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Sample records for aecl

  1. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-06-01

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

  7. Update on use of AECL's MACSTOR module at CANDU 6 stations

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, R.; Moussalam, G.; Kachef, I. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada)

    2006-07-01

    AECL has contributed to the technology development and implementation of dry spent fuel management facilities in Canada and internationally over the last three decades. During that period, AECL has designed a number of concrete canister models and the MACSTOR module; a medium size air-cooled vault. AECL's dry storage technology was used in Canada, Korea and Romania for the construction of eight large-scale above ground dry storage facilities for CANDU spent fuel. These projects add up to a constructed capacity in excess of 5,000 MgU, that represents a significant share of the total worldwide dry storage capacity. This paper describes basic research and technology developments made at AECL's facilities to develop those dry storage technologies for its own reactors and for the operating CANDU 6 reactors. The current operating status of the facilities using concrete canisters is provided. A description of the MACSTOR 200 modules each having a capacity of 228 MgU that is in use at the Gentilly 2 and Cernavoda stations is provided. The Cernavoda spent fuel management facility was commissioned in 2003. The organisational, licensing, equipment supply and construction aspects that were necessary to deliver this turnkey project by AECL and its Romanian partners in 25 months are described. The paper also provides an outline of the joint program between AECL and KHNP-NETEC to develop the new MACSTOR/KN-400 and provides a description of this module having a capacity of 456 MgU (thus twice the MACSTOR 200 capacity) to be deployed by 2007 at the Wolsong site in Korea. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Schankula, M.H.; Plaice, E.L. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Woodworth, L.G. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada)

    1998-04-01

    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)

  9. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  11. MACSTOR, an on-site, dry, spent-fuel storage system developed by AECL for use by U. S. utilities

    Energy Technology Data Exchange (ETDEWEB)

    Durante, R.; Feinroth, H.; Pattanyus, P. (AECL Technologies, Bethesda, MD (United States))

    1992-01-01

    The continuing delay in the U.S. Department of Energy's Yucca Mountain and monitored retrievable storage spent-fuel disposal and storage programs has prompted U.S. utilities to consider expanding on-site storage of spent reactor fuel. Long-term, on-site storage has certain advantages to U.S. utilities since it eliminates the need for costly and difficult shipping and puts control of the spent fuel completely under the direction of the owner-utility. AECL Technologies (AECL), through its research company and Canada deuterium uranium (CANDU) engineering services division, has been developing on-site storage for Canadian heavy water nuclear plants for almost 20 yr. AECL has developed a design for a dry storage unit, designated MACSTOR (modular air-cooled storage), that can accommodate U.S. light water reactor (LWR) fuel elements and could become a candidate for the U.S. market. This paper describes MACSTOR and its evolution from the original silos and CANSTOR system that was developed and used in Canada. These systems are subject to regulatory controls by the Atomic Energy Control Board of Canada and have proven to be safe, convenient, and cost effective.

  12. Simulating thermal behavior of AECL's spent fuel dry storage system with CATHENA

    Energy Technology Data Exchange (ETDEWEB)

    Sabourin, G. [Atomic Energy of Canada Limited, Montreal, PQ (Canada)

    1998-07-01

    This paper documents the comparisons between CATHENA predictions and temperature measurements taken at the Gentilly-2 NPP spent fuel dry storage facility and in a mock--up of a storage basket placed inside a storage cylinder. It also presents CATHENA temperature predictions related to the storage of spent fuel in MACSTOR modules as planned for Ignalina NPP, Lithuania. CATHENA has been chosen because it can simulate many noncondensable gases including air and helium, and because of its great flexibility in the representation of the MACSTOR module geometry. The results of the simulations show good agreement with the experimental measurements. The two comparisons indicate that CATHENA can be used to simulate heat transfer from the fuel to the external air circuit of the spent fuel dry storage system. For the Ignalina MACSTOR module, containing RBMK fuel having higher heat release than typical CANDU fuel, CATHENA predicts that the maximum fuel temperature is expected to be around 240 deg C, giving an acceptable margin below the maximum allowed temperature of 300 deg C. In conclusion, this paper shows that the thermalhydraulic code CATHENA can accurately predict the thermal behavior AECL's air cooled spent fuel dry storage system. (author)

  13. Determination of the dose rate to the center of the irradiation chamber of the Gamma cell 220 AECL; Determinacion de la razon de dosis al centro de la camara de irradiacion del Gammacell 220 AECL

    Energy Technology Data Exchange (ETDEWEB)

    Zuazua G, M.P

    1991-11-15

    To determine the dose rate at the center of the irradiation chamber of the Gamma cell 220 AECL, two different spectrophotometers for to measure the absorbency of the irradiated dosemeters were used. In the first one dosimetry, the absorbency of the irradiated Fricke solution was read in the Varian-UV-visible spectrophotometer Series 634 of the Applied Research Management. For the second dosimetry it was used the Shimadzu UV-visible spectrophotometer belonging to the Special Projects Department. The obtained results in this study are presented. (Author)

  14. 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.

  15. 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)

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

    1999-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-01-01

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

  18. Smectite-to-illite reaction. AECL research No. AECL-10842

    Energy Technology Data Exchange (ETDEWEB)

    Oscarson, D.W.; Hume, H.B.

    1993-01-01

    The smectite component of the buffer material in a nuclear fuel waste disposal vault could slowly transform over long periods of time to an inter-stratified illite/smectite material with important implications for the long-term effectiveness of the buffer material. The smectite-to-illite reaction was examined by treating Wyoming bentonite at 150C, 200C, and 250C for periods ranging from 90-194 days in five synthetic solutions having widely varying compositions. Progress of the smectite alteration reaction was determined by measuring the expandability of the reaction products by X-ray diffractometry after the exchange complex of the clay was saturated with potassium and solvated with ethylene glycol.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-05-01

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

  1. Radiation applications research and facilities in AECL research company

    Science.gov (United States)

    Iverson, S. L.

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

  2. Technetium diffusion in clay-based materials under oxic and anoxic conditions. AECL research No. AECL-11419

    Energy Technology Data Exchange (ETDEWEB)

    Hume, H.B.

    1995-12-31

    Describes experiments to determine diffusion coefficients for technetium in compacted clay-based material (soils) saturated with a synthetic groundwater solution whose principal ions were calcium, sodium, and chlorine. Tests were conducted in anoxic conditions established by conducting the experiments in a low- oxygen glove box and by mixing 0.5% by weight of powdered iron with the soils (Lake Agassiz clay and a 1:3 mix of dry mass of clay and crushed granite aggregate). Effective diffusion coefficients were also measured in oxic conditions in Avonlea bentonite, Lake Agassiz clay, and illite/smectite. Implications of the results for transport of radionuclides through backfill material and clay barriers used in underground disposal of nuclear fuel waste are discussed.

  3. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

  5. 1993 Annual progress report for subsidiary agreement No. 2 (1991--1996) between AECL and US/DOE for a radioactive waste management technical co-operative program

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-12-31

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

  6. N286.7-99, A Canadian standard specifying software quality management system requirements for analytical, scientific, and design computer programs and its implementation at AECL

    Energy Technology Data Exchange (ETDEWEB)

    Abel, R. [R and M Abel Consultants Inc. (Canada)

    2000-07-01

    Analytical, scientific, and design computer programs (referred to in this paper as 'scientific computer programs') are developed for use in a large number of ways by the user-engineer to support and prove engineering calculations and assumptions. These computer programs are subject to frequent modifications inherent in their application and are often used for critical calculations and analysis relative to safety and functionality of equipment and systems. N286.7-99(4) was developed to establish appropriate quality management system requirements to deal with the development, modification, and application of scientific computer programs. N286.7-99 provides particular guidance regarding the treatment of legacy codes.

  7. Qualification plan for the Genmod-PC computer program

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

  8. Integrated plant information technology design support functionality

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Seung; Kim, Dae Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Barber, P. W.; Goland, D. [Atomic Energy Canada Ltd., (Canada)

    1996-06-01

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

  9. Macstor dry spent fuel storage system

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F. E. [Atomic Energy of Canada Limited, Montreal (Canada)

    1996-04-15

    AECL, a Canadian Grown Corporation established since 1952, is unique among the world's nuclear organizations. It is both supplier of research reactors and heavy water moderated CANDU power reactors as well as operator of extensive nuclear research facilities. As part of its mandate, AECL has developed products and conceptual designs for the short, intermediate and long term storage and disposal of spent nuclear fuel. AECL has also assumed leadership in the area of dry storage of spent fuel. This Canadian Crown Corporation first started to look into dry storage for the management of its spent nuclear fuel in the early 1970's. After developing silo-like structures called concrete canisters for the storage of its research reactor enriched uranium fuel, AECL went on to perfect that technology for spent CANDU natural uranium fuel. In 1989 AECL teamed up with Trans nuclear, Inc.,(TN), a US based member of the international Trans nuclear Group, to extend its dry storage technology to LWR spent fuel. This association combines AECL's expertise and many years experience in the design of spent fuel storage facilities with TN's proven capabilities of processing, transportation, storage and handling of LWR spent fuel. From the early AECL-designed unventilated concrete canisters to the advanced MACSTOR concept - Modular Air-Cooled Canister Storage - now available also for LWR fuel - dry storage is proving to be safe, economical, practical and, most of all, well accepted by the general public. AECL's experience with different fuels and circumstances has been conclusive.

  10. Macstor system for spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

    Pattantyus, P. (Atomic Energy of Canada Ltd., Montreal, PQ (Canada). Power Projects)

    1993-01-01

    In 1989, Transnuclear Inc. and AECL jointly developed the conceptual design for the Modular Aircooled Canister Storage System (Macstor) for LWR fuel. The development effort has proceeded to the completion of successful full-scale thermal testing. In 1990, AECL adapted the Macstor System approach for use with Candu fuel. The adapted design, called Canstor, has also successfully completed full-scale thermal testing, and the final system design has been completed. (author) 1 fig.

  11. Annual report 1997-1998

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-07-01

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

  12. Challenges for remote monitoring and control of small reactors

    Energy Technology Data Exchange (ETDEWEB)

    Trask, D., E-mail: traskd@aecl.ca [Atomic Energy of Canada Limited, Fredericton, New Brunswick (Canada)

    2013-07-01

    This paper considers a model for small, unmanned, remotely located reactors and discusses the ensuing cyber security and operational challenges for monitoring and control and how these challenges might be overcome through some of AECL's research initiatives and experience. (author)

  13. Software Fault Tree Analysis of Concurrent Ada Processes

    Science.gov (United States)

    1994-09-01

    excerpt from the investigation into the Therac -25 accidents: It is clear from the AECL (Atomic Energy of Canada Limited) documentation on the...Publishing Co., Inc., New York, N.Y., 1988 25. Leveson, N.G and Turner, C.S., An Investigation of the Therac -25 Accidents, IEEE Computer, July 1993

  14. Advanced CANDU reactor pre-licensing progress

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.K.; West, J.; Snell, V.G.; Ion, R.; Archinoff, G.; Xu, C. [Atomic Energy of Canada Limited., Mississauga, Ontario (Canada)]. E-mail: popovn@aecl.ca

    2005-07-01

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

  15. Pre-licensing of the Advanced CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ion, R.; Popov, N.K.; Snell, V. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)]. E-mail: popovn@aecl.ca; West, J. [Candesco, Toronto, Ontario (Canada); Xu, C. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2006-07-01

    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 output of about 1200 MWe. AECL initiated pre-licensing reviews of the ACR reactor design in Canada, US and China, with an objective to take into account regulatory feedback early in the design process. The Canadian Nuclear Safety Commission (CNSC) is performing a pre-project pre-licensing assessment of the ACR design. The objective of the assessment is to issue a formal statement as to whether there are any fundamental barriers that would prevent the licensing of the new CANDU reactor design in Canada under the Nuclear Safety and Control Act. The CNSC review is being conducted in four phases. In Phase 1 (September 2003 to September 2004) CNSC performed a pre-licensing review of the ACR-700, and focused on the design process, methodology, design concepts and R and D. CNSC staff reviewed about 100 reports, and submitted to AECL questions and comments. In Phase 2 (September 2004 to August 2005) AECL provided responses and additional information to CNSC on their comments and questions in Phase 1. Phase 3 is the Transition Phase (September 2005 to May 2006), bridging the transition from the ACR-700 to the ACR-1000 design. Phase 3 focused on review of generic aspects of the ACR design, on the Safety

  16. Licensing the ACR-700 in the USA

    Energy Technology Data Exchange (ETDEWEB)

    Langman, V.; Ion, R. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Reid, C. [Bechtel Power Corporation, San Fransisco, California (United States); Snell, V. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2003-07-01

    Atomic Energy of Canada Ltd. (AECL), via its 100% owned US subsidiary AECL Technologies Inc., is performing a pre-application review of the Advanced CANDU Reactor (ACR) with the US Nuclear Regulatory Commission (USNRC). Completion of this pre-application review by mid-2004, is in support of an application to the USNRC for Standard Design Certification that is targeted for the fall of 2004. The intent of the pre-application review is to deal up-front with potential issues associated with the CANDU reactor genealogy of the ACR that are different from the Light Water Reactor (LWR) and Boiling Water Reactor (BWR) regulatory framework in the USA. The focus of the paper will be to describe the pre-application review process currently underway with the NRC staff. In the context of the pre-application review this paper will provide an overview of the licensing approach being used to introduce the ACR-700 to the USA. (author)

  17. Transport modeling of sorbing tracers in artificial fractures

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-02-01

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

  18. Annual report 1998-1999

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-07-01

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

  19. 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)

  20. Electron processing of fibre-reinforced advanced composites

    Science.gov (United States)

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

    1996-08-01

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

  1. The small (or large) modular CANDU

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    2002-07-01

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

  4. Guide to Canadian Aerospace Related Industries,

    Science.gov (United States)

    1986-02-28

    Simulator for the CF-18 fighter and an start, it has become recognized as a leading supplier of nanose - ’Automated Pilot Selection System" for the...research specifically with computer print outs, "Mechanical Properties" and hardness dealt with improved oxidation resistance and mechanical test...warehousing. Very rough area role in developing and commercializing zirconium alloys. of all facilities is 150,000m 2. 129 More recently, AECL has

  5. Testing the Capacity of the NBDRP EX30701

    Science.gov (United States)

    2009-11-01

    Energy of Canada Limited (AECL)) that are capable of providing radiation biological dose estimates using the dicentric chromosome assay (DCA). As...Plan. Samples were scored for the dicentric chromosome assay and the CBMN assay. Using the DCA, cells were analysed for either 50 cells or 30...assay in exercises. Although not radiation specific, this assay readily detects radiation-induced chromosomal damage, requires less training for

  6. Extending the world's uranium resources through advanced CANDU fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    De Vuono, Tony; Yee, Frank; Aleyaseen, Val; Kuran, Sermet; Cottrell, Catherine

    2010-09-15

    The growing demand for nuclear power will encourage many countries to undertake initiatives to ensure a self-reliant fuel source supply. Uranium is currently the only fuel utilized in nuclear reactors. There are increasing concerns that primary uranium sources will not be enough to meet future needs. AECL has developed a fuel cycle vision that incorporates other sources of advanced fuels to be adaptable to its CANDU technology.

  7. Post-irradiation examinations of a Zr2.5Nb pressure tube of the Karachi nuclear power plant (KANUPP)

    Science.gov (United States)

    Zaheer, Mohammed Sajjad; Akhtar, Javed Iqbal; Ahmad, Ejaz; Saleem, Muhammad; Hussain, Syed Mukarrum; Qureshi, Masroor Ahmad; Khan, Azmatullah; Ali, Rafaqat; Zafarullah, Muhammad

    1996-09-01

    The results of post-irradiation examinations of a pressure tube of fuel channel No. G-12 of KANUPP have been described. A detailed study was made in Canada by AECL. A parallel investigation on its seven rings of about 50 mm length each was also carried out at PINSTECH. Visual inspection showed normal oxidation effects. Gamma spectrometry showed the presence of 95Zr and 95Nb. Microstructural study revealed the characteristic alpha plus a transformed beta phase structure.

