and future activities of TRIGA RC-1 Reactor

    International Nuclear Information System (INIS)

    A summary of reactor activities is presented and discussed. The RC-1 reactor is used by ENEA's laboratories, research institutes and national industries for different aims: research, analysis materials behaviour under neutron flux, etc. To satisfy the requests increase it is important to signalize: - the realization of a new radiochemical laboratory for radioisotopes production, to be used in a medical and/or diagnostic field in general; - the realization of a tritium handling laboratory, to study tritium solubility, release and diffusion in different material (particularly in ceramic breeder as lithium aluminate) to support Italian programs on fusion technology; - a research activity on the reactors computerized control by a console of advanced conception. The aim of this activity is the development of an ergonomic control room that could be a reference point for the planning of the power reactor control rooms

  9. A radio-high-performance liquid chromatography dual-flow cell gamma-detection system for on-line radiochemical purity and labeling efficiency determination

    DEFF Research Database (Denmark)

    Lindegren, S; Jensen, H; Jacobsson, L

    2014-01-01

    In this study, a method of determining radiochemical yield and radiochemical purity using radio-HPLC detection employing a dual-flow-cell system is evaluated. The dual-flow cell, consisting of a reference cell and an analytical cell, was constructed from two PEEK capillary coils to fit into the w...

  10. To question of NPP power reactor choice for Kazakhstan

    International Nuclear Information System (INIS)

    in the regions of Republic of Kazakhstan occurs. Southern and western regions import electric power and capacity because of undeveloped circuit of networks. Moreover, power intensity of an industrial-agrarian complex of the country is limited transmission capacity of lines is insufficient; plenty of small consumers are removed from power supply lines. Thus, nuclear stations of medium and low power are the most acceptable for construction in Kazakhstan. Recommendations for the choice of maximum safe, reliable and economically competitive reactors for Kazakhstan have been made in result of the carried out projects' comparison of the power reactors according to 15 criteria of safety and economic competitiveness, with respect to condition and perspectives of Kazakhstan power complex development: Recommended power reactors of medium capacity: - P-600 - passive PWR, developed of the Westinghouse company, USA; - CANDU-6, developed by Atomic Energy of Canada, Limited (AECL), Canada. - MS-600 - Mitsubishi Company, Japan. Recommended reactors of low power: - IRIS - reactor of IV generation developed by the international corporation of 13 organizations from 7 countries; - NPP 'UNITERM' - development NIKIET, Moscow, Russia; - MRX - the Project of sea reactor MRX for civil applications, is developed by the Japanese Research Institute of Atomic Energy (JAERI). The most important advantages of recommended medium and low power reactors are given

  11. To question of NPP power reactor choice for Kazakhstan

    International Nuclear Information System (INIS)

    regions of Republic of Kazakhstan occurs. Southern and western regions import electric power and capacity because of undeveloped circuit of networks. Moreover, power intensity of an industrial-agrarian complex of the country is limited transmission capacity of lines is insufficient; plenty of small consumers are removed from power supply lines. Thus, nuclear stations of medium and low power are the most acceptable for construction in Kazakhstan. Recommendations for the choice of maximum safe, reliable and economically competitive reactors for Kazakhstan have been made in result of the carried out projects' comparison of the power reactors according to 15 criteria of safety and economic competitiveness, with respect to condition and perspectives of Kazakhstan power complex development: Recommended power reactors of medium capacity: - P-600 - passive PWR, developed of the Westinghouse company, USA; - CANDU-6, developed by Atomic Energy of Canada, Limited (AECL), Canada. - MS-600 - Mitsubishi Company, Japan. Recommended reactors of low power: - IRIS - reactor of IV generation developed by the international corporation of 13 organizations from 7 countries; - NPP 'UNITERM' - development NIKIET, Moscow, Russia; - MRX - the Project of sea reactor MRX for civil applications, is developed by the Japanese Research Institute of Atomic Energy (JAERI). The most important advantages of recommended medium and low power reactors are given

  12. 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

  13. Plutonium Consumption Program, CANDU Reactor Project final report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-31

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

  14. Radiochemical determination of Np-237 in soil samples contaminated with weapon grade plutonium

    International Nuclear Information System (INIS)

    The Palomares terrestrial ecosystem (Andalusia, southwestern Spain) is known to constitute a natural laboratory to study the distribution, behaviour and migration of certain actinides, such as plutonium, americium and neptunium. This scenario is partially contaminated with weapon grade plutonium since the burn-out and fragmentation of two of the four thermonuclear bombs accidentally dropped by a B-52 from the USA Air Force back in 1966. While performing radiometric measurements on the field, with the goal of gathering information about the surface contamination levels, the possible presence of 237Np was observed through its 29 keV gamma emission. To accomplish a more detailed characterization of the source term in the contaminated area using the isotopic ratios Pu-Am-Np, the radiochemical isolation and later quantification by alpha spectrometry of 237Np was initiated. As a first approach, a bibliographic study was undertaken, considering different radiochemical methods applied to a variety of samples such as soils, sediments, sea water, lichens, etc. The radiochemical procedure selected in our laboratory involves separation of neptunium from americium, uranium and plutonium with ionic resin (AG 1x2), given that in soil samples from Palomares 239+240Pu levels are several orders of magnitude higher than 237Np. Then neptunium is isolated using TEVA organic resin. After electro-deposition, quantification is performed by high resolution alpha spectrometry. Different analyses have been performed with blank solutions spiked with 236Pu, 237Np and 244Cm, solutions resulting from the total dissolution of isolated radioactive particles and soil samples. The lack of an appropriate tracer in our lab led to the determination of an average percentage of Np recovery using a certified solution of 237Np. Decontamination percentages obtained during the Pu-Np separation ranged from 98 % to 100 %. Some tests to investigate the effect of the addition or absence of NaNO2 (responsible for

  15. Reactor Physics

    International Nuclear Information System (INIS)

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

  16. Reactor Physics

    International Nuclear Information System (INIS)

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

  17. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

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

  18. 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

  19. 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)

  20. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp