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Sample records for belgian reactor 3

  1. Irradiation of Fuel Elements in the Belgian BR3 Reactor

    International Nuclear Information System (INIS)

    Under a contract concluded by EURATOM and CEN-BelgoNucléaire, fuel rods containing plutonium-enriched uranium were irradiated in the Belgian BR3 reactor with the object of evaluating the behaviour of plutonium fuel elements in power reactors. The first experiment consisted in introducing 12 fuel elements fabricated by vibration and compacting followed by swaging into a core assembly of the BR3 pressurized-water power reactor. Irradiation was carried out for a period corresponding to 4820 h at full power. Subsequent examination of the fuel rods showed that they had been unaffected by irradiation. A second series of experiments is being carried out in collaboration with the United Kingdom Atomic Energy Authority. These experiments involve irradiating an assembly of 37 plutonium-enriched fuel elements, some compacted and others of the pellet type, in the BR3/VN power reactor. The fabrication of the vibrocompacted elements and the thermal studies relating to the assembly are briefly described. (author)

  2. Lessons learned from the decommissioning of the Belgian pressurized water test reactor BR3

    International Nuclear Information System (INIS)

    The BR3 plant was operated with the main objective of testing advanced PWR fuels under irradiation conditions similar to those encountered in large commercial PWR plants. In 1989, it was selected as one of the pilot projects of the European Commission for its R and D programme on Decommissioning of nuclear installations. With the decommissioning of the BR3 reactor, the Belgian Nuclear Research Centre SCK·CEN gained a lot of experiences in the field of decommissioning. This paper describes the main phases carried out in the decommissioning project up till now and will discuss the main lessons to learn. (author)

  3. Mechanical testing and microstructural characterization of pressure vessel at decommissioned Belgian BR3 Plant

    International Nuclear Information System (INIS)

    The objective of this paper was a discussion of a proposal to perform mechanical testing and microstructural characterization of the annealed reactor vessel of the Belgian BR-3 reactor. Motivation for this effort was discussed, and a preliminary cost estimate for some of the tasks was also presented

  4. Qualification of non-destructive examination for belgian nuclear reactor pressure vessel inspection

    Energy Technology Data Exchange (ETDEWEB)

    Couplet, D. [TRACTEBEL, Brussels (Belgium); Francoise, T. [Intercontrole, 94 - Rungis (France)

    2001-07-01

    In Service Inspection (ISI) participates to the assessment of Nuclear Reactor Pressure Vessel Integrity. The performance of Non Destructive Examination (NDE) techniques must be demonstrated according to predefined objectives. The qualification process is essential to trust the reliability of the information provided by NDE. In Belgian Nuclear Power Plants, the qualification was conducted through a collaboration between the vendor and a technical group from the Electricity Utility. The important facts of this qualification will be presented: - the detailed definition of the inspection and qualifications objectives, based on a combination of the ASME code and the European Methodology for Qualification; - the systematic verification of the NDE performance and limitations, for each ISI objective, through an adequate combination of tests on blocks and technical justification; - the continuous improvement of the NDE procedure; - the feedback and the lessons learnt from site experience; - the necessary multi-disciplinary approach (NDE, degradation mechanisms, structural integrity)

  5. VISIT OF BELGIAN FIRMS AT CERN: 2 - 3 APRIL 2003

    CERN Multimedia

    2003-01-01

    14:00 to 17:30 hrs Wednesday 2nd April 09:00 to 17:30 hrsThursday 3rd April Individual interviews will take place in different conference rooms or in technicians' offices. The firms will contact relevant users/technicians but any user wishing to make contact with a particular firm is welcome to use the contact details which are available from each secretariat of division or from the Purchasing web pages at the following URL http://spl-div.web.cern.ch/spl-div/member_states/exhibitions_visits.htm List of Companies: 1. Advanco13. Gillam-Fei SA25. Opticable SA 2. Balteau SA14. G-Tec SA26. Orthodyne SA 3. Barco NV15. Groupe Hamon27. Polmans Atelier Mecanique 4. Blonde SA16. HTMS NV28. RESARM Engineering Plastics SA 5. Britte SA17. IMCORP Europe29. SAMTECH 6. Cablerie d'Eupen18. Inductive Systems Europe NV30. Schreder - Hazemeyer SA 7. Cegelec SA19. Link Software31. SYREG Sprl 8. Clever House20. MACQ Electronique SA32. Thales Communications Belgium 9. SA Coppee - Courtoy NV21. Mecanique de Precision pour Equipm...

  6. VISIT OF BELGIAN FIRMS AT CERN: 2 - 3 APRIL 2003

    CERN Multimedia

    2003-01-01

    14:00 to 17:30 hrs Wednesday 2nd 09:00 to 17:30 hrs Thursday 3rd Individual interviews will take place in different conference rooms or in technicians' offices. The firms will contact relevant users/technicians but any user wishing to make contact with a particular firm is welcome to use the contact details which are available from each secretariat of division or from the Purchasing web pages at the following URL http://spl-div.web.cern.ch/spl-div/member_states/exhibitions_visits.htm. List of Companies: 1. Advanco18. Link Software 2. Balteau SA19. MACQ Electronique SA 3. Barco NV20. Mecanique de Precision pour Equipments 4. Blonde SA21. Mecasoft SA 5. Britte SA22. Mockel SCA 6. Cablerie d'Eupen23. Notifier Benelux 7. Cegelec SA24. Opticable SA 8. SA Coppee - Courtoy NV25. Orthodyne SA 9. Denys NV26. Polmans Atelier Mecanique 10. DSI Sprl27. RESARM Engineering Plastics SA 11. Engetec SA28. SAMTECH 12. Gillam-Fei SA29. Schreder - Hazemeyer SA 13. G-Tec SA30. SYREG Sprl 14. Groupe Hamon31. Thales Communication...

  7. VISIT OF BELGIAN FIRMS AT CERN: 2 - 3 APRIL 2003

    CERN Multimedia

    2003-01-01

    14:00 to 17:30 hrs Wednesday 2nd 09:00 to 17:30 hrs Thursday 3rd Individual interviews will take place in different conference rooms or in technicians' offices. The firms will contact relevant users/technicians but any user wishing to make contact with a particular firm is welcome to use the contact details which are available from each secretariat of division or from the Purchasing web pages at the following URL http://spl-div.web.cern.ch/spl-div/member_states/exhibitions_visits.htm. List of Companies: 1. Advanco18. Link Software 2. Balteau SA19. MACQ Electronique SA 3. Barco NV20. Mecanique de Precision pour Equipments 4. Blonde SA21. Mecasoft SA 5. Britte SA22. Mockel SCA 6. Cablerie d'Eupen23. Notifier Benelux 7. Cegelec SA24. Opticable SA 8. SA Coppee - Courtoy NV25. Orthodyne SA 9. Denys NV26. Polmans Atelier Mecanique 10. DSI Sprl27. RESARM Engineering Plastics SA 11. Engetec SA28. SAMTECH 12. Gillam-Fei SA29. Schreder - Hazemeyer SA 13. G-Tec SA30. SYREG Sprl30. SYREG Sprl 14. Groupe Hamon31. Thale...

  8. A review of silicon neutron transmutation doping and its practice at French and Belgian research reactors

    International Nuclear Information System (INIS)

    The role of NTD of silicon in the semiconductor market for electrical power systems is a major incentive for the development of improved characteristics in terms of fabrication techniques and materials, in order to obtain, as economically as possible, components that are more compact, generate less waste and permit higher power ratings. The market demand for NTD silicon will be met as long as there are research reactors capable of offering a reliable irradiation service of adequate capacity and quality to the customer at a cost that is competitive with high quality chemically doped silicon. Production on a 'just in time' basis is becoming less and less compatible with the operation of research reactors, which are subject to increasingly stringent safety checks, particularly as many of them are more than 30 years old and therefore may be subject to ever increasing down times and longer outages for refurbishment. Consequently, cooperation between research reactors will become increasingly necessary to guarantee the continued availability of silicon doped by neutron transmutation for industrial customers

  9. Belgian Firms Visit CERN

    CERN Multimedia

    2001-01-01

    Fifteen Belgian firms visited CERN last 2 and 3 April to present their know-how. Industrial sectors ranging from precision machining to electrical engineering and electronics were represented. And for the first time, companies from the Flemish and Brussels regions of the country joined their Walloon compatriots, who have come to CERN before. The visit was organised by Mr J.-M. Warêgne, economic and commercial attaché at the Belgian permanent mission for the French-speaking region, Mr J. Van de Vondel, his opposite number for the Flemish region, and Mrs E. Solowianiuk, economic and commercial counsellor at the Belgian permanent mission for the Brussels-Capital region.

  10. The Belgian nuclear research centre

    International Nuclear Information System (INIS)

    The Belgian Nuclear Research Centre is almost exclusively devoted to nuclear R and D and services and is able to generate 50% of its resources (out of 75 million Euro) by contract work and services. The main areas of research include nuclear reactor safety, radioactive waste management, radiation protection and safeguards. The high flux reactor BR2 is extensively used to test fuel and structural materials. PWR-plant BR3 is devoted to the scientific analysis of decommissioning problems. The Centre has a strong programme on the applications of radioisotopes and radiation in medicine and industry. The centre has plans to develop an accelerator driven spallation neutron source for various applications. It has initiated programmes to disseminate correct information on issues of nuclear energy production and non-energy nuclear applications to different target groups. It has strong linkages with the IAEA, OECD-NEA and the Euratom. (author)

  11. A stop-gain in the laminin, alpha 3 gene causes recessive junctional epidermolysis bullosa in Belgian Blue cattle

    OpenAIRE

    Sartelet, Arnaud; Harland, Chad; Tamma, Nico; Karim, Latifa; Bayrou, Calixte; Li, Wanbo; Ahariz, Naïma; Coppieters, Wouter; Georges, Michel; Charlier, Carole

    2015-01-01

    Four newborn purebred Belgian Blue calves presenting a severe form of epidermolysis bullosa were recently referred to our heredo-surveillance platform. SNP array genotyping followed by autozygosity mapping located the causative gene in a 8.3-Mb interval on bovine chromosome 24. Combining information from (i) whole-genome sequencing of an affected calf, (ii) transcriptomic data from a panel of tissues and (iii) a list of functionally ranked positional candidates pinpointed a private G to A nuc...

  12. Decommissioning of a small reactor (BR3 reactor, Belgium)

    International Nuclear Information System (INIS)

    Since 1989, SCK-CEN has been dismantling its PWR reactor BR3 (Belgian Reactor No. 3). After gaining a great deal of experience in remote dismantling of highly radioactive components during the actual dismantling of the two sets of internals, the BR3 team completed the cutting of its reactor pressure vessel (RPV). During the feasibility phase of the RPV dismantling, a decision was made to cut it under water in the refuelling pool of the plant, after having removed it from its cavity. The RPV was cut into segments using a milling cutter and a bandsaw machine. These mechanical techniques have shown their ability for this kind of operations. Prior to the segmentation, the thermal insulation situated around the RPV was remotely removed and disposed of. The paper will describe all these operations. The BR3 decommissioning activities also include the dismantling of contaminated loops and equipment. After a careful sorting of the pieces, optimized management routes are selected in order to minimize the final amount of radioactive waste to be disposed of. Some development of different methods of decontamination were carried out: abrasive blasting (or sand blasting), chemical decontamination (Oxidizing-Reducing process using Cerium). The main goal of the decontamination program is to recycle most of the metallic materials either in the nuclear world or in the industrial world by reaching the respective recycling or clearance level. Overall the decommissioning of the BR3 reactor has shown the feasibility of performing such a project in a safe and economical way. Moreover, BR3 has developed methodologies and decontamination processes to economically reduce the amount of radwaste produced. (author)

  13. Preliminary LEU fuel cycle analyses for the Belgian BR2 reactor

    International Nuclear Information System (INIS)

    Fuel cycle calculations have been performed with reference HEU fuel and LEU fuel using Cd wires or boron as burnable absorbers. The 235U content in the LEU element has increased 20% to 480g compared to the reference HEU element. The number of fuel plates has remained unchanged while the fuel meat thickness has increased to 0.76 mm from 0.51 mm. The LEU meat density is 5.1 Mg U/m3. The reference fuel cycle was a 31 element core operating at 56 MW with a 19.8 day cycle length and eight fresh elements loaded per cycle. Comparable fuel cycle characteristics can be achieved using the proposed LEU fuel element with either Cd wires or boron burnable absorbers. The neutron flux for E/sub n/ > 1 eV changes very little (<5%) in LEU relative to HEU cores. Thermal flux reductions are 5 to 10% in non-fueled positions, and 20 to 30% in fuel elements

  14. Coolant chemistry studies at the Belgian PWRs, Doel 3 and Doel 4

    International Nuclear Information System (INIS)

    Collaborative studies of coolant chemistry on the Doel 3/4 reactors have benefitted from standardised sampling equipment to provide self-consistent and integrated data. Modern methods of analyses have been applied for the in-line assessment of particulates in coolant and soluble corrosion products. Refined techniques for the measurement of soluble and insoluble active species have provided useful comparative data at Doel 3 and 4. While our analyses of the data suggest that the soluble component of coolant-borne activity is most important in setting early doserate trends around the external coolant circuit, the contribution of particulates cannot be ignore in the longer-term. The properties and composition of particles in suspension in the coolant have been explored in depth during steady reactor operation, during transients and at shutdown and reactor start-up. At similar stages of operation our studies have covered the role of all metallic and active species carried by the coolant. In two similar reactor systems where the main variables were coolant pHT and Zircaloy or Inconel 718 gridded fuel, the advantages of higher alkalinity and the absence of an in-core component source of 60Co on external doserates were most evident. This work was funded by the CEGB in the UK. (author)

  15. Initial Operational Results of the First Belgian Nuclear Power Station BR3

    International Nuclear Information System (INIS)

    The BR3 Nuclear Power Plant, property of the Centre d'etude de l'energie nucleaire (CEN), is located at Mol-Donk on the site of the latter, and at a distance of 40 miles as the crow flies from Brussels. The operation of the plant is done by an Industrial Group. The reactor went critical, for the first time, on 29 August 1962. The paper gives the results of the preliminary tests and of the first months of operation. The reactor BR3 is of the pressurized, light-water moderated and cooled type, with a nominal thermal power of 40.9 MW. The core is heterogeneous, using UO2 fuel, clad with stainless steel and has two regions. The outer region is enriched to 4.43% by weight, the inner one to 3.7%. Main coolant pressure is 140 kg/cm2. Different ''primary'' circuits are in connection with the reactor. The steam generator has a nominal capacity of 70 t/h. The turbine has two stages and three extraction lines and a nominal power of 11 700 kW. The personnel was selected from the experienced personnel of the conventional power plants and received a thorough training in both the nuclear and the classical field. The paper gives the results of the practical determination of the plant characteristics: control-rod worth and boron worth determination, calibration of the control rods, temperature and pressure coefficients determination, determination of the dropping time of the control rods, etc. The excess reactivity was calculated for the clean core. These tests were done at zero power and after 441 effective full power hours (EFPH) burn-up: 750 MWJ). A resume is given of the load transient tests and the power coefficient determination. Different problems encountered during operation are treated: (a) Radiation activity in the plant: we may conclude from the results that the habitual work in the normal areas is not affected by the radiation intensities present. (b) Activity in the air of the building: the only problem which appeared until now is the presence of radon in the reactor

  16. Research Reactors: Decommissioning of a Small Reactor (BR3 Reactor, Belgium). Appendix III

    International Nuclear Information System (INIS)

    Research reactors are nuclear reactors that serve primarily as source of neutrons. They are less complex than power reactors and operate at lower temperatures. Research reactors need far less fuel, and far less fission products build up as the fuel is used. On the other hand, their fuel requires more highly enriched uranium, typically up to 20% 235U. More than 650 research reactors worldwide have been built or are under construction or in a planning phase; of which more than 350 have been shut down and partly or wholly decommissioned. Experience has shown that decommissioning can be undertaken in line with safety standards aimed at protecting human beings or the environment from harm, provided that decommissioning activities are undertaken in accordance with a properly formulated plan. The potential or actual radiological hazards associated with reactors may require the application of special techniques and procedures during decommissioning. The decommissioning of the BR3 reactor in Mol, Belgium, Belgian nuclear research centre SCK•CEN, provides an example of current good practice in decommissioning research reactors.13 Since 1991, the organization’s statutory mission gives priority to research on problems of societal concern such as the safety of nuclear installations, radiation protection, safe treatment and disposal of radioactive waste, fighting against uncontrolled proliferation of fissile materials, and education and training. BR3 was the first European pressurized water reactor (PWR) power plant and was put into service in 1962. It was in that industrial context that the BR3 has played its role as a demonstration unit for the development and improvement of decommissioning related techniques. While the BR3 power level was low (40 MW(th), 10.5 MW(e) net), it contains all the features of commercial PWR power plants. The reactor was used at the beginning of its lifetime as a training facility for future nuclear power plant operators. Later, it was also used

  17. Reactor Materials Research

    International Nuclear Information System (INIS)

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  18. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  19. A stop-gain in the laminin, alpha 3 gene causes recessive junctional epidermolysis bullosa in Belgian Blue cattle.

    Science.gov (United States)

    Sartelet, Arnaud; Harland, Chad; Tamma, Nico; Karim, Latifa; Bayrou, Calixte; Li, Wanbo; Ahariz, Naima; Coppieters, Wouter; Georges, Michel; Charlier, Carole

    2015-10-01

    Four newborn purebred Belgian Blue calves presenting a severe form of epidermolysis bullosa were recently referred to our heredo-surveillance platform. SNP array genotyping followed by autozygosity mapping located the causative gene in a 8.3-Mb interval on bovine chromosome 24. Combining information from (i) whole-genome sequencing of an affected calf, (ii) transcriptomic data from a panel of tissues and (iii) a list of functionally ranked positional candidates pinpointed a private G to A nucleotide substitution in the LAMA3 gene that creates a premature stop codon (p.Arg2609*) in exon 60, truncating 22% of the corresponding protein. The LAMA3 gene encodes the alpha 3 subunit of the heterotrimeric laminin-332, a key constituent of the lamina lucida that is part of the skin basement membrane connecting epidermis and dermis layers. Homozygous loss-of-function mutations in this gene are known to cause severe junctional epidermolysis bullosa in human, mice, horse, sheep and dog. Overall, our data strongly support the causality of the identified gene and mutation. PMID:26370913

  20. Strategic groups in the Belgian fishing fleet

    OpenAIRE

    Stouten, H.; A. HEENE; Gellynck, X.; Polet, H

    2011-01-01

    This study examines the heterogeneity of the Belgian fishing fleet based on “strategic groups”, a concept borrowed from the field of strategic management. Its objectives are: (1) to define strategic groups within the Belgian fishing fleet; (2) to examine the performance differences among these strategic groups; (3) to examine whether firms (i.e., vessels) move between strategic groups over time; and (4) to examine if firm-movement (i.e., vessel-movement) differs across strategic groups. In th...

  1. MOLIERE PROJECT - Belgian National Report

    OpenAIRE

    Naedenoen, Frédéric

    2014-01-01

    This document aims at introducing the restructuring phenomena in Belgium. It makes part of a larger project (called the MOLIERE project) undertaken by eleven European countries gathered to analyse restructurings. This national document follows a common reporting format: an introduction of the complex Belgian restructuring frameworks (1), a presentation of the main actors involved in the process (2), a synthesis of the measures created to anticipate change (3) and a synthesis of the measures ...

  2. Measurement of Urinary Biomarkers of Parabens, Benzophenone-3, and Phthalates in a Belgian Population

    OpenAIRE

    Lucas Dewalque; Catherine Pirard; Corinne Charlier

    2014-01-01

    Parabens, benzophenone-3 (BP3), and phthalates are commonly used as antimicrobial conservator, UV-filter, and plasticizer, respectively, and are thought to exhibit endocrine disrupting properties. These endocrine disrupting activities have been recently assumed to lead to cutaneous malignant melanoma. Humans are exposed to these chemicals through different sources such as food, personal care products, or cosmetics. In this study, we measured urinary levels of 4 parabens, BP3, and 7 metabolite...

  3. The utility of different reactor types for the research

    International Nuclear Information System (INIS)

    The report presents a general view of the use of the different belgian research reactor i.e. venus reactor, BR-1 reactor, BR-2 reactor and BR-3 reactor. Particular attention is given to the programmes which is in the interest of international collaboration. In order to reach an efficient utilization of such reactors they require a specialized personnel groups to deal with the irradiation devices and radioactive materials and post irradiation examinations, creating a complete material testing station. (A.J.)

  4. Fuel cycle transition - A Belgian implementation scenario

    International Nuclear Information System (INIS)

    At the end of 2002 the total installed electric power in Belgium was 16,200 MWe of which 40% (6485 MWe) corresponds to the seven nuclear power plants installed on the two Belgian sites of Doel (4 power plants) and Tihange (3 power plants) and the 25% participation in the two French Units B1 and B2 at Chooz at the Belgian-French border. The nuclear installed power in Belgium is 5800 MWe. In 2003, the government decided to phase out the nuclear energy progressively by closing the Belgian NPPs after 40 years of operation. This means that the first generation units (Doel 1, Doel 2 and Tihange 1) will be closed in 2015 and the four other remaining units in 2022-2025. Nevertheless, this phase out is subject to various conditions: the guarantee of energy independence should not be affected and the engagement to respect the Kyoto agreement (reducing the CO2 production by 7.5% in 2010 as compared to the 1990 production). Thus the phase-out decision can be re-opened if the above mentioned conditions are not met. The paper has the following contents: 1. Introduction; 2. Actual fuel cycle; 3. Transition fuel cycle; 4. Calculations; 4.1. PWR modelling; 4.2. ADS modelling; 4.3. Calculation code; 5. Results; 5.1. PWR/EPR; 5.2. ADS; 6. Conclusions. In conclusion it is shown that the evaluated stock pile of waste in Belgium (with no increase of electricity demand) coming from the thermal reactors park is 4380 tons (52 t Pu, 9 t MA, 217 t FP) with phase out (i.e. between 1975, first PWR and 2025, last PWR) and 7825 tons (84 t Pu, 20 t MA, 381 t FP) without phase out (i.e. between 1975, first PWR and 2075, last EPR). According to this study, Belgium should keep all its first generation Pu for the eventual starting of the self burning FR. Indeed, the Pu needed to start the self burning FR is evaluated between 60 t and 90 t (based on 10 t to 15 t per GWe). With an homogeneous core loading, 54% of the MA could be eliminated after 24 years in three 600 MWth industrial ADS (corresponding

  5. The Belgian Nuclear Higher Education Network

    International Nuclear Information System (INIS)

    Full text: BNEN, the Belgian Nuclear Higher Education Network has been created in 2001 by five Belgian universities and the Belgian Nuclear Research Centre (SCK-CEN) as a joint effort to maintain and further develop a high quality programme in nuclear engineering in Belgium. In a country where a substantial part of electricity generation will remain of nuclear origin for a number of years, there is a need for well educated and well trained engineers in this area. Public authorities, regulators and industry brought their support to this initiative. In the framework of the new architecture of higher education in Europe, the English name for this 60 ECTS programme is 'Master of Science in Nuclear Engineering'. To be admitted to this programme, students must already hold a university degree in engineering or equivalent. Linked with university research, benefiting from the human resources and infrastructure of SCK-CEN, encouraged and supported by the partners of the nuclear sector, this programme should be offered not only to Belgian students, but also more widely throughout Europe and the world. The master programme is a demanding programme where students with different high level backgrounds in engineering have to go through highly theoretical subjects like neutron physics, fluid flow and heat transfer modelling, and apply them to reactor design, nuclear safety and plant operation and control. At a more interdisciplinary level, the programme includes some important chapters of material science, with a particular interest for the fuel cycle. Radiation protection belongs also to the backbone of the programme. All the subjects are taught by academics appointed by the partner universities, whereas the practical exercises and laboratory sessions are supervised by researchers of SCK-CEN. The final thesis offers an opportunity for internship in industry or in a research laboratory. More information: http://www.sckcen.be/BNEN. (author)

  6. Prediction of neutron embrittlement in the reactor pressure vessel. Venus-1 and Venus-3 benchmarks

    International Nuclear Information System (INIS)

    The OECD/NEA Task Force on Computing Radiation Dose and Modelling of Radiation-Induced Degradation of Reactor Components (TFRDD) launched two international blind intercomparison exercises to examine the current computation techniques used in NEA Member countries for calculating neutron and gamma doses to reactor components. Various methodologies and different nuclear data were applied to predict dose rates in the Belgian VENUS-1 and three-dimensional VENUS-3 configurations for comparison with measured data. This report provides the detailed results from the two benchmarks.The exercise revealed that three-dimensional neutron fluence calculations provide results that are significantly more accurate than those obtained from two-dimensional calculations. Performing three-dimensional calculations is technically feasible given the power of today's computers. (author)

  7. Dynamics of TRIGA-3 Salazar Reactor

    International Nuclear Information System (INIS)

    The theoretical study of temporal behavior of a nuclear reactor is of great importance, since it allows to know, in advance, the conditions to which a reactor is going to be submitted. The reliability of two computer codes (AIREK-JEN and PLANKIN) designed to reproduce the temporal behavior of nuclear reactors, generally power reactors, when they are applied to reproduce the dynamic behavior of TRIGA-3 Salazar Reactor is analyzed. In the first chapters, the fundamental equations that solve this computer codes are deduced, and also the main characteristics of TRIGA-3 Salazar Reactor and the necessary data to run the programs are presented; later the results obtained with the computer codes and the experimental results reported in the operational logbook of the reactor are compared, with the result that such computer codes are applicable to the temporal study of TRIGA-3 Salazar Reactor. (Author)

  8. Reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Research and development activities related to reactor pressure vessel steels during 1997 are reported. The objectives of activities of the Belgian Nuclear Research Centre SCK/CEN in this domain are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate a methodology on a broad database; (3) to achieve regulatory acceptance and industrial use

  9. Annual report for the steering committee of the association Euratom-Belgian State for fusion 1999

    International Nuclear Information System (INIS)

    This report is prepared for the annual steering committee meting of the Association Euratom - Belgian State in the area of fusion reactor technology. The Belgian contribution focuses on the assessment of the first wall and blanket materials under radiation and coolant interaction and on developments for the remote handling in maintenance activities. The period October 1998 to September 1999 is reported on.The fusion technology work performed at the Belgian Nuclear Research Centre SCK/CEN, the Department of Metallurgy and Materials Engineering of the Louvain University (Belgium) and S.A. Gradel, a Luxemburg-based organisation, is described

  10. Annual report for the steering committee of the association Euratom-Belgian State for fusion 1998

    International Nuclear Information System (INIS)

    This report is prepared for the annual steering committee meting of the Association Euratom - Belgian State in the area of fusion reactor technology. The Belgian contribution focuses on the assessment of the first wall and blanket materials under radiation and coolant interaction and on developments for the remote handling in maintenance activities. The period October 1997 to September 1998 is reported on.The fusion technology work performed at the Belgian Nuclear Research Centre SCK/CEN, the Department of Metallurgy and Materials Engineering of the Louvain University (Belgium) and S.A. Gradel, a Luxemburg-based organisation, is described

  11. Annual report for the steering committee of the association Euratom-Belgian State for fusion 1999

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    1999-10-01

    This report is prepared for the annual steering committee meting of the Association Euratom - Belgian State in the area of fusion reactor technology. The Belgian contribution focuses on the assessment of the first wall and blanket materials under radiation and coolant interaction and on developments for the remote handling in maintenance activities. The period October 1998 to September 1999 is reported on.The fusion technology work performed at the Belgian Nuclear Research Centre SCK/CEN, the Department of Metallurgy and Materials Engineering of the Louvain University (Belgium) and S.A. Gradel, a Luxemburg-based organisation, is described.