  8. Computer-Based Procedure Systems: Technical Basis and Human Factors Review Guidance

    Science.gov (United States)

    2000-03-01

    computer-assisted system for fuel reloading while at power was designed for CANDU NPPs (Gertman et al., 1994). AECL has several aids under...for visualizing the interior of a reactor fuel channel to support the removal of stuck fuel bundles. This system is envisioned as a training aid...and vendor documents and correspondence; NRC correspondence and internal memoranda; bulletins and information no- tices; inspection and

  9. Computer code applicability assessment for the advanced Candu reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wren, D.J.; Langman, V.J.; Popov, N.; Snell, V.G. [Atomic Energy of Canada Ltd (Canada)

    2004-07-01

    AECL Technologies, the 100%-owned US subsidiary of Atomic Energy of Canada Ltd. (AECL), is currently the proponents of a pre-licensing review of the Advanced Candu Reactor (ACR) with the United States Nuclear Regulatory Commission (NRC). A key focus topic for this pre-application review is the NRC acceptance of the computer codes used in the safety analysis of the ACR. These codes have been developed and their predictions compared against experimental results over extended periods of time in Canada. These codes have also undergone formal validation in the 1990's. In support of this formal validation effort AECL has developed, implemented and currently maintains 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 discusses the SQA program used to develop, qualify and maintain the computer codes used in ACR safety analysis, including the current program underway to confirm the applicability of these computer codes for use in ACR safety analyses. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  13. Mechanical engineering utilizing advanced engineering tools for the CANDU 9 project

    Energy Technology Data Exchange (ETDEWEB)

    Nuzzo, F.; Yu, S.K.W.; Hedges, K.R. [Atomic Energy of Canada Limited (AECL), Ontario (Canada)

    1998-05-01

    To meet the increasing challenging project requirements such as cost and schedule reduction, AECL has incorporated a comprehensive suite of integrated, advanced engineering tools for CANDU project engineering and delivery. This paper provides a description of the advanced engineering tools developed and used by AECL in the pre-project engineering of the CANDU 9 product and the construction projects such as the construction of two CANDU 6 units in Qinshan, China. The advanced mechanical engineering tools described include: the Process and Instrument Diagram ; the mechanical/piping 3D models; the CADDS/piping analysis interface (PAI) tool; the pipe support design system (SDS) tool; and the powerful equipment database tool - TeddyBase. A description of the enhanced work process will also be provided. The work process improvement is a direct result of the implementation of advanced information technology and the integration of AECL tools with commercial engineering and project tools available in the market. The use of these advanced tools results in better design quality; enhanced presentation of the engineering deliverables to construction and commissioning staff; and potential support to future plant operations (Ref.1). (author). 2 refs., 3 figs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-11-01

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

  15. Tritium handling experience at Atomic Energy of Canada Limited

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

  16. Operating Experience of MACSTOR Modules at CANDU 6 Stations

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, Robert R. [Atomic Energy Canada Ltd., Chalk River (Canada)

    2005-11-15

    Over the last three decades, Atomic Energy of Canada Limited (AECL) has contributed to the technology development and implementation of dry spent fuel management facilities in Canada, Korea and Romania During that period, AECL has developed a number of concrete canister models and the MACSTOR200 module, a medium size air-cooled vault with a 228 MgU (Mega grams of Uranium) capacity. AECL's dry storage technologies were used for the construction of eight large-scale above ground dry storage facilities for CANDU spent fuel. As of 2005, those facilities have an installed capacity in excess of 5,000 MgU. Since 1995, the two newest dry storage installations built for CANDU 6 reactors at Gentilly 2 (Canada) and Cernavoda (Romania) used the MACSTOR 200 module. Seven such modules have been built at Gentilly 2 during the 1995 to 2004 period and one at Cernavoda in 2003. The construction and operating experience of those modules is reviewed in this paper. The MACSTOR 200 modules were initially designed for a 50-year service life, with recent units at Gentilly 2 licensed for a 100-year service life in a rural (non-maritime) climate. During the 1995-2005 period, six of the eight modules were loaded with fuel. Their operation has brought a significant amount of experience on loading operations, performance of fuel handling equipment, radiation shielding, heat transfer, monitoring of the two confinement boundaries and radiation dose to personnel. Heat dissipation performance of the MACSTOR 200 was initially licensed using values derived from full scale tests made at AECL's Whiteshell Research Laboratories, that were backed-up by temperature measurements made on the first two modules. Results and computer models developed for the MACSTOR 200 module are described. Korea Hydro and Nuclear Power (KHNP) and its subsidiary Nuclear Environment Technology Institute (NETEC), in collaboration with Hyundai Engineering Company Ltd. (HEC) and AECL, are developing a new dry storage

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

    Science.gov (United States)

    Ashmore, J Patrick; Gentner, Norman E; Osborne, Richard V

    2010-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Wren, D.J.; Popov, N.; Snell, V.G. [Atomic Energy of Canada Ltd, (Canada)

    2004-07-01

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

  19. ACR fuel storage analysis: finite element heat transfer analysis of dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Khair, K.; Baset, S.; Millard, J. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2006-07-01

    Over the past decade Atomic Energy of Canada Limited (AECL) has designed and licensed air-cooled concrete structures used as above ground dry storage containers (MACSTOR) to store irradiated nuclear fuel from CANDU plants. A typical MACSTOR 200 module is designed to store 12,000 bundles in 20 storage cylinders. MACSTOR 200 modules are in operation at Gentilly-2 in Canada and at Cernavoda in Romania. The MACSTOR module is cooled passively by natural convection and by conduction through the concrete walls and roof. Currently AECL is designing the Advanced Candu Reactor (ACR) with CANFLEX slightly enriched uranium fuel to be used. AECL has initiated a study to explore the possibility of storing the irradiated nuclear fuel from ACR in MACSTOR modules. This included work to consider ways of minimizing footprint both in the spent fuel storage bay and in the dry storage area. The commercial finite element code ANSYS has been used in this study. The FE model is used to complete simulations with the higher heat source using the same concrete structural dimensions to assess the feasibility of using the MACSTOR design for storing the ACR irradiated fuel. This paper presents the results of the analysis. The results are used to confirm the possibility of using, with minimal changes to the design of the storage baskets and the structure, the proven design of the MACSTOR 200 containment to store the ACR fuel bundles with higher enrichment and burnup. This has thus allowed us to confirm conceptual feasibility and move on to investigation of optimization. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Shoman, Nathan T., E-mail: nshoman@vols.utk.edu; Skutnik, Steven E., E-mail: sskutnik@utk.edu

    2016-06-15

    Highlights: • New ORIGEN libraries for CANDU 28 and 37-element fuel assemblies have been created. • These new reactor data libraries are based on modern ENDF/B-VII.0 cross-section data. • The updated CANDU data libraries show good agreement with radiochemical assay data. • Eu-154 overestimated when using ENDF-VII.0 due to a lower thermal capture cross-section. - Abstract: A new set of SCALE fuel lattice models have been developed for the 28-element and 37-element CANDU fuel assembly designs using modern cross-section data from ENDF-B/VII.0 in order to produce new reactor data libraries for SCALE/ORIGEN depletion analyses. These new libraries are intended to provide users with a convenient means of evaluating depletion of CANDU fuel assemblies using ORIGEN through pre-generated cross sections based on SCALE lattice physics calculations. The performance of the new CANDU ORIGEN libraries in depletion analysis benchmarks to radiochemical assay data were compared to the previous version of the CANDU libraries provided with SCALE (based on WIMS-AECL models). Benchmark comparisons with available radiochemical assay data indicate that the new cross-section libraries perform well at matching major actinide species (U/Pu), which are generally within 1–4% of experimental values. The library also showed similar or better results over the WIMS-AECL library regarding fission product species and minor actinoids (Np, Am, and Cm). However, a notable exception was in calculated inventories of {sup 154}Eu and {sup 155}Eu, where the new library employing modern nuclear data (ENDF/B-VII.0) performed substantially poorer than the previous WIMS-AECL library (which used ENDF-B/VI.8 cross-sections for these species). The cause for this discrepancy appears to be due to differences in the {sup 154}Eu thermal capture cross-section between ENDF/B-VI.8 and ENDF/B-VII.0, an effect which is exacerbated by the highly thermalized flux of a CANDU heavy water reactor compared to that of a

  1. Development and Performance of a Proton and Deuteron ECR Ion Source

    CERN Document Server

    Dunkel, Kai; Piel, Christian; Plitzko, J

    2005-01-01

    A 5mA proton and deuteron rf source is under development at ACCEL. This source will provide the front end of our superconducting proton/deuteron linear accelerator. The design of the source is based on the proven AECL design of a 100 mA proton source. The paper will describe the design of the source and the layout of the test bench currently set up at ACCEL to characterize the source. Results of the beam dynamic simulations performed to optimize the source geometry based on KOBRA 3D will be presented and compared with first measurement results.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  3. Deployment of advanced MACSTOR dry spent fuel storage technology in Korea - A joint development program

    Energy Technology Data Exchange (ETDEWEB)

    Cobanoglu, M. M.; Pattantyus, P. [Atomic Energy Canada Limited, Ottawa (Canada); Song, M. J.; Lee, H. Y. [KHNP/NETEC, Daejeon (Korea, Republic of)

    2002-04-15

    KHNP/NETEC's (K/N) and Atomic Energy of Canada Limited (AECL) are undertaking to jointly develop a high capacity dry storage structure made of reinforced concrete that uses the MACSTOR storage module concept. This effort is based on AECL's experience and on the successful deployment of concrete canisters at Wolsong and on the deployment of air-cooled MACSTOR modules at the Gentilly 2 reactor in Canada. The proposed approach addresses the conditions specific to the Wolsong site: large yearly fuel throughput, space limitations and the need for an economical dry storage structure that can store lifetime spent fuel inventories expected from the four CANDU units. The selected configuration is a 4-row MACSTOR module with a capacity of 24,000 bundles stored in 400 baskets, each holding 60 spent fuel bundles. The module is thus termed MACSTOR/KN-400 and is expected to offer a repetitive storage density increase by a factor of approximately 3, compared to concrete canisters presently used. The four Wolsong units generate spent fuel bundles that, with the high capacity factors achieved, are in the order of 20,000 bundles or more per year. At all Korean nuclear facilities, space limitations dictate the need for storage structures having high storage density. Storage density increases have to be accomplished while maintaining safety parameters during the full term storage of nuclear fuel. During the early 1990's AECL has proceeded with the development of a 2-row MACSTOR storage module that offered a higher storage density and a more economical solution compared to the stand alone concrete canister used at Wolsong 1. These modules are in use at Gentilly since the mid 1990's and operate at a capacity of 200 baskets. The selection of a MACSTOR module with 4 rows of storage cylinders is the natural evolution of the already deployed configuration. It can be developed without additional thermal testing as the fuel is maintained within the existing licensing

  4. Analysis specifications for the CC3 biosphere model biotrac

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Wright, E.D. [comp.

    1992-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Scriven, M.G.; Hummel, D.W.; Novog, D.R.; Luxat, J.C. [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

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

  7. The balanced scorecard advantage: Driving strategic change into Canada's nuclear laboratory site operations

    Energy Technology Data Exchange (ETDEWEB)

    Lafreniere, P. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Weeks, D. [Haggarty' s Cove Ventures (2001 Ltd.) St. Stephen, New Brunswick (Canada)

    2001-07-01

    The challenges presented by the size, diversity, complexity and history of the Facilities and Nuclear Operations (FNO) Group at AECL's Chalk River Laboratories (CRL) required a change to the traditional management approach. As a result, a strategy was adopted that focused on integrating contemporary business practices such as process mapping, activity based management and use of the Balanced Scorecard methodology into the operational culture at CRL. In addition, revitalization of the performance management methods process was undertaken to provide a tool for assessment of business and individual performance. performance. (author)

  8. Energy- and Intensity-Modulated Electron Beam for Breast Cancer Treatment

    Science.gov (United States)

    1999-10-01

    Dosimetry of small fields for Therac 20 electron beams Med. Phys. 11 697-702 Shiu A S, Tung S, Hogstrom K R, Wong J W, Gerber R L, Harms W B and Purdy...accelerator, was developed by Lovelock et al (1995). Sixel and Faddegon (1995) simulated a Therac -6 treatment head in radiosurgery mode using the cylindrically...simulated three GGR MeV/AECL accelerators, i.e. Therac 40 Sagittaire, Therac 20 Saturne and Therac 10 Neptune, using a Monte Carlo code based on the

  9. 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.

  10. Development of CANDU pressure tube integrity evaluation system

    Energy Technology Data Exchange (ETDEWEB)

    Kwac, S. L.; Kim, Y. J. [Sungkyunkwan Univ., Seoul (Korea, Republic of); Lee, J. S. [Kyonggi Univ., Suwon (Korea, Republic of); Park, Y. W. [KINS, Taejon (Korea, Republic of)

    1999-05-01

    The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw or contact with their calandria tubes is found during the periodic inspection, the integrity evaluation must be carried out, and the safety requirements must be satisfied for continued service. In order to perform the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the integrity evaluation process. For this reason, an integrity evaluation system was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL. The evaluation procedure includes the crack growth calculation both by DHC and by fatigue. It also provides the prediction of fracture initiation, plastic collapse and leak-before-break(LBB), blister formation and blister growth. This system provides various databases including the 3-D finite element analysis results on pressure tubes, inspection data and design specifications. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

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

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Moon Sung

    2000-11-01

    The BDL-443 CANFLEX-RU bundle AKW was fabricated at Korea Atomic Energy Research Institute (KAERI) for power ramp irradiation testing in NRU reactor. The bundle was fabricated with IDR and ADU fuel pellets in adjacent elements and contains fuel pellets enriched to 1.65 wt% {sup 235}U in the outer and intermediate rings and also contains pellets enriched to 2.00 wt% {sup 235}U in the inner ring. This bundle does not have a center element to allow for insertion on a hanger bar. KAERI produced the IDR pellets with the IDR-source UO{sub 2} powder supplied by BNFL. ADU pellets were fabricated and supplied by AECL. Bundle kits (Zircaloy-4 end plates, end plugs, and sheaths with brazed appendages) manufactured at KAERI earlier in 1996 were used for the fabrication of the bundle. The CANFLEX bundle was fabricated successfully at KAERI according to the QA provisions specified in references and as per relevant KAERI drawings and technical specification. This report covers the fabrication activities performed at KAERI. Fabrication processes performed at AECL will be documented in a separate report.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  13. Simulation of transient heat transfer in MACSTOR/KN-400 module storing irradiated CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sabourin, G. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada); Lee, K.-H.; Yoon, J.-H.; Choi, B.-I.; Lee, H.-Y.; Song, M.-J. [KHNP, Nuclear Environment Technology Inst., Taejon (Korea, Republic of)

    2004-07-01

    Korea Hydro and Nuclear Power (KHNP), in collaboration with Atomic Energy of Canada Limited (AECL), are developing a new module for the dry storage of spent fuel from the four CANDU 6 nuclear reactors at the Wolsong site in South Korea, the MACSTOR/KN-400. The simulation of transient conditions for AECL's spent fuel dry storage systems, presented in this paper, has not been performed before and is considered a major achievement of the present work. In a fist step, CATHENA was compared to MACSTOR-200 temperature measurements and the accuracy of the results were very good. In a second step, CATHENA was applied to the MACSTOR/KN-400. Four cases were performed for the MACSTOR/KN-400: Off-normal cases in summer and winter and reduced air flow cases in summer and winter. The maximum local concrete temperatures were predicted to be 63{sup o}C for the off-normal case and 65{sup o}C in the reduced air flow case. The maximum temperature gradients in the concrete are predicted to be 28{sup o}C for the off-normal case and 30{sup o}C in the reduced air flow case, incorporating a 3{sup o}C uncertainty. This paper shows that the maximum temperature for the module is expected to meet the temperature limitations of appropriate standards. (author)