  12. Reactor physics

    International Nuclear Information System (INIS)

    Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

  13. G2 and G3 reactors design

    International Nuclear Information System (INIS)

    The 'FRANCE ATOME' Manufacturers Party has been entrusted with the G2 and G3 reactors engineering by the french A.E.C., for the first-five-year french project. Although these reactors are essentially plutonium generators, everyone has been linked with a power station which is supposed to supply with 40 MW, 'Electricite de France' has taken the liability upon itself. The reactor core includes most of G1 reactor parts (central gap excluded): horizontal channels, graphite parallelepipedic bricks stacking, steel thermal shield. The cooling is provided with CO2 under a 15 atmospheres pressure. This pressure is kept steady in a press-stressed concrete packing-case which is a cylinder horizontally shaped. Steel strips tightened encircle the concrete cylinder; itself protected by sole-plates. The cylinder bottom has brought about unusual problems which have been solved by the choice of an hemispheric shape. Packing-case tightness is provided by a 30 mm iron-plate connected with the inner wall of concrete. One of the reactor's special characteristics is the possibility of loading and unloading while operating. On loading side, barrel locks, each weighting 50 tons, allow new cans, at a pressure of 15 atmospheres, to pass. The cans process almost in a steady way through the channel, and finally drop down through bent spouts, then through spiral toboggans into a new lock. The cooling CO2 flow is provided with 3 turbo-bellows, these are actuated by average pressure-steam, obtained from exchangers. Every reactor supplies 4 exchangers which have been very difficult to build and to set up. The secondary cycle is standard and contains 3 stages (pressure 10,3: 2 and 0,5 kg/cm2). Steam can be condensed in the event of a group turbo-generator stopping, with no modification for the normal operating conditions of the reactor. Auxiliary circuits have to assure the continuous purifying of cooling CO2, its storage and drain. 49 boron carbide rods are used to control the operating power

  14. 3. Halden Reactor Project Workshop

    International Nuclear Information System (INIS)

    A workshop was held in Halden 2nd-3rd March 2005 to discuss 'VR in the Future Industrial Workplace: Working Together - Regardless of Distance'. The workshop sessions and discussions focused on design, operations and maintenance, training, and engineering virtual reality systems, and provided useful insights into the current state of the art of research and development in the fields of virtual and augmented reality. (Author)

  15. Integral test of JENDL-3.3 for fast reactors

    International Nuclear Information System (INIS)

    An integral test of JENDL-3.3 was performed for fast reactors. Various types of fast reactors were analyzed. Calculation values of the nuclear characteristics were greatly especially affected by the revisions of the cross sections of U-235 capture and elastic scattering reactions. The C/E values were improved for ZPPR cross where plutonium is mainly fueled, but not for BFS cores where uranium is mainly fueled. (author)

  16. A Belgian Approach to Learning Disabilities.

    Science.gov (United States)

    Hayes, Cheryl W.

    The paper reviews Belgian philosophy toward the education of learning disabled students and cites the differences between American behaviorally-oriented theory and Belgian emphasis on identifying the underlying causes of the disability. Academic methods observed in Belgium (including psychodrama and perceptual motor training) are discussed and are…

  17. 3. Interindustry conference on reactor materials science

    International Nuclear Information System (INIS)

    This document contains abstracts on papers presented at the Third Interindustry Conference on Reactor Materials Science (Dimitrovgrad, 27-30 October 1992). The subject scope of the papers is a follows: fuel and fuel elements of power reactors; structural materials of fast breeder reactors and thermonuclear reactors; structural materials of WWER and RBMK type reactors; absorbers and moderators

  18. Systems analysis of the CANDU 3 Reactor

    International Nuclear Information System (INIS)

    This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ''significant to safety,'' and identification of key operator actions for the analyzed events

  19. 3 Investment Scenarios for Fast Reactors

    International Nuclear Information System (INIS)

    Results: • 4 families of scenarios: – In each of them, 3 options for national nuclear policy → 12 scenarios; – 3 favorable to FRs: - “climate constraint” with strong pro-nuclear policy - “climate constraint” with moderate pro-nuclear policy - “totally green” with strong pro-nuclear policy. • Business As Usual is not favorable to Fast Reactors; Fast reactors deployment: - Needs strong climate policy - Is viable in case of important renewable progress as long as climate policy is strong. International perspective: • Results are valid for Europe, other drivers being likely to be more important in other countries : high growth and demand (Asia); • With strong contrasts between European countries. Further research: • Finer modeling of drivers with unclear influence (clustered and excluded variables): Influence of weak signals

  20. History of the Belgian nuclear power controversy

    International Nuclear Information System (INIS)

    Partly because nuclear energy technology continues to provoke profound controversy, the Flemish institute for technology assessment (viWTA) took the initiative to order a study aimed at mapping out the historical dynamics of the societal debate on nuclear energy. This study was carried out by the Belgian Nuclear Research Centre (SCK-CEN, under the research programme PISA) together with the Free university of Brussels (VUB, research group MEKO) in 2004. In 2007, the report was updated and published by Acco (Leuven) under the title Kernenergie (on)besproken. This study had three main objectives: 1) to discuss the societal debate on nuclear energy in Belgium in relation to major events (Chernobyl, TMI, etc.); 2) to elucidate the role of social actors in the controversy on both a national and international level and 3) to discuss possible alternatives for a better structuring of the debate in the future, building on existing approaches

  1. The reactor dynamics code DYN3D

    International Nuclear Information System (INIS)

    The article provides an overview on the code DYN3D which is a three-dimensional core model for steady-state, dynamic and depletion calculations in reactor cores with quadratic or hexagonal fuel assembly geometry being developed by the Helmholtz-Zentrum Dresden-Rossendorf for more than 20 years. The current paper gives an overview on the basic DYN3D models and the available code couplings. The verification and validation status is shortly outlined. The paper concludes with the current developments of the DYN3D code. For more detailed information the reader is referred to the publications cited in the corresponding chapters.

  2. Integrated modelling of the Belgian coastal zone

    OpenAIRE

    Delhez, E. J. M.; Carabin, G.

    2001-01-01

    The management of the water resources in coastal or delta plains asks for an integrated modelling of the water system at a regional scale. In the SALMON project, detailed descriptions of the groundwater, river and marine domains are provided by coupling appropriate numerical models of these different sub-systems.The application of this three-fold model to the Scheldt and Belgian Coastal Zone reveals a marked river plume extending along the Belgian Coast with strong offshore gradients. This pl...

  3. An ecosystem approach towards Belgian coastal policy

    OpenAIRE

    Vanden Eede, S.

    2013-01-01

    The Belgian coastal zone hosts a complex of space-use and resource-use activities with a myriad of pressures impairing environmental conditions both on the coastline and on coastal waters Specifically at the beach zone, predictions on sea level rise, intensified storms accelerated erosion and flood risk for the North Sea have led to the drafting of the Belgian Integrated Coastal Safety Plan. The preferred coastal defence measure is beach nourishment as it safeguards the natural dynamics of th...

  4. From pralines to multinationals. The economic history of Belgian chocolate

    OpenAIRE

    Garrone, Maria; Pieters, Hannah; Swinnen, Jo

    2015-01-01

    Belgium is associated with high-quality chocolate products and Belgian companies play an important role in cocoa processing. However, in historical perspective the global success and reputation of Belgian chocolate is a relatively recent phenomenon. Especially since the 1980s exports of "Belgian chocolates" have grown exponentially. We document the growth of the sector and discuss its determinants. Today, the very concept of "Belgian chocolate" faces challenges, as successful companies have b...

  5. D-3He fuel cycles for neutron lean reactors

    International Nuclear Information System (INIS)

    The intrinsic potential of D-3He as a reactor fuel is investigated for a large range of 3He to D density ratios. A steady-state zero-dimensional reactor model is developed in which much care is attributed to a proper treatment of fast fusion products. Useful ranges of reactor parameters as well as temperature-density windows for driven and ignited operation are identified. Various figures of merit are calculated, such as power densities, net power production, neutron production, tritium load and radiative power. These results suggest several optimistic conclusions about the performance of D-3He as a reactor fuel

  6. Experiments of fuzzy logic control ion a nuclear research reactor

    International Nuclear Information System (INIS)

    The application of a fuzzy logic control scheme is presented in order to improve the power system stability of BR1 (Belgian Reactor 1) at the Belgian Nuclear Research Centre (SCK-CEN). The control scheme is developed based on OMRON's fuzzy hardware (C200H-FZ001) and the Fuzzy Support Software (FSS) because of their high performance and flexibility. The various possibilities are discussed to find the best or optimal fuzzy logic control scheme for controlling BR1. On basis of the previous researches 1,2,3, some experiments have been carried out in both the steady-state and dynamic operation conditions. The results reveal that the fuzzy logic control scheme has the potential to replace nuclear reactor operators in the control room. Hence, the entire control process can be automatic, simple and effective

  7. Operating reactors licensing actions summary. Vol. 3, No. 3

    International Nuclear Information System (INIS)

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regularory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program

  8. Prospects for Revitalising the Belgian University System.

    Science.gov (United States)

    Hecquet, Ignace

    1984-01-01

    Concerns of a researcher in a Belgian university regarding the trend toward policy implementation dominated by short- and medium-term budgetary considerations are outlined. The university system's heterogeneous, compartmentalized nature and potential are noted, along with the relationship to the government and the real danger of break-up of the…

  9. Performance communication of the Belgian Railway

    NARCIS (Netherlands)

    Gelders, Dave; Verckens, Jan Pieter; Galetzka, Mirjam; Seydel, Erwin

    2007-01-01

    Purpose – The purpose of this paper is to provide an insight into performance communication from an important public service, i.e. the Belgian Railway, towards its employees (internal) and stakeholders (external). Design/methodology/approach – A qualitative research approach was taken in the form o

  10. Helium-3 blankets for tritium breeding in fusion reactors

    Science.gov (United States)

    Steiner, Don; Embrechts, Mark; Varsamis, Georgios; Vesey, Roger; Gierszewski, Paul

    1988-01-01

    It is concluded that He-3 blankets offers considerable promise for tritium breeding in fusion reactors: good breeding potential, low operational risk, and attractive safety features. The availability of He-3 resources is the key issue for this concept. There is sufficient He-3 from decay of military stockpiles to meet the International Thermonuclear Experimental Reactor needs. Extraterrestrial sources of He-3 would be required for a fusion power economy.

  11. Reversed field pinch reactor study 3

    International Nuclear Information System (INIS)

    This report, the third of a series on the Reversed Field Pinch Reactor, describes a preliminary concept of the engineering design and layout of this pulsed toroidal reactor, which uses the stable plasma behaviour first observed in ZETA. The basic parameters of the 600 MW(e) reactor are taken from a companion study by Hancox and Spears. The plasma volume is 1.75m minor radius and 16m major radius surrounded by a 1.8m blanket-shield region - with the blanket divided into 14 removable segments for servicing. The magnetic confinement system consists of 28 toroidal field coils situated just outside the blanket and inside the poloidal and vertical field coils and all coils have normal copper conductors. The requirement to incorporate a conducting shell at the front of the blanket to provide a short-time plasma stability has a marked effect on the design. It sets the size of the blanket segment and the scale of the servicing operations, limits the breeding gain and complicates the blanket cooling and its integration with the heat engine. An extensive study will be required to confirm the overall reactor potential of the concept. (author)

  12. SIMULATE-3 core model for nuclear reactor training simulators

    International Nuclear Information System (INIS)

    This paper describes the adaptation of the Studsvik nuclear reactor analysis code, SIMULATE-3, to nuclear reactor training simulation. This adaption to real-time applications permits training simulation to be performed using the same 'engineering grade' core model used for core design, loading optimisation, safety analysis, and plant technical support. Use of SIMULATE-3R in training simulation permits simple initialisation of simulator core-models (without need for tuning) and facilitates application of cycle-specific core models. SIMULATE-3R permits training simulation of reactor cores with the accuracy normally associated with engineering analysis and enhances the simulator's 'plant analyser' functions. (author)

  13. Contamination incident at RA 3 research reactor

    International Nuclear Information System (INIS)

    The reactor was shutdown after 5 days of operation. While 5 people were checking a new pneumatic facility, it took place a contamination with activation products. The main product was Zn-65, the personnel contamination level fell between 20 and 50 Bq/cm2 and the floor contamination was around 50-70 Bq/cm2. There was no radiological consequences in the personnel. (author)

  14. Focal epilepsy in the Belgian shepherd

    DEFF Research Database (Denmark)

    Berendt, Mette; Gulløv, Christina Hedal; Fredholm, Merete

    2009-01-01

    OBJECTIVES: To establish the mode of inheritance and describe the clinical features of epilepsy in the Belgian shepherd, taking the outset in an extended Danish dog family (199 individuals) of Groenendael and Tervueren with accumulated epilepsy. METHODS: Epilepsy positive individuals (living and...... deceased) were ascertained through a telephone interview using a standardised questionnaire regarding seizure history and phenomenology. Living dogs were invited to a detailed clinical evaluation. Litters more than five years of age, or where epilepsy was present in all offspring before the age of five...... seizures. In seven dogs, seizures could not be classified. The mode of inheritance of epilepsy was simple Mendelian. CLINICAL SIGNIFICANCE: This study identified that the Belgian shepherd suffers from genetically transmitted focal epilepsy. The seizure phenomenology expressed by family members have a...

  15. Optical remote sensing of Belgian coastal waters

    OpenAIRE

    Ovidio, F.; K. Ruddick; Vasilkov, A.; Burenkov, V.

    2001-01-01

    This paper summarises the research conducted at MUMM in optical remote sensing of Belgian coastal waters during the period 1997-2000. The motivation for this research consists of the need to provide information for marine environmental management of coastal eutrophication and sediment transport related problems. The basic products provided by optical remote sensing are maps of chlorophyll concentration and total suspended matter. A key contribution has been made for the atmospheric correction...

  16. Advertising budgeting practices of Belgian industrial marketers.

    OpenAIRE

    François, Pierre

    2003-01-01

    The author reports on the results of a survey of a random sample of 102 belgian industrial companies, which measured which budget setting processes companies use, how they set budgets and the resulting budget composition. The objective of the study was first to compare the results with international practice, and second to try to explain their budgeting practices as a function of company, product and market characteristics measured in the same survey. The major conclusions are mixed : on the ...

  17. Course of operators of the RA-3 reactor

    International Nuclear Information System (INIS)

    Description of the fundamental principles of the nuclear reactors' control systems. The RA-3 reactor's control and measurement systems are principally described, without setting aside the basic criteria for the design of an appropriate instrumentation for the control of a nuclear reactor, as well as the theory on which the functioning of the several detectors and equipments used in a nuclear instrumentation are based. The main purpose of this course is that of serving, preferentially as a text, for the training of personnel which shall perform operation tasks in this reactor. The work includes three well-defined sections. The first two ones make an introduction to the subject, while the third one, extending to more than half-work, deals with the general description of the system in which the control and operation logic of RA-3 are included. (R.J.S)

  18. Neutronics design of upgraded JRR-3 research reactor

    International Nuclear Information System (INIS)

    The research reactor JRR-3 is currently planned to be upgraded by replacing the core and related cooling system. The proposed reactor is a water-moderated and -cooled pool type of 20 MW thermal output. The neutronics calculation was carried out on the core using 20% enriched U.Alsub(x) fuel. The results show that the core performances, such as reactivity, neutron flux, and burnup, are sufficient for beam experiments, material testing, and isotope production. (author)

  19. Conceptual design of D-3He FRC reactor 'ARTEMIS'

    International Nuclear Information System (INIS)

    A comprehensive design study of the D-3He fueled field-reversed configuration (FRC) reactor 'ARTEMIS' is carried out for the purpose of proving its attractive characteristics and clarifying the critical issues for a commercial fusion reactor. The FRC burning plasma is stabilized and sustained in a steady equilibrium by means of a preferential trapping of D-3He fusion-produced energetic protons. A novel direct energy converter for 15MeV protons is also presented. On the bases of a consistent scenario of the fusion plasma production and simple engineering, a compact and simple reactor concept is presented. The design of the D-3He FRC power plant definitely offers the most attractive prospect for energy development. It is environmentally acceptable in view of radio-activity and fuel resources; and the estimated cost of electricity is low compared to a light water reactor. Critical issues concerning physics or engineering for the development of the D-3He FRC reactor are clarified. (author)

  20. 3D computer visualization and animation of CANDU reactor core

    International Nuclear Information System (INIS)

    Three-dimensional (3D) computer visualization and animation models of typical CANDU reactor cores (Darlington, Point Lepreau) have been developed using world-wide-web (WWW) browser based tools: JavaScript, hyper-text-markup language (HTML) and virtual reality modeling language (VRML). The 3D models provide three-dimensional views of internal control and monitoring structures in the reactor core, such as fuel channels, flux detectors, liquid zone controllers, zone boundaries, shutoff rods, poison injection tubes, ion chambers. Animations have been developed based on real in-core flux detector responses and rod position data from reactor shutdown. The animations show flux changing inside the reactor core with the drop of shutoff rods and/or the injection of liquid poison. The 3D models also provide hypertext links to documents giving specifications and historical data for particular components. Data in HTML format (or other format such as PDF, etc.) can be shown in text, tables, plots, drawings, etc., and further links to other sources of data can also be embedded. This paper summarizes the use of these WWW browser based tools, and describes the resulting 3D reactor core static and dynamic models. Potential applications of the models are discussed. (author)

  1. Heterogeneous 3-D SN transport reactor calculations using Attila

    International Nuclear Information System (INIS)

    The Canadian Nuclear Safety Commission (CNSC) is preparing to license diverse reactor technologies (CANDU, LWR, small/research reactors). To this end, the CNSC has acquired the Attila SN transport code to serve as an independent tool in licensing reactor design evaluation. In this paper, we are presenting 3-D large scale parallel benchmark calculations of a small PWR with MOX using Attila SN transport code and their comparison to MCNP Monte Carlo. Our benchmark is that of Nam Zin Cho et al transformed into a new 3-D hexagonal geometry heterogeneous benchmark. It provides an evaluation of Attila code in complex calculations of power reactor core with MOX. In this benchmark, we computed using Attila the keff of the core with Control Rods and generated the assembly and pin powers choosing the pins placed in strong transport boundary layer effect zones. As a reference solution, the Monte Carlo MCNP calculations were obtained. The results show that the full core parallel heterogeneous 3-D SN transport calculations of a power reactor are feasible on a small workstation. Our keff results are within 0.8% (800 pcm) relative difference to MCNP reference result (0.9919) and assembly and pin power results are on the average about 2% and 3.6% different. These results validate the Attila code for nuclear design and licensing work. (author)

  2. EL3 reactor description and safety analysis report

    International Nuclear Information System (INIS)

    The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 1014 neutrons/cm2/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements

  3. Religiosity, Values, and Acculturation: A Study of Turkish, Turkish-Belgian, and Belgian Adolescents

    Science.gov (United States)

    Gungor, Derya; Bornstein, Marc H.; Phalet, Karen

    2012-01-01

    We address the understudied religious dimension of acculturation in acculturating adolescents who combine a religious Islamic heritage with a secularized Christian mainstream culture. The religiosity of 197 Turkish-Belgian adolescents was compared with that of 366 age-mates in Turkey (the heritage culture) and 203 in Belgium (the mainstream…

  4. Fusion reactor materials

    International Nuclear Information System (INIS)

    At the Belgian Nuclear Research Centre SCK-CEN, activities related to fusion focus on environmental tolerance of opto-electronic components. The objective of this program is to contribute to the knowledge on the behaviour, during and after neutron irradiation, of fusion-reactor materials and components. The main scientific activities for 1997 are summarized

  5. Belgian Photography: Towards a Minor Photography.

    OpenAIRE

    Jan Baetens; Hilde Van Gelder; Mieke Bleyen

    2011-01-01

    Abstract: This article investigates Belgian photography from within the national framework. Using the notion of "banal nationalism" (Billig), it explores how even in the case of a nation wi...

  6. Nuclear research reactor 0.5 to 3 MW

    International Nuclear Information System (INIS)

    This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MWTH, with a minimum thermal neutron flux of approx, 1013 n/cm2·sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor building has a ventilation system

  7. Dietary intake of artificial sweeteners by the Belgian population.

    Science.gov (United States)

    Huvaere, Kevin; Vandevijvere, Stefanie; Hasni, Moez; Vinkx, Christine; Van Loco, Joris

    2012-01-01

    This study investigated whether the Belgian population older than 15 years is at risk of exceeding ADI levels for acesulfame-K, saccharin, cyclamate, aspartame and sucralose through an assessment of usual dietary intake of artificial sweeteners and specific consumption of table-top sweeteners. A conservative Tier 2 approach, for which an extensive label survey was performed, showed that mean usual intake was significantly lower than the respective ADIs for all sweeteners. Even consumers with high intakes were not exposed to excessive levels, as relative intakes at the 95th percentile (p95) were 31% for acesulfame-K, 13% for aspartame, 30% for cyclamate, 17% for saccharin, and 16% for sucralose of the respective ADIs. Assessment of intake using a Tier 3 approach was preceded by optimisation and validation of an analytical method based on liquid chromatography with mass spectrometric detection. Concentrations of sweeteners in various food matrices and table-top sweeteners were determined and mean positive concentration values were included in the Tier 3 approach, leading to relative intakes at p95 of 17% for acesulfame-K, 5% for aspartame, 25% for cyclamate, 11% for saccharin, and 7% for sucralose of the corresponding ADIs. The contribution of table-top sweeteners to the total usual intake (sucralose: 3.08 versus 3.03, expressed as mg kg(-1) bodyweight day(-1) at p95) showed that the latter group was not exposed to higher levels. It was concluded that the Belgian population is not at risk of exceeding the established ADIs for sweeteners. PMID:22088137

  8. Measurement of 3D distance from nuclear reactors to detectors in Daya Bay reactor neutrino experiment

    International Nuclear Information System (INIS)

    Background: The Daya Bay neutrino experiment is designed to measure the mixing angle θ13 using anti-neutrinos produced by the reactors of the Daya Bay Nuclear Power Plant (NPP) and the Ling Ao NPP. The distance from nuclear reactors to experiment detectors is needed. Purpose: The aim is to introduce the way of building and measuring the control network during the distance surveying. Methods: The 3D distance was obtained by GPS, total station and laser tracker, and several software and different instruments were used for the combination of data adjustment and coordinate transformation, as well as the correctness checking. Results: Through the actual measurement and data processing, the accuracy of the distance is better than the designed requirement of ±40 mm. Conclusion: The success of the Daya Bay reactor neutrino experiment demonstrates that the result of 3D distance measurement is reasonable and correct, and the survey work makes an important contribution to the Daya Bay reactor neutrino experiment. (authors)

  9. 3D simulation of CANDU reactor regulating system

    International Nuclear Information System (INIS)

    Present paper shows the evaluation of the performance of the 3-D modal synthesis based reactor kinetic model in a closed-loop environment in a MATLAB/SIMULINK based Reactor Regulating System (RRS) simulation platform. A notable advantage of the 3-D model is the level of details that it can reveal as compared to the coupled point kinetic model. Using the developed RRS simulation platform, the reactor internal behaviours can be revealed during load-following tests. The test results are also benchmarked against measurements from an existing (CANDU) power plant. It can be concluded that the 3-D reactor model produces more realistic view of the core neutron flux distribution, which is closer to the real plant measurements than that from a coupled point kinetic model. It is also shown that, through a vectorization process, the computational load of the 3-D model is comparable with that of the 14-zone coupled point kinetic model. Furthermore, the developed Graphical User Interface (GUI) software package for RRS implementation represents a user friendly and independent application environment for education training and industrial utilizations. (authors)

  10. Vitamin D inadequacy in Belgian postmenopausal osteoporotic women

    Directory of Open Access Journals (Sweden)

    Collette Julien

    2007-04-01

    Full Text Available Abstract Background Inadequate serum vitamin D [25(OHD] concentrations are associated with secondary hyperparathyroidism, increased bone turnover and bone loss, which increase fracture risk. The objective of this study is to assess the prevalence of inadequate serum 25(OHD concentrations in postmenopausal Belgian women. Opinions with regard to the definition of vitamin D deficiency and adequate vitamin D status vary widely and there are no clear international agreements on what constitute adequate concentrations of vitamin D. Methods Assessment of 25-hydroxyvitamin D [25(OHD] and parathyroid hormone was performed in 1195 Belgian postmenopausal women aged over 50 years. Main analysis has been performed in the whole study population and according to the previous use of vitamin D and calcium supplements. Four cut-offs of 25(OHD inadequacy were fixed : Results Mean (SD age of the patients was 76.9 (7.5 years, body mass index was 25.7 (4.5 kg/m2. Concentrations of 25(OHD were 52.5 (21.4 nmol/L. In the whole study population, the prevalence of 25(OHD inadequacy was 91.3 %, 87.5 %, 43.1 % and 15.9% when considering cut-offs of 80, 75, 50 and 30 nmol/L, respectively. Women who used vitamin D supplements, alone or combined with calcium supplements, had higher concentrations of 25(OHD than non-users. Significant inverse correlations were found between age/serum PTH and serum 25(OHD (r = -0.23/r = -0.31 and also between age/serum PTH and femoral neck BMD (r = -0.29/r = -0.15. There is a significant positive relation between age and PTH (r = 0.16, serum 25(OHD and femoral neck BMD (r = 0.07. (P Vitamin D concentrations varied with the season of sampling but did not reach statistical significance (P = 0.09. Conclusion This study points out a high prevalence of vitamin D inadequacy in Belgian postmenopausal osteoporotic women, even among subjects receiving vitamin D supplements.

  11. Belgian NPPS fit passive autocatalytic recombiners

    International Nuclear Information System (INIS)

    Because hydrogen production during severe accidents can endanger the containment or safety-graded equipment, the Belgian Utility requested Belgatom in 1989 to perform a comparative study between the possible mitigative measures. From the outcome of that study, the Utility decided to install Passive Autorcatalytic Recombiners (PARs) in the seven Belgian units and mandated Belgatom to carry out the project. The author presents the successive steps from the principle decision to the installation on site. A first section is devoted to the sizing of the catalytic surface and its distribution inside the containment. The results of the method are presented for the 5 units equipped before October 96. Besides the functional requirements, the catalyst must also resist to poisoning agents which can be released not only during severe accidents but also during normal operation or shutdown; the relevant qualification criteria are reminded in the second section. Because the installation of the PARs requires also the design of supports, recommendations are drawn to minimize the design effort and the assembling work. At last, the effectiveness of the PARs entirely relies on a catalyst material of which the potential has to be periodically controlled. A section briefly describes the in-service inspection and the related test procedures

  12. BRAMS --- the Belgian RAdio Meteor Stations

    Science.gov (United States)

    Lamy, H.; Ranvier, S.; Martinez Picar, A.; Gamby, E.; Calders, S.; Anciaux, M.; De Keyser, J.

    2014-07-01

    BRAMS is a new radio observing facility developed by the Belgian Institute for Space Aeronomy (BISA) to detect and characterize meteors using forward scattering. It consists of a dedicated beacon located in the south-east of Belgium and in 25 identical receiving stations spread over the Belgian territory. The beacon transmits a pure sinusoidal wave at a frequency of 49.97 MHz with a power of 150 watts. A complete description of the BRAMS network and the data produced will be provided. The main scientific goals of the project are to compute fluxes, retrieve trajectories of individual objects, and determine physical parameters (speed, ionization, mass) for some of the observed meteor echoes. All these goals require a good knowledge of the radiation patterns of the transmitting and receiving antennas. Simulations have been made and will be validated with in-situ measurements using a UAV/drone equipped with a transmitter flying in the far-field region. The results will be provided. Each receiving station generates around 1 GB of data per day with typical numbers of sporadic meteor echoes of 1500--2000. An automatic detection method of these meteor echoes is therefore mandatory but is complicated by spurious echoes mostly due to airplanes. The latest developments of this automatic detection method will be presented and compared to manual counts for validation. Strong and weak points of the method will be presented as well as a possible alternative method using neural networks.

  13. Characteristics of suicide hotspots on the Belgian railway network.

    Science.gov (United States)

    Debbaut, Kevin; Krysinska, Karolina; Andriessen, Karl

    2014-01-01

    In 2004, railway suicide accounted for 5.3% of all suicides in Belgium. In 2008, Infrabel (Manager of the Belgian Railway Infrastructure) introduced a railway suicide prevention programme, including identification of suicide hotspots, i.e., areas of the railway network with an elevated incidence of suicide. The study presents an analysis of 43 suicide hotspots based on Infrabel data collected during field visits and semi-structured interviews conducted in mental health facilities in the vicinity of the hotspots. Three major characteristics of the hotspots were accessibility, anonymity, and vicinity of a mental health institution. The interviews identified several risk and protective factors for railway suicide, including the training of staff, introduction of a suicide prevention policy, and the role of the media. In conclusion, a comprehensive railway suicide prevention programme should continuously safeguard and monitor hotspots, and should be embedded in a comprehensive suicide prevention programme in the community. PMID:24020492

  14. Experiences with a continuous vibration monitoring system on rotating machines in the Belgian power plants

    International Nuclear Information System (INIS)

    Since the availability of turbogenerators is very important, especially for nuclear plants, some Belgian power plants asked Laborelec to develop a better vibration monitoring with analysis capabilities and the possibility to record data. The main goal was to have a better insight in the vibrational behaviour of some critical machines and to limit the outages due to vibrational problems. The first prototype of a computerised monitoring system was installed in 1985 to survey the Doel 4 turbine from the beginning. Meanwhile 39 turbines have been equipped ranging from 125 MW, 25 years old, fossil fuel plants to large 1000 MW nuclear plants. Some reactor coolant pumps were instrumented too. (orig.)

  15. 1997 Scientific Report[1997 Scientific Report of the Belgian Nuclear Research Centre

    Energy Technology Data Exchange (ETDEWEB)

    Govaerts, P.