  14. Large block migration experiments: INTRAVAL phase 1, Test Case 9

    Energy Technology Data Exchange (ETDEWEB)

    Gureghian, A.B.; Noronha, C.J. (Battelle, Willowbrook, IL (USA). Office of Waste Technology Development); Vandergraaf, T.T. (Atomic Energy of Canada Ltd., Ottawa, ON (Canada))

    1990-08-01

    The development of INTRAVAL Test Case 9, as presented in this report, was made possible by a past subsidiary agreement to the bilateral cooperative agreement between the US Department of Energy (DOE) and Atomic Energy of Canada Limited (AECL) encompassing various aspects of nuclear waste disposal research. The experimental aspect of this test case, which included a series of laboratory experiments designed to quantify the migration of tracers in a single, natural fracture, was undertaken by AECL. The numerical simulation of the results of these experiments was performed by the Battelle Office of Waste Technology Development (OWTD) by calibrating an in-house analytical code, FRACFLO, which is capable of predicting radionuclide transport in an idealized fractured rock. Three tracer migration experiments were performed, using nonsorbing uranine dye for two of them and sorbing Cs-137 for the third. In addition, separate batch experiments were performed to determine the fracture surface and rock matrix sorption coefficients for Cs-137. The two uranine tracer migration experiment were used to calculate the average fracture aperture and to calibrate the model for the fracture dispersivity and matrix diffusion coefficient. The predictive capability of the model was then tested by simulating the third, Cs-137, tracer test without changing the parameter values determined from the other experiments. Breakthrough curves of both the experimental and numerical results obtained at the outlet face of the fracture are presented for each experiment. The reported spatial concentration profiles for the rock matrix are based solely on numerical predictions. 22 refs., 12 figs., 8 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  17. Multi-purpose hydrogen isotopes separation plant design

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

  18. Licensing the ACR-700 in the USA; progress on the pre-application review

    Energy Technology Data Exchange (ETDEWEB)

    Ion, R. [Atomic Energy of Canada Ltd., Mississauga, Ontario (Canada); Langman, V. [AECL Technologies, Gaithersburg, Maryland (United States); Snell, V. [Atomic Energy of Canada Ltd., Mississauga, Ontario (Canada); Reid, C. [Bechtel Power Corp., San Francisco, California (United States)

    2004-07-01

    The ACR-700 is an evolutionary reactor design, which incorporates the inherent safety features of the CANDU products, as well as the successful operating experience of the current CANDU 6 reactors. The improvements to the ACR-700 from CANDU 6 result in significant reductions in capital and operating costs as well as enhanced safety. AECL Technologies (AECLT, a wholly owned US subsidiary of Atomic Energy of Canadian Limited) is the proponent for a pre-application review of the ACR-700 with the US Nuclear Regulatory Commission (NRC) in the United States. This pre-application review will be completed shortly and will support an application to the US NRC for Standard Design Certification (SDC). AECL Technologies' overall objective for the pre-application review of the ACR-700 is to obtain an understanding, from interactions with the USNRC, of the scope, cost, and the schedule to obtain a Standard Design Certification for the ACR-700. The pre-application review will address licensing issues associated with the CANDU reactor technologies in ACR-700 that depart from the light water reactor, pressure-vessel based regulatory framework in the USA. Therefore, during the course of the pre-application review, any major US NRC issues with the ACR-700 design will be identified early and the scope of the work required to address these concerns, along with associated completion schedules, will be formulated and ultimately agreed upon with the US NRC. AECLT has been informed by the NRC staff that the results of their pre-application review will be documented in a Safety Assessment Report (SAR), which will state whether there are any major impediments to licensing the ACR in the United States. In particular, the SAR should provide confirmation of the licensing criteria applicable to the ACR, provide an assessment of the completeness of AECL 's Research and Development (R and D) programs that exist or are planned in support of the ACR, provide an assessment of the

  19. Pre-licensing of the Advanced CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ion, R.; Popov, N.; Snell, V.; Xu, C. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); West, J. [Candesco Co., Toronto, Ontario (Canada)

    2006-09-15

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  1. Integrated plant for treatment of liquid radwaste

    Energy Technology Data Exchange (ETDEWEB)

    Sen Gupta, S.K. [Chalk River Laboratories, Ontario (Canada)

    1995-05-01

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

  2. BEAM 1.7: development for modelling fuel element and bundle buckling strength

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, G.; Xu, S.; Xu, Z.; Paul, U.K. [Atomic Energy of Canada, Mississauga, Ontario (Canada)

    2010-07-01

    This paper describes BEAM, an AECL developed computer program, used to assess mechanical integrity of CANDU fuel bundles. The BEAM code has been developed to satisfy the need for buckling strength analysis of fuel bundles. Buckling refers to the phenomenon where a compressive axial load is large enough that a small lateral load can cause large lateral deflections. The buckling strength refers to the critical compressive axial load at which lateral instability is reached. The buckling strength analysis has practical significance for the design of fuel bundles, where the buckling strength of a fuel element/bundle is assessed so that the conditions leading to bundle jamming in the pressure tube are excluded. This paper presents the development and qualification of the BEAM code, with emphasis on the theoretical background and code implementation of the newly developed fuel element/bundle buckling strength model. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Sugita, Yutaka [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan). Tokai Works

    2002-09-01

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

  4. Dosimetry of the JS-6500 industrial irradiator for the irradiation of the PVC graduated flasks; Dosimetria del irradiador industrial JS-6500 para la irradiacion de probetas de PVC

    Energy Technology Data Exchange (ETDEWEB)

    Castaneda F, A.; Carrasco A, H.; Martinez P, M.E. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2002-07-01

    The dosimetry of the JS-6500 AECL irradiator was realized, outside of the industrial transportation rails to know the dose distribution, as well as its dose speed. This one with the intention of exposing to gamma radiation; plastified PVC graduated flasks and evaluating their interweavement or degradation or both. This study of dosimetry was carried out by means of a theoretical and experimental evaluation in air atmosphere. The results allow to know the irradiation conditions of the PVC graduated flasks as well as those results prove that has not a significant difference among the obtained result as theoretical as experimentally due to that the obtained result in the theoretical evaluation is 2.62 KGy/h and the result for the case of the experimental evaluation is 2.74 KGy/h. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  6. Waste management issues and their potential impact on technical specifications of CANDU fuel materials

    Energy Technology Data Exchange (ETDEWEB)

    Tait, J.C.; Johnson, L.H. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada)

    1997-07-01

    The technical specifications for the composition of nuclear fuels and materials used in Canada's CANDU reactors have been developed by AECL and materials manufacturers, taking into account considerations specific to their manufacture and the effect of minor impurities on fuel behaviour in reactor. Nitrogen and chlorine are examples of UO{sub 2} impurities, however, where there is no technical specification limit. These impurities are present in the source materials or introduced in the fabrication process and are neutron activated to {sup 14}C and {sup 36}C1, which after {sup 129}I , are the two most significant contributors to dose in safety assessments for the disposal of used fuel. For certain impurities, environmental factors, particularly the safety of the disposal of used fuels, should be taken into consideration when deriving 'allowable' impurity limits for nuclear fuel materials. (author)

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

    CERN Document Server

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

    1999-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

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

  10. The smallest SMRs

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K. [ACSION (Canada)

    2013-07-01

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

  11. Evolution of the MACSTOR{trademark} dry spent fuel storage system

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F.E.; Joubert, W.M. [Atomic Energy of Canada Ltd., Montreal, Quebec (Canada)

    1995-12-31

    The MACSTOR{trademark} (Modular Air-Cooled Canister Storage) system was developed by Atomic Energy of Canada Limited (AECL) for the interim storage of spent fuel discharged by light water reactors. It is a hybrid system which combines the operational economies of metal cask technology with the capital economies of concrete technology. The system includes all the necessary equipment to transfer spent fuel from a storage pool to an independent interim dry spent fuel storage site. After presenting a description of the system and a brief history of its development, the paper addresses its thermal performance and modeling for various design configurations. Finally, a brief summary of the experience being gained during the implementation of a MACSTOR{trademark} system modified for CANDU spent fuel at the Gentilly-2 NPP in Quebec is presented. It includes progress made in licensing activities and in public hearings pertinent to the initiation of the project.

  12. Modelling iodine behaviour using LIRIC 3.0

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Smellie, J. [Conterra AB (Sweden); Karlsson, Fred [Swedish Nucelear Fuel and Waste Management Co., Stockholm (Sweden)

    1996-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-04-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

  19. Test requirement for PIE of HANARO irradiated fuel rod

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Phalippou, C.; Delaune, X.

    1996-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

  2. ACR-1000 design provisions for severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.K.; Santamaura, P.; Shapiro, H.; Snell, V.G. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)]. E-mail: popovn@aecl.ca

    2006-07-01

    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 a 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 enhanced safety features, shorter construction schedule, high plant capacity factor, improved operations and maintenance, and increased operating life. 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 output of about 1200 MWe. The ACR-1000 design meets Canadian regulatory requirements and follows established international practice with respect to severe accident prevention and mitigation. This paper presents the ACR-1000 features that are designed to mitigate limited core damage and severe core damage states, including core retention within vessel, core damage termination, and containment integrity maintenance. While maintaining existing structures of CANDU reactors that provide inherent prevention and retention of core debris, the ACR-1000 design includes additional features for prevention and mitigation of severe accidents. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel is designed for debris retention by minimizing penetrations at the bottom periphery and by accommodating thermal and weight loads of the core debris. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes

  3. MACSTOR{trademark}: Dry spent fuel storage for the nuclear power industry

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F.E.; Pattantyus, P. [AECL Candu, Montreal, Quebec (Canada); Hanson, A.S. [Transnuclear, Inc., Hawthorne, NY (United States)

    1993-12-31

    Safe storage of spent fuel has long been an area of critical concern for the nuclear power industry. As fuel pools fill up and re-racking possibilities become exhausted, power plant operators will find that they must ship spent fuel assemblies off-site or develop new on-site storage options. Many utility companies are turning to dry storage for their spent fuel assemblies. The MACSTOR (Modular Air-cooled Canister STORage) concept was developed with this in mind. Derived from AECL`s successful vertical loading, concrete silo program for storing CANDU nuclear spent fuel, MACSTOR was developed for light water reactor spent fuel and was subjected to full scale thermal testing. The MACSTOR Module is a monolithic, shielded concrete vault structure than can accommodate up to 24 spent fuel canisters. Each canister holds 12 PWR or 32 PWR previously cooled spent fuel assemblies with burn-up rates as high as 45,000 MWD/MTU. The structure is passively cooled by natural convection through an array of inlet and outlet gratings and galleries serving a central plenum where the (vertically) stored canisters are located. The canisters are continuously monitored by means of a pressure monitoring system developed by TNI. The MACSTOR system includes the storage module(s), an overhead gantry system for cask handling, a transfer cask for moving fuel from wet to dry storage and a cask transporter. The canister and transfer cask designs are based on Transnuclear transport cask designs and proven hot cell transfer cask technology, adapted to requirements for on-site spent fuel storage. This Modular Air Cooled System has a number of inherent advantages: efficient use of construction materials and site space; cooling is virtually impossible to impede; has the ability to monitor fuel confinement boundary integrity during storage; the fuel canisters may be used for both storage and transport and canisters utilize a flanged, ASME-III closure system that allows for easy inspection.

  4. Qualification of a Human-System Interface to Meet IEC 61513

    Energy Technology Data Exchange (ETDEWEB)

    Malcom, Scott; Kim, Sun Ho [Atomic Energy of Canada Limited, Mississauga (Canada); Macdonald, Marienna [Atomic Energy of Canada Limited, Fredericton (Canada)

    2011-08-15

    This paper describes the steps Atomic Energy of Canada Limited (AECL{sup TM}) undertook to qualify its Advanced Control Centre Information System (ACCIS{sup TM}) to meet the requirements of IEC 61513. It will address the different strategies used for software versus hardware. As well, the paper will discuss the steps that have been taken to qualify third-party commercial-off-the-shelf products that are used in conjunction with a qualified product. ACCIS is a display, monitoring and supervisory control system that is designed to be readily configurable and scalable to satisfy the display requirements for single functions or complex industrial plant systems. The ACCIS software services are configured and deployed across a distributed computing environment to meet the needs of a given implementation. From small single-node applications to large, complex, multi-node configurations, system behaviour is largely configured via data specifications. This design reduces the costs associated with development of custom software and allows the user to have greater control of behavioral attributes of the system, including data sampling and storage rates, the appearance and behaviour of displays, alarm annunciation features, and the configuration of system health checks. Utilities and regulators are demanding that these computer-based systems are developed and maintained with an appropriate amount of engineering rigor. To meet this challenge, AECL is qualifying its ACCIS HSI, which is intended for use in all future CANada Nuclear Deuterium (CANDU{sup TM}) nuclear power plants, to meet the requirements of the International Electrotechnical Commission's (IEC) standard for instrumentation and control systems important to safety, IEC 61513. Transitioning to the IEC standards has not been without its challenges. While AECL previously used a software development model very similar to the IEC model, absorbing the volume of the IEC standards and understanding how they should be

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

    Energy Technology Data Exchange (ETDEWEB)

    Slater, I.J

    2003-07-01

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

  6. Estimation of longitudinal and transverse dispersivities in the Twin Lake natural gradient tracer tests

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Seiji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Moltyaner, G.L.

    1998-06-01

    The field scale tracer tests were carried out with a non-reactive tracer of {sup 131}I at Twin Lake on the Chalk River Laboratory (CRL) site in Atomic Energy of Canada Limited (AECL). The natural gradient dispersion test such as the Twin Lake tracer test is very few and useful for evaluating mass transport parameters and testing groundwater flow and transport models. In this report, mass transport parameters, velocity, longitudinal and transverse dispersivities, were estimated from the Twin Lake 40 m tracer tests. This estimation was performed to provide dispersion data for 3-D transport modelling in the Lake 233 site scale (600 m in east-west direction and 1400 m in north-south direction). Two different methods were applied to the measured breakthrough curves of {sup 131}I in order to evaluate velocity and longitudinal dispersivity. The first method is the fitting procedure of the 1-D advection-dispersion solution, and the second one is the temporal moments analysis. The effect of applying these methods to field data on transport parameters was discussed in this study. The vertical profiles of {sup 131}I were used in the estimation of transverse dispersivity by fitting the 3-D advection-dispersion solution. This report refereed to the effect of variable velocity on the estimated dispersivities. The correlation between magnitude of both dispersivities and the travel distance up to 40 m was also investigated. (author)

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

    Science.gov (United States)

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

    2007-11-01

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

  8. Argentina: Nuclear power development and Atucha 2

    Energy Technology Data Exchange (ETDEWEB)

    Nogarin, Mauro

    2015-08-15

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

  9. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Kitazuru, E.R. E-mail: erkitazu@ipen.br; Moreira, A.V.B.; Mancini-Filho, J. E-mail: jmamcini@usp.br; Delincee, H.; Villavicencio, A.L.C.H. E-mail: villavic@ipen.br

    2004-10-01

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

  11. Analog information and the Canadian concept for disposal of nuclear fuel waste

    Energy Technology Data Exchange (ETDEWEB)

    Cramer, J.J. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada)

    1996-07-01

    AECL, with support from Ontario Hydro under auspices of the Candu Owners Group, has assessed a concept for the safe disposal of nuclear fuel waste in Canada. The disposal concept is to place nuclear fuel waste in corrosion-resistant containers and emplace the containers with sealing materials in an engineered vault at depths of 500 to 1000m in plutonic rock of the Canadian Shield. Humans and the environment would be protected from contaminants in the waste by several barriers; the waste itself, the container, the sealing materials, and the rock. This disposal concept permits a great deal of flexibility in its implementation, which means that a wide range of circumstances could be accommodated. Studies of natural analogues provide important information for evaluating and improving our knowledge and understanding of the disposal concept. Analogue information is used to develop the scenarios and conceptual models, to provide input to databases, and to test models, thereby enhancing the level of confidence in the safety predictions from the assessment models. In addition, natural analogues are valuable illustrative tools when presenting information on the disposal concept to the non-expert and the public.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

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

  14. Development of Zirconium alloys (for pressure tubes)

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

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

  15. Radiation sensitivity of different citric pectins

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won Young; Min, Byung Joo

    2004-03-15

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

  18. Recent findings on the oxidation of UO{sub 2} fuel under nominally dry storage conditions

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, P.; McEachern, R.J.; Sunder, S.; Wasywich, K.M.; Miller, N.H.; Wood, D.D

    1995-07-01

    This paper is an overview of fuel-storage demonstration experiments, supporting research on UO{sub 2} oxidation, and associated model development, in progress at AECL's Whiteshell Laboratories. The work is being performed to determine the time/temperature limits for safe storage of irradiated CANDU fuel in dry air. The most significant recent experimental finding has been the detection of small quantities of U{sub 3}O{sub 8}, formed over periods of one to several years in a variety of experiments at 150-170 deg C. Another important trading is the slight suppression of U{sub 3}O{sub 8} formation in SIMFUEL and other doped U0{sub 2} formulations. The development of a nucleation-and-growth model for U{sub 3}O{sub 8} formation is discussed, along with available activation energy data. These provide a basis for predicting U{sub 3}O{sub 8} formation rates under dry-storage conditions, and hence optimizing fuel storage strategies. (author)

  19. Wet sipping system at Wolsong-1

    Energy Technology Data Exchange (ETDEWEB)

    Park, J.Y.; Shin, J.C.; Kim, Y.C.; Park, C.H.; Choi, T.Y.; Park, C.J., E-mail: jyoulpark@knfc.co.kr [Korea Nuclear Fuel Co. Ltd. (KNF), Yousong, Daejeon (Korea, Republic of); Manger, A.M. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2010-07-01

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

  20. Safety assessment for the CANFLEX-NU fuel bundles with respect to the 37-element fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Suk, H. C.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-11-01

    The KAERI and AECL have jointly developed an advanced CANDU fuel, called CANFLEX-NU fuel bundle. CANFLEX 43-element bundle has some improved features of increased operating margin and enhanced safety compared to the existing 37-element bundle. Since CANFLEX fuel bundle is designed to be compatible with the CANDU-6 reactor design, the behaviour in the thermalhydraulic system will be nearly identical with 37-element bundle. But due to different element design and linear element power distribution between the two bundles, it is expected that CANFLEX fuel behaviour would be different from the behaviour of the 37-element fuel. Therefore, safety assessments on the design basis accidents which result if fuel failures are performed. For all accidents selected, it is observed that the loading of CANFLEX bundle in an existing CANDU-6 reactor would not worsen the reactor safety. It is also predicted that fission product release for CANFLEX fuel bundle generally is lower than that for 37-element bundle. 3 refs., 2 figs., 2 tabs. (Author)

  1. The effects of gamma radiation on soybean isoflavones contents

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

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

    Directory of Open Access Journals (Sweden)

    Dae-Hyun Hwang

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

    De Grosbois, J.; Hepburn, G. A.; Olmstead, R. [Atomic Energy Canada Ltd., 2251 Speakman Drive, Mississauga, Ont. L5K 1B2 (Canada); Goble, W. [Exida, 64 N. Main St., Sellersville, PA 18960 (United States); Kumar, V. [Carleton Univ., 1125 Colonel By Drive, Ottawa, Ont. K1S 5B6 (Canada)

    2006-07-01

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

  5. Replacement of Cobalt base alloys hardfacing by NOREM alloy; EDF experience and development, some metallurgical considerations. Valves application (CLAMA, RAMA)

    Energy Technology Data Exchange (ETDEWEB)

    Carnus, M. [EDF DPN UTO Direction Expertise Technique, Noisy le Grand (France); Confort, X. [VELAN SAS, Lyon (France)

    2011-07-01

    Cobalt base alloys, such as Stellite 6 and 21, are used extensively in applications where superior resistance to wear and corrosion are required. However the use of Cobalt alloys hardfacing materials, especially on valves, is a major contributor to the level of radioactive contamination of nuclear facilities. NOREM alloys, an iron base and cobalt free materials, have been developed through an Electric Power Research Institute (EPRI) long running program during the eighties as an alternative of Stellite. This alloy has relatively good weldability properties, it was developed initially for repairing Stellite hardfacing (deposit over existing hardfacing alloys). This alloy has good corrosion resistance properties associated with elevated hardness (HRC 36-42). Technological properties (such as galling resistance, wear resistance) have been evaluated through different testing programs led by EPRI, AECL(Atomic Energy of Canada Limited), Valves manufacturers, EDF and others during the nineties. More recently EDF (for replacement of globe valves) has carried out testing program focused on weld deposit chemistry and mechanical properties. NOREM is a candidate for replacement of stellite hardfacing on valves. However this alloy is not so versatile as stellite alloys regarding technological properties (such as wear resistance) at elevated temperature and under high contact pressure. As a consequence some limits have to be considered for application on valves operating at elevated temperature and under high contact pressure (> 20 Mpa). Examples of application on valves, from VELAN manufacturer, for EDF PWR equipment are given. The industrial feedback from installed equipment (CLAMA, RAMA) since 2006 on EDF PWR has been good

  6. Proceedings of the WIN-Global 2008 conference

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-07-01

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

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

    Science.gov (United States)

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

    2004-09-01

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

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

    Science.gov (United States)

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

    2004-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-08-15

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

  12. Uncertainty analysis guide

    Energy Technology Data Exchange (ETDEWEB)

    Andres, T.H

    2002-05-01

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

  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. Gamma radiation influence on technological characteristics of wheat flour

    Science.gov (United States)

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

    2012-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-11-15

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

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

    Science.gov (United States)

    Hill, David J. T.; Choi, Bong-Ku; Ahn, Hung-Kun; Choi, E.-Joon

    2001-07-01

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

  17. Lessons learned from the NRU vessel leak repair and return to service projects

    Energy Technology Data Exchange (ETDEWEB)

    Heeney, P.; Turcotte, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2011-07-01

    In May 2009 the National Research Universal (NRU) reactor was shut down due to a small leak detected from the reactor vessel into the annulus surrounding the reactor. What ensued was a challenging, yet successful, 15 month long Repair and Return to Service Outage. This Repair and Return to Service Outage presented many first-of-a-kind challenges that provide learning opportunities which have been incorporated into subsequent planned outages. These lessons learned are invaluable tools to be used in the planning and execution of future outages. Following the repair of the NRU vessel, AECL was required to conduct annual inspections of the vessel wall. These inspections require an annual Extended Outage (up to 4 weeks in length). A planned Extended Outage was conducted in May/June 2011 and provided an opportunity to implement some of the lessons learned during the Repair and Return to Service Outage. Lessons learned from that Extended Outage have been incorporated in the subsequent monthly maintenance outages, with lessons learned sessions being held after each outage to ensure that the execution of outages is constantly improving. (author)

  18. The race for megavoltage. X-rays versus telegamma.

    Science.gov (United States)

    Robison, R F

    1995-01-01

    Roentgen's discovery was announced in January, 1896, and x-ray therapy trials followed in 1897. Becquerel rays and radioactive minerals were identified during 1896 through 1898. Radium was used for therapy by 1901, even though a pure standard was not achieved until 1910-1912. Quantities of radium finally became available after 1919, and for 20 years telegamma therapy machines underwent progressive development. Their megavoltage beam was much preferred over the standard 200-250 KV x-ray units of that time. Nuclear physicists during the Great Depression modified electron accelerators into giant 600-900 KV medical x-ray therapy machines and achieved one MV by 1937-1939. These were huge, complex, expensive, and unique to major academic and/or metropolitan centers. During World War II nuclear reactors superseded cyclotrons as efficient factories for few new radioisotopes, including "artificial radium". Few seemed interested in the latter for use in telegamma therapy until 1949-1951, when three competing teams from Canada and the USA designed telecobalt machines. From this competition, among then unknown innovators, emerged three future giants in radiation therapy: A.E.C.L., H. Johns, and G.H. Fletcher. The clinical application of telecobalt therapy was to revolutionize cancer care in community hospitals worldwide.

  19. Mobile robots for hazardous environments

    Energy Technology Data Exchange (ETDEWEB)

    Bains, N.; Scott, D.A.; Tran, K.; Campbell, T. (Atomic Energy of Canada Ltd., Mississauga, Ontario (Canada))

    1992-01-01

    This paper describes the development of a mobile robot ARK-2 (Autonomous Robot for Known Environments) that utilizes a number of sensors for navigation in a known relatively structured indoor environment. At present, there are robots that can be preprogrammed and that move along a specified path, but they use dead-reckoning to evaluate position at any point along their paths, and this can lead to major error accumulation through wheel slippage and running over unforeseen objects on the floor. The ARK-2 robot will have the intelligence to determine its position utilizing natural landmarks at any point along its path; it is this feature that gives ARK-2 its uniqueness as well as its ability to operate in an industrial environment. The project started in September 1991 and will last 4 yr. There are five organizations involved in the project: Ontario Hydro, Atomic Energy of Canada Limited (AECL) CANDU, US Nuclear Regulatory Commission (NRC), University of Toronto, and York University. Funding is provided by the organizations involved as well as the federal and provincial governments and PRECARN Associates, which is a nonprofit precompetitive research consortium made up of 38 members.

  20. Gamma radiation effects on the viscosity of green banana flour

    Energy Technology Data Exchange (ETDEWEB)

    Uehara, Vanessa B.; Inamura, Patricia Y.; Mastro, Nelida L. Del [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)], e-mail: vanessa.uehara@usp.br, e-mail: patyoko@yahoo.com, e-mail: nlmastro@ipen.br

    2009-07-01

    Banana (Musa sp) is a tropical fruits with great acceptability among consumers and produced in Brazil in a large scale. Bananas are not being as exploited as they could be in prepared food, and research could stimulate greater interest from industry. The viscosity characteristics and a product consistency can determine its acceptance by the consumer. Particularly the starch obtained from green banana had been studied from the nutritional point of view since the concept of Resistant Starch was introduced. Powder RS with high content of amylose was included in an approved food list with alleged functional properties in Brazilian legislation. Ionizing radiation can be used as a public health intervention measure for the control of food-borne diseases. Radiation is also a very convenient tool for polymer materials modification through degradation, grafting and crosslinking. In this work the influence of ionizing radiation on the rheological behavior of green banana pulp was investigated. Samples of green banana pulp flour were irradiated in a {sup 60}Co Gammacell 220 (AECL) with doses of 0 kGy,1 kGy, 3 kGy, 5 kGy and 10 kGy in glass recipients. After irradiation 3% and 5% aqueous dilution were prepared and viscosity measurements performed in a Brooksfield, model DVIII viscometer using spindle SC4-18 and SC4-31. There was a reduction of the initial viscosity of the samples as a consequence of radiation processing, being the reduction inversely proportional to the flour concentration. The polysaccharide content of the banana starch seems to be degraded by radiation in solid state as shown by the reduction of viscosity as a function of radiation dose. (author)

  1. Scenarios for the transmutation of actinides in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hyland, Bronwyn, E-mail: hylandb@aecl.ca [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada); Gihm, Brian, E-mail: gihmb@aecl.ca [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (Canada)

    2011-12-15

    With world stockpiles of used nuclear fuel increasing, the need to address the long-term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes to reduce the decay heat period or consumed as fuel in a CANDU(R) reactor. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle simplifies the fabrication and handling of active fuels. Online refuelling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio. This paper provides a summary of actinide transmutation schemes that have been studied in CANDU reactors at AECL, including the works performed in the past. The schemes studied include homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel. The transmutation schemes that are presented reflect several different partitioning schemes. Separation of americium, often with curium, from the other actinides enables targeted destruction of americium, which is a main contributor to the decay heat 100-1000 years after discharge from the reactor. Another scheme is group-extracted transuranic elements, in which all of the transuranic elements, plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm) are extracted together and then transmuted. This paper also addresses ways of utilizing the recycled uranium, another stream from the separation of spent nuclear fuel, in order to drive the transmutation of other actinides.

  2. The reduction of I{sub 2} by H{sub 2}O{sub 2} in aqueous solution

    Energy Technology Data Exchange (ETDEWEB)

    Ball, J.M.; Hnatiw, J.B. [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.; Sims, H.E. [AEA Technology, Harwell Laboratory, Didcot (United Kingdom)

    1996-12-01

    The reduction of iodine by hydrogen peroxide is an important process which leads to a lower amount of molecular iodine in irradiated solutions of iodide as the pH is increased. There is quite a large amount of information on the reaction now but no consensus in the literature on the mechanisms for reaction and the generally accepted mechanism does not appear to be correct. A number of studies of the kinetics of the reaction in the pH range 2-7 have been carried out where the iodine reduction process exhibited a 1/[H{sup +}]{sup 2} dependence consistent with the proposed mechanism which were attributed primarily to the reaction of H{sub 2}O{sub 2} with IO{sup -}. Deviations were observed in the pH range 6-7 and were explained by incorporating the reaction of I{sub 2}OH{sup -} with H{sub 2}O{sub 2}. In some other experiments it was suggested that the failure to maintain a 1/[H{sup +}]{sup 2} dependence at high pH was due to the iodine hydrolysis being rate determining. Data from an experimental program performed at AECL described in this paper confirms that the 1/[H{sup +}]{sup 2} dependence does not hold at high pH. These studies were carried out as a function of acid, iodide, peroxide and buffer concentration for three buffers, barbital, citrate and phosphate. This paper discuss two mechanisms which involve the formation of an HOOI intermediate in the rate determining step and which adequately describe the experimental data. (author) 4 figs., 2 tabs., 23 refs.

  3. Investigation of Thermal Hydraulics of a Nuclear Reactor Moderator

    Science.gov (United States)

    Sarchami, Araz

    A three-dimensional numerical modeling of the thermo hydraulics of Canadian Deuterium Uranium (CANDU) nuclear reactor is conducted. The moderator tank is a Pressurized heavy water reactor which uses heavy water as moderator in a cylindrical tank. The main use of the tank is to bring the fast neutrons to the thermal neutron energy levels. The moderator tank compromises of several bundled tubes containing nuclear rods immersed inside the heavy water. It is important to keep the water temperature in the moderator at sub-cooled conditions, to prevent potential failure due to overheating of the tubes. Because of difficulties in measuring flow characteristics and temperature conditions inside a real reactor moderator, tests are conducted using a scaled moderator in moderator test facility (MTF) by Chalk River Laboratories of Atomic Energy of Canada Limited (CRL, AECL). MTF tests are conducted using heating elements to heat tube surfaces. This is different than the real reactor where nuclear radiation is the source of heating which results in a volumetric heating of the heavy water. The data recorded inside the MTF tank have shown levels of fluctuations in the moderator temperatures and requires in depth investigation of causes and effects. The purpose of the current investigation is to determine the causes for, and the nature of the moderator temperature fluctuations using three-dimensional simulation of MTF with both (surface heating and volumetric heating) modes. In addition, three dimensional simulation of full scale actual moderator tank with volumetric heating is conducted to investigate the effects of scaling on the temperature distribution. The numerical simulations are performed on a 24-processor cluster using parallel version of the FLUENT 12. During the transient simulation, 55 points of interest inside the tank are monitored for their temperature and velocity fluctuations with time.