    1998-07-01

    The 1997 Scientific Report of the Belgian Nuclear Research Centre SCK-CEN describes progress achieved in nuclear safety, radioactive waste management, radiation protection and safeguards. In the field of nuclear research, the main projects concern the behaviour of high-burnup and MOX fuel, the embrittlement of reactor pressure vessels, the irradiation-assisted stress corrosion cracking of reactor internals, and irradiation effects on materials of fusion reactors. In the field of radioactive waste management, progress in the following domains is reported: the disposal of high-level radioactive waste and spent fuel in a clay formation, the decommissioning of nuclear installations, the study of alternative waste-processing techniques. For radiation protection and safeguards, the main activities reported on are in the field of site and environmental restoration, emergency planning and response and scientific support to national and international programmes.

  16. Nuclear reactor control with fuzzy logic approaches - strengths, weakness, opportunities, and threats

    International Nuclear Information System (INIS)

    As part of the special track on 'Lessons learned from computational intelligence in nuclear applications' at the forthcoming FLINS 2004 conference on Applied Computational Intelligence (Blankenberge, Belgium, September 1-3, 2004), research experiences on fuzzy logic techniques in applications of nuclear reactor control operation are critically reviewed in this presentation. Assessment of four real fuzzy control applications at the MIT research reactor in the US, the FUGEN heavy water reactor in Japan, the BR1 research reactor in Belgium, and a TRIGA Mark III reactor in Mexico will be examined thought a SWOT analysis (strengths, weakness, opportunities, and threats). Special attention will be paid to the current cooperation between the Belgian Nuclear Research Centre (SCK-CEN) and the Mexican Nuclear Centre (ININ) on the fuzzy logic control for nuclear reactor control project under the partial support of the National Council for Science and Technology of Mexico (CONACYT). (Author)

  17. Belgian Workshop (November 2003) - Executive Summary and International Perspective

    International Nuclear Information System (INIS)

    The fourth workshop of the OECD/NEA Forum on Stakeholder Confidence (FSC) was hosted by ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste Management and enriched fissile materials. The central theme of the workshop was 'Dealing with interests, values and knowledge in managing risk' within the Belgian context of local partnerships for the long term management of low-level, short-lived radioactive waste. The four-day workshop started with a half-day session in Brussels giving a general introduction on the Belgian context and the local partnership methodology. This was followed by community visits to three local partnerships, PaLoFF in Fleurus-Farciennes, MONA in Mol, and STOLA in Dessel. After the visits, the workshop continued with two full-day sessions in Brussels. One hundred and nineteen registered participants, representing 13 countries, attended the workshop or participated in the community visits. About two thirds were Belgian stakeholders; the remainder came from FSC member organisations. The participants included representatives of municipal governments, civil society organisations, government agencies, industrial companies, the media, and international organisations as well as private citizens, consultants and academics. The four-day meeting was structured as follows: Day 1 morning was devoted to introductory presentations. Information was given on the general radioactive waste management context in Belgium. Regarding the management of LLW, and in particular the search for a disposal facility site, the workshop heard about the local partnership methodology developed by university researchers of the University of Antwerp and the Fondation Universitaire Luxembourgeoise (FUL). These partnerships between the potential host municipalities and the radwaste agency have the mission to develop an integrated facility proposal adapted to local conditions. Community visits took place on Day 1 afternoon and Day 2. Visits offered an opportunity for

  18. Introduction to D-He(3) fusion reactors

    Science.gov (United States)

    Vlases, G. C.; Steinhauer, L. C.

    1989-01-01

    A review and evaluation of D-He(3) fusion reactor technology is presented. The advantages and disadvantages of the D-He(3) and D-T reactor cycles are outlined and compared. In addition, the general design features of D-He(3) tokamaks and field reversed configuration (FRC) reactors are described and the relative merits of each are compared. It is concluded that both tokamaks and FRC's offer certain advantages, and that the ultimate decision as to which to persue for terrestrial power generation will depend heavily on how the physics performance of each of them develops over the next few years. It is clear that the D-He(3) fuel cycle offers marked advantages over the D-T cycle. Although the physics requirements for D-He(3) are more demanding, the overwhelming advantages resulting from the two order of magnitude reduction of neutron flux are expected to lead to a shorter time to commercialization than for the D-T cycle.

  19. Preparations for the shipment of RA-3 reactor irradiated fuel

    International Nuclear Information System (INIS)

    During the last quarter of 2000, in the Radioactive Waste Management Area of the Argentine National Commission of Atomic Energy (CNEA), located at Ezeiza Atomic Center (CAE), activities associated to the shipment of 207 MTR spent fuels containing high enrichment uranium were carried out within the Foreign Research Reactor/Domestic Research Reactor Receipt Program launched by the US Department of Energy (DOE). The MTR spent fuel shipped to Savannah River Site (SRS) was fabricated in Argentina with 90% enriched uranium of US origin and it was utilized in the operation of the research and radioisotope production reactor RA-3 from 1968 until 1987. After a cooling period at the reactor, the spent fuel was transferred to the Central Storage Facility (CSF) located in the waste management area of CAE for interim storage. The spent fuel (SF) inventory consisted of 166 standard assemblies (SA) and 41 control assemblies (CA). Basically, the activities performed were the fuel conditioning operations inside the storage facility (remote transference of the assemblies to the operation pool, fuel cropping, fuel re-identification, loading in transport baskets, etc.) conducted by CNEA. The loading of the filled baskets in the transport casks (NAC-LWT) by means of intermediate transfer systems and loaded casks final preparations were conducted by NAC personnel (DOE's contractor) with the support of CNEA personnel. (author)

  20. The Belgian experience on the backfitting and safety upgrading of old operating nuclear power plants

    International Nuclear Information System (INIS)

    The paper describes the methodology for backfitting and safety upgrading during the reevaluation of the Belgian NPP's: first generation (Doel-1, Doel-2, Tihange-1) and second generation plants (Doel-3, Doel-4, Tihange-2 and Tihange-3). A list of essential safety subjects and topics is given. The experience has proved the feasibility of a safety upgrading of operating NPP without injury to its availability, the benefit of a close cooperation between owner, engineering company and safety authorities throughout the project. A global approach to solving numerous specific deficiencies along with the optimization of the investments regarding the safety improvement of the NPP is suggested. Further increase of the know-how will be achieved through the present Belgian programme along with similar activities abroad. (R.I.)

  1. Economic evaluation of D-T, D-3He, and catalyzed D-D fusion reactors

    International Nuclear Information System (INIS)

    Because the D-3He reaction generates no neutrons and the D-D reaction can use abundant fuel resources, these reactions are expected to be used in advanced fuel fusion reactors. Economic considerations and engineering problems are important for realizing such reactors as commercial plants. Therefore, we estimate and compare the cost of electricity (COE) from D-T, D-3He, and catalyzed D-D (cat D-D) fusion reactors. D-3He and cat D-D reactors have a low neutron wall load. Therefore, the D-3He reactor has no wall replacement cost. In addition, no tritium breeding system is needed for the D-3He reactor, but 3He gas is rare. Because the reaction rates of the D-3He and D-D reactions are less, D-3He and D-D reactors require highly efficient confinement properties and operation at high ion temperatures. Furthermore, the power densities of D-3He and D-D reactors are smaller than that of the D-T reactor; thus, D-3He and D-D reactors require a large plasma volume. Assuming a high ion temperature (= 60 keV) and high normalized beta (= 7-8), the COE of a D-3He reactor is expected to be similar to that of a D-T reactor. In terms of cost, cat D-D is disadvantageous in comparison with D-3He and D-T reactors. (author)

  2. 3. International conference on catalysis in membrane reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    The 3. International Conference on Catalysis in Membrane Reactors, Copenhagen, Denmark, is a continuation of the previous conferences held in Villeurbanne 1994 and Moscow 1996 and will deal with the rapid developments taking place within membranes with emphasis on membrane catalysis. The approx. 80 contributions in form of plenary lectures and posters discuss hydrogen production, methane reforming into syngas, selectivity and specificity of various membranes etc. The conference is organised by the Danish Catalytic Society under the Danish Society for Chemical Engineering. (EG)

  3. The trigonal nodal SP3 method of the reactor code DYN3D

    International Nuclear Information System (INIS)

    DYN3D is a 3D nodal diffusion code for steady-state and transient analyses of Light-Water Reactors (LWRs) with square and hexagonal fuel assembly geometries. Currently several versions of the DYN3D code are available including a multi-group diffusion and a simplified P3 (SP3) neutron transport option. In this work, the multi-group SP3 method based on trigonal-z geometry was developed. The method is applicable to the analysis of reactor cores with hexagonal fuel assemblies and allows flexible mesh refinement, which is of particular importance for VVER-type Pressurized Water Reactors (PWRs) as well as for innovative reactor concepts including block type High-Temperature Reactors (HTRs) and Sodium Fast Reactors (SFRs). In this paper, the theoretical background for the trigonal SP3 methodology is outlined and the results of a preliminary verification analysis are presented by means of two VVER-440 single assembly test examples with different material compositions. The accordant cross sections and reference solutions were produced by the Monte Carlo code SERPENT. The DYN3D results are shown for 2 and 8 energy groups, respectively, and are in good agreement with the reference solutions. The deviation in the nodal power distribution is about 1%. (author)

  4. Decommissioning of the BR3 reactor: status and perspectives

    International Nuclear Information System (INIS)

    The BR3 plant at Mol in Belgium built at the end of the fifties was the first PWR plant built outside the USA. The reactor had a small net power output (10 MWe) but comprised all the loops and features of a commercial PWR plant. The BR3 plant was operated with the main objective of testing advanced PWR fuels under irradiation conditions similar to those encountered in large commercial PWR plants. The reactor was started in 1962 and shut down in 1987 after 25 years of continuous operation. Since 1989, SCK.CEN is decommissioning the BR3 PWR research reactor. The dismantling of the metallic components including reactor pressure vessel and internals is completed and extensively reported in the literature. The dismantling of auxiliary components and the decontamination of parts of the infrastructure are now going on. The decommissioning progress is continuously monitored and costs and strategy are regularly reassessed. The first part of the paper describes the main results and lessons learned from the reassessment exercises performed in 1994, 1999, 2004 and 2007. Impacts of changes in legal framework on the decommissioning costs will be addressed. These changes concern e.g. licensing aspects, clearance levels, waste management... The middle part of the paper discusses the management of activated and/or contaminated concrete. The costing exercise performed in 1995 highlighted that the management of activated and contaminated concrete is the second main cost item after the dismantling of the reactor pressure vessel and internals. Different possible solutions were studied. These are evacuation as radioactive waste with or without supercompaction, recycling this 'radioactive' grout or concrete for conditioning of radioactive waste e.g. conditioning of metallic waste. The paper will give the results of the cost-benefit analysis made to select the solution retained. The last part of the paper will discuss the end goal of the decommissioning of the BR3. In the final

  5. SIMULATE-3K linkage with reactor systems codes

    International Nuclear Information System (INIS)

    SIMULATE-3K is Studsvik Scandpower's best-estimate three-dimensional core kinetics code. SIMULATE-3K has been coupled to several best-estimate reactor systems codes including, RELAP5-3D, RELAP5-3.3, TRACE V5.0, and RETRAN-3D. The coupled codes can be applied to existing reactors and to advanced reactor designs. The S3K linkage to each of the systems codes is a direct, explicit coupling of the two codes on a synchronous time-step basis. The coupling provides an execution method for the S3K three-dimensional neutronic model using the Nuclear Steam Supply System (NSSS) boundary conditions calculated by the systems code. Also, it allows the S3K calculated total core power and core power distributions to drive the system model core. Detailed calculations from the component codes result in a methodology for analyzing limiting transients such as steam line breaks, rod drops/ejections, and ATWS scenarios. These transient events require detailed three- dimensional core data and information about the behavior of NSSS components. A coupled analysis of these transients is important because the core behavior is closely tied to the NSSS system. For example, to capture the timing and characteristics of the important thermal-hydraulic phenomena and/or operations events, such as valve closures, safety injection, or control system interactions, requires a detailed plant model. The Peach Bottom 2 turbine trip transient is used to assess the accuracy of the coupled code calculations. Comparisons of the important plant parameters to results from RELAP5-3D, RELAP5-3.3, and TRACE V5.0 calculations are shown and discussed. The MSLB benchmark is also used to demonstrate the capabilities of the coupled code systems. Comparisons of the calculated reactor power to the reference data are shown can discussed. The comparisons demonstrate the applicability of S3K, either standalone or coupled with a system analysis code, to properly model system response during accident scenarios. (author)

  6. Decommissioning of the BR3 PWR

    International Nuclear Information System (INIS)

    The dismantling and the decommissioning of nuclear installations at the end of their life-cycle is a new challenge to the nuclear industry. Different techniques and procedures for the dismantling of a nuclear power plant on an existing installation, the BR-3 pressurized-water reactor, are described. The scientific program, objectives, achievements in this research area at the Belgian Nuclear Research Centre SCK-CEN for 1997 are summarized

  7. PR-EDB: Power Reactor Embrittlement Database Version 3

    International Nuclear Information System (INIS)

    The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. 'User-friendly' utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

  8. PR-EDB: Power Reactor Embrittlement Database - Version 3

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Subramani, Ranjit [ORNL

    2008-03-01

    The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

  9. Joint Estimation of Mark-up and Bargaining Power Parameters for Belgian Manufacturing

    OpenAIRE

    Dobbelaere, Sabien

    2002-01-01

    This paper applies several extensions of Hall's (1988) methodology to analyse imperfections in both the product and the labour market for firms in the Belgian manufacturing industry over the period 1988-1995. We investigate (1) the heterogeneity in mark-up and bargaining power parameters among 17 sectors within the manufacturing industry, (2) whether higher bargaining power parameters are associated with higher mark-ups and (3) whether both parameters are influenced by cyclical and competitio...

  10. Helium-3 induced enhancement of tritium production for fusion reactors

    International Nuclear Information System (INIS)

    This report provides the results of an inquiry into the feasibility of enhancing tritium production levels through the activation of helium-3 following its external addition to the moderator system of a CANDU reactor. The physical basis for the scheme lies in the fact that the cross section for the activation of helium-3 to tritium is several orders of magnitude larger than the cross section for deuterium activation. The imminent introduction of a centralized facility for the removal, immobilization, and storage of tritium ensures a supply of helium-3, the product of tritium decay

  11. Uranium-fuel thermal reactor benchmark testing of CENDL-3

    International Nuclear Information System (INIS)

    CENDL-3, the new version of China Evaluated Nuclear Data Library are being processed, and distributed for thermal reactor benchmark analysis recently. The processing was carried out using the NJOY nuclear data processing system. The calculations and analyses of uranium-fuel thermal assemblies TRX-1,2, BAPL-1,2,3, ZEEP-1,2,3 were done with lattice code WIMSD5A. The results were compared with the experimental results, the results of the '1986'WIMS library and the results based on ENDF/B-VI. (author)

  12. New reactor concepts

    International Nuclear Information System (INIS)

    The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost

  13. Exporting Ethnic Divisions? The Political Participation of Belgian Citizens Abroad

    OpenAIRE

    Lafleur, Jean-Michel

    2011-01-01

    The Belgian emigrants’ political participation in home country politics is an issue that has undergone several important developments since the end of the 1990s but has surprisingly been subject to very little research. The absence of the issue of emigration from the Belgian political agenda combined with the limited engagement of this population with the home country thus stands for the absence of research on the topic. The lack of academic interest, in turn, has favoured the development of ...

  14. Dietary habits during adolescence - results of the Belgian Adolux Study

    OpenAIRE

    Paulus, Dominique; Saint-Remy, Annie; JeanJean, Michel

    2001-01-01

    STUDY: To analyse the usual dietary habits of Belgian adolescents from a high cardiovascular risk population. METHODS: A food frequency questionnaire (57 items) was administered to the whole sample. Complementary questions specified some types of food (eg fat content). A subgroup of 234 adolescents gave detailed information on portion size (picture book and food samples). SETTING: Twenty-four secondary schools in the Belgian province of Luxembourg. SUBJECTS: A total of 1,526 adolesce...

  15. Survival Among Belgian Centenarians (1870-1894 Cohorts)

    OpenAIRE

    Liu Yuzhi; Zhang Chunyuan; Foulon, M.; D. Chambre; Poulain, M.

    2001-01-01

    Poulain Michel, Chambre Dany, Foulon Michel.- Survival Among Belgian Centenarians (1870-1894 Cohorts) Calculating the probability of dying among post-centenarians is problematic and often flawed by a high risk of error. This is partly due to the unreliability of statistical data on centenarians, and partly to the small populations concerned. The Belgian centenarian database of over 4,000 centenarians in the 1870 to 1894 birth cohorts used here endeavours to compensate for these two failings. ...

  16. Psychosocial predictors of actual turnover among Belgian health care workers

    OpenAIRE

    Derycke, Hanne; Vlerick, Peter; Clays, Els; D'hoore, William; Braeckman, Lutgart

    2010-01-01

    Background: Turnover of nursing staff is a major challenge for healthcare settings and for healthcare in general, urging the need to improve retention. Aim: The aim was to explore the prospective relations between personal and psychosocial work-related factors and actual turnover among Belgian healthcare workers. Methods: Predictors of actual turnover were assessed using the longitudinal Belgian data from the Nurses Early Exit Study (NEXT). Two self-administered questionnaires with a time...

  17. BWR [boiling water reactor] shutdown margin model in SIMULATE-3

    International Nuclear Information System (INIS)

    Boiling water reactor (BWR) technical specifications require that the reactor be kept subcritical (by some prescribed margin) when at room temperature rodded conditions with any one control rod fully withdrawn. The design of an acceptable core loading pattern may require hundreds or thousands of neutronic calculations in order to predict the shutdown margin for each control rod. Direct, full-core, three-dimensional calculations with the SIMULATE-3 two-group advanced nodal code require 3 to 6 CPU min (on a SUN-4 workstation) for each statepoint/control rod that is computed. Such computing and manpower requirements may be burdensome, particularly during the early core design process. These requirements have been significantly reduced by the development of a fast, accurate shutdown margin model in SIMULATE-3. The SIMULATE-3 shutdown margin model achieves a high degree of accuracy and speed without using axial collapsing approximations inherent in many models. The mean difference between SIMULATE-3 one-group and two-group calculations is approximately - 12 pcm with a standard deviation of 35 pcm. The SIMULATE-3 shutdown margin model requires a factor of ∼15 less CPU time than is required for stacked independent two-group SIMULATE-3 calculations

  18. The Belgian nuclear education network, 5th academic year

    International Nuclear Information System (INIS)

    Full text: In a country where a substantial part of the electricity generation will remain of nuclear origin for a number of years, there is a need for well educated and well trained engineers in this area. Public authorities, regulators and industry brought their support to this initiative. In 2001, the Belgian Nuclear Research Centre SCK CEN and five Belgian universities signed a consortium agreement to set up an education programme in nuclear engineering. This academic year, a sixth university, ULB, joined the programme. The universities involved are now: KUL (Leuven), UG (Ghent), VUB (Brussels), UCL (Louvain-la-Neuve), ULg (Liege) and ULB (Brussels). These seven partners have engaged themselves anew to provide students and young-professionals with a high-standard nuclear engineering programme. The BNEN academic programme is a one-year (60 ECTS) Master-after-Master programme open for holders of a Master degree in engineering. The programme consists of ten courses to be followed mandatory (41 ECTS), the opportunity to select a number of advanced courses at will (up to 4 ECTS worth) and a Master thesis (15 ECTS). The subjects of the courses range form nuclear physics, nuclear reactor theory, nuclear thermo hydraulics to reactor plant operation and control, radiation protection and safeguards and nuclear materials. It also includes courses on nuclear energy and the nuclear fuel cycle. All courses are given in a modular fashion, i.e. the students get a course in the duration of one up to three weeks of continuous lectures and lab sessions. Attention is indeed paid to the fact that most courses are not only theoretical ones, but many of them have exercise sessions and laboratory sessions associated with them. These sessions are organised and thought by the scientific staff at SCK CEN. The number of students enrolling for the BNEN has seen a serious growth since the start of the initiative. The programme does not serve only 'full-time' students, i.e. people having

  19. Mechanical degradation processes: The Belgian experience

    International Nuclear Information System (INIS)

    Design life is merely used in Belgium as a requirement in the 'Design Specification' of some components subjected to known degradation processes, such as stress induced fatigue, embrittlement (irradiation or other), various types of corrosion, wear, erosion, thermal aging (electrical insulation, ...), etc. Design life is in no way directly related to the duration of the plant operation. In that sense design life for the Belgian NPP components includes the values of 20, 30 and 40 years. The oldest plant (20 years design life) has been decommissioned in 1991. The most recent units (40 years design life) have still a good time to go. The intermediate units (30 years design life) started around 1975. Consequently components of these plants need be looked at to determine whether or not deteriorations have occurred. The paper presents the various known mechanical degradation processes and how they affect various components. Emphasis is laid on prevention, mitigation or repair measures that have been or are being taken to avoid that the 'Equipment design life' be the limiting factor in the duration of the plant operation. (author)

  20. Description of the RA-3 research reactor as a model facility

    International Nuclear Information System (INIS)

    The Argentine RA-3 reactor is described as a model facility for the information to be provided to the IAEA in accordance with the requirements of the Model Additional Protocol. RA-3 reactor was designed as a 5 MW swimming pool reactor, moderated and cooled with light water. Its fuel was 90% enriched uranium. The reactor started its operation in 1967, has been modified and improved in many components, including the core, that now is fueled with moderately enriched uranium

  1. Atomization of U3Si2 for research reactor fuel

    International Nuclear Information System (INIS)

    Rotating disk atomization technique is applied to KMRR (Korea Multi-purpose Research Reactor) fuel fabrication. A rotating disk atomizer is designed and manufactured locally and U-4.0 wt. % Si alloy powders are produced. The atomized powders are heat-treated to transform into U3Si and the mixture of U3Si and Al are extruded to fuel meat. Most of the atomized powders are spherical in shape. The microstructure of the powder is fine due to the rapid solidification. The time required for peritectoid reaction is reduced due to the fine microstructures and the resultant U3Si grain size is finer than ever obtained from ingot process. The mechanical properties of the fuel meat are improved: yield strength about 30 %, tensile strength 10% and elongation 250 % increased. (author)

  2. Operating reactors licensing actions summary. Volume 5, No. 3

    International Nuclear Information System (INIS)

    The ''Operating Reactors Licensing Actions Summary'' is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

  3. RA reactor exploitation, task 3.08/01

    International Nuclear Information System (INIS)

    During 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report

  4. Different compositions of pharmaceuticals in Dutch and Belgian rivers explained by consumption patterns and treatment efficiency

    NARCIS (Netherlands)

    Laak, ter T.L.; Kooij, P.J.F.; Tolkamp, H.; Hofman, J.

    2014-01-01

    In the current study, 43 pharmaceuticals and 18 transformation products were studied in the river Meuse at the Belgian-Dutch border and four tributaries of the river Meuse in the southern part of the Netherlands. The tributaries originate from Belgian, Dutch and mixed Dutch and Belgian catchments. I

  5. The Resilience Scale for Adults: Construct Validity and Measurement in a Belgian Sample

    Science.gov (United States)

    Hjemdal, Odin; Friborg, Oddgeir; Braun, Stephanie; Kempenaers, Chantal; Linkowski, Paul; Fossion, Pierre

    2011-01-01

    The Resilience Scale for Adults (RSA) was developed and has been extensively validated in Norwegian samples. The purpose of this study was to explore the construct validity of the Resilience Scale for Adults in a French-speaking Belgian sample and test measurement invariance between the Belgian and a Norwegian sample. A Belgian student sample (N =…

  6. Inventory of nuclear liabilities - The Belgian perspective

    International Nuclear Information System (INIS)

    Like all countries that use radioactive materials for producing electricity or for other peaceful purposes, Belgium is faced with an important challenge: the safe management of all these materials, in both the short and long term. Of course there is a price to pay for this management, which in accordance with the ethical principle of inter-generational fairness should be borne mainly by the current generations. However, it is possible that when the moment has come, the financial resources to cover the costs of decommissioning and remediation of these installations, prove to be insufficient or even completely non-existent: this then results in a nuclear liability. This kind of situation can have several causes, such as an underestimation of the actual costs by the operator or the owner of the nuclear installation or by the holder or the owner of the radioactive materials, negligence, transfer of ownership of the nuclear installation or the nuclear site without transfer of the corresponding provisions, a reduction in the operating time, a bankruptcy as well as ignorance. Because it wishes to avoid the occurrence of new nuclear liabilities, the Belgian legislator, by virtue of article 9 of the programme law of 12.12.97, charged ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, with collecting all the elements that are necessary in order to examine to which degree the decommissioning and remediation costs can be actually covered when the time comes. ONDRAF/NIRAS was specifically charged with ascertaining all facts of a technical and financial nature which should enable the minister responsible for energy to verify whether every operator or owner of a nuclear installation and every holder or owner of radioactive materials have provided in time for the requisite financial resources to cover the future costs of decommissioning and remediation. This evaluation of course also serves to enable the government to take the necessary

  7. RELAP5/MOD 3.3 analysis of Reactor Coolant Pump Trip event at NPP Krsko

    International Nuclear Information System (INIS)

    In the paper the results of the RELAP5/MOD 3.3 analysis of the Reactor Coolant Pump (RCP) Trip event at NPP Krsko are presented. The event was initiated by an operator action aimed to prevent the RCP 2 bearing damage. The action consisted of a power reduction, that lasted for 50 minutes, followed by a reactor and a subsequent RCP 2 trip when the reactor power was reduced to 28 %. Two minutes after reactor trip, the Main Steam Isolation Valves (MSIV) were isolated and the steam dump flow was closed. On the secondary side the Steam Generator (SG) pressure rose until SG 1 Safety Valve (SV) 1 opened. The realistic RELAP5/MOD 3.3 analysis has been performed in order to model the particular plant behavior caused by operator actions. The comparison of the RELAP5/MOD 3.3 results with the measurement for the power reduction transient has shown small differences for the major parameters (nuclear power, average temperature, secondary pressure). The main trends and physical phenomena following the RCP Trip event were well reproduced in the analysis. The parameters that have the major influence on transient results have been identified. In the paper the influence of SG 1 relief and SV valves on transient results was investigated more closely. (author)

  8. SIMULATE-3K: Enhancements and Application to Boiling Water Reactor Transients

    International Nuclear Information System (INIS)

    The SIMULATE-3K (S-3K) reactor analysis code has been applied to a variety of pressurized water reactor (PWR) and boiling water reactor (BWR) transients since 1993. Over the years, many changes have occurred in the S-3K channel hydraulics and ex-core component modeling. This paper summarizes those changes and outlines the status of existing vessel and steam line models. Examples are given for BWR transients that can be analyzed with S-3K

  9. Neutron scattering facilities at the research reactor DR3

    International Nuclear Information System (INIS)

    DR3 is a heavy-water-moderated 10 MW thermal neutron research reactor. The 26 fuel elements contain 2.5-3.5 kg uranium enriched to less than 20% 235U. Neutron beams emerge from four horizontal through-tubes tangential to the reactor core. Two of the horizontal tubes are used for neutron scattering experiments in the field of materials research. The vertical tubes are predominantly used for isotope production and materials testing. The thermal neutron flux is about 3.5x1013 n/cm2/s in the centre of the 7-inch diameter horizontal through-tubes. The thermal neutron flux in equilibrium with the D2O moderator (50 deg. C) has a nearly Maxwellian distribution peaking at 1.1 A. At the maximum flux position in the two horizontal through-tubes used for materials research are installed scatterers designed with a considerably higher scattering power for thermal neutrons than for fast neutrons and gamma-rays. The scatterer is either a 10 mm slab of light water, providing a nearly thermal Maxweellian spectrum at the beam port, or a chamber filled with supercritical hydrogen gas at 16 atmospheres and 38 K, a so-called cold neutron source. The spectrum from a cold source has a considerable flux enhancement in the long wavelength region when compared to the thermal water scatterer. Neutron beams are available for materials research from two thermal and two cold beam ports in the Reactor Hall. One of the cold beams is shared with a 20 meter long cold-neutron guide-tube which provides three beam ports in a separate building, the Neutron House, with could neutrons. Only neutrons that have undergone total reflection from the Ni-coated glass plates in the bent guide-tube arrive at the end of the guide tube in the Neutron House. The angle of total reflection is proportional to the neutron wavelength. Therefore almost no neutrons of wavelength below a certain ''critical'' wavelength are transmitted through the guide-tube, and the experimental equipment installed in the Neutron Houyse

  10. A CO2-strategy for BTC [Belgian Development Agency

    Energy Technology Data Exchange (ETDEWEB)

    Bailly, J. [Prospect C and S, Brussels (Belgium); Hanekamp, E. [Partners for Innovation, Amsterdam (Netherlands)

    2008-09-15

    The CO2 footprint is determined the CO2 strategy is developed for the Belgian Technical Cooperation (BTC). BTC is the Belgian agency for development cooperation, and finances development projects in 23 partner countries. The CO2 footprint covered BTC's activities in 2007 in all their offices worldwide. Footprint and strategy were finalised and adopted by the Executive Board at the end of 2008. Meanwhile, the BTC began with the introduction of the proposed strategy. Partners for Innovation and Prospect were asked to support the introduction of the strategy and to determine the CO2 footprint of 2008.

  11. Coastal flooding risk calculations for the Belgian coast

    OpenAIRE

    Verwaest, T.; Van der Biest, K.; Vanpoucke, Ph.; Reyns, J.; Vanderkimpen, P.; de Vos, L.; De Rouck, J.; Mertens, T.

    2009-01-01

    Coastal flooding risk calculations are carried out for the entire Belgian coastal zone to support the management ofthe coastal defence system. The floodprone low-lying coastal area has an average width of 20 km and is locatedon average 2 m below the surge level of an annual storm. The natural sea defences are sandy beaches anddunes, which have been strengthened by revetments in the coastal towns. The Belgian standard of coastalprotection is to be safe against a surge level with a return perio...

  12. Which future for 3. and 4. generation reactors?

    International Nuclear Information System (INIS)

    After having briefly recalled some characteristics of energy producing nuclear reactors by presenting their three main components (fuel, heat transfer fluid, moderator), and outlined that about twenty types of reactors have been historically tested as prototypes in the USA, Russia, UK and France, the author addresses third generation reactors. He states that these reactors do not display an important technological break with respect to PWRs which are presently exploited in France, but that technical advances are such that one can say they belong to a new generation. He states that the EPR (European pressurized reactor) is amongst the best reactors presently on the market. He outlines its technological advances: safety, increased containment, performance, adaptation to various fuel types, availability, reduction of workers exposure, easier maintenance). Of course, the author evokes construction delays and costs for the Finnish and French reactors. Then, he addresses fourth-generation reactors which comprise six types of system: supercritical water reactors, very high temperature reactors (for non electricity generation applications), and four fast neutron systems. These systems have already been experimented in the past and some will be operated in India and Russia. However, due to the relatively low price of uranium and to the high level of uranium reserves, these fast breeders are not really needed on the short or on the medium term. The author outlines France's commitment in the field of fast breeders

  13. Operating reactors licensing actions summary. Vol. 3, No. 6

    International Nuclear Information System (INIS)

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  14. Saddle point condition for D - 3He tokamak fusion reactor

    International Nuclear Information System (INIS)

    In this paper the concept of a generalized ignition contour map, showing bar PhtT2E, NTE, and T, is used to study the ignition criterion for a D-3He fusion reactor with plasma temperature and density profiles. Direct heating scenarios to the D - 3He ignition regime without the help of deuterium-tritium burning are considered. The machine size and enhancement factor for the confinement time required to reach D - 3He ignition can be simple determined by comparing the height of the operation path with Goldston L-mode scaling and the height of the generalized saddle point. A confinement enhancement factor of 2 to 3 is required in the case of a large plasma current (30 to 80 MA) in a small-aspect-ratio tokamak. On the other hand, for a small plasma current (approx-lt 10 MA), large-aspect-ratio tokamak, an enhancement factor of 5 to 6 is necessary to reach ignition. Fuel dilution effects by fusion products and impurities, the confinement degradation effect due to 14-MeV protons, and the operation paths are also considered. To lower the height of the saddle point, and hence the auxiliary heating power, we optimize the fuel composition and examine operation in the hot ion mode

  15. Safety culture in a Belgian nuclear research centre from a social science point of view

    International Nuclear Information System (INIS)

    This paper is the result of a reflection within the framework of a Ph.D. research at SCK-CEN (Belgian Nuclear Research Centre) in collaboration with the University of Liege. The starting point of the work was the 'safety culture' model presented in the IAEA report 75-INSAG-4. This model is applied to the working organization of the SCK-CEN, also considering the safety culture as an open concept given its multi dimensionality. The methodology is based on three methods: observations, focus groups and interviews. The fieldwork was limited to two main installations: a research reactor, and a dismantling site. The preliminary findings are based on the data resulting from 4 Focus Groups. The most prominent components of a safety culture and the multiplicity of safety cultures in a large organization such as SCK-CEN will be discussed. (author)

  16. EC initiatives promise mixed blessings: a Belgian utility perspective

    International Nuclear Information System (INIS)

    The potential effects on nuclear power of European Community initiatives are analysed from the viewpoint of a Belgian utility. The initiatives fall under the three broad headings of: East-West co-operation; completing the internal market; and carbon dioxide emission. (Author)

  17. Belgian support programme to the IAEA for safeguards implementation

    International Nuclear Information System (INIS)

    The objective of the Belgian Support Programme (BEL SP) is to contribute to the optimization of safeguards measures in accordance with INFCIRC/153 type agreements. This optimization has to take into account cost and effectiveness for the Agency, for the State System of Accounting and Control, for the plant operator and for EURATOM, which is considered an inherent partner. The Belgian support programme has undertaken a series of tasks covering the following domains: Non-destructive Measurement Technology (high resolution gamma spectroscopy, neutron measurements on powders, pins, assemblies and waste, test of Phonid, calorimetry on Pu samples, input tank calibration in a reprocessing plant, combined neutron and gamma measurements on irradiated fuel assemblies, allowing the estimation of burnup, plutonium content and cooling time, measurements of coincidence neutrons from fresh MOX fuel assemblies stored under water), System Studies, Analytical Measurements, Containment and surveillance, Training. Numerous field tests in Belgian in Belgian installations have led to a valuable contribution to the Support Programme to the IAEA, either on an individual basis or through a collaboration effort. In the past, the choice of tasks was based mainly on the availability of nuclear facilities; currently, the trend has shifted to technical contributions, system studies, possibly in a joint effort with other programmes. This approach follows the trend of on-going internationalisation of R and D

  18. The Dutch-Belgian beamline at the ESRF.

    Science.gov (United States)

    Borsboom, M; Bras, W; Cerjak, I; Detollenaere, D; Glastra Van Loon, D; Goedtkindt, P; Konijnenburg, M; Lassing, P; Levine, Y K; Munneke, B; Oversluizen, M; Van Tol, R; Vlieg, E

    1998-05-01

    A brief description is given of the design principles and layout of the Dutch-Belgian beamline at the ESRF. This beamline optimizes the use of the available bending-magnet radiation fan by splitting the beam into two branches, each accommodating two experimental techniques. PMID:15263564

  19. Continuing Vocational Training in Belgian Companies: An Upward Tendency

    Science.gov (United States)

    Buyens, Dirk; Wouters, Karen

    2005-01-01

    Purpose: As part of the European continuing vocational training survey, this paper aims to give an overview of the evolutions in continuing vocational training (CVT) in Belgian companies, by comparing both the results of the survey of 1994 and those of 2000/2001. Design/methodology/approach: In Belgium 1,129 companies took part in the survey of…

  20. The attitudes of Belgian adolescents towards peers with disabilities

    NARCIS (Netherlands)

    Bossaert, Goele; Colpin, Hilde; Pijl, Sip Jan; Petry, Katja

    2011-01-01

    This study aimed to explore Belgian adolescents' attitudes towards peers with disabilities and to explore factors associated with these attitudes. Based on the theory of persuasive communication, this study focused on receiver variables (the "whom"), characteristics of students with disabilities ("c

  1. Open Access to the Belgian Nuclear higher Education Network

    International Nuclear Information System (INIS)

    Under the name of the Belgian Nuclear higher Education Network, five Belgian universities, Universite Catholique de Louvain, Universiteit Gent, Universite de Liege, Vrije Universiteit Brussel have established in 2002, in collaboration with the Belgian Nuclear Research Centre SCK-CEN, a common Belgian Interuniversity Programme of the third cycle leading to the academic degree of Master of Science in Nuclear Engineering. Under the lead of the SCK-CEN a project to use and share the acquired experience of the Consortium BNEN - in order to support the realization of a common European Education Programme in Nuclear Engineering - has been accepted by the European Commission for funding under the EU's Sixth Research Framework Programme.The project wants to contribute actively to the development of a more harmonised approach for education in nuclear sciences and engineering in Europe. It brings the European higher Education Area closer to realization and helps to safeguard the necessary competence and expertise for the continued safe use of nuclear energy and other uses of radiation in industry and medicine in Europe. The project foresees input and participation from stakeholders from different countries of the enlarged European Union (EU-25) and will therefore contribute to the integration of the new member states into the European Research Area and thus to the enlargement of Europe. The set-up of the project foresees an active role for female experts with the intention to reinforce the place and role of women in science

  2. Development of telerobotic systems for reactor decommissioning, (3)

    International Nuclear Information System (INIS)

    This paper describes the telerobotic system for reactor decommissioning in the scope of engineering demonstration of dismantling radioactive reactor internals of an experimental boiling water power reactor JPDR. The total system consists of a telerobotic manipulator system equipped with a multi-functional amphibious slave manipulator with a load capacity of 25 daN, a chain-driven transport system, and a computer-assisted monitoring and control system. Preceding to the application of the telerobotic system to actual dismantling operation, a mockup test was performed of dismantling the simulated reactor internals of actual-size by the method of underwater plasma arc cutting in order to study the performance of the telerobotic system in a realistic environment. The system was then successfully applied to dismantling the actual reactor internals according to the JPDR decommissioning program. (author)

  3. Minimum core configuration with IRT-3M fuel in the VR-1 reactor

    International Nuclear Information System (INIS)

    The present paper shortly describes advances of the RERTR program in the Czech Republic. The minimum core configuration B2 with 9 fuel elements IRT-3M and Beryllium reflector was performed on the training reactor VR-1 Sparrow. The paper presents results of reactor calculations and experimental measurements on the core configuration B2, their evaluation as well as the operation experiences with the Russian fuel assemblies IRT-3M on the reactor VR-1. (author)

  4. Safety regulations of fuzzy-logic control to nuclear reactors

    OpenAIRE

    RUAN, Da

    2000-01-01

    We present an R&D project on fuzzy-logic control applications to the Belgian Nuclear Reactor 1 (BR1) at the Belgian Nuclear Research Centre (SCK•CEN). The project started in 1995 and aimed at investigating the added value of fuzzy logic control for nuclear reactors. We first review some relevant literature on fuzzy logic control in nuclear reactors, then present the state-of-the-art of the BR1 project, with an understanding of the safety requirements for this real fuzzy-logic control ...

  5. RELAP5/MOD3 model and transport analyses for Maria research reactor in Poland

    International Nuclear Information System (INIS)

    RELAP5/MOD3 input data model of the Maria research reactor has been developed to provide the capability for the analysis of the reactor core under loss of flow and reactivity insertion transients. The model was qualified against the reactor data at steady state conditions and, additionally, against the existing reliable experimental data for the transient initiated by the reactor scram. The results obtained with the code agree well with the experimental data. The RELAP transient simulation was performed for loss of forced flow accidents including two scenarios with with protected and unprotected (no scram) reactor core. Calculations allow estimating time margin for reactor scram initiation and reactivity feedbacks contribution to the results. The presented input data model should be treated as a first step for developing of the model including the whole primary cooling circuit of the reactor. (author)

  6. A survey of bacteria found in Belgian dairy farm products

    Directory of Open Access Journals (Sweden)

    N'Guessan, E.

    2015-01-01

    Full Text Available Description of the subject. Due to the potential hazards caused by pathogenic bacteria, farm dairy production remains a challenge from the point of view of food safety. As part of a public program to support farm diversification and short food supply chains, farm dairy product samples including yogurt, ice cream, raw-milk butter and cheese samples were collected from 318 Walloon farm producers between 2006 and 2014. Objectives. Investigation of the microbiological quality of the Belgian dairy products using the guidelines provided by the European food safety standards. Method. The samples were collected within the framework of the self-checking regulation. In accordance with the European Regulation EC 2073/2005, microbiological analyses were performed to detect and count Enterobacteriaceae, Listeria monocytogenes, Salmonella spp., Escherichia coli and Staphylococcus aureus. Results. Even when results met the microbiological safety standards, hygienic indicator microorganisms like E. coli and S. aureus exceeded the defined limits in 35% and 4% of butter and cheese samples, respectively. Unsatisfactory levels observed for soft cheeses remained higher (10% and 2% for S. aureus and L. monocytogenes respectively than those observed for pressed cheeses (3% and 1% and fresh cheeses (3% and 0% (P ≥ 0.05. Furthermore, the percentages of samples outside legal limits were not significantly higher in the summer months than in winter months for all mentioned bacteria. Conclusions. This survey showed that most farm dairy products investigated were microbiologically safe. However, high levels of hygiene indicators (e.g., E. coli in some products, like butter, remind us of applying good hygienic practices at every stage of the dairy production process to ensure consumer safety.

  7. New trends in the Belgian programme on nuclear air cleaning technology

    International Nuclear Information System (INIS)

    In the Belgian Programme on nuclear air cleaning technology the Mercurex process has been developed to trap iodine compounds from dissolver off-gases. Krypton is removed with the help of a cryogenic distillation unit. The various gas cleaning units have been integrated in a gas purification test loop for dissolver off-gas at a through put of 25 m3 gas h-1. The separation of tritium from liquid reprocessing effluents is being developed according to the ELEX-process. New research is started on the capture of semi-volatile ruthenium compounds

  8. Nuclear Reactor RA Safety Report, Vol. 3, Building and installations

    International Nuclear Information System (INIS)

    RA reactor building is built of concrete and bricks as an enclosed building with limited number of controlled openings, and limited number of doors and windows. It is made of three parts: central; circular annex in the central part; sanitary corridor. The largest part of the RA reactor building is the reactor hall. This volume includes detailed description, figure and diagrams showing building characteristics, power supply systems, water supply systems, ventilation and heating systems, gas and compressed air installation as well as fire prevention system

  9. Reactors

    International Nuclear Information System (INIS)

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

  10. Belgian experience in applying the {open_quotes}leak-before-break{close_quotes} concept to the primary loop piping

    Energy Technology Data Exchange (ETDEWEB)

    Gerard, R.; Malekian, C.; Meessen, O. [Tractebel Energy Engineering, Brussels (Belgium)

    1997-04-01

    The Leak Before Break (LBB) concept allows to eliminate from the design basis the double-ended guillotine break of the primary loop piping, provided it can be demonstrated by a fracture mechanics analysis that a through-wall flaw, of a size giving rise to a leakage still well detectable by the plant leak detection systems, remains stable even under accident conditions (including the Safe Shutdown Earthquake (SSE)). This concept was successfully applied to the primary loop piping of several Belgian Pressurized Water Reactor (PWR) units, operated by the Utility Electrabel. One of the main benefits is to permit justification of supports in the primary loop and justification of the integrity of the reactor pressure vessel and internals in case of a Loss Of Coolant Accident (LOCA) in stretch-out conditions. For two of the Belgian PWR units, the LBB approach also made it possible to reduce the number of large hydraulic snubbers installed on the primary coolant pumps. Last but not least, the LBB concept also facilitates the steam generator replacement operations, by eliminating the need for some pipe whip restraints located close to the steam generator. In addition to the U.S. regulatory requirements, the Belgian safety authorities impose additional requirements which are described in details in a separate paper. An novel aspect of the studies performed in Belgium is the way in which residual loads in the primary loop are taken into account. Such loads may result from displacements imposed to close the primary loop in a steam generator replacement operation, especially when it is performed using the {open_quote}two cuts{close_quotes} technique. The influence of such residual loads on the LBB margins is discussed in details and typical results are presented.

  11. Calculation of fundamental parameters for the dynamical study of TRIGA-3-Salazar reactor (Mixed reactor core)

    International Nuclear Information System (INIS)

    Kinetic parameters for dynamic study of two different configurations, 8 and 9, both with standard fuel, 20% enrichment and Flip (Fuel Life Improvement Program with 70% enrichment) fuel, for TRIGA Mark-III reactor from Mexico Nuclear Center, are obtained. A calculation method using both WIMS-D4 and DTF-IV and DAC1 was established, to decide which of those two configurations has the best safety and operational conditions. Validation of this methodology is done by calculate those parameters for a reactor core with new standard fuel. Configuration 9 is recommended to be use. (Author)

  12. The French authority of nuclear safety (ASN) authorizes the restart-up of Cattenom's reactor 3

    International Nuclear Information System (INIS)

    On August 31, 2001, the French authority of nuclear safety (ASN) gave permission to Electricite de France (EdF) to restart the reactor no 3 of Cattenom's power plant. This reactor encountered important degradations of its fuel assemblies during its previous operating cycle which led to a level 1 incident on the INES scale. Thus the ASN has imposed to EdF a reinforced surveillance of the primary circuit of the reactor. This document brings together the different press releases, information notes, and rulings that have been written by the ASN about the defects that occurred on the fuel rods of Cattenom's reactor 3. (J.S.)

  13. Level 3 decommissioning of Triton - Nereide research reactor

    International Nuclear Information System (INIS)

    The French Atomic Energy Commission Center located at Fontenay-Aux-Roses has launched an extensive programme of site cleanup and decommissioning of nuclear facilities. This programme includes the level 3 decommissioning of the Triton and Nereide piles. These pool type research reactors were constructed in the late 1950's, primarily for R and D activities related to neutron physics studies, radiological shielding experiments and radioelement production. As of 1982, a level 2 decommissioning was achieved and over the the last twenty years, no activities were carried out in the facility. During 2001, there has been extensive investigation work carried out to acquire a better knowledge of the radiological status of the facility, in order to set up dismantling scenarios and to reduce the volume of generated radioactive waste. Indeed, one of the first and main operations to be carried out for dismantling Triton and Nereide piles is waste zoning, by using the facility layout, operating conditions and history, as well as the present radiological inventory. The paper describes the investigations and studies carried out to implement waste zoning. The paper also describes the preliminary dismantling operations undertaken on equipment and studies conducted to optimize the dismantling and cleanup of the facility. Finally, the paper presents the outline of the preferred dismantling and decommissioning options and the progress of the work to date. (author)

  14. The Belgian laboratory for standard dosimetry calibrations used in radiotherapy

    International Nuclear Information System (INIS)

    Starting from the end of the year 2008, the RDC (Radiation Protection dosimetry and Calibrations) expertise group of SCK CEN took over the calibration and research activities at the Laboratory for Standard Dosimetry Ghent. The laboratory runs under a collaboration between SCK CEN and the University of Ghent, with the support of Federal Agency for Nuclear Control (FANC). The calibrations in Ghent were stopped at the beginning of 2008 and then restarted at the end of 2008. A new 60Co source was installed at Ghent, a Theratron 780 unit. All the calibration setups installed in the past to the old 60Co source had to move to the new source and measurement history had to be acquired. The calibration of cylindrical and plane-parallel ionization chambers in terms of absorbed dose to water was defined as the first priority, since there was an urgent need from the Belgian hospitals. These calibrations are presently done in Ghent as secondary standard calibrations, traceable to the water calorimeter of VSL, Delft, The Netherlands and following the recommendations from TRS-398 protocol. The second priority was restarting the calibrations of cylindrical ionization chambers in terms of air kerma. A cylindrical graphite ionization chamber of type CC01 is used for the absolute measurement of air kerma. Both setups are fully operational. Special efforts were done to implement the SCK CEN quality assurance (QA) system regarding ISO 17025 accreditation. The activity at the laboratory in Ghent was integrated as part of the Laboratory for Nuclear Calibrations (LNK-from the Dutch translation) of the SCK-CEN. Most of the activities of the LNK are already accredited by Belgian Accreditation Body (BELAC) with respect to the ISO-17025 standards. The quality assurance procedures were prepared and are routinely followed for the two new setups mentioned above: calibrations in terms of absorbed dose to water and air kerma in 60Co beam. During the preparation of the quality assurance procedures

  15. The SPES3 Experimental Facility Design for the IRIS Reactor Simulation

    OpenAIRE

    Alessandro Alemberti; Fabio Berra; Davor Grgic; Graydon Yoder; Stefano Monti; Paride Meloni; Davide Papini; Fosco Bianchi; Marco Ricotti; Roberta Ferri; Cinzia Congiu; Gustavo Cattadori; Andrea Achilli; Bojan Petrovic; Andrea Maioli

    2009-01-01

    IRIS is an advanced integral pressurized water reactor, developed by an international consortium led by Westinghouse. The licensing process requires the execution of integral and separate effect tests on a properly scaled reactor simulator for reactor concept, safety system verification, and code assessment. Within the framework of an Italian R&D program on Nuclear Fission, managed by ENEA and supported by the Ministry of Economic Development, the SPES3 facility is under design and will be bu...

  16. Neutron flux determinations in the reactors G2 and G3 during operation

    International Nuclear Information System (INIS)

    After demonstrating the sensitivity of the distribution of power in a production reactor to a deformation caused by dissymmetries of reactivity in the reactor, the authors describe the method of neutron flux determination devised for the reactors G2 and G3 under working conditions; the detector used is a tungsten or nickel wire, the γ activity of which is measured with an ionisation chamber. Several flux determinations are given as examples to illustrate the sensitivity of the method. (author)

  17. Non destructive examination of Reactor DR-3. Reactor wall, horisontal experimental tubes, up- and down comers

    International Nuclear Information System (INIS)

    The initial scope of work was to perform thickness/corrosion measurements of one up-comer and one down-comer, perform thickness/corrosion measurements in selected areas of the reactor wall and horizontal experimental pipes inside the reactor. Furthermore the lower circumferential weld and the connected longitudinal weld should be inspected to the extent possible, without major changes of the manipulator. Eddy current was performed in the same areas. Also hardness tests were carried out in four positions inside the reactor. Due to the outcome of the above examinations, additional metallurgical and dye penetrant examinations (PT) were carried out. The examination of the up- and down comers showed no sign of serious service induced defects. The eddy current testing did not reveal any inner surface breaking defects. The thickness/corrosion ultrasonic measurement showed only minor local indications with small or no reductions of original nominal wall thickness. The examination of the horizontal tubes showed no sign of serious service induced defects. The eddy current testing did not reveal any inner surface breaking defects. The thickness/corrosion ultrasonic measurement showed only minor local indications with small or no reductions of original nominal wall thickness. The hardness test showed increased hardness compared to calibration values. The examination of the reactor wall base material revealed several indications located in different depths in the plate. Some indications have been proved to be connected to the inner surface, while most indications appear to be either inclusions or areas corroded from the outside reactor wall. Minimum measured wall thickness is between 4.2 and 11.0 mm. There is, however, no evidence that these values are caused by corrosion at the outer reactor surface. The ET showed no signs of service induced cracks. The hardness test showed values close to calibration values. The extensive number of indications has resulted in additional

  18. Extension and application of the reactor dynamics code DYN3D for Block-type High Temperature Reactors

    International Nuclear Information System (INIS)

    The reactor code DYN3D was developed at the Helmholtz-Zentrum Dresden-Rossendorf to study steady state and transient behavior of Light Water Reactors. Concerning the neutronics part, the multigroup diffusion or SP3 transport equation based on nodal expansion methods is solved both for hexagonal and square fuel element geometry. To deal with Block-type High Temperature Reactor cores DYN3D was extended to a version DYN3D-HTR. A 3D heat conduction model was introduced to include 3D effects of heat transfer and heat conduction and the detailed structure of the fuel element. Homogenized neutronic cross sections were generated by applying a Monte Carlo approach with resolution of each individual TRISO fuel particle. Results of coupled steady state and transient calculations with 12 energy groups are presented. Transient case studies are control rod insertion, a change of the inlet coolant temperature and a change of the coolant gas mass flow rate. It is shown that DYN3D-HTR is an appropriate code system to simulate steady states and short time transients. Furthermore the necessity of the 3D heat conduction model is demonstrated

  19. Belgian nuclear forum - launching the public debate on nuclear energy

    International Nuclear Information System (INIS)

    In the past decades, public opinion on nuclear power was dominated by a 'sleeping', indifferent majority. Nothing moved until (a minority of) opponents began to stir. Their activism strongly contrasted with the low-profile attitude of the nuclear players and pushed a considerable part of the indifferent majority towards a more negative attitude. A 2007 opinion poll (IFOP) confirmed this trend. The poll also revealed a major lack of objective and factual information. Incorrect and incomplete arguments tended to demonize nuclear energy, and 'nuclear' became a brand polarizing public opinion. This had a negative impact on decision-makers and culminated in the Belgian phase-out law of 2003. Based on the opinion poll, the members of the Belgian Nuclear Forum decided to launch a public information campaign, which they would jointly finance, with these goals: - In 3 to 4 years time, create greater public awareness on energy matters and move public opinion towards a more positive attitude. - Gain recognition of nuclear energy's legitimate place in the mix, and of the importance of peaceful nuclear applications. - Attract attention to the Belgian know-how and the importance of the industry on the scientific and economical level. - Optimize conditions for important nuclear issues such as long-term operation of NPPs, new nuclear research projects (MYRRHA),.. A 'push-pull' approach was adopted: push communication to the public (campaign) to pull (involve) decision-makers and get nuclear back on the political agenda. The Forum also opted for a sustained, long-term effort based on public campaigning, public relations and public affairs. Considering its long-time absence from the public debate, the Forum and its agency Saatchi and Saatchi agreed upon the following principles to underpin the campaign: - No 'pro-campaign'; that would be highly unrealistic and have a negative effect; - No taboos: bring up both the pros and cons; - No emotions: bring reason into a mainly emotional

  20. Exploring Pupils' Perceptions of Teacher Racism in Their Context: A Case Study of Turkish and Belgian Vocational Education Pupils in a Belgian School

    Science.gov (United States)

    Stevens, Peter A. J.

    2008-01-01

    This article employs ethnographic data gathered from one Belgian (Flemish) secondary school to explore the meaning Belgian and Turkish-speaking minority pupils enrolled in technical and vocational education attach to teacher racism and racial discrimination, and to explore variations between pupils in making claims of teacher racism. A symbolic…

  1. Did the « Pax Electrica » agreements reduce the Suez’ power relationship on Elia, the Belgian electricity grid manager ?

    OpenAIRE

    Levy, Marc

    2009-01-01

    The company Elia, which manages the grid system of electricity in Belgium, is mainly controlled by the principal producer of electricity, Electrabel, and by political powers. The Belgian state has recently constrained the French group to yield 3% of its stake in Elia as foreseen in the Pax Electrica agreements. This article studies the impact of this transfer by means of the Banzhaf index that measures the intensity of control. The results show that theoretically these agreements change the p...