  4. SU-E-T-315: The Change of Optically Stimulated Luminescent Dosimeters (OSLDs) Sensitivity by Accumulated Dose and High Dose

    Energy Technology Data Exchange (ETDEWEB)

    Han, S; Jung, H; Kim, M; Ji, Y; Kim, K [University of Science and Technology, Daejeon (Korea, Republic of); Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of); Choi, S; Park, S; Yoo, H [Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of); Yi, C [Korea Research Institute of Standards and Science, Daejeon (Korea, Republic of)

    2014-06-01

    Purpose: The objective of this study is to evaluate radiation sensitivity of optical stimulated luminance dosimeters (OSLDs) by accumulated dose and high dose. Methods: This study was carried out in Co-60 unit (Theratron 780, AECL, and Canada) and used InLight MicroStar reader (Landauer, Inc., Glenwood, IL) for reading. We annealed for 30 min using optical annealing system which contained fluorescent lamps (Osram lumilux, 24 W, 280 ∼780 nm). To evaluate change of OSLDs sensitivity by repeated irradiation, the dosimeters were repeatedly irradiated with 1 Gy. And whenever a repeated irradiation, we evaluated OSLDs sensitivity. To evaluate OSLDs sensitivity after accumulated dose with 5 Gy, We irradiated dose accumulatively (from 1 Gy to 5 Gy) without annealing. And OSLDs was also irradiated with 15, 20, 30 Gy to certify change of OSLDs sensitivity after high dose irradiation. After annealing them, they were irradiated with 1Gy, repeatedly. Results: The OSLDs sensitivity increased up to 3% during irradiating seven times and decreased continuously above 8 times. That dropped by about 0.35 Gy per an irradiation. Finally, after 30 times irradiation, OSLDs sensitivity decreased by about 7%. For accumulated dose from 1 Gy to 5 Gy, OSLDs sensitivity about 1 Gy increased until 4.4% after second times accumulated dose compared with before that. OSLDs sensitivity about 1 Gy decreased by 1.6% in five times irradiation. When OSLDs were irradiated ten times with 1Gy after irradiating high dose (10, 15, 20 Gy), OSLDs sensitivity decreased until 6%, 9%, 12% compared with it before high dose irradiation, respectively. Conclusion: This study certified OSLDs sensitivity by accumulated dose and high dose. When irradiated with 1Gy, repeatedly, OSLDs sensitivity decreased linearly and the reduction rate of OSLDs sensitivity after high dose irradiation had dependence on irradiated dose.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  6. Spatial Distribution and Dynamics of Carbon-14 in a Wetland Ecosystem

    Energy Technology Data Exchange (ETDEWEB)

    Yankovich, Tamara L. [International Atomic Energy Agency, P.O. Box 100, 1400 Vienna (Austria); Carr, James; King-Sharp, K.; Doug Killey, R.W. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada); Robertson, Erin [201 21st Street East, Saskatoon, SK S7K 0B8 (Canada); Beresford, Nicholas A. [NERC Centre for Ecology and Hydrology, Lancaster Environment Center, Bailrigg, Lancaster, LA14AP (United Kingdom); School of Environment and Life Sciences, University of Salford, Manchester, M44WT (United Kingdom); Wood, Michael D. [School of Environment and Life Sciences, University of Salford, Manchester, M44WT (United Kingdom)

    2014-07-01

    There is significant interest in assessing the impact of {sup 14}C releases from nuclear facilities, radioactive waste management areas, and geologic disposal facilities. As a result, there is a general need to gain understanding of {sup 14}C dynamics, especially in complex interface ecosystems, such as wetlands. This paper summarizes the key findings of two studies undertaken in Duke Swamp, a circa 0.1 km{sup 2} area of wetland consisting of marsh, fen and swamp habitats, on the Atomic Energy of Canada Limited (AECL)'s Chalk River Laboratories Site. The swamp receives radionuclides, such as {sup 14}C and tritium, from an up-gradient waste management area. The first study was an extensive field sampling campaign, involving collection of surface vegetation at 69 locations on a 50 m x 50 m grid, to evaluate the spatial distribution of {sup 14}C in Duke Swamp. Representative receptor plants and animals, and corresponding environmental media (including air, soil, and plant) samples were then collected, as part of a second study, at a subset of six locations with {sup 14}C specific activities that spanned the range present in Duke Swamp and also represented the different wetland habitats occurring there. The highest specific activity concentrations in surface vegetation were highly localized, representing a surface area of only about 150 m{sup 2}. The spatial distribution of {sup 14}C in the swamp seemed to be at least partly accounted for by the physical attributes of the Duke Swamp habitat. In general, it was found that specific activities of {sup 14}C in biota tissues reflected those measured in surface vegetation collected from the same sampling location. Such information provides needed insight for biosphere assessments, as well as for the development of monitoring programs that demonstrate protection of biota in areas where exposure to {sup 14}C is elevated. (authors)

  7. Transmit-receive eddy current probes for defect detection and sizing in steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Obrutsky, L.S.; Cecco, V.S.; Sullivan, S.P. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1997-02-01

    Inspection of steam generator tubes in aging Nuclear Generating Stations is increasingly important. Defect detection and sizing, especially in defect prone areas such as the tubesheet, support plates and U-bend regions, are required to assess the fitness-for-service of the steam generators. Information about defect morphology is required to address operational integrity issues, i.e., risk of tube rupture, number of tubes at risk, consequential leakage. A major challenge continues to be the detection and sizing of circumferential cracks. Utilities around the world have experienced this type of tube failure. Conventional in-service inspection, performed with eddy current bobbin probes, is ineffectual in detecting circumferential cracks in tubing. It has been demonstrated in CANDU steam generators, with deformation, magnetite and copper deposits that multi-channel probes with transmit-receive eddy current coils are superior to those using surface impedance coils. Transmit-receive probes have strong directional properties, permitting probe optimization according to crack orientation. They are less sensitive to lift-off noise and magnetite deposits and possess good discrimination to internal defects. A single pass C3 array transmit-receive probe developed by AECL can detect and size circumferential stress corrosion cracks as shallow as 40% through-wall. Since its first trial in 1992, it has been used routinely for steam generator in-service inspection of four CANDU plants, preventing unscheduled shutdowns due to leaking steam generator tubes. More recently, a need has surfaced for simultaneous detection of both circumferential and axial cracks. The C5 probe was designed to address this concern. It combines transmit-receive array probe technology for equal sensitivity to axial and circumferential cracks with a bobbin probe for historical reference. This paper will discuss the operating principles of transmit-receive probes, along with inspection results.

  8. Residual Stress Measurements After Proof and Flight: ETP-0403

    Science.gov (United States)

    Webster, Ronald L..

    1997-01-01

    The intent of this testing was to evaluate the residual stresses that occur in and around the attachment details of a case stiffener segment that has been subjected to flight/recovery followed by proof loading. Not measured in this test were stresses relieved at joint disassembly due to out-of-round and interference effects, and those released by cutting the specimens out of the case segment. The test article was lightweight case stiffener segment 1U50715, S/N L023 which was flown in the forward stiffener position on flight SRM 14A and in the aft position on flight SRM24A. Both of these flights were flown with the 3 stiffener ring configuration. Stiffener L023 had a stiffener ring installed only on the aft stub in its first flight, and it had both rings installed on its second flight. No significant post flight damage was found on either flight. Finally, the segment was used on the DM-8 static test motor in the forward position. No stiffener rings were installed. It had only one proof pressurization prior to assignment to its first use, and it was cleaned and proof tested after each flight. Thus, the segment had seen 3 proof tests, two flight pressurizations, and two low intensity water impacts prior to manufacturing for use on DM-8. On DM-8 it received one static firing pressurization in the horizontal configuration. Residual stresses at the surface and in depth were evaluated by both the x-ray diffraction and neutron beam diffraction methods. The x-ray diffraction evaluations were conducted by Technology for Energy Corporation (TEC) at their facilities in Knoxville, TN. The neutron beam evaluations were done by Atomic Energy of Canada Limited (AECL) at the Chalk River Nuclear Laboratories in Ontario. The results showed general agreement with relatively high compressive residual stresses on the surface and moderate to low subsurface tensile residual stresses.

  9. The effect of gamma irradiation on the microbiological analysis on commercial functional Brazilian green banana flour

    Energy Technology Data Exchange (ETDEWEB)

    Taipina, Magda S.; Lamardo, Leda C.A.; Santos, Josefina S.; Silva Junior, Eneo A. da [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Balian, Simone C., E-mail: balian@usp.b [Universidade de Sao Paulo (USP), SP (Brazil). Fac. de Medicina Veterinaria e Zootecnia

    2011-07-01

    In Brazil, although it is qualified as a major world producers, however, the production losses are high. Nevertheless, these losses can be reduced by processing the fruit 'unsuitable' for consumption into products based on green banana (pulp, rind and flour). The green banana flour shows enhanced nutrition value, with higher contents of mineral, dietary fiber, resistant starch, and total phenolics, for use in Brazilian irradiated ready - to eat foods, such as bread, macaroni, among others. Food irradiation has been identified as safe technology to reduce risk of foodborne illness as part of high-quality food production, processing, handling and preparation. Food irradiation utilizes a source of ionizing energy that passes through food to destroy harmful bacteria and other organisms. Often referred to as 'cold pasteurization', food irradiation offers negligible loss of nutrients or sensory qualities in food as it does not substantially raise the temperature of the food during processing. The object of this work was to determine the effect of gamma irradiation on microbiological analyses of the: the number of mesophiles, total coliforms at 35 deg C, coliforms at 45 deg C, Staphylococcus aureus and Salmonella spp of the green banana flour, commercially found in the Brazilian market. The microbiological analyses were carried out in conformity with the methodologies described at the Faculty of Veterinary Medicine, according to the current legislation. Irradiation was performed in a {sup 60}Co Gammacell 220 (AECL) source, with dose of 3kGy at IPEN/CNEN-SP. In samples of Brazilian green banana flour, irradiated at 3 kGy, the growth of all microorganisms (mesophiles, total coliforms at 35 deg C, coliform at 45 deg C and Staphylococcus coagulase positive) were reduced. As a result, the application of the irradiation technique may be recommended to enhance the food safety. (author)

  10. Parametric studies of radiolytic oxidation of iodide solutions with and without paint: comparison with code calculations

    Energy Technology Data Exchange (ETDEWEB)

    Poletiko, C.; Hueber, C. [Inst. de Protection et de Surete Nucleaire, C.E. Cadarache, St. Paul-lez-Durance (France); Fabre, B. [CISI, C.E. Cadarache, St. Paul-lez-Durance (France)

    1996-12-01

    In case of severe nuclear accident, radioactive material may be released into the environment. Among the fission products involved, are the very volatile iodine isotopes. However, the chemical forms are not well known due to the presence of different species in the containment with which iodine may rapidly react to form aerosols, molecular iodine, hydroiodic acid and iodo-organics. Tentative explanations of different mechanisms were performed through benchscale tests. A series of tests has been performed at AEA Harwell (GB) to study parameters such as pH, dose rate, concentration, gas flow rate, temperature in relation to molecular iodine production, under dynamic conditions. Another set of tests has been performed in AECL Whiteshell (CA) to study the behaviour of painted coupons, standing in gas phase or liquid phase or both, with iodine compounds under radiation. The purpose of our paper is to synthesize the data and compare the results to the IODE code calculation. Some parameters of the code were studied to fit the experimental result the best. A law, concerning the reverse reaction of iodide radiolytic oxidation, has been proposed versus: pH, concentrations and gas flow-rate. This law does not apply for dose rate variations. For the study of painted coupons, it has been pointed out that molecular iodine tends to be adsorbed or chemically absorbed on the surface in gas phase, but the mechanism should be more sophisticated in the aqueous phase. The iodo-organics present in liquid phase tend to be partly or totally destroyed by oxidation under radiation (depending upon the dose delivered). These points are discussed. (author) 18 figs., 3 tabs., 15 refs.

  11. Monte Carlo study of MLC fields for cobalt therapy machine

    Directory of Open Access Journals (Sweden)

    Komanduri M Ayyangar

    2014-01-01

    Full Text Available An automated Multi-Leaf Collimator (MLC system has been developed as add-on for the cobalt-60 teletherapy machines available in India. The goal of the present computational study is to validate the MLC design using Monte Carlo (MC modeling. The study was based on the Kirloskar-supplied Phoenix model machines that closely match the Atomic Energy of Canada Limited (AECL theratron-80 machine. The MLC is a retrofit attachment to the collimator assembly, with 14 non-divergent leaf pairs of 40 mm thick, 7 mm wide, and 150 mm long tungsten alloy plates with rounded edges and 20 mm tongue and 2 mm groove in each leaf. In the present work, the source and collimator geometry has been investigated in detail to arrive at a model that best represents the measured dosimetric data. The authors have studied in detail the proto-I MLC built for cobalt-60. The MLC field sizes were MC simulated for 2 × 2 cm 2 to 14 × 14 cm 2 square fields as well as irregular fields, and the percent depth dose (PDD and profile data were compared with ROPS† treatment planning system (TPS. In addition, measured profiles using the IMATRIXX system‡ were also compared with the MC simulations. The proto-I MLC can define radiation fields up to 14 × 14 cm΂ within 3 mm accuracy. The maximum measured leakage through the leaf ends in closed condition was 3.4% and interleaf leakage observed was 7.3%. Good agreement between MC results, ROPS and IMATRIXX results has been observed. The investigation also supports the hypothesis that optical and radiation field coincidence exists for the square fields studied with the MLC. Plots of the percent depth dose (PDD data and profile data for clinically significant irregular fields have also been presented. The MC model was also investigated to speed up the calculations to allow calculations of clinically relevant conformal beams.