  2. Belgian nuclear forum - launching the public debate on nuclear energy

    Energy Technology Data Exchange (ETDEWEB)

    Leclere, Robert [Belgian Nuclear Forum, Gulledelle, 1200 Brussels (Belgium); Van Landeghem, Yves [Saatchi and Saatchi Belgium, Avenue Rogier, 1030 Brussels (Belgium)

    2010-07-01

    In the past decades, public opinion on nuclear power was dominated by a 'sleeping', indifferent majority. Nothing moved until (a minority of) opponents began to stir. Their activism strongly contrasted with the low-profile attitude of the nuclear players and pushed a considerable part of the indifferent majority towards a more negative attitude. A 2007 opinion poll (IFOP) confirmed this trend. The poll also revealed a major lack of objective and factual information. Incorrect and incomplete arguments tended to demonize nuclear energy, and 'nuclear' became a brand polarizing public opinion. This had a negative impact on decision-makers and culminated in the Belgian phase-out law of 2003. Based on the opinion poll, the members of the Belgian Nuclear Forum decided to launch a public information campaign, which they would jointly finance, with these goals: - In 3 to 4 years time, create greater public awareness on energy matters and move public opinion towards a more positive attitude. - Gain recognition of nuclear energy's legitimate place in the mix, and of the importance of peaceful nuclear applications. - Attract attention to the Belgian know-how and the importance of the industry on the scientific and economical level. - Optimize conditions for important nuclear issues such as long-term operation of NPPs, new nuclear research projects (MYRRHA),.. A 'push-pull' approach was adopted: push communication to the public (campaign) to pull (involve) decision-makers and get nuclear back on the political agenda. The Forum also opted for a sustained, long-term effort based on public campaigning, public relations and public affairs. Considering its long-time absence from the public debate, the Forum and its agency Saatchi and Saatchi agreed upon the following principles to underpin the campaign: - No 'pro-campaign'; that would be highly unrealistic and have a negative effect; - No taboos: bring up both the pros and cons; - No

  3. Fast reactor 3D core and burnup analysis using VESTA

    Energy Technology Data Exchange (ETDEWEB)

    Luciano, N.; Shamblin, J.; Maldonado, I. [Nuclear Engineering Dept., Univ. of Tennessee, Knoxville, TN 37996-2300 (United States)

    2012-07-01

    Burnup analyses using the VESTA code have been performed on a MOX-fuelled fast reactor model as specified by an IAEA computational benchmark. VESTA is a relatively new code that has been used for burnup credit calculations and thermal reactor models, but not typically for fast reactor applications. The detailed input and results of the IAEA benchmark provides an opportunity to gauge the use of VESTA in a fast reactor application. VESTA employs an ultra-fine multi-group binning approach that accelerates Monte Carlo burnup calculations. Using VESTA to compute the end of cycle (EOC) power fractions by enrichment zone showed agreement with the published values within 5%. When comparing the ultra-fine multi-group binning approach to the tally-based approach, EOC isotopic masses also agree within 5%. Using the ultra-fine multi-group binning approach, we obtain a wall-time speedup factor of 35 when compared to the tally-based approach for computing a k{sub eff} eigenvalue with burnup problem. The authors conclude the use of VESTA's ultra-fine multi-group binning approach with Monte Carlo transport performs accurate depletion calculations for this fast reactor benchmark. (authors)

  4. RELAP5/MOD3.3 analysis of reactor trip event in nuclear power plant

    International Nuclear Information System (INIS)

    Measured plant data from various abnormal events or incidents are of great importance for assessing large system thermal-hydraulic computer codes like RELAP5. In the present study the reactor trip, which occurred at Krsko Nuclear Power Plant (NPP) on April 10, 2005, has been analyzed. The purpose of the analysis was to assess the RELAP5/MOD3.3 Patch 03 computer code against plant measured data and validate the RELAP5 input model for Krsko NPP, which is a two-loop Westinghouse pressurized water reactor. The RELAP5 input model delivered by Krsko NPP was used. The event analyzed was a malfunction, which occurred during a power reduction sequence when regular periodic testing of the turbine valves was performed. This caused plant trip. The calculation agrees very well with the plant measured data when operator actions are modelled properly. It was found out that the long term transient evolution is very sensitive to the steam flows from the steam generators after the reactor trip and only proper modelling of these flows gives good quantitative agreement. (author)

  5. Development of facilities to irradiate materials in the RA1 and RA3 experimental reactors

    International Nuclear Information System (INIS)

    To study the properties of the materials under irradiation, devices and facilities were designed to work at experimental reactors of National Atomic Energy Commission. The radiological protection of the operators and the influence of the irradiated materials on the radiological inventory of the reactors were the most important aspects considered during the design stage. In the present work devices to operate in the argentine reactor 'Reactor Argentino (RA)', RA1 and RA3 experimental reactors are shown. These devices are dedicated to the study of the radiation damage by measuring property changes related to dimensional integrity and embrittlement of materials in zirconium alloys, steels and other materials used in nuclear reactors. The emphasis is on the previsions adopted to minimize the activation of their components and the criteria applied to guarantee the safety of the operators during their performance and after their subsequent dismantling. (author)

  6. Proceedings of 2. Yugoslav symposium on reactor physics, Part 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    This Volume 3 of the Proceedings of 2. Yugoslav symposium on reactor physics includes three papers describing the following: model for spatial synthesis of automated control system of the GCR type reactor; model for analysis of hydrodynamic processes at the BHWR type reactors; mathematical model for safety analysis of heavy water power reactor

  7. THESEE-3, Orgel Reactor Performance and Statistic Hot Channel Factors

    International Nuclear Information System (INIS)

    1 - Nature of physical problem solved: The code applies to a heavy-water moderated organic-cooled reactor channel. Different fuel cluster models can be used (circular or hexagonal patterns). The code gives coolant temperatures and velocities and cladding temperatures throughout the channel and also channel performances, such as power, outlet temperature, boiling and burn-out safety margins (see THESEE-1). In a further step, calculations are performed with statistical values obtained by random retrieval of geometrical in- put data and taking into account construction tolerances, vibrations, etc. The code evaluates the mean value and standard deviation for the more important thermal and hydraulic parameters. 2 - Method of solution: First step calculations are performed for nominal values of parameters by solving iteratively the non-linear system of equations which give the pressure drops in subchannels of the current zone (see THESEE-1). Then a Gaussian probability distribution of possible statistical values of the geometrical input data is assumed. A random number generation routine determines the statistical case. Calculations are performed in the same way as for the nominal case. In the case of several channels, statistical performances must be adjusted to equalize the normal pressure drop. A special subroutine (AVERAGE) then determines the mean value and standard deviation, and thus probability functions of the most significant thermal and hydraulic results. 3 - Restrictions on the complexity of the problem: Maximum 7 fuel clusters, each divided into 10 axial zones. Fuel bundle geometries are restricted to the following models - circular pattern 6/7, 18/19, 36/67 rods, with or without fillers. The fuel temperature distribution is not studied. The probability distribution of the statistical input is assumed to be a Gaussian function. The principle of random retrieval of statistical values is correct, but some additional correlations could be found from a more

  8. The Mandate System for the Belgian Public Prosecution

    Directory of Open Access Journals (Sweden)

    Bruno BROUCKER

    2009-12-01

    Full Text Available The law of 22 December 1998 introduced the mandate system for the heads of the Public Prosecution offices, which were appointed permanent before that. Theoretically, such a system needs to enhance, within the organization, effectiveness, efficiency, responsabilisation, and goal-orientation. However, the mandate system within the Belgian Public Prosecution was introduced prematurely, for dubious reasons and in a precipitate manner. In the current situation, the position of the mandate holder is uncertain, with a bounded autonomy and a low wage increase. Moreover, it remains impossible to intervene in the policy of appointed heads of office (during their mandate, the efficiency and effectiveness is only increased in some prosecution offices and a contract containing actual management responsibilities is absent. In sum: there is a large gap between the theoretical principles of mandate systems and the way it is introduced in the Belgian Public Prosecution.

  9. EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1969-02-01

    The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99

  10. Shielding analyses for design of the upgraded JRR-3 research reactor, 1

    International Nuclear Information System (INIS)

    Shielding analyses for design of the upgraded JRR-3 research reactor have been performed. In the report described are the design principles and the overall analytical procedures. In addition, described are the results of shielding analyses of reactor, canal, spent fuel storage pond and so on. (author)

  11. Benchmark for a 3D Monte Carlo boiling water reactor fluence computational package - MF3D

    International Nuclear Information System (INIS)

    A detailed three dimensional model of a quadrant of an operating BWR has been developed using MCNP to calculate flux spectrum and fluence levels at various locations in the reactor system. The calculational package, MF3D, was benchmarked against test data obtained over a complete fuel cycle of the host BWR. The test package included activation wires sensitive in both the fast and thermal ranges. Comparisons between the calculational results and test data are good to within ten percent, making the MF3D package an accurate tool for neutron and gamma fluence computation in BWR pressure vessel internals. (orig.)

  12. Management and Program Effectiveness in Belgian Sports Clubs

    OpenAIRE

    A. BALDUCK; M. BUELENS; Maes, M.

    2009-01-01

    This study investigated management and program effectiveness using the competing values approach as theoretical framework. The sample consisted of 823 board and sports members of Belgian sports clubs. Two scales were developed. Factor analysis revealed 12 management and 9 program effectiveness dimensions. Reliability scores were acceptable. Results showed that both board and sports members rated the dimension atmosphere at management and program level as the most effective factor in sports cl...

  13. Characteristics and challenges of the modern Belgian veal industry

    OpenAIRE

    Pardon, Bart; CATRY, Boudewijn; Boone, Randy; Theys, Hubert; De Bleecker, Koen; Dewulf, Jeroen; Deprez, Piet

    2014-01-01

    In this paper, the modern Belgian veal industry is situated in a European context, and an overview is provided of the major past, present and future challenges for veal production. The production of white veal requires a specific diet and housing conditions to assure a controlled iron anemic state resulting in pale carcasses. In response to the increasing public concern about animal welfare, legal limits for hemoglobin (in 1990), the provision of a minimum quality of solid feed to assure rumi...

  14. A survey of bacteria found in Belgian dairy farm products

    OpenAIRE

    N'Guessan, E.; Godrie, T.; de Laubier, J.; di Tanna, S.; Ringuet, M.; Sindic, M.

    2015-01-01

    Description of the subject. Due to the potential hazards caused by pathogenic bacteria, farm dairy production remains a challenge from the point of view of food safety. As part of a public program to support farm diversification and short food supply chains, farm dairy product samples including yogurt, ice cream, raw-milk butter and cheese samples were collected from 318 Walloon farm producers between 2006 and 2014. Objectives. Investigation of the microbiological quality of the Belgian dairy...

  15. Development path and capital structure of belgian biotechnology firms

    OpenAIRE

    Véronique Bastin; Albert Corhay; Georges Hübner; Pierre-Armand Michel

    2002-01-01

    This study investigates the relationship between the evolution of real options values and associated financing policies for Belgian companies in the sector of bio-industries. Each firm's situation regarding the relevant types of real options is stylistically represented through a scenario tree. The consumption of a time-to-build or a growth option is respectively considered as a success or a failure in company development. Empirically, several variables enable us to locate each company along ...

  16. A Belgian traveler who acquired yellow fever in The Gambia

    OpenAIRE

    Colebunders, R; Mariage, J. L.; Coche, J. C.; Pirenne, B; Kempinaire, S.; Hantson, P.; Gompel, A; Niedrig, M; Van Esbroeck, M.; Bailey, R; Drosten, C.; Schmitz, H

    2002-01-01

    A 47-year-old Belgian woman acquired yellow fever during a 1-week vacation in The Gambia; she had never been vaccinated against yellow fever. She died of massive gastrointestinal bleeding 7 days after the onset of the first symptoms. This dramatic case demonstrates that it is important for persons to be vaccinated against yellow fever before they travel to countries where yellow fever is endemic, even if the country, like The Gambia, does not require travelers to be vaccinated.

  17. Auditor Choice in the Belgian Nonprofit Sector: a Behavioral Perspective

    OpenAIRE

    REHEUL, Anne-Mie; Van Caneghem, Tom; Verbruggen, Sandra

    2011-01-01

    This study investigates auditor choice in Belgian nonprofit organizations from a behavioral perspective. We investigate whether auditor choice in favor of an auditor with a high (versus low) level of sector specialization is associated with the importance that nonprofit organizations attach to six auditor attributes: competence/integrity/deontology, working relationship with management, audit fee, practical execution of the audit, client oriented analysis and suggestions, and sector expertise...

  18. Internal finance and corporate investment: Belgian evidence with panel data

    OpenAIRE

    Barran, Fernando; Peeters, Marga

    1998-01-01

    In this paper the corporate investment decision under financial restrictions is investigated with Belgian firm data from 1984 to 1992. An investment Euler equation is derived from a dynamic optimization model with debt ceilings and an elastic credit supply. The model is estimated by GMM for different firm groups. An important aspect is that the sample is split according to a firm’s association with coordination centers. These centers have become the major external funding source of corpora...

  19. Dietary Intake of Artificial Sweeteners by the Belgian Population

    OpenAIRE

    Huvaere, Kevin; Vandevijvere, Stefanie Marie; Hasni, Moez; Vinkx, Christine; Van Loco, Joris

    2011-01-01

    Abstract In this study it was investigated whether the Belgian population older than 15 years was at risk of exceeding ADI levels of acesulfame-K, saccharin, cyclamate, aspartame, and sucralose through assessment of usual dietary intake of artificial sweeteners and specific consumption of table-top sweeteners. The conservative Tier 2 approach, for which an extensive label survey was performed, showed that mean usual intake was significantly lower than the respective ADIs for all sw...

  20. The Attitudes of Belgian Adolescents towards Peers with Disabilities

    OpenAIRE

    Bossaert, Goele; Colpin, Hilde; Pijl, Sip Jan; Petry, Katja

    2011-01-01

    This study aimed to explore Belgian adolescents' attitudes towards peers with disabilities and to explore factors associated with these attitudes. Based on the theory of persuasive communication, this study focused on receiver variables (the "whom"), characteristics of students with disabilities ("concerning who") and channel ("how"). An online survey was created and published on several popular websites for youngsters. Attitudes were assessed by means of the CATCH questionnair...

  1. Assessment of marine debris on the Belgian Continental Shelf

    OpenAIRE

    Van Cauwenberghe, L.; Claessens, M.; Vandegehuchte, M.B.; Mees, J.; Janssen, C. R.

    2013-01-01

    A comprehensive assessment of marine litter in three environmental compartments of Belgian coastal waters was performed. Abundance, weight and composition of marine debris, including microplastics, was assessed by performing beach, sea surface and seafloor monitoring campaigns during two consecutive years. Plastic items were the dominant type of macrodebris recorded: over 95% of debris present in the three sampled marine compartments were plastic. In general, concentrations of macrodebris wer...

  2. A survey of bacteria found in Belgian dairy farm products

    OpenAIRE

    N'Guessan, Elise; Godrie, Thérèse; De Laubier, Juliette; Ringuet, Mélanie; Sindic, Marianne

    2015-01-01

    Description of the subject. Due to the potential hazards caused by pathogenic bacteria, farm dairy production remains a challenge from the point of view of food safety. As part of a public program to support farm diversification and short food supply chains, farm dairy product samples including yogurt, ice cream, raw-milk butter and cheese samples were collected from 318 Walloon farm producers between 2006 and 2014. Objectives. Investigation of the microbiological quality of the Belgian da...

  3. Sustainable groundwater extraction in coastal areas: a Belgian example

    OpenAIRE

    Vandenbohede, A.; Van Houtte, E.; Lebbe, L.

    2009-01-01

    Water extractions in coastal areas have to deal with salt water intrusion and lowering of hydraulic heads in valuable ecosystems. Therefore, sustainable management of fresh water resources in these areas is crucial. This is illustrated here with two water extractions in the western Belgian coastal plain which extract groundwater from a phreatic dune aquifer. One water extraction faced problems with salt water intrusion, while lowering of hydraulic heads was an issue for both. To remedy the sa...

  4. Crisis behind the figures? Belgian trade unions between strength, paralysis and revitalisation

    OpenAIRE

    Faniel, Jean

    2012-01-01

    Unlike most of the trade unions in European countries, Belgian unions managed to preserve a high and stable union density, and strong institutional positions. However, their situation is not blissful and the condition of both the workforce and the unions has been worsening for three decades. This article looks at the strengths and weaknesses of Belgian unions and presents four initiatives of union revitalisation recently developed. The argument is that Belgian unions do not fully size the sco...

  5. Risk analysis of marine activities in the Belgian part of the North Sea (RAMA): final report

    OpenAIRE

    Le Roy, D; Volckaert, A; Vermoote, S.; Wachter, B; Maes, F.; Coene, J.; Calewaert, J.-B.

    2006-01-01

    RAMA is a 2-year project (04/2004 - 04/2006) executed by two Belgian partners, Ecolas NV (Environmental Consultancy Agency) and the Maritime Institute (University of Ghent), and financed by the SPSD II research program, specific actions, of the Belgian Science Policy (BELPSO). RAMA aims to assess the environmental risks of spills by commercial shipping activities on the Belgian Part of the North Sea. Shipping patterns, transports of dangerous goods, probability of risks and the potential impa...

  6. Planning and implementation of the Belgian nuclear programme

    International Nuclear Information System (INIS)

    In the first part of the paper, the authors recall Belgian conditions, initially as regards primary energy (high degree of energy consumption and high degree of dependence on other countries), and then as regards electricity (divided up according to energy sources and types of producer). In the second part, the method used in Belgium for planning electrical power production is explained. Particular emphasis is placed on both the economic and technical assumptions made (trends in fuel costs, method of calculating investment costs, etc.). The development required, for the period 1982-92, of the means of production is stated in the light of the assumptions made. Fuel cycle planning (front and back ends) is also described with a review of the principal stages, namely supply of natural uranium, enrichment, reprocessing, treatment of irradiated fuel, and geological storage of wastes. The third and last part of the paper looks back at events in the implementation of the Belgian nuclear programme in chronological order. The beginnings of nuclear development in Belgium are recalled, as is the decision to construct the first three units (Doel 1, Doel 2 and Tihange 1), which were completed and put into service in 1975. The programme now under way is also briefly described, together with the characteristics of Belgian power stations, especially those concerned with safety. In conclusion, the paper outlines the main advantages of the nuclear option for a country as vulnerable where energy is concerned, as Belgium. (author)

  7. Artificial intelligence in nuclear reactor operation

    International Nuclear Information System (INIS)

    Assessment of four real fuzzy control applications at the MIT research reactor in the US, the FUGEN heavy water reactor in Japan, the BR1 research reactor in Belgium, and a TRIGA Mark III reactor in Mexico will be examined through a SWOT analysis (strengths, weakness, opportunities, and threats). Special attention will be paid to the current cooperation between the Belgian Nuclear Research Centre (SCK·CEN) and the Mexican Nuclear Centre (ININ) on AI-based intelligent control for nuclear reactor operation under the partial support of the National Council for Science and Technology of Mexico (CONACYT). (authors)

  8. Extension of the reactor dynamics code MGT-3D for pebblebed and blocktype high-temperature-reactors

    International Nuclear Information System (INIS)

    The High Temperature Gas cooled Reactor (HTGR) is an improved, gas cooled nuclear reactor. It was chosen as one of the candidates of generation IV nuclear plants [1]. The reactor can be shut down automatically because of the negative reactivity feedback due to the temperature's increasing in designed accidents. It is graphite moderated and Helium cooled. The residual heat can be transferred out of the reactor core by inactive ways as conduction, convection, and thermal radiation during the accident. In such a way, a fuel temperature does not go beyond a limit at which major fission product release begins. In this thesis, the coupled neutronics and fluid mechanics code MGT-3D used for the steady state and time-dependent simulation of HTGRs, is enhanced and validated [2]. The fluid mechanics part is validated by SANA experiments in steady state cases as well as transient cases. The fuel temperature calculation is optimized by solving the heat conduction equation of the coated particles. It is applied in the steady state and transient simulation of PBMR, and the results are compared to the simulation with the old overheating model. New approaches to calculate the temperature profile of the fuel element of block-type HTGRs, and the calculation of the homogeneous conductivity of composite materials are introduced. With these new developments, MGT-3D is able to simulate block-type HTGRs as well. This extended MGT-3D is used to simulate a cuboid ceramic block heating experiment in the NACOK-II facility. The extended MGT-3D is also applied to LOFC and DLOFC simulation of GT-MHR. It is a fluid mechanics calculation with a given heat source. This calculation result of MGT-3D is verified with the calculation results of other codes. The design of the Japanese HTTR is introduced. The deterministic simulation of the LOFC experiment of HTTR is conducted with the Monte-Carlo code Serpent and MGT-3D, which is the LOFC Project organized by OECD/NEA [3]. With Serpent the burnup

  9. Application of the SSYST-3 program system to WWER type nuclear reactors Pt. 1

    International Nuclear Information System (INIS)

    A computer code was developed for the simulation of reactor physical, thermohydraulical and chemical processes taking place in WWER-1000 type nuclear reactors. Two versions of this code, the SSYST-2 and SSYST-3 were compared with special attention to their data handling capabilities. The MULTRAN module of the SSYST-3 used for the calculation of Zircaloy fuel cladding oxidation was tested in detail. Some problems concerning the adaptation of SSYST-3 modules to WWER-type reactors were analyzed. 8 refs.; 4 tabs

  10. «One Difference Is Enough»: Towards a History of Disability in Belgian Congo (1908-1960

    Directory of Open Access Journals (Sweden)

    Evelyne Verhaegen

    2015-11-01

    Full Text Available This article aims to investigate the educational initiatives provided for Congolese people with disabilities during the Belgian colonization, 1908-1960. We found out disability strongly influenced the foundation of the Belgian colony and that it can be assumed that a significant number of Congolese in the Belgian colony were disabled. Yet no historical research about this subject can be found. The subject seemed to be hardly neglected and overlooked. It is this particular contradiction or silence in historiography that this article wants to elucidate. For this purpose, various and sometimes conflicting sources have been consulted. In addition to basic literature on the Belgian colonization and more specific literature on disability in relation to culture, various archives, such as audiovisual material and oral witnesses of this particular period have been included in this research. Our main finding is that in most of the colonial period little or no educational initiatives were provided for Congolese people with disabilities. This we explain by the very limited differentiation which was made between the Congolese themselves. We argue that the black man as such was considered as a rather alien figure and consequently the additional factor of disability remained hardly unnoticed. In the last years of the colonization an increased amount of educational initiatives emerged, which this article explains by the probable increased differentiation between blacks towards the end of the colonization. How to reference this article Verhaegen, E., Verstraete, P., & Depaepe, M. (2016. «One Difference Is Enough»: Hacia una historia de la discapacidad en el Congo Belga (1908-1960. Espacio, Tiempo y Educación, 3(1, 407-420. doi: http://dx.doi.org/10.14516/ete.2016.003.001.19

  11. On the role of radiologists in the Belgian PROject on CAncer of the REctum, PROCARE.

    Science.gov (United States)

    Penninckx, F; Danse, E

    2006-01-01

    Radiologists are involved at all stages of the treatment of patients with rectum cancer: in the preoperative staging, in the diagnosis of postoperative complications, in the detection of recurrent or metastatic disease during follow-up, in the monitoring of the therapeutic effect of palliative therapy. PROCARE is a Belgian national project to improve outcome in all patients with rectum cancer. Guidelines were made by a multidisciplinary workgroup and are available on the web. Decentralised implementation of guidelines is organised by the scientific and professional organisations. It is planned that a central review committee of radiologists, delegated by the Royal Belgian Society of Radiology, will survey the quality of preoperative staging. Overall quality of care will be assured by registration in a specific national database starting in 2006. Participating teams will receive annual feedback. Radiologists should provide data on cTNM staging and cCRM. Differentiation between cT2 and cT3, cN0 and cN+, and measurement of the cCRM in mm are crucial as they have a relevant impact on treatment strategy. While spiral abdominal CT is adequate for cM staging, high-resolution MRI is highly recommended and, in fact, a necessity for locoregional staging because its adequacy is superior to that of CT-scan and EUS. However, EUS is mandatory when local excision is considered, i.e. for cT1N0 lesions. PMID:16607873

  12. RA Reactor operation and maintenance (I-IX), Part IV, Task 3.08/04, Refurbishment of the RA reactor

    International Nuclear Information System (INIS)

    This volume contains reports describing maintenance and repair work of the RA reactor instrumentation, equipment of the reactor dosimetry control system, and equipment for regulation and control systems

  13. UPGRADE OF INSTRUMENTATION FOR PURDUE REACTOR PUR-1, PHASE 3

    International Nuclear Information System (INIS)

    The major objective of this program is to upgrade and replace instruments and equipment that significantly improve the performance, control and operational capability of the Purdue University nuclear reactor (PUR-1). Under this major objective one project on design and installation of interface cards for channel four detector was considered. This report is the final report and gives the efforts and progress achieved on these projects from August 2002 to July 2004

  14. Summary of the 3rd workshop on the reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi [eds.

    2000-06-01

    The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

  15. Design and implementation of the control system for the new console of TRIGA-3-Salazar Reactor

    International Nuclear Information System (INIS)

    TRIGA-3-Salazar Reactor was set in operation in 1968 and the aging of its components has cause the increasing in the maintenance. In the presence of this, it becomes necessary to replace the reactor console using new technologies, considering the incorporation of a personal computer. The aim of this work is the design and construction of the equipment interfaces as well as the digital computer program for the automation and control of the TRIGA-3-Salazar Reactor by means of a personal computer. (Author)

  16. 3D AGENT methodology validation for prismatic high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The Generation IV of nuclear reactors includes as highly competitive the design of a Very High Temperature Reactor (VHTR). This type of reactors can be of a prismatic block, or pebble-bed type. An example of a prismatic block nuclear reactor is the High Temperature Test Reactor (HTTR) operated by Japan Atomic Energy Agency; the reactor reached its full power of 30 MWth for the first time in 1999. The primary coolant is helium at the pressure of ∼4 MPa, with inlet-outlet temperatures of 395°C and 850 – 950°C, respectively. The fuel is 6% enriched uranium, and the moderator is made of graphite. Using the literature available data, a comprehensive validation study is performed to benchmark and assess the AGENT (Arbitrary GEometry Neutron Transport) methodology capabilities in predicting and capturing reactor physics details affected by double heterogeneity of the fuel. Using AGENT with explicit modeling of the fuel double heterogeneity, the HTTR neutronics parameters are compared to NEWT and KENO VI, as well as to experimental data as found in literature. Detailed analysis of spatial steady-state reaction rates and flux spatial maps are provided. The AGENT methodology is based on the method of characteristics and the only one in the world as applied to reactor systems, the R-function based reactor solid modeler, in providing an accurate deterministic solution for 3D steady-state reactor physics. The R-functions modeler presents no limits to reactor geometry and materials types with their distributions. (author)

  17. Analysis of innovative water reactor for flexible fuel cycle in FCA using JENDL-3.3

    International Nuclear Information System (INIS)

    To obtain experimental data and to evaluate the prediction accuracy for the core characteristics in the design study of Innovative Water Reactor for Flexible fuel cycle (FLWR), critical experiments were carried out using a series of mock-up cores at FCA. Three mockup cores with different void fractions of the moderator were constructed to obtain experimental data in wide range of neutron spectra. Major items of the experiment are criticality, reaction rate ratios, moderator void reactivity worth and the Doppler effect. Conventional deterministic calculation systems were used to analyze the experiment with the use of the JENDL-3.2 and JENDL-3.3 libraries. The ratios of calculated and experimental (C/E) values were compared between both the JENDL libraries. The current analysis method showed good prediction accuracy in most of the experiments and no significant differences were observed in the C/E values between the libraries in this study. (author)

  18. 12th Annual conference of the Nuclear Society of Russia. Research reactors: science and high technologies. Proceedings. Vol. 2. Part 3. Research reactors - present and future (Physics and engineering of research reactors)

    International Nuclear Information System (INIS)

    The part 3 of the volume 2 of the Proceedings of the 12th Annual conference of the Nuclear Society of Russia - Research reactors: science and high technologies (June 25-29, 2001, Dimitrovgrad, SRC RIAR) is presented. The scope of reports published in this part includes: projects of the pool-type reactor for heat supply and the neutron therapy centrum on the basis of the MEPhI reactor; water-chemical mode by storage of the research reactors (RR) spent fuel elements; possibility of the RR service life prolongation; material science problems on the operational control of the RR reactor core elements and heat exchangers

  19. Embrittlement and annealing of reactor pressure vessel steels: comparison of BR3 surveillance and vessel plates to the surrogate plates representative of the Yankee Rowe vessel

    International Nuclear Information System (INIS)

    The sister pressure vessels at the BR3 and Rowe Yankee PWR plants were operated at a lower-than-usual temperature (260 degrees Celsius) and their plates were austenitized at higher-than-usual temperature (970 degrees Celsius). A heat tratemement leading to a coarser microstructure than typical for the fine grain plates that are considered in development of USNRC Regulatory guide 1.99. This material displayed outlier behaviour charackterized by a 41J CVN-shift significantly larger than predicted by Regulatory Guide 1.99. Because lower radiation temperature and nickell alloying are generally considered detrimental to irradiation sensitivity, there was a major concern that the nickel-modified lower Rowe plate and the nickel-modified BR3 plate may become too embrittled to satisfy the toughness requirements enbodied in the PTS screening criterion. This paper compares three complementary studies undertaken to clarify these uncertainties: 1) the accelerated irradiation and test program launched in 1990 by Yankee Atomic Electric Company using typical vessel plate materials containing 0.24% copper at two nickel levels: YA1, 0.63% (A533-B) and YA9, 0.19% (A302-B). These were heat-treated to produce the coarse and fine grain microstructures representative of the Yankee/BR3 and the Regulatory Guide plates, respectively, 2) the BR3 surveillance and vessel testing program: this vessel was wet-annealed in 1984, relicensed for operation till the plant shutdown in 1987, ANCL was trepanned in early 1995, 3) the accelerated irradiations in the Belgian BR2 test reactor of the Yankee coarse grain plates YA1 and YA9 together with BR3 vessel specimens extracted at nozzle elevation, a location with negligible radiation exposure. It is shown that the PTS screening criterion was never attained by the BR3 and Rowe plates, and that the BR3 vessel anneal was neither necessary nor sufficient. Finally, the sensitivity of embrittlement, annealing and post-annealing reembrittlement to irradiation

  20. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 3. Appendix A. Equipment list

    International Nuclear Information System (INIS)

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system and was prepared by the General Electric Company. Core scoping studies were performed which evaluated the effects of annular and cylindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations. Volume 3 is an Appendix containing the equipment list for the plant and was also prepared by United Engineers and Constructors, Inc. It tabulates the major components of the plant and describes each in terms of quantity, type, orientation, etc., to provide a basis for cost estimation

  1. Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

    International Nuclear Information System (INIS)

    G1 (46 MWth), G2 (250 MWth) and G3 (250 MWth) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide (14C, 36Cl, 63Ni, 60Co,3H) from graphite, (iv) chlorine diffusion through graphite. (authors)

  2. The TRIAD3 code for NRU reactor simulation

    International Nuclear Information System (INIS)

    The NRU research reactor is difficult to simulate due to its extreme heterogeneity. The simulation code system used for the last 25 years has given deteriorating results as more complex rod assemblies have been installed in the core. To the remedy this situation, various options for improving the code were studied. These included modelling the axial dimension, an improved representation of the lattice split, and using a better cell code (WIMS) to derive the homogenized cell parameters; but perhaps the most significant was the introduction of discontinuity factors, which account for the fact that the homogenized cell fluxes are not continuous across a cell interface. This improved diffusion-theory methodology gave much better comparison with careful measurements, taken in a zero-power reactor for certain NRU assemblies, than did the usual diffusion theory. A new, modular code system was written to incorporate these improvements, and it is now in a commissioning phase. The code is operated through an interactive monitor. Results so far are encouraging, although differences from the measured power distribution in NRU are still significant

  3. Application of stable adaptive schemes to nuclear reactor systems, (3)

    International Nuclear Information System (INIS)

    Stable parameter identification and adaptive control schemes are considered for a reactor model embodying two temperature feedbacks-slow and fast. This reactor model is liable to see its feedback coefficients change sign in the course of long periods of operation, resulting in nonlinear oscillations of neutron flux, which cannot be described by a linearized model. This nonlinear system is expressed in terms of memoryless nonlinear elements in the feedback loop of a linear system, with the aid of linear and nonlinear transformations, and the nonlinear elements are here treated without being linearized. A new system representation is introduced, using which, stable parameter identification and adaptive control schemes are developed in the pattern of the Model Reference Adaptive System (MRAS) with use made of the Lyapunov method. Both schemes are shown to be stable, and furthermore globally stable if the input has frequencies sufficiently varied to permit all the excited modes to be considered linearly independent. It is thus shown that the estimated parameters converge to the true values for the parameter identification, and that, for the adaptive control, the output error between the plant and the model tends toward zero. (author)

  4. The Influence of the 1974 Oil Crisis on Sectoral Growth Rates in the Belgian Economy

    OpenAIRE

    F.BOSSIER; D. DUWEIN

    1981-01-01

    This paper briefly presents and analyses the behaviour of the different sectors of the Belgian economy during the period 1965-1978. Special attention is paid to the influence of the 1974 oil crisis on sectors of the Belgian economy. It is shown that the 1974 shock had different consequences according to the energy components of the sector

  5. Differences between Belgian and Brazilian group A Streptococcus epidemiologic landscape.

    Directory of Open Access Journals (Sweden)

    Pierre Robert Smeesters

    Full Text Available BACKGROUND: Group A Streptococcus (GAS clinical and molecular epidemiology varies with location and time. These differences are not or are poorly understood. METHODS AND FINDINGS: We prospectively studied the epidemiology of GAS infections among children in outpatient hospital clinics in Brussels (Belgium and Brasília (Brazil. Clinical questionnaires were filled out and microbiological sampling was performed. GAS isolates were emm-typed according to the Center for Disease Control protocol. emm pattern was predicted for each isolate. 334 GAS isolates were recovered from 706 children. Skin infections were frequent in Brasília (48% of the GAS infections, whereas pharyngitis were predominant (88% in Brussels. The mean age of children with GAS pharyngitis in Brussels was lower than in Brasília (65/92 months, p<0.001. emm-typing revealed striking differences between Brazilian and Belgian GAS isolates. While 20 distinct emm-types were identified among 200 Belgian isolates, 48 were found among 128 Brazilian isolates. Belgian isolates belong mainly to emm pattern A-C (55% and E (42.5% while emm pattern E (51.5% and D (36% were predominant in Brasília. In Brasília, emm pattern D isolates were recovered from 18.5% of the pharyngitis, although this emm pattern is supposed to have a skin tropism. By contrast, A-C pattern isolates were infrequently recovered in a region where rheumatic fever is still highly prevalent. CONCLUSIONS: Epidemiologic features of GAS from a pediatric population were very different in an industrialised country and a low incomes region, not only in term of clinical presentation, but also in terms of genetic diversity and distribution of emm patterns. These differences should be taken into account for designing treatment guidelines and vaccine strategies.