  12. Continental glaciation and its potential impact on a used-fuel disposal vault in the Canadian Shield

    Energy Technology Data Exchange (ETDEWEB)

    Ates, Y.; Bruneau, D.; Ridgway, W.R

    1997-09-01

    AECL has been assessing the concept of nuclear fuel waste disposal in a vault excavated at a depth ranging between 500 m and 1000 m in a plutonic rock mass of the Canadian Shield. Glaciation is a natural process that has occurred in the past, and is likely to occur in the future, thus causing changes in the loading conditions on the rock mass hosting the disposal vault. Because the rock mass is a natural barrier to the migration of radionuclides, it is important to evaluate its integrity under load changes caused by the glaciation process. Assuming that the magnitude and extent of the future glaciation will be similar to those of the past, we have reviewed published data pertaining to the last continental ice sheet that covered a large area of North America. Estimates have been madefor the magnitude of stresses due to ice sheet loading for a vault located at depths of 500 to 1000 m. These analyses have shown that the uniform loading of a continental ice sheet would reduce the deviatoric stresses in the Canadian Shield, creating more favourable conditions than those existing at the present time, namely, high horizontal stresses. The effects of surface erosion and increase in the in-situ shear stresses have also been examined. Based on the existing data and structural modelling studies, there would be no significant structural effect on a disposal vault located at 1000-m depth in a plutonic rock. At its maximum size, an ice sheet comparable to the Laurentide ice sheet could reactivate the faults and fracture zones along the perimeter areas. Our analyses have been based on fully drained conditions only. At a potential disposal site, it would be important also to consider the potential for excess pore pressure in the analyses. (author)

  13. Dry spent fuel storage with the MACSTOR system

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F. [Atomic Energy of Canada Ltd., Montreal, PQ (Canada). CANDU Operations

    1996-10-01

    Atomic Energy of Canada Limited (AECL), and Transnuclear Inc. (TNI) began in 1989 the development of the concrete spent fuel storage system, called MACSTOR (Modular Air-Cooled Canister STORage) for use with LWR spent fuel assemblies. It is a hybrid system which combines the operational economies of metal cask technology with the capital economies of concrete technology. The MACSTOR Module is a monolithic, shielded concrete vault structure that can accommodate up to 20 spent fuel canisters. Each canister typically holds up to 21 PWR or 44 BWR spent fuel assemblies with a nominal fuel burn up rate of 40,000 MWD/MTU and a 7 year minimum cooling period. The structure is passively cooled by natural convection through an array of inlet and outlet gratings and galleries serving a central plenum where the (vertically) stored canisters are located. The canisters are continuously monitored by means of a pressure monitoring system developed by TNI. Thus, the utility can be assured of both positive cooling of the fuel and verification of the integrity of the fuel confinement boundary. The structure is seismically designed and is capable of withstanding site design basis accident events. The MACSTOR system includes the storage module(s), an overhead gantry system for cask handling, a transfer cask for moving fuel from wet to dry storage and a cask transporter. The canister and transfer cask designs are based on Transnuclear transport cask designs and proven hot cell transfer cask technology, adapted to requirements for on-site spent fuel storage. The MACSTOR system can economically address a wide range of storage capacity requirements. The modular concept allows for flexibility in determining each module`s capacity. Starting with 8 canisters, the capacity can be increased by increments of 4 up to 20 canisters. The MACSTOR system is also flexible in accommodating the various spent fuel types from such reactors as VVER-440, VVER-1000 and RBMK 1500. (J.P.N.)

  14. Corrosion Tests of Steel Bar in Concrete under High Temperature by Salt Solution

    Energy Technology Data Exchange (ETDEWEB)

    Lee, ChangMin; Lee, YoonHee; Lee, KunJai [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Lee, KyungHo; Jang, HyunKie; Kim, JeongMook [Nuclear Engineering and Technology Institute, Daejeon (Korea, Republic of)

    2007-07-01

    The saturation of South Korea's at-reactor (AR) spent fuel storage pools has created necessity for additional spent fuel storage capacity. The utility company (Korea Hydro and Nuclear Power Company) is planning to construct a dry storage facility, which offers advantages such as no generation of second time radioactive waste, relatively low operational cost, and a short construction period. Spent nuclear fuel from CANDU will be stored in MACSTOR-400. MACSTOR-400 developed by KHNP and AECL is a new dry storage module to replace Korea's existing concrete silo. This module composed of reinforced concrete has a capacity of 446MgU, twice the MACSTOR 200. Concrete has been used in the construction of nuclear facilities because of two primary properties, its structural strength and its ability to shield radiation. The use of concrete in nuclear facilities for containment and shielding of radiation and radioactive materials has made its performance crucial for the safe operation of the facility. Corrosion of reinforcing bars deteriorates the concrete structures and reduces their service life. Because spent fuel dry storage will be constructed near seashore, the reinforced concrete components and structures must withstand the damage due to salt attack under high temperature that is emitted by spent fuel. It can be noted that the temperature considerably affects degradation of reinforced concrete structure. However, there are very few examination examples to make clear the influence of the temperature. To obtain the basic material properties at high temperature and evaluate life time of spent fuel dry storage facility, the following test is now in progress.

  15. The buffer/container experiment design and construction report

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, N.A.; Wan, A.W.L.; Roach, P.J

    1998-03-01

    The Buffer/Container Experiment was a full-scale in situ experiment, installed at a depth of 240 m in granitic rock at AECL's Underground Research Laboratory (URL). The experiment was designed to examine the performance of a compacted sand-bentonite buffer material under the influences of elevated temperature and in situ moisture conditions. Buffer material was compacted in situ into a 5-m-deep, 1.24-m-diameter borehole drilled into the floor of an excavation. A 2.3-m long heater, representative of a nuclear fuel waste container, was placed within the buffer, and instrumentation was installed to monitor changes in buffer moisture conditions, temperature and stress. The experiment was sealed at the top of the borehole and restrained against vertical displacement. Instrumentation in the rock monitored pore pressures, temperatures and rock displacement. The heater was operated at a constant power of 1200 W, which provided a heater skin temperature of approximately 85 degrees C. Experiment construction and installation required two years, followed by two and a half years of heater operation and two years of monitoring the rock conditions during cooling. The construction phase of the experiment included the design, construction and testing of a segmental heater and controller, geological and hydrogeological characterization of the rock, excavation of the experiment room, drilling of the emplacement borehole using high pressure water, mixing and in situ compaction of buffer material, installation of instrumentation in the rock, buffer and on the heater, and the construction of concrete curb and steel vertical restraint system at the top of emplacement borehole. Upon completion of the experiment, decommissioning sampling equipment was designed and constructed and sampling methods were developed which allowed approximately 2000 samples of buffer material to be taken over a 12-day period. Quality assurance procedures were developed for all aspects of experiment

  16. Low LET radiolysis escape yields for reducing radicals and H2 in pressurized high temperature water

    Science.gov (United States)

    Sterniczuk, Marcin; Yakabuskie, Pamela A.; Wren, J. Clara; Jacob, Jasmine A.; Bartels, David M.

    2016-04-01

    Low Linear Energy Transfer (LET) radiolysis escape yields (G values) are reported for the sum (G(radH)+G(e-)aq) and for G(H2) in subcritical water up to 350 °C. The scavenger system 1-10 mM acetate/0.001 M hydroxide/0.00048 M N2O was used with simultaneous mass spectroscopic detection of H2 and N2 product. Temperature-dependent measurements were carried out with 2.5 MeV electrons from a van de Graaff accelerator, while room temperature calibration measurements were done with a 60Co gamma source. The concentrations and dose range were carefully chosen so that initial spur chemistry is not perturbed and the N2 product yield corresponds to those reducing radicals that escape recombination in pure water. In comparison with a recent review recommendation of Elliot and Bartels (AECL report 153-127160-450-001, 2009), the measured reducing radical yield is seven percent smaller at room temperature but in fairly good agreement above 150 °C. The H2 escape yield is in good agreement throughout the temperature range with several previous studies that used much larger radical scavenging rates. Previous analysis of earlier high temperature measurements of Gesc(radOH) is shown to be flawed, although the actual G values may be nearly correct. The methodology used in the present report greatly reduces the range of possible error and puts the high temperature escape yields for low-LET radiation on a much firmer quantitative foundation than was previously available.

  17. Neutron spectrometry and dosimetry study at two research nuclear reactors using Bonner sphere spectrometer (BSS), rotational spectrometer (ROSPEC) and cylindrical nested neutron spectrometer (NNS).

    Science.gov (United States)

    Atanackovic, J; Matysiak, W; Hakmana Witharana, S S; Aslam, I; Dubeau, J; Waker, A J

    2013-01-01

    Neutron spectrometry and subsequent dosimetry measurements were undertaken at the McMaster Nuclear Reactor (MNR) and AECL Chalk River National Research Universal (NRU) Reactor. The instruments used were a Bonner sphere spectrometer (BSS), a cylindrical nested neutron spectrometer (NNS) and a commercially available rotational proton recoil spectrometer. The purposes of these measurements were to: (1) compare the results obtained by three different neutron measuring instruments and (2) quantify neutron fields of interest. The results showed vastly different neutron spectral shapes for the two different reactors. This is not surprising, considering the type of the reactors and the locations where the measurements were performed. MNR is a heavily shielded light water moderated reactor, while NRU is a heavy water moderated reactor. The measurements at MNR were taken at the base of the reactor pool, where a large amount of water and concrete shielding is present, while measurements at NRU were taken at the top of the reactor (TOR) plate, where there is only heavy water and steel between the reactor core and the measuring instrument. As a result, a large component of the thermal neutron fluence was measured at MNR, while a negligible amount of thermal neutrons was measured at NRU. The neutron ambient dose rates at NRU TOR were measured to be between 0.03 and 0.06 mSv h⁻¹, while at MNR, these values were between 0.07 and 2.8 mSv h⁻¹ inside the beam port and <0.2 mSv h⁻¹ between two operating beam ports. The conservative uncertainty of these values is 15 %. The conservative uncertainty of the measured integral neutron fluence is 5 %. It was also found that BSS over-responded slightly due to a non-calibrated response matrix.

  18. Fabrication of Zr-2.5Nb pressure tubes to minimize the harmful effects of trace elements

    Energy Technology Data Exchange (ETDEWEB)

    Theaker, J.R.; Coleman, C.E. [AECL Research, Chalk River, Ontario (Canada). Chalk River Labs.; Choubey, R. [AECL Research, Pinawa, Manitoba (Canada). Whiteshell Labs.; Moan, G.D. [AECL CANDU, Mississauga, Ontario (Canada); Aldridge, S.A. [Nu-Tech Precision Metals Inc., Arnprior, Ontario (Canada); Davis, L.; Graham, R.A. [Teledyne Wah Chang Albany, OR (United States)

    1994-12-31

    Trace elements can reduce the fracture resistance of Zr-2.5Nb pressure tubes. The effects of hydrogen as hydrides and oxygen as an alloy-strengthening agent are well known, but the contributions of carbon, phosphorus, chlorine, and segregated oxygen have only recently been recognized. Carbides and phosphides are brittle particles, while chlorine segregates to form planes of weakness that produce fissures on the fracture face of test specimens. A high density of fissures is associated with low toughness. With long hold times in the ({alpha} + {beta}) region, oxygen partitions into the {alpha}-grains; such grains are hard and, if they survive fabrication, may reduce the toughness of the finished tube. Through a cooperative program involving AECL and the manufacturers, a series of manufacturing innovations and controls has been introduced that minimizes these harmful effects. Hydrogen is present in the zirconium sponge as water, can be absorbed at each stage of tube fabrication, and needs to be carefully controlled, particularly during ingot breakdown and subsequent forging. Hydrogen concentrations in finished tubes have been reduced by a factor of three through the optimization of manufacturing processes and the implementation of new technology. Multiple vacuum arc melting, use of selected raw materials, and intermediate ingot surface conditioning have resulted in much improved fracture toughness through the reduction of chlorine and phosphorus concentrations. Optimum distribution of oxygen may be achieved through changes to the extrusion process cycle. An understanding of the Zr-2.5Nb-C phase diagram, particularly the solubility of carbon at low concentrations, has resulted in the specification of a lower carbon concentration.

  19. Assessment of ASSERT-PV for prediction of critical heat flux in CANDU bundles

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-15

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for CHF prediction. • CANDU 28-, 37- and 43-element bundle CHF experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of CHF model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against five full-scale CANDU bundle experiments conducted in 1990s and in 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element (CANFLEX) bundles. A total of 15 CHF test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for CANDU reactors. Code predictions of channel dryout power and axial and radial CHF locations were compared against measurements from the SL CHF tests to quantify the code prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, the sensitivity studies evaluated the contribution of each CHF model change or enhancement to the improvement in CHF prediction. Overall, the assessment demonstrated significant improvement in prediction of channel dryout power and axial and radial CHF locations in horizontal fuel channels containing CANDU bundles.

  20. Assessment of ASSERT-PV for prediction of post-dryout heat transfer in CANDU bundles

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for PDO sheath temperature prediction. • CANDU 28-, 37- and 43-element bundle PDO experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of PDO model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles.

  1. Irradiation of mangoes as a quarantine treatment

    Energy Technology Data Exchange (ETDEWEB)

    Bustos R, M.E.; Enkerlin H, W.; Toledo A, J.; Reyes F, J.; Casimiro G, A

    1991-06-15

    This research project was conducted following guidelines of research protocols for post-harvest treatments developed by the United States Department of Agriculture CUSA. Laboratory bioassays included the irradiation of mangoes infested with third instar larvae of Anastrepha serpentina (Wied), A. ludens (Loew), A. obliqua (Macquart) and Ceratitis capitata (Wied) , at doses from 10 to 250 Gy. Irradiation doses were applied using a Co-60 AECL Model JS-7400 irradiator. The design was chosen to obtain a maximum to minimum ratio equal to, or less than, 1.025. C. capitata was the species most tolerant to irradiation. A dose of 60 Gy applied to third instar fruit fly larvae sterilized this species and prevented emergence of adults of the other three species. A dose of 250 Gy was required to prevent emergence of C. capitata. In fertility tests using emerged adults of A . Iudens, and A. obliqua a dose of 30 Gy gave 45 % and 27 % fertility, respectively. Adults of A. serpentina that emerged, died before reaching sexual maturity. The confirmatory tests, at probit-9 security level, were done at 100 Gy for the three species of Anastrepha and at 150 Gy for C. capitata. The quality of mangoes irradiated up to 1000 Gy was evaluated by chemical, physiological, and sensorial tests. The determination of vitamin C indicated that there was no loss of the nutritive value of the fruit. It also was observed that fruit metabolism was not accelerated since no significant increase in respiration or transpiration was registered and consumers accepted both treated and untreated fruit in the same way. (Author)

  2. ChemAND - a system health monitor for plant chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Turner, C.W.; Mitchel, G.R.; Tosello, G.; Balakrishnan, P.V.; McKay, G.; Thompson, M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Dundar, Y.; Bergeron, M.; Laporte, R. [Hydro-Quebec, Groupe Chimie, Centrale Nucleaire Gentilly-2, Gentilly, Quebec (Canada)

    2001-03-01

    Effective management of plant systems throughout their lifetime requires much more than data acquisition and display - it requires that the plant's system health be continually monitored and managed. AECL has developed a System Health Monitor called ChemAND for CANDU plant chemistry. ChemAND, a Chemistry ANalysis and Diagnostic system, monitors key chemistry parameters in the heat transport system, moderator-cover gas, annulus gas, and the steam cycle during full-power operation. These parameters can be used as inputs to models that calculate the effect of current plant operating conditions on the present and future health of the system. Chemistry data from each of the systems are extracted on a regular basis from the plant's Historical Data Server and are sorted according to function, e.g., indicators for condenser in-leakage, air in-leakage, heavy water leakage into the annulus gas, fuel failure, etc. Each parameter is conveniently displayed and is trended along with its alarm limits. ChemAND currently includes two analytical models developed for the balance-of-plant. The first model, ChemSolv, calculates crevice chemistry conditions in the steam generator (SG) from either the SG blowdown chemistry conditions or from a simulated condenser leak. This information can be used by plant staff to evaluate the susceptibility of the SG tubes to crevice corrosion. ChemSolv also calculates chemistry conditions throughout the steam-cycle system as determined by the transport of volatile species such as ammonia, hydrazine, morpholine, and oxygen. The second model, SLUDGE, calculates the deposit loading and distribution in the SG as a function of time, based on concentrations of corrosion product in the final feedwater for both normal and start-up conditions. Operations personnel can use this information to predict where to inspect and when to clean. (author)

  3. Doses from aquatic pathways in CSA-N288.1: deterministic and stochastic predictions compared

    Energy Technology Data Exchange (ETDEWEB)