  6. Operating experience with diesel generators in Belgian nuclear power plants

    International Nuclear Information System (INIS)

    Various problems have occurred on the diesel generators in the Belgian nuclear power plants, independently of the D.G. manufacturer or from the operating crew. Furthermore no individual part of the D.G. can be incriminated as being the main cause of the incidents. The incidents reported in this paper are chosen because of the importance for the safety or for the long repair period. The unavailability of a D.G. can only be detected by periodic tests and controls. Combined with a good preventive maintenance, the risks of incidents can be reduced. (author)

  7. Internal capital market efficiency of Belgian holding companies

    OpenAIRE

    Gautier, Axel; Malika HAMADI

    2004-01-01

    In this paper, we raise the following two questions. (1) Do Belgian holding companies operate an internal capital market to transfer financial resources amongst their subsidiaries? And if yes, (2) is the internal capital market efficient? To answer the first question, we check if group cash flow is a determinant of the group members investment spending. The answer is positive if the holding company’s subsidiary is affiliated to a coordination center and negative otherwise. To ans...

  8. Migration and americanisation: The special case of Belgian economics.

    OpenAIRE

    Maes, Ivo; Buyst, Erik

    2005-01-01

    One of the distinguishing features of Belgian economics is that, from the early 1920s, so many of Belgium's best economists pursued postgraduate studies at top American universities, a case of ‘temporary’ migration. This was made possible by the fellowships granted by the Commission for Relief in Belgium, a legacy of the First World War. After a stay in the US of a few years, most returned to Belgium. However, they maintained strong links with the US. Also, they tried to recreate in Belgium t...

  9. Health effects[1997 Scientific Report of the Belgian Nuclear Research Centre

    Energy Technology Data Exchange (ETDEWEB)

    Mahieu, L.

    1998-07-01

    The objectives of the research in the field of epidemiology , performed at the Belgian Nuclear Research Centre SCK-CEN are (1) to study cancer mortality and morbidity in nuclear workers in Belgium; (2) to document the feasibility of retrospective cohort studies in Belgium; (3) to participate in the IARC study. For radiobiology, the main objectives are: (1) to elucidate the mechanisms of the effects of ionizing radiation on the mammalian embryo during the early phase of its development, (2) to assess the genetic risks of maternal exposure to ionizing radiation, (3) to elucidate the mechanisms by which damage to the brain and mental retardation are caused in man after prenatal irradiation. The main achievements in these domains for 1997 are presented.

  10. Proceedings of the 3th national conference on nuclear reactor instrument

    International Nuclear Information System (INIS)

    The 3th National Conference on Nuclear Research Increment was held in Haikou, Province Hainan on April 2003. Proceedings published both by China Electronic Society and China Nuclear Society. 19 articles were collected. The contents including; Steady external neutron source system of driven sub-critical reactor for ADS; A rapid micro-current amplifier used for measure the dynamic parameter of the sub-critical reactor and simulation analysis of power regulating system of CARR and etc.

  11. Safety in the ARIES-III D-3He tokamak reactor design

    International Nuclear Information System (INIS)

    The ARIES-3 reactor study is an extensive examination of the viability of a D-3He-fueled commercial tokamak power reactor. Because neutrons are produced only through side reactions, the reactor has the significant advantages of reduced activation of the first wall and shield, low afterheat and Class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. A ferritic steel shield behind the first wall protects the magnets from gamma and neutron heating and from radiation damage. The ARIES-3 reactor uses an organic coolant to cool the first wall, shield and divertor. The organic coolant has a low vapor pressure at the operating temperature required for good thermal efficiency. Radiation damage requires processing the coolant to remove and crack radiolytic products that would otherwise foul cooling surfaces. The cracking process produces waste, which must be disposed of through incineration or burial. We estimated the offsite doses due to incineration at five candidate locations. The plasma confinement requirements for a D-3He reactor are much more challenging than those for a D-T reactor. Thus, the demands on the divertor are more severe, particularly during a disruption. We explored the potential for isotopically tailoring the 4 mm tungsten layer on the divertor in order to reduce the offsite doses should a tungsten aerosol be released from the reactor after an accident. We also modeled a loss-of-cooling accident in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. We analyzed the disposition of the 20 g/day of tritium that is produced by D-D reactions and removed by the vacuum pumps. For our reference design, the tritium will be burned in the plasma. These results re-emphasize the need for low activation materials and advanced divertor designs, even in reactors using advanced fuels

  12. Prestressed concrete nuclear reactor containment structures. Revision 3

    International Nuclear Information System (INIS)

    A discussion of the techniques and procedures used for the design of prestressed concrete nuclear reactor containment structures is presented. A physical description of Bechtel designed containment structures is presented. The design bases and load combinations are given for anticipated conditions of service. Reference design documents which include industry codes, specifications, AEC Regulatory Guides, Bechtel Topical Reports and additional criteria as appropriate to containment design are listed. Stepwise procedures typically followed by Bechtel for design of containments is discussed and design examples are presented. A description of currently used analytical methods and the practical application of these methods for containment design is also presented. The principal containment construction materials are identified and codes of practice pertaining to construction procedures are listed. Preoperational structural testing procedures and post-operational surveillance programs are furnished along with results of tests on completed containment structures. (U.S.)

  13. A federal audit of the Belgian radiotherapy departments in breast cancer treatment

    International Nuclear Information System (INIS)

    Background: The Belgian Federal College of Radiotherapy carried out an external audit of breast cancer patient documentation in the 26 Belgian radiotherapy centres. The objective was to assess compliance with the recommendations regarding minimal requirements for documentation of radiotherapy prescription and administration. All centres volunteered to take part in this audit. Methods: Two experienced radiation oncologists site-visited the departments over a 6 month period (Sept. 2003-Feb. 2004), with a list of items to be verified, including details on the surgery, the pathological report, details on systemic treatments, details on the radiotherapy prescription (and consistency with therapeutic guidelines) and delay surgery/radiotherapy. Findings: Three hundred and eighty-nine patients files were reviewed, for a total of 399 breast cancers (10 patients with bilateral cancer). Mean age was 57.8 y (range 29-96). Breast conservative surgery (BCS) was used in 71%; radical mastectomy in 29%. A complete pathological report was present in all files but 2 (99.5% conformity). 5.2% were treated for DCIS, 61.6% for pT1, 28.2% for pT2 and 5% for pT3-4. Data regarding resection margins were specified to be free in 76.2%, tangential in 12% (within 2 mm) and positive for DCIS in 3.8% or invasive cancer in 1.5% (no information, on margins in 6.5%). The pT stage was always specified, and consistent with the macroscopic and microscopic findings. Hormonal receptors were routinely assessed (94.7%), as well as Her2neu (87.4%). Axillary surgery was carried out in 92%, either by sentinel node biopsy or by complete clearance, in which case the median number of nodes analysed was 12 for all centres together (7-17). All radiotherapy prescriptions were in line with evidence-based standards of therapy (i.e., irradiation of breast after BCS or after mamectomy (in case of pN+), but one. The mean delay between surgery and radiotherapy was 5.5 weeks (SD 11days). Conclusion: There was a high

  14. NEHEX-3D, 3-D Neutron Diffusion for Fast Reactors and WWER in Hexagonal Geometry

    International Nuclear Information System (INIS)

    1 - Description of program or function: Neutronics calculation of fast and WWER type reactors with hexagonal assemblies (determination of keff value, group neutron fluxes and thermal power). 2 - Method of solution: The method is based on the following ideas: - nodal approach; - transverse integration technique; - expansion of one-dimensional neutron flux inside the node in polynomials up to the third order; - formulation of nodal equations in the form of response matrix equation; - solution of resulting nodal algebraic equations by means of iterative method. 3 - Restrictions on the complexity of the problem: Geometry: 30 deg. reflectional and 60 deg. rotational symmetry; Maximum number of subassemblies: NN = 100; Maximum number of energy groups: NG = 4; Maximum number of axial layers: NZ = 20; Maximum number of different materials : NM = 50

  15. VISIT OF BELGIAN FIRMS AT CERN

    CERN Multimedia

    2001-01-01

    2 - 3 APRIL 2001 14h00 to 17h30 Monday 2nd 09h00 to 17h30 Tuesday 3rd The firms will be in the Main Building - Pas Perdus and Mezzanine List of Companies: Alcatel ETCA Amos S.A. Asco Industries N.V. Barco N.V. Belgatom Capaul S.A. Comelec S.A. Groupe Hamon HTMS (High Tech Metal Seals) N.V. Kelatron S.A. Mécanique de Précision pour Equipements Mecasoft S.A. Pauwels International N.V. Penders & Vanherle Elektrotechniek N.V. Resarm Engineering Plastics S.A. Simonis Plastic S.A. Techspace Aero S.A. Verhaert Design & Development N.V. Detailed information on the companies is also available on the Web: http://spl-purchasing.web.cern.ch/spl-purchasing/exhibitions_visits.htm For further information please contact Mrs C.L. Jullien-Woringer (tel. 73722-76360)

  16. A study of in-package nuclear criticality in possible Belgian spent nuclear fuel repository designs

    OpenAIRE

    Wantz, Olivier

    2005-01-01

    About 60 percent of the electricity production in Belgium originates from nuclear power plants. Belgium owns 7 nuclear pressurized water reactors, which are located in two sites: 4 reactors in Doel and 3 reactors in Tihange. Together they have a capacity of approximately 5900 MWe. All these reactors use classical uranium oxide fuel assemblies. Two of them (Doel3, Tihange2) have also accepted a limited number of mixed (uranium and plutonium) oxide fuel assemblies. These mixed fuel assemblies c...

  17. Salmonella surveillance and control at post-harvest in the Belgian pork meat chain.

    Science.gov (United States)

    Delhalle, L; Saegerman, C; Farnir, F; Korsak, N; Maes, D; Messens, W; De Sadeleer, L; De Zutter, L; Daube, G

    2009-05-01

    Salmonella remains the primary cause of reported bacterial food borne disease outbreaks in Belgium. Pork and pork products are recognized as one of the major sources of human salmonellosis. In contrast with the primary production and slaughterhouse phases of the pork meat production chain, only a few studies have focussed on the post-harvest stages. The goal of this study was to evaluate Salmonella and Escherichia coli contamination at the Belgian post-harvest stages. E. coli counts were estimated in order to evaluate the levels of faecal contamination. The results of bacteriological analysis from seven cutting plants, four meat-mincing plants and the four largest Belgian retailers were collected from official and self-monitoring controls. The prevalence of Salmonella in the cutting plants and meat-mincing plants ranged from 0% to 50%. The most frequently isolated serotype was Salmonella typhimurium. The prevalence in minced meat at retail level ranged from 0.3% to 4.3%. The levels of Salmonella contamination estimated from semi-quantitative analysis of data relating to carcasses, cuts of meat and minced meat were equal to -3.40+/-2.04 log CFU/cm(2), -2.64+/-1.76 log CFU/g and -2.35+/-1.09 log CFU/g, respectively. The E. coli results in meat cuts and minced meat ranged from 0.21+/-0.50 to 1.23+/-0.89 log CFU/g and from 1.33+/-0.58 to 2.78+/-0.43 log CFU/g, respectively. The results showed that faecal contamination still needs to be reduced, especially in specific individual plants. PMID:19269567

  18. Initial operation and utilization of the Bangladesh 3 Mw TRIGA reactor

    International Nuclear Information System (INIS)

    A 3 Mw TRIGA MK-II pulsing type research reactor fuelled with low enrichment uranium having 19.7% U-235 and 20 wt % Uranium, 0.47% Erbium and Zirconium Hydride, has been installed at the Atomic Energy Research Establishment, savar in the last week of October, 1986. This multi-purpose reactor, capable of both steady-state and pulsing operation, has been put into service in several disciplines since its commissioning and presently in operation without any major problem. The paper describes the initial operating experience and the reactor utilization made in several areas including the operator training conducted for the formation of the initial crew for the reactor. (author)

  19. RISCOM Applied to the Belgian Partnership Model: More and Deeper Levels

    Energy Technology Data Exchange (ETDEWEB)

    Bombaerts, Gunter; Bovy, Michel; Laes, Erik [SCKCEN, Mol (Belgium). PISA

    2006-09-15

    Technology participation is not a new concept. It has been applied in different settings in different countries. In this article, we report a comparing analysis of the RISCOM model in Sweden and the Belgian partnership model for low and intermediate short-lived nuclear waste. After a brief description of the partnerships and the RISCOM model, we apply the latter to the first and come to recommendations for the partnership model. The strength of the partnership approach is at the community level. In one of the villages, up to one percent of the population was motivated to discuss at least once a month for four years the nuts and bolts of the repository concept. The stress on the community level and the lack of a guardian includes a weakness as well. First of all, if communities come into competition, the inter-community discussions can start resembling local politics and can become less transparent. Local actors are concerned actors but actors at the national level are concerned as well. The local decisions influence how the waste will be transported. The local decisions also determine an extra cost of electricity. We therefore recommend a broad (in terms of territory) public debate on the participation experiments preceding and concluding the local participation process in which this local process maintains an important position. The conclusions of our comparative analysis are: (1) The guardian of the process at the national level is missing. Since the Belgian nuclear regulator plays a controlling role after the process, we recommend a technology assessment institute at the federal level. (2) We state that stretching in the partnership model can happen more profoundly and recommend a 'counter institute' at the European level. The role of non-participative actors should be valued. (3) Recursion levels can be taken as a point of departure for discussion about the problem framing. If people accept them, there is no problem. If people clearly mention issues

  20. RISCOM Applied to the Belgian Partnership Model: More and Deeper Levels

    International Nuclear Information System (INIS)

    Technology participation is not a new concept. It has been applied in different settings in different countries. In this article, we report a comparing analysis of the RISCOM model in Sweden and the Belgian partnership model for low and intermediate short-lived nuclear waste. After a brief description of the partnerships and the RISCOM model, we apply the latter to the first and come to recommendations for the partnership model. The strength of the partnership approach is at the community level. In one of the villages, up to one percent of the population was motivated to discuss at least once a month for four years the nuts and bolts of the repository concept. The stress on the community level and the lack of a guardian includes a weakness as well. First of all, if communities come into competition, the inter-community discussions can start resembling local politics and can become less transparent. Local actors are concerned actors but actors at the national level are concerned as well. The local decisions influence how the waste will be transported. The local decisions also determine an extra cost of electricity. We therefore recommend a broad (in terms of territory) public debate on the participation experiments preceding and concluding the local participation process in which this local process maintains an important position. The conclusions of our comparative analysis are: (1) The guardian of the process at the national level is missing. Since the Belgian nuclear regulator plays a controlling role after the process, we recommend a technology assessment institute at the federal level. (2) We state that stretching in the partnership model can happen more profoundly and recommend a 'counter institute' at the European level. The role of non-participative actors should be valued. (3) Recursion levels can be taken as a point of departure for discussion about the problem framing. If people accept them, there is no problem. If people clearly mention issues that are

  1. Plan for the safe decommissioning of the BAEC 3MW TRIGA MARK-II research reactor

    International Nuclear Information System (INIS)

    The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production (131I, 99mTc, 46Sc), various R and D activities, and manpower training. The reactor has been operated successfully since it's commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power remained suspended for about 4 years. However, the reactor operation was continued during this period at a power level of 250 kW to cater the needs of various R and D groups, which required lower neutron flux for their experiments. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The reactor was made operational again at full power after successful replacement of the damaged decay tank in August 2001. At present the reactor is operated 5 days a week at a full power level of 3 MW for production of I-131 and R and D purposes. Up to December 2005 total burn-up of the core stands at about 358 Megawatt Days (MWDs). BAEC has planned to increase the production of 131I and as such, the core burn-up is expected to be increased very significantly in the years to come. There is a declaration from the US DOE that all US origin research reactor spent fuel generated within 2006 will be taken away to the USA at their own cost within 2009. But the fuel burn up of the BAEC research reactor is about 6%. So the reactor can operate for about 10-20 years more. An initial decommissioning plan for the BAEC TRIGA reactor and relevant facilities should be established as early as possible as recommended in the IAEA Safety Standards Series No.WS-G-2.1 (Decommissioning of Nuclear Power Plants and Research Reactors - Safety Standards Series No.WS-G-2.1, IAEA, Vienna, 1999). During the design and construction

  2. Pyrrolizidine alkaloids in food and feed on the Belgian market.

    Science.gov (United States)

    Huybrechts, Bart; Callebaut, Alfons

    2015-01-01

    Pyrrolizidine alkaloids (PAs) are widely distributed plant toxins with species dependent hepatotoxic, carcinogenic, genotoxic and pneumotoxic risks. In a recent European Food Safety Authority (EFSA) opinion, only two data sets from one European country were received for honey, while one feed data set was included. No data are available for food or feed samples from the Belgian market. We developed an LC-MS/MS method, which allowed the detection and quantification of 16 PAs in a broad range of matrices in the sub ng g(-1) range. The method was validated in milk, honey and hay and applied to honey, tea (Camellia sinensis), scented tea, herbal tea, milk and feed samples bought on the Belgian market. The results confirmed that tea, scented tea, herbal tea and honey are important food sources of pyrrolizidine alkaloid contamination in Belgium. Furthermore, we detected PAs in 4 of 63 commercial milk samples. A high incidence rate of PAs in lucerne (alfalfa)-based horse feed and in rabbit feed was detected, while bird feed samples were less contaminated. We report for the first time the presence of monocrotaline, intermedine, lycopsamine, heliotrine and echimidine in cat food. PMID:26373269

  3. The attitudes of Belgian adolescents towards peers with disabilities.

    Science.gov (United States)

    Bossaert, Goele; Colpin, Hilde; Pijl, Sip Jan; Petry, Katja

    2011-01-01

    This study aimed to explore Belgian adolescents' attitudes towards peers with disabilities and to explore factors associated with these attitudes. Based on the theory of persuasive communication, this study focused on receiver variables (the "whom"), characteristics of students with disabilities ("concerning who") and channel ("how"). An online survey was created and published on several popular websites for youngsters. Attitudes were assessed by means of the CATCH questionnaire among 167 adolescents between 11 and 20 years old. Univariate and multivariate regression analyses were conducted. Belgian adolescents had fairly tolerant attitudes towards peers with disabilities. Factors associated with more positive attitudes were being female, and viewing a video introduction of a peer with a disability before assessing attitudes. Factors such as having a parent, sibling or good friend with a disability and frequent contact with persons with disabilities did not remain significant in the overall model. The way in which students with disabilities are presented to their peers is very important. Further research is needed among larger samples, including more diverse variables, concerning the former mentioned categories, and also concerning the source (the "who") and message (the "what"). PMID:21257288

  4. Visit of Belgian firms at CERN

    CERN Multimedia

    Caroline Laignel - FI

    2006-01-01

    3 - 4 APRIL 2006 9.00 a.m. to 5.00 p.m. Monday 3 April 9.00 a.m. to 5.00 p.m. Tuesday 4 April Individual interviews will take place in technicians' offices. The firms will contact relevant users/technicians but any user wishing to make contact with a particular firm is welcome to use the contact details which are available from each departmental secretariat or from the Purchasing web pages at the following URL http://fi-dep.web.cern.ch/fi-dep/structure/memberstates/exhibitions_visits.htm List of Companies: Alcatel Alenia Space Etca ALM ATS Vanwers Cegelec SA Cherokee Europe Cincom Systems International SA DB Engineering BVBA FOS & S BVBA GDM Electronics NV Gemaco SA LMS International NV Management Centre Europe MD Technology Mecasoft SA N.E.T. Research Polmans SA Resarm Engineering Plastics SA Sogeti Belux Syreg SA Tein Telecom SA Groupe Vincott For further information please contact Mrs C. Laignel FI-DI 73722.

  5. Validation of RELAP/SCDAPSIM/MOD3.4 for research reactor applications

    International Nuclear Information System (INIS)

    The RELAP/SCDAPSIM/MOD3.4 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of the international SCDAP Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD3.4 uses publicly available RELAP/MOD3.3 and SCDAP/RELAP5/MOD3.2 models, developed by the US Nuclear Regulatory Commission, in combination with (a) new models for fission product transport and deposition, fuel assembly behavior, and in-vessel melt retention, (b) advanced programming and numerical techniques, and (c) integrated graphics displays. The RELAP5 and SCDAP models have been validated for a wide range of accident conditions using a variety of experiments and plant data. However, the validation of the models for research reactor applications has been much more limited. As a result, a group within SDTP, including the Nuclear Research Institute, Rez (Czech Republic), National Nuclear Energy Agency of Indonesia, China Institute of Atomic Energy, and Bangladesh Atomic Energy Research Establishment, has started work to validate the code for a variety of research reactor designs including TRIGA and other unique research reactors. This paper describes the reactor designs currently being considered in the study, the development and qualification of input models for the facilities, and the transients being analyzed as part of this effort. (authors)

  6. Reconstruction of JRR-3, revitalized domestically manufactured No. 1 research reactor

    International Nuclear Information System (INIS)

    JRR-3 is called domestically manufactured No. 1 reactor, because the whole reactor except fuel and heavy water as moderator and coolant was made in Japan. The JRR-3 attained the initial criticality on September 12, 1962, and was operated for 21 years till it was shut down in March, 1983, for reconstruction. The JRR-3 was not able to sufficiently meet the recent needs, accordingly, the improvement of its performance and the expansion of its utilization have become the important problems. The research reactor rearrangement plan was decided in May, 1980, and the reconstruction of JRR-3 has been advanced following this plan. So far, the conceptural design, the detailed design, the mock-up test and so on have been carried out, and the safety examination by the government was performed from April to November, 1984. The permission of installation was obtained in December, 1984, and the reconstruction work is started at the beginning of fiscal year 1985. The present status of research reactors, the change in the utilization of research reactors and future prospect, the reconstruction of JRR-3 and the construction plan hereafter are reported. (Kako, I.)

  7. Radiological optimization[1997 Scientific Report of the Belgian Nuclear Research Centre

    Energy Technology Data Exchange (ETDEWEB)

    Zeevaert, T.

    1998-07-01

    Radiological optimization is one of the basic principles in each radiation-protection system and it is a basic requirement in the safety standards for radiation protection in the European Communities. The objectives of the research, performed in this field at the Belgian Nuclear Research Centre SCK-CEN, are: (1) to implement the ALARA principles in activities with radiological consequences; (2) to develop methodologies for optimization techniques in decision-aiding; (3) to optimize radiological assessment models by validation and intercomparison; (4) to improve methods to assess in real time the radiological hazards in the environment in case of an accident; (5) to develop methods and programmes to assist decision-makers during a nuclear emergency; (6) to support the policy of radioactive waste management authorities in the field of radiation protection; (7) to investigate existing software programmes in the domain of multi criteria analysis. The main achievements for 1997 are given.

  8. Soil-structure interaction of JPDR reactor, 3

    International Nuclear Information System (INIS)

    In order to develop the aseismatic design method for the nuclear power stations constructed on Quaternary ground, examination was carried out for the Japan Power Demonstration Reactor in Tokai Village by the cooperation of the Central Research Institute of Electric Power Industry and the Japan Atomic Energy Research Institute. In this paper, the dynamic characteristics of the JPDR structure and the surrounding ground obtained by the observation of earthquake motion and the results of examination on the influence of soil resistance acting on the embedded side walls are reported in comparison of the observed data with the numerical analysis performed before. Six good records were obtained from June to September, 1982, including that of the earthquake with magnitude of 7.0. Its incident angle, the amplification factor of the maximum amplitude between the base and the surface of ground and that on the structure are reported. The transfer functions among the earth observational points for horizontal motion were evaluated with the Fourier's spectral ratio. The characteristics of the transfer functions were well simulated by simple multiple reflection theory. The soil resistance acting on the embedded side walls was larger than that in the past, and played important role. (Kako, I.)

  9. International tokamak reactor. Phase Two A, Part 3. Vol.2

    International Nuclear Information System (INIS)

    Volume 2 of the report of Phase 2A, Part III of the International Tokamak Reactor Workshop focuses on a critical analysis of five INTOR-like designs: INTOR, NET, FER, TIBER, and OTR. Topics covered are a report of IAEA Specialists' Meeting on INTOR-like designs, a comparison of these designs, and a systems analysis of the designs. In this volume, programmatic and technical objectives are discussed and compared. The resulting physics design features are based on Q (power out/power in), burn duration, impurity control, plasma configuration, startup, operational scenario, duty factor, plasma heating, and frequency of disruptions. Design constraints (confinement scaling, plasma beta, density limit, geometry, plasma parameters, operating mode, disruption, magnetic field, heating, and startup) are compared for the various machines. Systems analyses are used to determine sensitivities to the various physical design features, and those parameters with the most sensitivity are determined. Relative costs of the various designs are compared, ranging from TIBER at 35 per cent less than INTOR, to OTR at 25 per cent more than INTOR. Refs, figs and tabs

  10. Visit of Belgian Firms at CERN

    CERN Multimedia

    FP Department

    2009-01-01

    25 – 26 MAY 2009 09h00 to 17h00 Monday 25 May 09h00 to 17h00 Tuesday 26 May Individual interviews will take place in technicians’ offices. The firms will contact relevant users/technicians but any user wishing to make contact with a particular firm is welcome to use the contact details which are available from each secretariat of department or from the GS Department web pages at the following URL: http://gs-dep.web.cern.ch/gs-dep/groups/sem/ls/Industrial_Exhibitions.htm List of Companies: 1. Automation Services and Consulting BVBA 2. Burrick NV, (PLC) 3. Cissoid 4. DB Engineering 5. Design, Drafting & Services BVBA 6. Entelec Control Systems 7. GILLAM-Fei S.A. 8. HPC 9. ICSENSE 10. IWT – Enterprise Europe Flanders 11. Jema SA 12. Mecasoft SA 13. SA Polmans 14. Rapid-Torc 15. Resarm Engineering Plastics SA 16. Sentera Europa NV 17. SLC BVBA 18. Stocker Industrie SA 19. Technord 20. Tecnubel 21. Winlock BVBA For further information please contact Caroline Laignel GS-DI 737...

  11. Belgian experience with radiation technologies for sterilization

    International Nuclear Information System (INIS)

    Belgium pioneered in non energetic applications of nuclear science. In 1970, the National Institute for Radioisotopes (IRE) was founded on the ground of the experience acquired at the CEN/SCK, for developing nuclear techniques oriented to the well-being of population. In 1978 IRE started operation of 2 γ-irradiation units with a 2.25 million Ci Co60, source having a capacity for sterilization of more than 100 m3 of product per day. this installation is currently operated by Griffith-Mediris (Group Griffith Micro Science) and has accumulated 20 years of experience for irradiation of foodstuffs, medical appliances and pharmaceuticals. In 1986, IRE was producing radioisotopes from accelerators and joined UCL for founding Ion Beam Applications (IBA) company which shortly became the world leader for design and delivery of cyclotrons. More recently, on the basis of French CEA patent, IBA developed the Rhodotron, an e-beam accelerator, with an X-ray option, which is now facing commercial success i area of ionization of food and medical appliances. IRE developed also expertise for the back-end of these activities i.e. final disposal of radioactive sealed source and dismantling of the installations. (author)

  12. Recent historical changes on the Belgian Meuse

    International Nuclear Information System (INIS)

    When a nuclear power station was installed on the Meuse in central Belgium, the impact of thermal, radioactive, and chemical waste on the water of the Neuse and on its biocenoses was studied. Three successive periods of development of the channel bed and the flood plain in Belgium have occurred, and their hydrological, physicochemical, and ecological consequences have been examined. Since the last century, the ecosystem of the Meuse has undergone, due to the increasing activity of man, modifications of increasing importance: marked reduction of the water flow, a drastic increase in the suspended material being transported, a degree of eutrophication of the water, and the disturbance of the original floral and faunal communities. The causes of this evolution of the Meuse can be itemized as different types of human interference in descending order of importance: (1) occupation of the catchment area; (2) encroachment on the flood plain; (3) encroachment on the channel bed; (4) destruction of habitats; (5) water pollution; (6) overexploitation of fish-breeding stocks; and (7) introduction of foreign species. Thought should be given to restoring damaged sectors by recreating shallow riverside zones suitable for aquatic macrophytes, for the macroinvertebrates which are linked to them, and for the reproduction of many species of fish. The example of human interference on the Meuse. 47 refs., 9 figs., 5 tabs

  13. Methane combustion by moving bed fuel reactor with Fe2O3/Al2O3 oxygen carriers

    International Nuclear Information System (INIS)

    Highlights: • Moving bed reactor employed to methane combustion using iron-based oxygen carrier. • Fe2O3/Al2O3 oxygen carriers was prepared and provided with applicable performance. • Carbon formation was enhanced with increased retention time at 900 °C. • Full CH4 conversion was reached without carbon formation by moving bed operation. • FeO and FeAl2O4 were formed in the reacted oxygen carriers out of the reactor. - Abstract: Fe2O3/Al2O3 composite oxygen carriers were prepared for chemical looping combustion (CLC) with methane in a lab-scale moving bed fuel reactor provided with reasonable crush strength, reactivity and recyclability. Carbon formation was observed during the combustion process in the empty bed at 900 °C through methane decomposition reaction, and was enhanced for experiments conducted with increased retention time. Carbon formation was obviously reduced for experiments conducted in the moving bed fuel reactor with oxygen carrier-to-fuel ratio (ϕ) higher than 1.14. The oxygen carriers that moving out of the moving bed reactor were composed of mainly FeO and FeAl2O4, characterized by X-ray diffraction (XRD) analysis. The formation of FeO and FeAl2O4 indicated that further utilization of oxygen in iron-based oxygen carriers can be achieved by moving bed operation

  14. Organ procurement after euthanasia: Belgian experience.

    Science.gov (United States)

    Ysebaert, D; Van Beeumen, G; De Greef, K; Squifflet, J P; Detry, O; De Roover, A; Delbouille, M-H; Van Donink, W; Roeyen, G; Chapelle, T; Bosmans, J-L; Van Raemdonck, D; Faymonville, M E; Laureys, S; Lamy, M; Cras, P

    2009-03-01

    Euthanasia was legalized in Belgium in 2002 for adults under strict conditions. The patient must be in a medically futile condition and of constant and unbearable physical or mental suffering that cannot be alleviated, resulting from a serious and incurable disorder caused by illness or accident. Between 2005 and 2007, 4 patients (3 in Antwerp and 1 in Liège) expressed their will for organ donation after their request for euthanasia was granted. Patients were aged 43 to 50 years and had a debilitating neurologic disease, either after severe cerebrovascular accident or primary progressive multiple sclerosis. Ethical boards requested complete written scenario with informed consent of donor and relatives, clear separation between euthanasia and organ procurement procedure, and all procedures to be performed by senior staff members and nursing staff on a voluntary basis. The euthanasia procedure was performed by three independent physicians in the operating room. After clinical diagnosis of cardiac death, organ procurement was performed by femoral vessel cannulation or quick laparotomy. In 2 patients, the liver, both kidneys, and pancreatic islets (one case) were procured and transplanted; in the other 2 patients, there was additional lung procurement and transplantation. Transplant centers were informed of the nature of the case and the elements of organ procurement. There was primary function of all organs. The involved physicians and transplant teams had the well-discussed opinion that this strong request for organ donation after euthanasia could not be waived. A clear separation between the euthanasia request, the euthanasia procedure, and the organ procurement procedure is necessary. PMID:19328932

  15. Shielding analyses for design of the upgraded JRR-3 research reactor, 2

    International Nuclear Information System (INIS)

    Shielding analyses of neutron beam holes have been presented for the shield design of the upgraded JRR-3 research reactor. Description is given about the calculational procedures and results for the standard beam hole, the beam hole for neutron radiography and the guide tunnels. The streaming analyses are made by using the MORSE-CG and DOT 3.5 codes. (author)

  16. Spent Fuel Management Program in the 3MW TRIGA MARK-II Research Reactor of Bangladesh

    International Nuclear Information System (INIS)

    Bangladesh Atomic Energy Commission (BAEC) has been operating a 3 MW TRIGA MARK II research reactor since 1986. The reactor was installed in the campus of the Atomic Energy Research Establishment (AERE) at Savar, Dhaka. It is one of the main nuclear research facilities in the country. The reactor uses TRIGA LEU fuel with uranium content of 20% by weight. The enrichment level of the fuel is 19.7%. The reactor has so far been operated for 7834 hours with a total cumulative burn up of 15898 MWh (662.5 MWd). The total burn up life of the present core is 1200 MWd. The main areas of use are: training of man-power for nuclear power plant applications, radioisotope (RI) production, neutron activation analysis (NAA), neutron radiography (NR) and neutron scattering. The government of Bangladesh has taken decision to establish nuclear power programme in the country. There is an ADP (Annual Development Project) to accomplish necessary activities for construction of medium size nuclear power plant (NPP) in the western zone of the country. Now, with regard to the safe management, storage of spent fuel and disposal of radioactive waste arising from operation of the research reactor and also from the proposed NPP expected to be constructed in future, BAEC is drawing up short and long-term plans and programs. At present, there does not exist any spent fuel element in the reactor facility. It is to be mentioned that Bangladesh is aware of the US DOE’s ‘Take Back Program’ in connection with the research reactor spent fuel of US origin, and is very much interested to take part in this program. The paper presents the current status of handling and storage facilities available for spent fuel and strategy for the safe management of spent fuel to be generated from the research reactor in near future. (author)

  17. Carbon Dioxide, the Coolant Gas of the G2/G3 Reactors: Leaks, Analysis, Activity

    International Nuclear Information System (INIS)

    This paper first describes very briefly the coolant gas circuits of the G2 and G3 reactors. The first part of the paper contains a detailed study of gas leaks together with relevant operating results. This study was carried out with the purpose of reducing still further the daily leakage rate and thus the operating costs of the reactors. The classification of CO2 circuits is based on the kinematics of the gas in the main reactor vessel and in the storage and feed systems. In this way a distinction is made between the feed circuits, the heat-exchange circuits and the leakage zones. While it cannot be claimed that this work alone has reduced by one-half the daily leakage rate between the initial operating period of the reactors and the present period of normal operation, it has certainly contributed to a greater understanding of the dynamics of the coolant gas and helped in the investigation of leaks. This study is virtually indispensable when the leakage rate is very low (3t/day) in relation to reactor capacity and it helps in the investigation of very small leaks. The second part of the paper describes various installations for chemical analysis and for measuring the activity of the coolant gas of the G2 and G3 reactors, presenting operating results for a period of several years. It also contains an outline of the various methods employed, of modifications and improvements to the original installations made with a view to providing more information about the sensitive and vital points in the installations and about the measurement parameters affecting the behaviour of the reactors (humidity, argon content, etc.). (author)

  18. Development of the fast reactor group constant set JFS-3-J3.2R based on the JENDL-3.2

    CERN Document Server

    Chiba, G

    2002-01-01

    It is reported that the fast reactor group constant set JFS-3-J3.2 based on the newest evaluated nuclear data library JENDL3.2 has a serious error in the process of applying the weighting function. As the error affects greatly nuclear characteristics, and a corrected version of the reactor constant set, JFS-3-J3.2R, was developed, as well as lumped FP cross sections. The use of JFS-3-J3.2R improves the results of analyses especially on sample Doppler reactivity and reaction rate in the blanket region in comparison with those obtained using the JFS-3-J3.2.

  19. Risk assessment for furan contamination through the food chain in Belgian children.

    Science.gov (United States)

    Scholl, Georges; Huybrechts, Inge; Humblet, Marie-France; Scippo, Marie-Louise; De Pauw, Edwin; Eppe, Gauthier; Saegerman, Claude

    2012-08-01

    Young, old, pregnant and immuno-compromised persons are of great concern for risk assessors as they represent the sub-populations most at risk. The present paper focuses on risk assessment linked to furan exposure in children. Only the Belgian population was considered because individual contamination and consumption data that are required for accurate risk assessment were available for Belgian children only. Two risk assessment approaches, the so-called deterministic and probabilistic, were applied and the results were compared for the estimation of daily intake. A significant difference between the average Estimated Daily Intake (EDI) was underlined between the deterministic (419 ng kg⁻¹ body weight (bw) day⁻¹) and the probabilistic (583 ng kg⁻¹ bw day⁻¹) approaches, which results from the mathematical treatment of the null consumption and contamination data. The risk was characterised by two ways: (1) the classical approach by comparison of the EDI to a reference dose (RfD(chronic-oral)) and (2) the most recent approach, namely the Margin of Exposure (MoE) approach. Both reached similar conclusions: the risk level is not of a major concern, but is neither negligible. In the first approach, only 2.7 or 6.6% (respectively in the deterministic and in the probabilistic way) of the studied population presented an EDI above the RfD(chronic-oral). In the second approach, the percentage of children displaying a MoE above 10,000 and below 100 is 3-0% and 20-0.01% in the deterministic and probabilistic modes, respectively. In addition, children were compared to adults and significant differences between the contamination patterns were highlighted. While major contamination was linked to coffee consumption in adults (55%), no item predominantly contributed to the contamination in children. The most important were soups (19%), dairy products (17%), pasta and rice (11%), fruit and potatoes (9% each). PMID:22632631

  20. Validation of the TRIAD3 code used for the neutronic simulation of the NRU reactor

    International Nuclear Information System (INIS)

    The neutronic simulation of the NRU research reactor at Chalk River is performed by the TRIAD3 code. TRIAD3 is a three-dimensional code using a modified neutron diffusion theory in two-energy groups. The modification is the use of cell discontinuity factors (cdf) to improve the radial neutron leakage calculation between adjacent cells. This paper describes three validation exercises performed over the past few years. It describes methods of obtaining the flux, power and reactivity measurements from the NRU reactor and presents comparisons between these measurement data and code simulation results. (author)

  1. Validation of the TRIAD3 code used for the neutronic simulation of the NRU reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

    The neutronic simulation of the NRU research reactor at Chalk River is performed by the TRIAD3 code. TRIAD3 is a three-dimensional code using a modified neutron diffusion theory in two-energy groups. The modification is the use of cell discontinuity factors (cdf) to improve the radial neutron leakage calculation between adjacent cells. This paper describes three validation exercises performed over the past few years. It describes methods of obtaining the flux, power and reactivity measurements from the NRU reactor and presents comparisons between these measurement data and code simulation results. (author)

  2. Characteristics and facilities of a 3MW LEU fuelled TRIGA reactor

    International Nuclear Information System (INIS)

    A 3 MW TRIGA reactor fuelled with low enriched uranium having 19.7 % U-235 and 20 wt% Uranium and Zirconium Hydride, has been installed and recently made critical at a research laboratory of the Bangladesh Atomic Energy Commission. This paper describes the basic design, low and high power test results and the facilities of the reactor. The details of the core configuration of the initial criticality with 50 elements and the final core with 100 elements at 3 MW power are discussed. The available experimental facilities are also described briefly. (author)

  3. Application of COREMELT-3D code at analysis of severe fast reactor accidents

    International Nuclear Information System (INIS)

    The code COREMELT for calculations of initial and transition stages of severe accident is considered. It is used to conduct connected calculations of nonstationary neutronic and thermohydraulic processes in sodium fast reactor core. The code has some versions depending on dimensions of solving problem and consists of thermohydraulic module COREMELT and neutronic module RADAR. Using the code COREMELT-3D connected calculations of core disassembly accidents of ULOF and UTOP type have been conducted for sodium fast reactors safety analysis. The main problem of code COREMELT-3D use is duration of calculation, speeding of the code is possible when calculating algorithms are parallelized

  4. Operation experience with the 3 MW TRIGA Mark-II research reactor of Bangladesh

    International Nuclear Information System (INIS)

    The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production (131I, 99mTc, 46Sc), various R and D activities and manpower training. The reactor has been operated successfully since it's commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power under forced-convection mode remained suspended for about 4 years. During that time, the reactor was operated at a power level of 250 kW so as to carry out experiments that require lower neutron flux. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50 g of TeO2 powder got melted inside the DCT. The vial was melted due to high heat generation on its surface while the reactor was operated for 8 hours at 3 MW for trial production of Iodine-131 (131I). A Wet Central Thimble (WCT) was used to replace the damaged DCT in June 2002 such that the reactor operation could be resumed. The WCT was again replaced by a new DCT in June 2003 such that radioisotope production could be continued. A total of 873 irradiation requests (IRs) have been catered for different reactor uses. Out of these, 114 IRs were for radioisotope (RI) production and 759 IRs for different experiments. The total amount of RI produced stands at about 2100 GBq. The total amount of burn-up-fuel is about 6158 MWh. Efforts are on to undertake an ADP project so as to convert the analog console and I and C system of the reactor into digital one. The paper summarizes the reactor operation experiences focusing on troubleshooting, rectification, modification, RI production, various R and D activities

  5. Assessment of the acrylamide intake of the Belgian population and the effect of mitigation strategies

    OpenAIRE

    Claeys, Wendie Liliane; Baert, Katleen; Mestdagh, Frédéric; Vercammen, Jan; Daenens, Paul; Meulenaer, Bruno De; Maghuin-Rogister, Guy; Huyghebaert, André

    2010-01-01

    Abstract The acrylamide (AA) intake of the Belgian consumer was calculated based on AA monitoring data of the Belgian Federal Agency for the Safety of the Food Chain (FASFC) and consumption data of the Belgian food consumption survey coordinated by the Scientific Institute for Public Health (3214 participants of 15 years or older). The average AA exposure, calculated probabilistically, was 0.4 ?g/kg bw/day (P97.5 = 1.6 ?g/kg bw/day) with as main contributors to the average intake c...

  6. Severe Accident Management Measures Introduced in Belgian NPP's

    International Nuclear Information System (INIS)

    In response to the Belgian Safety Authorities' request to address the severe accident issue within a decennial safety review, Tractebel, on behalf of the Belgian Utility, Electrabel, examined in detail specific severe accident topics and provided the Utility with several measures that could be implemented to reduce the risk associated with beyond-design accidents. The present paper summarizes the key elements of the approach applied in Belgium: - Presentation of plant-specific studies related to severe accident issues; - Use of PSA results; - Inputs of international R and D projects; - Selection and justification of severe accident measures; - Comparative study between possible mitigative measures; - Definition and justification of implemented severe accident management strategies. The vulnerability to severe accidents as well as the potential causes of containment failures have been identified leading to the study of possible countermeasures taking into account the combination of conservative design and post-TMI measures already implemented . A section of the paper will also be devoted to the specific study made for the selection, the sizing and the implementation of hydrogen control means. After the description of the selected measures implemented, the paper also describes the content of the 'Severe Accident Management Guidelines' developed by Tractebel for the Tihange NPPs and for the Doel NPPs. This project aimed at providing the operators with procedures or guidelines enabling to deal with complex situations not formally considered in the standard Emergency Response Guidelines, including accidents in which a significant portion of the core melts. The objective of these SAMG's programs is to indicate actions that must bring the plant to a controlled stable state and, above all, mitigate any challenges to the fission product barriers. The plant personnel must use the available plant information to determine the best severe accident management measures. Obviously

  7. Concept definition of an FRC/DD-3He advanced fusion reactor

    International Nuclear Information System (INIS)

    Posibilities of advanced fusion fuel cycle reactors are investigated. Characteristics of various D - D fusion fuel cycles are clarified and which magnetic confinement method can fit the most efficient advanced fuel cycle reactor is examined. A concept definition is considered for an advanced fusion reactor with DD - 3He fuel cycle in which the plasma is confined in a field-reversed configuration or field-reversed mirror. The concept definition is developed with emphasis on the feasibility of a steady-state self-ignited DD - 3He plasma with temperatures of 100 keV, the production method, the formation of ambipolar potential in the ambient plasma and the design of plasma energy direct convertor. (author)

  8. Experience on the refurbishment of the cooling system of the 3 MW TRIGA Mark II research reactor of Bangladesh and the modernization plan of the reactor control console

    International Nuclear Information System (INIS)

    The 3 MW TRIGA Mark II research reactor of the Bangladesh Atomic Energy Commission (BAEC) achieved its first criticality on 14 September 1986. Since then, the reactor has been used for manpower training, radioisotope production, and various R and D activities in the field of neutron activation analysis (NAA), neutron radiography (NR), and neutron scattering. Full power reactor operations remained suspended from 1997-2001 when a corrosion leakage problem in the 16N decay tank threatened the integrity of the primary cooling loop. The new tank was installed in 2001 and some modification and upgrades were carried out in the reactor cooling system such that the operational safety of the reactor could be strengthened. The cooling system upgrade mainly included replacement of the fouled shell and tube-type heat exchanger by a new plate-type one, modification of the cooling system piping layout, installation of isolation valves, installation of a chemical injection system for the secondary cooling system, modification of the Emergency Core Cooling System (ECCS), etc. After successful completion of all these modifications, the reactor was made operational again at full power of 3 MW in August 2001. BAEC, the operating organization, is now implementing a government-funded project to replace the old analogue control console of the research reactor with a digital control console. This paper focuses on the modification of the cooling system as well as the I and C system and the upcoming control console upgrade of the 3 MW TRIGA Mark II research reactor of Bangladesh. It also presents short descriptions of major incidents encountered so far in the reactor facility. (author)

  9. Digital radiographic equipment in the Belgian dental office.

    Science.gov (United States)

    Gijbels, F; Debaveye, D; Vanderstappen, M; Jacobs, R

    2005-01-01

    A survey was performed among Belgian dentists to evaluate the use and management of digital radiographic equipment. The majority of respondents work as general dental practitioners. One out of eight sets of equipment for extraoral exposures is digital. For intraoral radiography, 30% of the equipment is digital. While exposure time is reduced by about 50% for digital intraoral radiography compared with conventional radiography, no differences can be found between different conventional film speed classes. Appropriate collimation of the radiation beam is only sparingly used. Beam aiming devices to hold the film and position the radiation beam are not used by the majority of dentists. While 25% of the respondents stand behind a protective wall during exposure, 8% of dentists remain next to the patient during exposure while assisting in holding the film inside the mouth. A minority of the latter practitioners wear lead aprons. PMID:16461489

  10. Can Belgian nuclear power stations work for longer?

    International Nuclear Information System (INIS)

    In 2008 Paul Magnette, former Minister of Climate and Energy, requested the GEMIX Commission - a team of Belgian and international energy specialists - to examine the energy future of Belgium. Within this framework, it was planned to examine ideal energy mixes to ensure the energy supplies of Belgium, to secure our competitive position and to ensure that environmental and climate objectives are achieved. In the framework of this study, the GEMIX Commission asked SCK-CEN to evaluate the lifetime of the commercial nuclear power plants at Doel and Tihange. In particular, the Commission wanted to know whether it is technically feasible and safe to keep these power plants open for longer than 40 years, the lifetime stipulated in the 2003 Nuclear Energy Extrication Act. The article gives a summary overview of the expert opinion of SCK-CEN to the Gemix Commission.

  11. Belgian nuclear power life extension and fuss about nuclear rents

    International Nuclear Information System (INIS)

    Nuclear decision-making is embedded in slowly evolving political, economic and financial institutions. Belgium houses extended nuclear activities, mostly under French control, for example: SUEZ-GDF and EDF own all Belgian nuclear power plants. But a 2003 law mandates the closure of Belgium's nuclear power plants at a service age of 40 years; only force majeure could lift the strict obligation. Opposition to the law argued with climate change danger, financial losses, and loss-of-load risks. The financial issue got interwoven with a fuzzy debate on the definition, height and appropriation of “nuclear rents”. As plausible hypothesis is adopted: the prospected transfer of hundreds millions of euro from power companies to the public interest will create public support for life extension. But the nuclear rents discussion had faded in July 2012 when the Belgian government admitted a 10-year life extension for TIHANGE I (962 MW) and imposed the closure of the 2×433 MW DOEL I and II. Loss-of-load risk was the government's only public argument. The opacity of the decision process and its “fifty–fifty” outcome do not allow proper testing of the hypothesis. The case illustrates that politicians cannot bind their followers except through the deployment of alternative power sources. - Highlights: • Nuclear phase-out is only successful when alternative supplies are deployed. • Politicians cannot bind their successors by words or by lawgiving. • The phase-out law exemplifies the disruption of a strong nuclear lock-in. • Life extension exemplifies the disruption of the phase-out law. • The impact of imprecise nuclear rents on life extension could not be tested

  12. The BR1 Reactor:. a Versatile Tool for Fission Experiments

    Science.gov (United States)

    Wagemans, J.

    2008-04-01

    The BR1 reactor located at the Belgian Nuclear Research Centre SCK·CEN in Mol, Belgium, is a research reactor with a variety of irradiation possibilities. Thanks to its large reactor core, its flexible operation and its different irradiation facilities, this reactor is particularly suited for in-core and ex-core neutron physics experiments. This paper gives a general description of the BR1 reactor, with special emphasis on the available irradiation possibilities. Then some examples of fission experiments that have been performed in the past will be referred to and two ongoing projects related to fission will be presented.

  13. Revised cost estimate for the decommissioning of the reactor DR3

    DEFF Research Database (Denmark)

    2001-01-01

    The report describes a revision of the cost estimate for the decommissioning of the research Reactor DR 3 as described in the report Risø-R-1250(EN). Decommissioning of the Nuclear Facilities at Risø National Laboratory. Edited by Kurt Lauridsen. Therevision has been performed by the planning group...

  14. Opinion on serviceability of Bugey 3 reactor steam generators until their replacement foreseen in September 2010

    International Nuclear Information System (INIS)

    This document briefly reports the damage characterization of tubular bundles in steam generators of the Bugey 3 reactor, discusses the actions which are foreseen to prevent a tube failure risk, and discusses the risk of leakage during operation. Recommendations are formulated about investigation on the corrosion, and about prediction computation to be performed

  15. FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

    International Nuclear Information System (INIS)

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative

  16. Simplified 3D model of a PWR reactor vessel using fluid dynamics code ANSYS CFX computational

    International Nuclear Information System (INIS)

    This paper presents the results from the calculation of the steady state simulation with model of CFD (computational fluid dynamic) operating under conditions of operation at full power (Hot Full Power). Development and the CFD model results show the usefulness of these codes for calculating 3D of the variable thermohydraulics of these reactors.

  17. Direct energy conversion and neutral beam injection for catalyzed D and D-3He tokamak reactors

    International Nuclear Information System (INIS)

    The calculated performance of single stage and Venetian blind direct energy converters for Catalyzed D and D-3He Tokamak reactors are discussed. Preliminary results on He pumping are outlined. The efficiency of D and T neutral beam injection is reviewed

  18. Problem dependent nuclear data libraries in activation and dose calculations for ''RB-3'' reactor decommissioning

    International Nuclear Information System (INIS)

    A method used for the radioactive inventory calculation and dose evaluation of the CIRENE fuel clusters and ECO fuel elements irradiated in the RB-3 reactor is discussed in this paper with particular regard to the processed problem dependent nuclear data libraries. 8 refs, 1 fig., 1 tab

  19. Revised cost estimate for the decommissioning of the Reactor DR 3

    International Nuclear Information System (INIS)

    The report describes a revision of the cost estimate for the decommissioning of the research Reactor DR 3 as described in the report Risoe-R-1250(EN) Decommissioning of the Nuclear Facilities at Risoe National Laboratory Edited by Kurt Lauridsen. The revision has been performed by the planning group in the Risoe Decommissioning Department, and has been carried out as a discussion and evaluation of procedures methods and necessary resources to overcome the different phases of the decommissioning task of the Reactor. (au)

  20. Dosimetry analyses of the Ringhals 3 and 4 reactor pressure vessels

    International Nuclear Information System (INIS)

    A comprehensive series of neutron dosimetry measurements consisting of surveillance capsules, reactor pressure vessel cladding samples, and ex-vessel neutron dosimetry has been analyzed and compared to the results of three-dimensional, cycle-specific neutron transport calculations for the Ringhals Unit 3 and Unit 4 reactors in Sweden. The comparisons show excellent agreement between calculations and measurements. The measurements also demonstrate that it is possible to perform retrospective dosimetry measurements using the 93Nb (n,n') 93mNb reaction on samples of 18-8 austenitic stainless steel with only trace amounts of elemental niobium. (authors)

  1. RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

    International Nuclear Information System (INIS)

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point

  2. RELAP5-3D code for supercritical-pressure, light-water-cooled reactors

    International Nuclear Information System (INIS)

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point. (author)

  3. Power measurement of the RA-3 reactor using the neutron noise technique and 16N

    International Nuclear Information System (INIS)

    This work describes a measurement method based on the neutron noise technique which is used for determining the relation between the power and the currents of two ionization chambers. These chambers are sensitive to the gamma radiation from the 16N decay produced in the RA-3 reactor core. The power during operation is obtained from the calibration factors by measuring those currents. As this calibration factors depend on the cooler flow that circulates in the reactor core and in the 16N measuring system, an estimator, that is a function of the ratio of this currents, is proposed in order to detect flow changes. (author)

  4. Study on the thermohydraulic characteristics of marine nuclear reactors, 3

    International Nuclear Information System (INIS)

    Critical heat flux(CHF) of Freon-113 flowing vertically upward under periodically oscillating gravity field were obtained experimentally. The test section is a 10 x 22 x 1300mm annular channel, and the heater is uniformly heated by electric current. Pressure was 3kg/cm2 abs, mass flux was 1.83 x 106 or 3.51 x 106kg/m2h, inlet subcooling 1 to 270C, and nominal acceleration amplitude 0 to 0.5g. In general, as acceleration amplitude increase, CHF decrease smoothly. When the test loop is moved, CHF phenomena become indistinct due to oscillating behavior of CHF detector output. This oscillation is regarded as a result of time variation of the dry-out boundary. (auth.)