    Chouhan, S.L.; Davis, P

    2002-04-01

    The conservatism and uncertainty in the Canadian Standards Association (CSA) model for calculating derived release limits (DRLs) for aquatic emissions of radionuclides from nuclear facilities was investigated. The model was run deterministically using the recommended default values for its parameters, and its predictions were compared with the distributed doses obtained by running the model stochastically. Probability density functions (PDFs) for the model parameters for the stochastic runs were constructed using data reported in the literature and results from experimental work done by AECL. The default values recommended for the CSA model for some parameters were found to be lower than the central values of the PDFs in about half of the cases. Doses (ingestion, groundshine and immersion) calculated as the median of 400 stochastic runs were higher than the deterministic doses predicted using the CSA default values of the parameters for more than half (85 out of the 163) of the cases. Thus, the CSA model is not conservative for calculating DRLs for aquatic radionuclide emissions, as it was intended to be. The output of the stochastic runs was used to determine the uncertainty in the CSA model predictions. The uncertainty in the total dose was high, with the 95% confidence interval exceeding an order of magnitude for all radionuclides. A sensitivity study revealed that total ingestion doses to adults predicted by the CSA model are sensitive primarily to water intake rates, bioaccumulation factors for fish and marine biota, dietary intakes of fish and marine biota, the fraction of consumed food arising from contaminated sources, the irrigation rate, occupancy factors and the sediment solid/liquid distribution coefficient. To improve DRL models, further research into aquatic exposure pathways should concentrate on reducing the uncertainty in these parameters. The PDFs given here can he used by other modellers to test and improve their models and to ensure that DRLs

  4. The International Cooperation and Partnership, Keystones for Engineering and Procurement for Cernavoda NPP Unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Stiopol, Mihaela; Vatamanu, Mariana; Bucur, Cristina [Nuclearelectrica SA, 65 Polona Street, 010494 Bucharest (Romania)

    2008-07-01

    protection and industrial safety achievements. According to the contract signed in May 2001, between SNN S.A. and AECL and ANSALDO, a 'Management Team', formed by representatives and specialists from Canada - AECL, Italy - ANSALDO and Romania - 'Nuclearelectrica' was in charge of the engineering, procurement, construction and commissioning processes for Cernavoda NPP Unit 2. More than 1600 workers were employed by 'Management Team' of which 110 are AECL experts from Canada, 80 ANSALDO employees from Italy and 700 are 'Nuclearelectrica' permanent employees. A similar organization was applied successfully in Unit 1 Project. After the turnover of Unit 2 from 'Management Team' to 'Nuclearelectrica', in 2007, the U2 operation was integrated into a 'Two-Units' organization (together with U1). The present paper will present the prerequisites of the successful operation of Unit 1, and will develop the main steps in the Cernavoda NPP-Unit 2 completion process as well as the corporate organization in the view of an integrated two units structure on Cernavoda site. (authors)

  5. Characterization of hydrides and delayed hydride cracking in zirconium alloys

    Science.gov (United States)

    Fang, Qiang

    This thesis tries to fill some of the missing gaps in the study of zirconium hydrides with state-of-art experiments, cutting edge tomographical technique, and a novel numerical algorithm. A new hydriding procedure is proposed. The new anode material and solution combination overcomes many drawbacks of the AECLRTM hydriding method and leads to superior hydriding result compared to the AECL RTM hydriding procedure. The DHC crack growth velocity of as-received Excel alloy and Zr-2.5Nb alloy together with several different heat treated Excel alloy samples are measured. While it already known that the DHC crack growth velocity increases with the increase of base metal strength, the finding that the transverse plane is the weaker plane for fatigue crack growth despite having higher resistance to DHC crack growth was unexpected. The morphologies of hydrides in a coarse grained Zircally-2 sample have been studied using synchrotron x-rays at ESRF with a new technique called Diffraction Contrast Tomography that uses simultaneous collection of tomographic data and diffraction data to determine the crystallographic orientation of crystallites (grains) in 3D. It has been previously limited to light metals such as Al or Mg (due to the use of low energy x-rays). Here we show the first DCT measurements using high energy x-rays (60 keV), allowing measurements in zirconium. A new algorithm of a computationally effcient way to characterize distributions of hydrides - in particular their orientation and/or connectivity - has been proposed. It is a modification of the standard Hough transform, which is an extension of the Hough transform widely used in the line detection of EBSD patterns. Finally, a basic model of hydrogen migration is built using ABAQUS RTM, which is a mature finite element package with tested modeling modules of a variety of physical laws. The coupling of hydrogen diffusion, lattice expansion, matrix deformation and phase transformation is investigated under

  6. A Investigation of Radiotherapy Electron Beams Using Monte Carlo Techniques

    Science.gov (United States)

    Ding, George X.

    1995-01-01

    Radiotherapy electron beams are more complicated than photon beams due to variations in the beam production, the scattering of low-energy electrons, and the presence contaminant photons. The detailed knowledge of a radiotherapy beam is essential to an accurate calculation of dose distribution for a treatment planning system. This investigation aims to enhance our understanding of radiotherapy beams by focusing on electron beams used in radiotherapy. It starts with a description of the Monte Carlo simulation code, BEAM, and a detailed simulation of an accelerator head to obtain realistic radiotherapy beams. The simulation covers electron beams from various accelerators, including the NRC research accelerator, the NPL (UK), accelerator, A Varian Clinac 2100C, a Philips SL75-20, a Siemens KD2, an AECL Therac 20, and a Scanditronix MM50. The beam energies range from 4 to 50 MeV. The EGS4 user code, BEAM, is extensively benchmarked against experiment by comparing calculated dose distributions with measured dose distributions in water. The simulated beams are analyzed to obtain the characteristics of various electron beams from a variety of accelerators. The simulated beams are also used as inputs to calculate the following parameters: the mean electron energy, the most probable energy, the energy-range relationships, the depth-scaling factor to convert depths in plastic to water-equivalent depths, the water-to-air stopping-power ratios, and the electron fluence correction factors used to convert dose measured in plastics to dose in water. These parameters are essential for electron beam dosimetry. The results from this study can be applied in cancer clinics to improve the accuracy of the absolute dosimetry. The simulation also provides information about the backscatter into the beam monitor chamber, and predicts the influence on the beam output factors. This investigation presents comprehensive data on the clinical electron beams, and answers many questions which could

  7. Aspects of the physics and chemistry of water radiolysis by fast neutrons and fast electrons in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    McCracken, D.R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Tsang, K.T. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Laughton, P.J

    1998-09-01

    Detailed radiation physics calculations of energy deposition have been done for the coolant of CANDU reactors and Pressurized Water Reactors (PWRs). The geometry of the CANDU fuel channel was modelled in detail. Fluxes and energy-deposition rates for neutrons, recoil ions, photons, and fast electrons have been calculated using MCNP4B, WIMS-AECL, and specifically derived energy-transfer factors. These factors generate the energy/flux spectra of recoil ions from fast-neutron energy/flux spectra. The energy spectrum was divided into 89 discrete ranges (energy bins).The production of oxidizing species and net coolant radiolysis can be suppressed by the addition of hydrogen to the coolant of nuclear reactors. It is argued that the net dissociation of coolant by gamma rays is suppressed by lower levels of excess hydrogen than when dissociation is by ion recoils. This has consequences for the modelling of coolant radiolysis by homogeneous kinetics. More added hydrogen is required to stop water radiolysis by recoil ions acting alone than if recoil ions and gamma rays acted concurrently in space and time. Homogeneous kinetic models and experimental data suggest that track overlap is very inefficient in providing radicals from gamma-ray tracks to recombine molecular products in ion-recoil tracks. An inhomogeneous chemical model is needed that incorporates ionizing-particle track structure and track overlap. Such a model does not yet exist, but a number of limiting cases using homogeneous kinetics are discussed. There are sufficient uncertainties and contradictions in the data relevant to the radiolysis of reactor coolant that the relatively high CHC's (critical hydrogen concentration) observed in NRU reactor experiments (compared to model predictions) may be explainable by errors in fundamental data and understanding of water radiolysis under reactor conditions. The radiation chemistry program at CRL has been focused to generate quantitative water-radiolysis data in a

  8. The near boiling reactor : conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

    Energy Technology Data Exchange (ETDEWEB)

    Cole, C.J.P

    2005-07-01

    Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the 'Victoria' Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96{sup o}C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has

  9. The near boiling reactor: Conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

    Science.gov (United States)

    Cole, Christopher J. P.

    Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the Victoria Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96°C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has highlighted additional areas

  10. Evaluation of safety margins during dry storage of CANDU fuel in MACSTOR/KN-400 module

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, R.; Shill, R. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada); Lee, K.-H.; Chung, S.-H.; Yoon, J.-H.; Choi, B.-I.; Lee, H.-Y.; Song, M.-J. [KHNP, Nuclear Environment Technology Inst., Taejon (Korea, Republic of)

    2004-07-01

    This paper covers an evaluation of the available safety margin against fuel bundle degradation during dry storage of CANDU spent fuel bundles in a MACSTOR/KN-400 module, considering normal, off-normal and postulated accidental conditions. Korea Hydro and Nuclear Power (KHNP), in collaboration with Atomic Energy of Canada Limited (AECL), are developing a new module for the dry storage of spent fuel from the four CANDU 6 nuclear reactors at the Wolsong site in South Korea. The module provides the benefit of occupying significantly less area than the concrete canisters presently used. The modules are designed for a minimum service life of 50 years. During that period, the spent fuel bundles shall be safely stored. This imposes that failure of a fuel bundle element or unacceptable degradation of an existing defect (from reactor operation) does not occur during the dry storage period. The fuel bundles are stored in an air-filled fuel basket that releases 365 Watts on average and a maximum of 390 Watts when rare fuel loading conditions are postulated. In addition, specific accidental air flow cooling conditions are postulated that consist of 100% blockage of all air inlets on one side of the module. These conditions can generate a peak daily fuel temperature of up to 155{sup o}C during a reference hot summer day during the first year of operation. The fuel temperature decreases over the years and also fluctuates due to daily and seasonal temperature variations. At this temperature, fuel elements with intact Zircaloy sheathing will not experience damage. However, for the few fuel bundle elements that are non-leaktight (less than 1 per 37,000), some re-oxidation of UO{sub 2} into higher oxides such as U{sub 3}O{sub 7} / U{sub 4}O{sub 9} and U{sub 3}O{sub 8} will occur. This latter form of Uranium oxide is undesirable due to its lower density that results in a volumetric increase of the pellet that can overstress the fuel element sheathing. The level of fuel pellet

  11. CORBA and MPI-based 'backbone' for coupling advanced simulation tools

    Energy Technology Data Exchange (ETDEWEB)

    Seydaliev, M.; Caswell, D., E-mail: marat.seydaliev@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2014-12-01

    There is a growing international interest in using coupled, multidisciplinary computer simulations for a variety of purposes, including nuclear reactor safety analysis. Reactor behaviour can be modeled using a suite of computer programs simulating phenomena or predicting parameters that can be categorized into disciplines such as Thermalhydraulics, Neutronics, Fuel, Fuel Channels, Fission Product Release and Transport, Containment and Atmospheric Dispersion, and Severe Accident Analysis. Traditionally, simulations used for safety analysis individually addressed only the behaviour within a single discipline, based upon static input data from other simulation programs. The limitation of using a suite of stand-alone simulations is that phenomenological interdependencies or temporal feedback between the parameters calculated within individual simulations cannot be adequately captured. To remove this shortcoming, multiple computer simulations for different disciplines must exchange data during runtime to address these interdependencies. This article describes the concept of a new framework, which we refer to as the 'Backbone', to provide the necessary runtime exchange of data. The Backbone, currently under development at AECL for a preliminary feasibility study, is a hybrid design using features taken from the Common Object Request Broker Architecture (CORBA), a standard defined by the Object Management Group, and the Message Passing Interface (MPI), a standard developed by a group of researchers from academia and industry. Both have well-tested and efficient implementations, including some that are freely available under the GNU public licenses. The CORBA component enables individual programs written in different languages and running on different platforms within a network to exchange data with each other, thus behaving like a single application. MPI provides the process-to-process intercommunication between these programs. This paper outlines the different

  12. Speciation of iodine (I-127) in the natural environment around Canadian CANDU sites

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, S.J.; Kotzer, T.G.; Chant, L.A

    2001-06-01

    In Canada, very little data is available regarding the concentrations and chemical speciation of iodine in the environment proximal and distal to CANDU Nuclear Power Generating Stations (NPGS). In the immediate vicinity of CANDU reactors, the short-lived iodine isotope {sup 131}I (t{sub 1/2} = 8.04 d), which is produced from fission reactions, is generally below detection and yields little information about the environmental cycling of iodine. Conversely, the fission product {sup 129}I has a long half-life (t{sub 1/2} = 1.57x10{sup 7} y) and has had other anthropogenic inputs (weapons testing, nuclear fuel reprocessing) other than CANDU over the past 50 years. As a result, the concentrations of stable iodine ({sup 127}I) have been used as a proxy. In this study, a sampling system was developed and tested at AECL's Chalk River Laboratories (CRL) to collect and measure the particulate and gaseous inorganic and organic fractions of stable iodine ({sup 127}I) in air and associated organic and inorganic reservoirs. Air, vegetation and soil samples were collected at CRL, and at Canadian CANDU Nuclear Power Generating Stations (NPGS) at OPG's (Ontario Power Generation) Pickering (PNGS) and Darlington NPGS (DNGS) in Ontario, as well as at NB Power's Pt. Lepreau NPGS in New Brunswick. The concentrations of particulate and inorganic iodine in air at CRL were extremely low, and were often found to be below detection. The concentrations are believed to be at this level because the sediments in the CRL area are glacial fluvial and devoid of marine ionic species, and the local atmospheric conditions at the sampling site are very humid. Concentrations of a gaseous organic species were comparable to worldwide levels. The concentrations of particulate and inorganic iodine in air were also found to be low at PNGS and DNGS, which may be attributed to reservoir effects of the large freshwater lakes in southern Ontario, which might serve to dilute the atmospheric iodine

  13. Ametryne degradation by ionizing radiation

    Energy Technology Data Exchange (ETDEWEB)

    Andrade, Debora Cristina de; Mori, Manoel Nunes; Duarte, Celina Lopes [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: deboracandrade@globo.com; mnmori@ipen.br; clduarte@ipen.br; Melo, Rita Paiva [Technological and Nuclear Institute (ITN), Sacavem (Portugal)]. E-mail: ritamelo@itn.pt

    2007-07-01

    Ametryne may be released to the environment during its manufacture, transport, storage, formulation and use as selective herbicide for the control of annual broadleaf and grass weeds. It is applied as an aqueous suspension for preemergence or post-directed applications on crops. Depending on the pesticide formulation and type of application, ametryne residues may be detectable in water, soil and on the surfaces for months or years. The herbicide used to this study was Ametryne (commercial name, Gesapax 500), commonly used on field crops and on corn and commercialized since 1975. Ametryne was analyzed by gas chromatography (GC Shimadzu 17A), after extraction with hexane/dichloromethane (1:1 v/v) solution. The calibration curve was obtained with a regression coefficient of 0.9871. In addition, the relative standard deviation was lower than 10%. The radiation-processing yield was evaluated by the destruction G-value (Gd) (Eq. 1), that is defined by the number of destroyed molecules by absorption of 100 eV of energy from ionizing radiation. Different concentrations of the herbicide (11.4 mol L{sup -1}; 22.7 mol L{sup -1}; 34.1 mol L{sup -1} and 45.5 mol L{sup -1}) were irradiated at the AECL 'Gammacell 220' {sup 60}Co source, with 1 kGy, 3 kGy, 6 kGy, 9 kGy, 12 kGy, 15 kGy and 30 kGy absorbed doses. After irradiation processing, the ametryne highest reduction rate occurs at low doses of radiation: at 6 kGy more than 85-90% of all ametryne compounds were removed. Two products of incomplete degradation of ametryne were identified as s-triazyne isomers. However, further work is needed in order to fully understand the ametryne degradation mechanisms the degradation yield of ametryne depends on its initial concentration and the process seems to be more efficient at higher concentrations. (author)

  14. A collaboration on extended INPRO case study of the DUPIC fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. H.; Yang, M. S.; Ko, W. I. (and others)

    2007-05-15

    Since 1992, KAERI, AECL, United States Department of States(USDOS) and IAEA have performed the DUPIC fuel cycle development activities as an international cooperative research program, which has now been chosen as a target nuclear system for an INPRO case study. This study will focus on a further improvement and modification of the basic principles, user requirements and acceptance limits, which are defined in the IAEA-TECDOC-1434 for an evaluation of its proliferation-resistance through a proliferation-resistance assessment of the whole fuel cycle of DUPIC based on the INPRO methodology. In order to further develop an evaluation method for a proliferation-resistance based on the INPRO methodology, the basic principles, user requirements and acceptance limits of a proliferation-resistance was reviewed and quantified. Then the evaluation model (material flow, facility scale, reference fuel, etc.) of the DUPIC fuel cycle was developed and a proliferation-resistance assessment of the DUPIC fuel cycle including the PWR fuel cycle was performed by using the revised INPRO methodology in the area of a proliferation resistance. Also, the recommendations for a further improvement of INPRO methodology were suggested through examining the INPRO methodology for a proliferation resistance assessment. Through the proliferation resistance assessment of the whole fuel cycle of DUPIC including the PWR fuel cycle, the proliferation-resistance methodology was updated and re-established. And based on its experience, The research results can be used not only to evaluate and determine the future domestic proliferation-resistant fuel cycles which were derived from the GEN{sub I}V or INPRO programs but also to improve a system design to enhance its proliferation resistance. The present results will be utilized for the development of an INPRO User's Manual which is being developed as an important issue by IAEA. The credibility of the research results were ensured by the IAEA

  15. Characterizing fractured plutonic rocks of the Canadian shield for deep geological disposal of Canada`s radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Lodha, G.S.; Davison, C.C.; Gascoyne, M. [Atomic Energy of Canada Ltd. , Pinawa, MB (Canada). Whiteshell Labs.

    1998-09-01

    Since 1978 AECL has been investigating plutonic rocks of the Canadian Shield as a potential medium for the disposal of Canada`s nuclear fuel waste. During the last two years this study has been continued as part of Ontario Hydro`s used fuel disposal program. Methods have been developed for characterizing the geotechnical conditions at the regional scale of the Canadian Shield as well as for characterizing conditions at the site scale and the very near-field scale needed for locating and designing disposal vault rooms and waste emplacement areas. The Whiteshell Research Area (WRA) and the Underground Research Laboratory (URL) in southeastern Manitoba have been extensively used to develop and demonstrate the different scales of characterization methods. At the regional scale, airborne magnetic and electromagnetic surveys combined with LANDSAT 5 and surface gravity survey data have been helpful in identifying boundaries of the plutonic rocks , overburden thicknesses, major lineaments that might be geological structures, lithological contacts and depths of the batholiths. Surface geological mapping of exposed rock outcrops, combined with surface VLF/EM, radar and seismic reflection surveys were useful in identifying the orientation and depth continuity of low-dipping fracture zones beneath rock outcrops to a depth of 500 to 1000 m. The surface time-domain EM method has provided encouraging results for identifying the depth of highly saline pore waters. The regional site scale investigations at the WRA included the drilling of twenty deep boreholes (> 500 m) at seven separate study areas. Geological core logging combined with borehole geophysical logging, TV/ATV logging, flowmeter logging and full waveform sonic logging in these boreholes helped to confirm the location of hydro geologically important fractures, orient cores and infer the relative permeability of some fracture zones. Single-hole radar and crosshole seismic tomography surveys were useful to establish the

  16. Development of a System Dynamics Model for Evaluating the Economics of an Advanced CANDU Fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-15

    Since the early 1990's, the Korea Atomic Energy Research Institute (KAERI) and the Atomic Energy of Canada Limited (AECL) have cooperated to develop, verify, and demonstrate the advanced CANDU fuel, so called CANFLEX-NU (Natural Uranium). The CANFLEX-NU fuel bundle consists of 43 fuel elements and has the buttons on the outer surface of the fuel elements for improving the CHF (Critical-Heat-Flux) characteristics. Because of this features of CANFLEXNU fuel, it offers higher operating and safety margins than current 37-element fuel. Recently, the interest for a CANFLEX-NU has been increased because of the power de-rating due to aging of CANDU reactors. Wolsong Unit 1 CANDU reactor has been operated over 25 years and the operating power at the present time is less than 90% of a full power because of a reduction of the margin of ROP trip set point. The most appropriate way to overcome such a power de-rating due to a crept pressure tube is the introduction of a CANFLEX-NU fuel into a CANDU reactor. Now, a CANFLEX-NU fuel is ready to be commercialized in a CANDU-6 reactor because the design and demonstration irradiation have been completed in both Korea and Canada. Economic evaluation for commercializing a CANFLEX-NU fuel in Wolsong Units was carried out by calculating the unit prime cost of electricity production. Throughout the economic evaluation, it was found that the introduction of CANFLEX-NU fuel into Wolsong Units would have much economic benefits due to a better operating performance. However, the amount of economic profit due to introducing CANFLEX-NU fuel depends on several parameters such as the required time to get license from regulatory institute before commercializing, licensing cost, failure probability of commercializing etc. Therefore, it is necessary to determine the optimum condition to get the highest economic profit. In this paper, an economic evaluation was carried out based on the starting year of the licensing study with considering the

  17. Feature Detection, Characterization and Confirmation Methodology: Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Karasaki, Kenzi; Apps, John; Doughty, Christine; Gwatney, Hope; Onishi, Celia Tiemi; Trautz, Robert; Tsang, Chin-Fu

    2007-03-01

    This is the final report of the NUMO-LBNL collaborative project: Feature Detection, Characterization and Confirmation Methodology under NUMO-DOE/LBNL collaboration agreement, the task description of which can be found in the Appendix. We examine site characterization projects from several sites in the world. The list includes Yucca Mountain in the USA, Tono and Horonobe in Japan, AECL in Canada, sites in Sweden, and Olkiluoto in Finland. We identify important geologic features and parameters common to most (or all) sites to provide useful information for future repository siting activity. At first glance, one could question whether there was any commonality among the sites, which are in different rock types at different locations. For example, the planned Yucca Mountain site is a dry repository in unsaturated tuff, whereas the Swedish sites are situated in saturated granite. However, the study concludes that indeed there are a number of important common features and parameters among all the sites--namely, (1) fault properties, (2) fracture-matrix interaction (3) groundwater flux, (4) boundary conditions, and (5) the permeability and porosity of the materials. We list the lessons learned from the Yucca Mountain Project and other site characterization programs. Most programs have by and large been quite successful. Nonetheless, there are definitely 'should-haves' and 'could-haves', or lessons to be learned, in all these programs. Although each site characterization program has some unique aspects, we believe that these crosscutting lessons can be very useful for future site investigations to be conducted in Japan. One of the most common lessons learned is that a repository program should allow for flexibility, in both schedule and approach. We examine field investigation technologies used to collect site characterization data in the field. An extensive list of existing field technologies is presented, with some discussion on usage and limitations

  18. The role of peer review in responsive decisions - a case study of clean-up and safe long-term management of historic waste at Port Hope and Clarington

    Energy Technology Data Exchange (ETDEWEB)

    Hardy, D. [Hardy Stevenson and Associates Inc., Toronto, Ontario (Canada)

    2006-07-01

    'Full text:' The Municipalities of Port Hope and Clarington ('Municipalities') are committed to leading the clean- up and safe long-term management of historic low- level radioactive and hazardous wastes deposited in Port Granby (in the Municipality of Clarington) and several locations in the Municipality of Port Hope. They are supported by the Government of Canada through Atomic Energy of Canada Limited's (AECL) Low- Level Radioactive Waste Management Office (LLRWMO). The wastes are the result of radium and uranium processing in Port Hope by Eldorado since the 1930s. To establish the parameters of the clean- up, including the ability to veto the project, the local municipalities negotiated and signed a Legal Agreement with the Government of Canada in 2001. As a Responsible Authority, Natural Resources Canada has defined and approved the scope of the two projects. The LLRWMO is designated as the proponent. Accordingly, the LLRWMO is conducting the Environmental Assessment (EA) Studies and seeking approval of a preferred method of conducting the clean up for each municipality. The municipalities recognized that these large and complex environmental assessment studies would challenge the resources of citizens, municipal professional staff and politicians. Thus, the Legal Agreement specified that both municipalities would have funded staff to work together to coordinate and expedite the project. A Peer Review Team (PRT) would be retained by the municipalities and funded by the Government of Canada. The PRT is made up of experienced professionals led by Hardy Stevenson and Associates Limited in disciplines appropriate to the peer review tasks on hand. The PRT has brought a unique approach to the peer review. The PRT is headed by planners and social scientists trained to be sensitive to the 'people aspects' of the EA process as a major priority. They are supported by engineers and technical specialists. The team includes a physician

  19. Modelisation de l'instabilite fluidelastique d'un faisceau de tubes soumis a un ecoulement diphasique transverse

    Science.gov (United States)

    Sawadogo, Teguewinde

    This study focuses on the modeling of fluidelastic instability induced by two-phase cross-flow in tube bundles of steam generators. The steam generators in CANDU type nuclear power plants for e.g., designed in Canada by AECL and exploited worldwide, have thousands of tubes assembled in bundles that ensure the heat exchange between the internal circuit of heated heavy water coming from the reactor core and the external circuit of light water evaporated and directed toward the turbines. The main objective of this research project is to extend the theoretical models for fluidelastic instability to two-phase flow, validate the models and develop a computer program for simulating flow induced vibrations in tube bundles. The quasi-steady model has been investigated in scope of this research project. The time delay between the structure motion and the fluid forces generated thereby has been extensively studied in two-phase flow. The study was conducted for a rotated triangular tube array. Firstly, experimental measurements of unsteady and quasi-static fluid forces (in the lift direction) acting on a tube subject to two-phase flow were conducted. Quasi-static fluid force coefficients were measured at the same Reynolds number, Re = 2.8x104, for void fractions ranging from 0% to 80%. The derivative of the lift coefficient with respect to the quasi-static dimensionless displacement in the lift direction was deduced from the experimental measurements. This derivative is one of the most important parameters of the quasi-steady model because this parameter, in addition to the time delay, generates the fluid negative damping that causes the instability. This derivative was found to be positive in liquid flow and negative in two-phase flow. It seemed to vanish at 5% of void fraction, challenging the ability of the quasi-steady model to predict fluidelastic instability in this case. However, stability tests conducted at 5% void fraction clearly showed fluidelastic instability

  20. Surface chemistry interventions to control boiler tube fouling - Part II

    Energy Technology Data Exchange (ETDEWEB)

    Turner, C.W.; Guzonas, D.A.; Klimas, S.J

    2004-06-15

    This is the third in a series of reports from an investigation co-funded by the Electric Power Research Institute (EPRI) and by Atomic Energy of Canada Limited (AECL) into the effectiveness of alternative amines for controlling the rate of tube-bundle fouling under steam generator (SG) operating conditions. The objectives of this investigation are to determine whether the fouling rate depends on the amine used for pH control, to identify those factors that influence the effectiveness, and use this information to optimize the selection of an amine for chemistry control and deposit control in the steam cycle and steam generator, respectively. Work to date has demonstrated that the rate of particle deposition under steam generator operating conditions is strongly influenced by surface chemistry (Turner et al., 1997; Turner et al., 1999). This dependence upon surface chemistry is illustrated by the difference between the deposition rates measured for hematite and magnetite, and by the dependence of the particle deposition rate on the amine used for pH control. Deposition rates of hematite were found to be more than 10 times greater than those for magnetite under similar test conditions (Turner et al., 1997). At 270{sup o}C and pH{sub T} 6.2, the surfaces of hematite and magnetite are predicted to be positively charged and negatively charged, respectively (Shoonen, 1994). Measurements of the point of zero charge (PZC) of magnetite at temperatures from 25{sup o}C to 290{sup o}C by Wesolowski et al. (1999) have confirmed that magnetite is negatively charged at the stated conditions. A PZC of 4.2 was measured for Alloy 600 at 25{sup o}C (Balakrishnan and Turner, un-published results), and its surface is expected to remain negatively charged for alkaline chemistry over the temperature range of interest. Therefore, there will be a repulsive force between the surfaces of magnetite particles and Alloy 600 at 270{sup o}C and pH{sub T} 6.2 that is absent for hematite particles

  1. Feature Detection, Characterization and Confirmation Methodology: Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Karasaki, Kenzi; Apps, John; Doughty, Christine; Gwatney, Hope; Onishi, Celia Tiemi; Trautz, Robert; Tsang, Chin-Fu

    2007-03-01

    This is the final report of the NUMO-LBNL collaborative project: Feature Detection, Characterization and Confirmation Methodology under NUMO-DOE/LBNL collaboration agreement, the task description of which can be found in the Appendix. We examine site characterization projects from several sites in the world. The list includes Yucca Mountain in the USA, Tono and Horonobe in Japan, AECL in Canada, sites in Sweden, and Olkiluoto in Finland. We identify important geologic features and parameters common to most (or all) sites to provide useful information for future repository siting activity. At first glance, one could question whether there was any commonality among the sites, which are in different rock types at different locations. For example, the planned Yucca Mountain site is a dry repository in unsaturated tuff, whereas the Swedish sites are situated in saturated granite. However, the study concludes that indeed there are a number of important common features and parameters among all the sites--namely, (1) fault properties, (2) fracture-matrix interaction (3) groundwater flux, (4) boundary conditions, and (5) the permeability and porosity of the materials. We list the lessons learned from the Yucca Mountain Project and other site characterization programs. Most programs have by and large been quite successful. Nonetheless, there are definitely 'should-haves' and 'could-haves', or lessons to be learned, in all these programs. Although each site characterization program has some unique aspects, we believe that these crosscutting lessons can be very useful for future site investigations to be conducted in Japan. One of the most common lessons learned is that a repository program should allow for flexibility, in both schedule and approach. We examine field investigation technologies used to collect site characterization data in the field. An extensive list of existing field technologies is presented, with some discussion on usage and limitations

  2. Test with different stress measurement methods in two orthogonal bore holes in Aespoe HRL

    Energy Technology Data Exchange (ETDEWEB)

    Janson, Thomas; Stigsson, Martin [Golder Associates AB, Stockholm (Sweden)

    2002-12-01

    Within the scope of work, to provide the necessary rock mechanics support for the site investigations, SKB has studied some available pieces of equipment for in situ stress measurements in deep boreholes. A project with the objective to compare three different pieces of equipment for in situ stress measurements under similar conditions has been carried out. The main objective for the project is to compare the three different pieces of equipment for in situ stress measurements and find a strategy for SKB's Site Investigations to determine the state of stress in the rock mass. Two units of equipment use the overcoring method while the third uses the hydraulic fracturing method. The overcoring was performed by AECL, using Deep Door stopper Gauge System (DDGS), and SwedPower, using their triaxial strain measuring instrument (Borre Probe). MeSy Geo Systeme GmbH performed the hydraulic fracturing. The DDGS system is a new method to SKB while the experience of the SwedPower overcoring and the hydraulic fracturing methods are long. The tests were performed in the same orthogonal boreholes at Aespoe Hard Rock Laboratory (HRL), Oskarshamn, Sweden. The measured results have been verified against known conditions at the Aespoe HRL. The results from the three in situ stress measurement methods rose more questions than answers. Which illustrate the complexity to determine the in situ stresses in a rock mass. To understand the difference in results and answer the questions, it was necessary to do deeper investigations such as laboratory tests and theoretical calculations such as geological structure model, analysis of the influence of a nearby fracture, P-wave measurements, uniaxial tests on small cores from the HQ-3 core, theoretical and numerical analyses of the hole bottom (theoretical strains, stress concentrations and microcracking), auditing of DDGS measurements results and assumptions in the DDGS analyse and microscopy investigations on the cores. The following