  5. Advanced reactors transition fiscal year 1995 multi-year program plan WBS 7.3

    International Nuclear Information System (INIS)

    This document describes in detail the work to be accomplished in FY-1995 and the out years for the Advanced Reactors Transition (WBS 7.3). This document describes specific milestones and funding profiles. Based upon the Fiscal Year 1995 Multi-Year Program Plan, DOE will provide authorization to perform the work outlined in the FY 1995 MYPP. Following direction given by the US Department of Energy (DOE) on December 15, 1993, Advanced Reactors Transition (ART), previously known as Advanced Reactors, will provide the planning and perform the necessary activities for placing the Fast Flux Test Facility (FFTF) in a radiologically and industrially safe shutdown condition. The DOE goal is to accomplish the shutdown in approximately five years. The Advanced Reactors Transition Multi-Year Program Plan, and the supporting documents; i.e., the FFTF Shutdown Program Plan and the FFTF Shutdown Project Resource Loaded Schedule (RLS), are defined for the life of the Program. During the transition period to achieve the Shutdown end-state, the facilities and systems will continue to be maintained in a safe and environmentally sound condition. Additionally, facilities that were associated with the Office of Nuclear Energy (NE) Programs, and are no longer required to support the Liquid Metal Reactor Program will be deactivated and transferred to an alternate sponsor or the Decontamination and Decommissioning (D and D) Program for final disposition, as appropriate

  6. Removal of reactor cooling system facilities and others in reconstruction works of JRR-3

    International Nuclear Information System (INIS)

    The home-manufactured No.1 reactor 'JRR-3' in Tokai Research Establishment, Japan Atomic Energy Research Institute, stopped its operation in March, 1983, with the results of cumulative operation time 47,135.5 hours and cumulative power output 419,073 MWh. Since then, the reconstruction works to construct a research reactor with higher performance has been advanced, and the new reactor has attained the initial criticality on March 22, 1990. The removal of the facilities which are not used after the reconstruction was carried out since August, 1985, and in this report, the removal of the parts with relatively high dose equivalent such as reactor cooling system facilities is outlined, except the reactor proper. The range of the removed facilities, the planning of removal works, the progress of removal works and the results are reported. About 110 t of large equipments were preserved as they are, and about 400 t of pipings and others were dismantled, removed, cut and put in containers. (K.I.)

  7. Calculation of SPERT Reactor benchmarks using 3D diffusion code DIREN

    International Nuclear Information System (INIS)

    The three dimensional diffusion code DIREN was developed at Institute for Nuclear Research (INR) Pitesti for reactor physics calculations for natural uranium and advanced CANDU reactors. Cell codes used are WIMS (from NEA library) and DRAGON (available in open source system). The latter is used also for super cell modeling of reactor control devices. These codes and the auxiliary programs were linked together in a calculation system. In order to apply WIMS-DRAGON-DIREN system to LWR, first the reactor SPERT benchmarks problems were calculated. The core including the control rods was modeled in three dimensional geometry. Following the calculations of the critical height (Hcrit), three dimensional power and flux distributions were obtained. The standard procedure used for CANDU reactor calculations (incremental cross sections for reactivity devices) underestimated the worth of control rods. A simple procedure to obtain the internal boundary conditions was developed using the super cell code DRAGON. Also the DIREN 3D diffusion code was modified to apply inner boundary conditions at control rods assigned volumes. Applying the inner boundary conditions yielded results closer to the measured values (e.g. the measured Hcrit was 49.53 cm as compared to 53.15 cm, the calculated one on 7 groups for nominal temperature). The reactivity coefficients for temperature and density required in transient's simulations were also calculated. The sample test problem T83 (hot stand-by, fast transient) was simulated using the RELAP code. (authors)

  8. Assessment of doses received by the Belgian population due to the Chernobyl releases

    International Nuclear Information System (INIS)

    The consequences of the exposure during the first year and beyond the first year after the Chernobyl accident in terms of radiation effects on the Belgian population are discussed as well as some uncertainties in these evaluations. (A.F.)

  9. The Labour Market as the Driving Force of Belgian Higher Education.

    Science.gov (United States)

    Wielemans, Willy

    1988-01-01

    An examination of internal and external forces on Belgian higher education suggests that the system is too closely controlled by economic and political forces in the labor market, which threatens to distort university life and higher education in general. (MSE)

  10. Simulation of the neutron flux in the irradiation facility at RA-3 reactor

    International Nuclear Information System (INIS)

    A facility for the irradiation of a section of patients' explanted liver and lung was constructed at RA-3 reactor, Comisión Nacional de Energía Atómica, Argentina. The facility, located in the thermal column, is characterized by the possibility to insert and extract samples without the need to shutdown the reactor. In order to reach the best levels of security and efficacy of the treatment, it is necessary to perform an accurate dosimetry. The possibility to simulate neutron flux and absorbed dose in the explanted organs, together with the experimental dosimetry, allows setting more precise and effective treatment plans. To this end, a computational model of the entire reactor was set-up, and the simulations were validated with the experimental measurements performed in the facility.

  11. RA Reactor operation and maintenance (I-IX), part VI, Task 3.08/04, Refurbishment of the RA reactor

    International Nuclear Information System (INIS)

    During the period planned for maintenance and refurbishment of the RA reactor the gas reactor system including the ventilation system was inspected and tested, the components were cleaned. This report describes detailed instructions and actions concerning repair and decontamination of the gas and ventilation systems components

  12. MERIS imagery of Belgian coastal waters: mapping of suspended particulate matter and chlorophyll-a

    OpenAIRE

    Ruddick, K.; PARK, Y.; B. Nechad

    2003-01-01

    This paper describes a first application-oriented analysis of MERIS products for Suspended Particulate Matter (SPM) and Chlorophyll-a (CHL) concentration in Belgian coastal waters. Regional algorithms designed for Belgian waters have been implemented and compared with the standard MERIS products, termed Total Suspended Matter and Algal2 respectively. The standard and regional SPM products seem robust and give similar data. Notwithstanding a more complete match-up validation analysis, these pr...

  13. Working Paper 13-09 - Qualitative Employment Multipliers for the Belgian Environmental Industry

    OpenAIRE

    Adja Awa Sissoko; Bart Van den Cruyce

    2009-01-01

    The present paper computes cumulative employment generated by the Belgian environmental industry. Relying on Belgian input-output tables for the year 2000 and on detailed employment data (SAM sub ]matrix), we investigate the patterns of the employment in the environmental industry, by considering the worker types differentiated by gender, educational attainment or a combination of these characteristics. The employment multiplier analysis of environmental employment reveals some interesting di...

  14. Compliance of Companies with Corporate Governance Codes: Case Study on Listed Belgian

    OpenAIRE

    Sven H. De Cleyn

    2014-01-01

    Listed and large companies become increasingly subject to internal and external pressure to comply with ethical and social standards. This article focuses on one aspect of this matter, namely the corporate governance issue. Within the framework of recent corporate scandals, this paper investigates whether and to which extent Belgian publicly listed SMEs comply with the Belgian Code on Corporate Governance after its first year of introduction, which has been constituted in the framework of the...

  15. A nationwide Hospital Survey on Patient Safety Culture in Belgian Hospitals: Analysis and Benchmarking

    OpenAIRE

    Vlayen, Annemie; Hellings, Johan; Claes, Neree; Schrooten, Ward

    2010-01-01

    Objective To measure patient safety culture in Belgian hospitals and to examine the homogeneous grouping of underlying safety culture dimensions. Methods The Hospital Survey on Patient Safety Culture was distributed organisation-wide in 180 Belgian hospitals participating in the federal program on quality and safety between 2007 and 2009. Participating hospitals were invited to submit their data to a comparative database. Homogeneous groups of underlying safety culture dimensions were sou...

  16. Annual report for the steering committee of the association Euratom-Belgian State for fusion 1996

    Energy Technology Data Exchange (ETDEWEB)

    Moons, F.; Bogaerts, W.; Decreton, M.; Biver, E.; Coenen, S.; Benoit, Ph.; Coheur, L.; Deboodt, P.; Andreev, D.

    1996-09-01

    This report is prepared for the annual steering committee meting of the Association Euratom - Belgian State for Fusion. The period October 1995 to September 1996 is reported on.The fusion technology work performed at the Belgian Nuclear Research Centre SCK/CEN, the Department of Metallurgy and Materials Engineering of the Louvain University (Belgium) and S.A. Gradel, a Luxemburg company, is described.

  17. impact of six multimodal country-wide campaigns to promote hand hygiene in belgian hospitals

    OpenAIRE

    Fonguh, Sylvanus; Hammami, N; Catry, B; Simon, Anne; ICPIC

    2015-01-01

    Six campaigns sponsored by the Belgian federal government were organized to promote hand hygiene(HH) in Belgian hospitals between 2005 and 2015. The campaigns combined educational sessions for healthcare workers (HCWs), promotion of alcohol-based hand rubs, patient awareness and audits with performance feedback. Each campaign consisted of a pre-campaign data collection period, an awareness period with training and a post-campaign data collection period.

  18. BNAIC 2008: Proceedings 20th Belgian-Netherlands Conference on Artificial Intelligence

    OpenAIRE

    Nijholt, Anton; Pantic, Maja; Poel, Mannes; Hondorp, Hendri

    2008-01-01

    This book contains the proceedings of the 20th edition of the Belgian-Netherlands Conference on Artificial Intelligence. The conference was organized by the Human Media Interaction group of the University of Twente. As usual, the conference was under the auspices of the Belgian-Dutch Association for Artificial Intelligence (BNVKI) and the Dutch Research School for Information and Knowledge Systems (SIKS). The conference aims at presenting an overview of state-of-the-art research in artificial...

  19. Annual report for the steering committee of the association Euratom-Belgian State for fusion 1996

    International Nuclear Information System (INIS)

    This report is prepared for the annual steering committee meting of the Association Euratom - Belgian State for Fusion. The period October 1995 to September 1996 is reported on.The fusion technology work performed at the Belgian Nuclear Research Centre SCK/CEN, the Department of Metallurgy and Materials Engineering of the Louvain University (Belgium) and S.A. Gradel, a Luxemburg company, is described

  20. Application of Raptor-M3G to reactor dosimetry problems on massively parallel architectures - 026

    International Nuclear Information System (INIS)

    The solution of complex 3-D radiation transport problems requires significant resources both in terms of computation time and memory availability. Therefore, parallel algorithms and multi-processor architectures are required to solve efficiently large 3-D radiation transport problems. This paper presents the application of RAPTOR-M3G (Rapid Parallel Transport Of Radiation - Multiple 3D Geometries) to reactor dosimetry problems. RAPTOR-M3G is a newly developed parallel computer code designed to solve the discrete ordinates (SN) equations on multi-processor computer architectures. This paper presents the results for a reactor dosimetry problem using a 3-D model of a commercial 2-loop pressurized water reactor (PWR). The accuracy and performance of RAPTOR-M3G will be analyzed and the numerical results obtained from the calculation will be compared directly to measurements of the neutron field in the reactor cavity air gap. The parallel performance of RAPTOR-M3G on massively parallel architectures, where the number of computing nodes is in the order of hundreds, will be analyzed up to four hundred processors. The performance results will be presented based on two supercomputing architectures: the POPLE supercomputer operated by the Pittsburgh Supercomputing Center and the Westinghouse computer cluster. The Westinghouse computer cluster is equipped with a standard Ethernet network connection and an InfiniBandR interconnects capable of a bandwidth in excess of 20 GBit/sec. Therefore, the impact of the network architecture on RAPTOR-M3G performance will be analyzed as well. (authors)

  1. The 4th surveillance testing for Kori unit 3 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-10-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

  2. Nuclear group constant set FUSION-J3 for fusion reactor nuclear calculations based on JENDL-3

    International Nuclear Information System (INIS)

    Based on evaluated nuclear data file JENDL-3, published in April 1990, we produced a nuclear group constant set 'FUSION-J3' for fusion reactor nuclear calculation by ANISN code instead of GICX40 produced in 1977. The set FUSION-J3 is the coupled group constant set with neutron 125 and gamma-ray 40 group structure, and has the maximum order of 5 as Legendre expansion in scattering cross section. Forty nuclides included in FUSION-J3 can be used in fusion reactor nuclear calculations. Considering mobility in two-dimensional calculations and fixed group structure in induced activity calculation code system as the GICX40 structure, we composed also FUSION-40 group constant set with neutron 42 group and gamma-ray 21 group structure. The set FUSION-40 includes the same maximum order of the Legendre expansion and the same nuclides as FUSION-J3. From the results in experimental analysis and benchmark calculations, it became proved that JENDL-3 is at higher level of accuracy than ENDF/B-IV and -V. The set FUSION-J3 can be clear applicable to fusion reactor nuclear calculations. (author)

  3. Verification of RBMK-1500 reactor main circulation circuit model with Cathare V1.3L

    International Nuclear Information System (INIS)

    Among other computer codes, French code CATHARE is also applied for RBMK reactor calculations. In this paper results of such application for Ignalina NPP reactor (RBMK-1500 type) main circulation circuit are presented. Three transients calculations were performed: all main circulation pumps (MCP) trip, trip of one main circulation pump and trip of one main circulation pump without a closure of check valve on the pump line. Calculation results were compared to data from the Ignalina NPP, where all these transients were recorded in the years 1986, 1996 and 1998. The presented studies prove the capability of the CATHARE code to treat thermal-hydraulic transients with a reactor scram in the RBMK, in case of single or multiple pump trips. However, the presented model needs further improvements in order to simulate loss of coolant accidents. For this reason, emergency core cooling system should be included in the model. Additional model improvement is also needed in order to gain more independent pressure behavior in both loops. Also, flow rates through the reactor channels should be modeled by dividing channels into several groups, referring to channel power (in RBMK power produced in a channel, located in different parts of the core is not the same). The point-neutron kinetic model of the CATHARE code is not suitable to predict transients when the reactor is operating at a nominal power level. Such transients would require the use of 3D-neutron kinetics model to describe properly the strong space-time effect on the power distribution in the reactor core

  4. A robot-automated work site for repair of the Chinon A3 reactor

    International Nuclear Information System (INIS)

    In 1982, following degradation due to corrosion of low-carbon steel by carbon dioxide gas, the utility undertook to repair some of the support structures at Chinon A3. This involved consolidation and reinforcing thermocouples and gas monitor pipeworks supports. A welding process was selected and the use of robots became indispensable because of the large number of components to be replaced (200 per outage). Two robots, supplied with tool heads and replacement components from outside the reactor were used. The robots and their servers were coordinated by a central computer and monitored by a closed circuit television system. Each repair operation was performed after ''training'' on a full-scale mockup of the top of the reactor reconstructed from telemetry of the real reactor dimensions. Since becoming operational in June 1986, the robots have accumulated over 20 000 hours of operation and seventy parts have been welded to the reactor. A 3D CAD system has been adapted to simulate the robots and analyse long trajectories in order to reduce robot learning time

  5. Analysis of the particular spill characteristics observed by the Belgian aerial surveillance program during the Tricolor incident

    International Nuclear Information System (INIS)

    This presentation described the Tricolor oil spill incident, the remote sensing equipment used to monitor the spill, the observed spill characteristics and the flight data assessment. The spill occurred on December 14, 2002 following a collision between the carrier Tricolor and the container vessel Kariba in French waters in the Zone of Joint Responsibility, close to the Belgian and English borders. The Tricolor sank and 3 more vessels collided with the wreck in the five weeks following the collision, spilling several 100 tons of mostly heavy fuel oil into the sea. The remote sensing equipment aboard Belgian surveillance aircraft noted that freshly spilled oil formed a network of widespread dark oil trails surrounded by light oil fractions. The spill volumes were estimated to be high because of the large extent of the polluted area. Nine months following the spill, the emulsified oil trails had a density close to that of seawater. It was assumed that a cold and thick emulsion had formed and became trapped inside the wreck. Upon release, the emulsion could submerse and resurface. The incident demonstrated that early stage oil sample analysis could help interpret slick behaviour by means of remote sensing. 9 refs., 3 tabs., 1 fig

  6. DT and DHe3 tokamak test reactor concepts using advanced, high field superconductors

    International Nuclear Information System (INIS)

    If practical high temperature superconducting ceramic magnets can be developed, there could be a significant impact on reactor design. Potential advantages include a simpler, more robust magnet design, the possibility of demountable superconducting toroidal field coils and reduced shielding requirements. The high temperature superconductors can also have very high critical fields and could provide super high field operation. This could substantially increase eta tau/sub E/ values, reduce β requirements, and improve prospects for ohmic heating to ignition. The combination of moderately high β and super high field could make DHe3 operation possible in a JET size tokamak. In this paper we discuss possibilities for test reactor designs using high temperature high field superconductors. An illustrative design has a field at the plasma of 15 T. This reduces the required β to less than 2% for DT operation. The required plasma current is 5 MA. For a reactor size of R0 = 3.4m and a = 0.6m, the neutron wall loading is 3.3 MW/m2 at β = 1.5% for DT operation and an equal amount of fusion power is produced at β = 10% for DHe3 operation. One possible mode of operation is to use ohmic heating to ignition in a DT plasma followed by thermal runaway to DHe3 temperatures. 7 refs., 1 fig., 2 tabs

  7. Authorisation to irradiate two U3Si2 prototype fuel elements in the RA-3 research reactor

    International Nuclear Information System (INIS)

    The Argentine RA-3 research reactor (with U3O8-Al LEU MTR fuel type elements and in the process to increase its power from 5 MW to 10 MW since October 2002) is operated by the National Atomic Energy Commission (CNEA). After evaluating the safety studies submitted by CNEA, the Argentine Nuclear Regulatory Authority (ARN) authorised the irradiation of two U3Si2-Al prototype fuel elements named P-06 and P-07, in the RA-3, in September 2000 and September 2001 respectively. These irradiations are part of the qualification programme as manufacturer (CNEA). This report contains a brief description of the regulatory process the ARN carries out in order to approve major modifications or new experiments in research reactors; it also summarises conditions in force for operable core configurations, considers a definition of prototype fuel element and lists the main characteristics of the RA-3 research reactor. Then, the process involved in the authorisation to irradiate the prototypes P-06 and P-07 is particularly analysed. Finally, and in case an authorisation to irradiate a prototype fuel element of a non-qualified material (such as U-Mo) in the RA-3 is requested, the regulatory aspects that should be taken into consideration are analysed. This would be the first time a question of this nature is submitted to the ARN. (author)

  8. Effect of HEXBU-3D feedback and boundary condition models on the calculated reactor core characteristics of Loviisa NPP

    International Nuclear Information System (INIS)

    This paper discusses the effect of feedback and boundary condition models used in the Loviisa core design code HEXBU-3D on the calculated reactor core characteristics. Two alternative feedback models and two boundary condition models are shortly demonstrated and their effects on the calculated reactor physical properties are considered. Computational results by these alternative models will also be compared with measured reactor physical properties of the recent Loviisa-1 and Loviisa-2 cores. (authors)

  9. Development of a version of the reactor dynamics code DYN3D applicable for High Temperature Reactors; Entwicklung einer Version des Reaktordynamikcodes DYN3D fuer Hochtemperaturreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Rohde, Ulrich; Apanasevich, Pavel; Baier, Silvio; Duerigen, Susan; Fridman, Emil; Grahn, Alexander; Kliem, Soeren; Merk, Bruno

    2012-07-15

    Based on the reactor dynamics code DYN3D for the simulation of transient processes in Light Water Reactors, a code version DYN3D-HTR for application to graphitemoderated, gas-cooled block-type high temperature reactors has been developed. This development comprises: - the methodical improvement of the 3D steady-state neutron flux calculation for the hexagonal geometry of the HTR fuel element blocks - the development of methods for the generation of homogenised cross section data taking into account the double heterogeneity of the fuel element block structure - the implementation of a 3D model for heat conduction and heat transport in the graphite matrix. The nodal method for neutron flux calculation based on SP3 transport approximation was extended to hexagonal fuel element geometry, where the hexagons are subdivided into triangles, thus the method had finally to be derived for triangular geometry. In triangular geometry, a subsequent subdivision of the hexagonal elements can be considered, and therefore, the effect of systematic mesh refinement can be studied. The algorithm was verified by comparison with Monte Carlo reference solutions, on the node-wise level, as well as also on the pin-wise level. New procedures were developed for the homogenization of the double-heterogeneous fuel element structures. One the one hand, the so-called Reactivity equivalent Physical Transformation (RPT), the two-step homogenization method based on 2D deterministic lattice calculations, was extended to cells with different temperatures of the materials. On the other hand, the progress in development of Monte Carlo methods for spectral calculations, in particular the development of the code SERPENT, opened a new, fully consistent 3D approach, where all details of the structures on fuel particle, fuel compact and fuel block level can be taken into account within one step. Moreover, a 3D heat conduction and heat transport model was integrated into DYN3D to be able to simulate radial

  10. Tritium production, management and its impact on safety for a D3He fusion reactor

    International Nuclear Information System (INIS)

    This paper reports that about three percent of the fusion energy produced by the D-3He reactor is in the form of neutrons. Those neutrons are generated by D-D and D-T reactions, with the tritium produced by the D-D fusion. The neutrons will react with structural steel, deuterium, 3He and shielding material to produce tritium. about half of the tritium generated by the D-D reaction will not burn in the plasma and will exit as a part of the plasma exhaust. Thus, there is enough tritium produced in a D-3He reactor and careful management will be required. The tritium produced in the shield and plasma can be managed with an acceptable effect on cost and safety

  11. Development of Ignalina NPP RBMK-1500 reactor RELAP5-3D model

    International Nuclear Information System (INIS)

    This paper deals with the development of an integrated thermal-hydraulics-neutronics model for RBMK-1500 reactors for the analysis of specific plant transients in which the neutronic response of the core is important. A successful best estimate coupled RELAP5-3D model of Ignalina nuclear power plant (NPP) has been developed. The validation of the thermal-hydraulic model has been performed using operational transients from Ignalina NPP. The results of the calculations obtained with the RELAP5-3D model compare reasonably with the real plant data. The RELAP5-3D nodal kinetics model provides reasonable agreement with Ignalina NPP reactor power and coolant density profiles. The eigenvalue is close to unity, indicating that reasonable values are calculated for the neutron fluxes

  12. H-3 and Li-6 poisoning of the Maria reactor beryllium matrix

    International Nuclear Information System (INIS)

    This report discusses methods used to evaluate Li-6 and He-3 poison concentrations, initiated by Be-9(n, α) reaction in the beryllium blocks of the Maria reactor. The results based on ENDF/B-VI neutron cross sections, 3D diffusion neutron fluxes, and solutions to the differential equations which describe the time-dependent poison concentrations as function of reactor operation and shutdown periods. MCNP Monte Carlo calculations were used to verify calculated poison levels for observed critical configurations. Previous evaluations used somewhat less refined methods based on asymptotic solutions for the poison concentrations. It was found that Li-6 and He-3 in the beryllium blocks limit the available excess reactivities and alter flux and power distributions. Based on analyses of critical cores, it was determined that poison concentrations need be evaluated for an in-core region and for an excore region and not for each beryllium block. (author)

  13. The design of a fuel element for the RA-3 reactor (Ezeiza Atomic Center)

    International Nuclear Information System (INIS)

    Some features of the mechanical design of the low enrichment fuel element for the RA-3 reactor are described, with emphasis in those aspects of the original design that have been modified considering the experience acquired in the design of other fuel elements. The proposed modification is based fundamentally on the replacement of all welded joints by screwed joints, which facilitates the manufacture of the fuel element, avoiding the distortions produced by the welds used at present and contributing to the fulfillment of the foreseen tolerances. A basic characteristic of this design is a careful manufacture of the fuel element's structural components in order to assure an assembling of the fuel element that fulfills the tolerances intrinsically required. The fuel is designed for the RA-3 reactor and uses U3O8 or U3Si2 as carrying phase of the fissile material with an enrichment of 19.70% of 235U. The design verification was performed by analytical and numerical methods, and is supported by testing of materials in laboratory, hydrodynamics tests and performance evaluations of the fuel elements in the RA-3 reactor. (author)

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

    International Nuclear Information System (INIS)

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

  15. 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.

  16. The radiological impact of the Belgian phosphate industry

    International Nuclear Information System (INIS)

    The Belgian phosphate industry processes huge amounts of phosphate ore (1.5 to 2 Mton/year) for a wide range of applications, the most important being the production of phosphoric acid, fertilizers and cattle food. Marine phosphate ores show high specific activities of the natural uranium decay series (usually indicated by Ra-226) (e.g. 1200 to 1500 Bq/kg for Moroccan ore). Ores of magmatic origin generally contain less of the uranium and more of the thorium decay series (up to 500 Bq/kg). These radionuclides turn up in by-products, residues or product streams depending on the processing method and the acid used for the acidulation of the phosphate rock. Sulfuric acid is the most widely used, but also hydrochloric acid and nitric acid are applied in Belgium. For Flanders, the northern part of Belgium, we already have a clear idea of the production processes and waste streams. The five Flemish phosphate plants, from 1920 to 2000, handled 54 million ton of phosphate ore containing 65 TBq of radium-226 and 2.7 TBq of thorium- 232. The total surface area of the phosphogypsum and calcium fluoride sludge deposits amounts to almost 300 ha. There is also environmental contamination along two small rivers receiving the waste waters of the hydrochloric production process: the Winterbeek (> 200 ha) and the Grote Laak (12 ha). The data on the impact of the phosphate industry in the Walloon provinces in Belgium is less complete. A large plant produced in 2004 0.8 Mton of phosphogypsum, valorizing about 70 % of the gypsum in building materials (plaster, cement), in fertilizers, and in other products such as paper. The remainder was stored on a local disposal site. The radiological impact of the Belgian phosphate industry on the local population will be discussed. At present most contaminated areas are still recognizable as waste deposits and inaccessible to the population. However as gypsum deposits and other contaminated areas quickly blend in with the landscape, it is

  17. 3D Neutronic Analysis in MHD Calculations at ARIES-ST Fusion Reactors Systems

    Science.gov (United States)

    Hançerliogulları, Aybaba; Cini, Mesut

    2013-10-01

    In this study, we developed new models for liquid wall (FW) state at ARIES-ST fusion reactor systems. ARIES-ST is a 1,000 MWe fusion reactor system based on a low aspect ratio ST plasma. In this article, we analyzed the characteristic properties of magnetohydrodynamics (MHD) and heat transfer conditions by using Monte-Carlo simulation methods (ARIES Team et al. in Fusion Eng Des 49-50:689-695, 2000; Tillack et al. in Fusion Eng Des 65:215-261, 2003) . In fusion applications, liquid metals are traditionally considered to be the best working fluids. The working liquid must be a lithium-containing medium in order to provide adequate tritium that the plasma is self-sustained and that the fusion is a renewable energy source. As for Flibe free surface flows, the MHD effects caused by interaction with the mean flow is negligible, while a fairly uniform flow of thick can be maintained throughout the reactor based on 3-D MHD calculations. In this study, neutronic parameters, that is to say, energy multiplication factor radiation, heat flux and fissile fuel breeding were researched for fusion reactor with various thorium and uranium molten salts. Sufficient tritium amount is needed for the reactor to work itself. In the tritium breeding ratio (TBR) >1.05 ARIES-ST fusion model TBR is >1.1 so that tritium self-sufficiency is maintained for DT fusion systems (Starke et al. in Fusion Energ Des 84:1794-1798, 2009; Najmabadi et al. in Fusion Energ Des 80:3-23, 2006).

  18. Oxidation behavior of ammonium in a 3-dimensional biofilm-electrode reactor.

    Science.gov (United States)

    Tang, Jinjing; Guo, Jinsong; Fang, Fang; Chen, Youpeng; Lei, Lijing; Yang, Lin

    2013-12-01

    Excess nitrogenous compounds are detrimental to natural water systems and to human health. To completely realize autohydrogenotrophic nitrogen removal, a novel 3-dimensional biofilm-electrode reactor was designed. Titanium was electroplated with ruthenium and used as the anode. Activated carbon fiber felt was used as the cathode. The reactor was separated into two chambers by a permeable membrane. The cathode chamber was filled with granular graphite and glass beads. The cathode and cathode chamber were inhabited with domesticated biofilm. In the absence of organic substances, a nitrogen removal efficiency of up to 91% was achieved at DO levels of 3.42 +/- 0.37 mg/L when the applied current density was only 0.02 mA/cm2. The oxidation of ammonium in biofilm-electrode reactors was also investigated. It was found that ammonium could be oxidized not only on the anode but also on particle electrodes in the cathode chamber of the biofilm-electrode reactor. Oxidation rates of ammonium and nitrogen removal efficiency were found to be affected by the electric current loading on the biofilm-electrode reactor. The kinetic model of ammonium at different electric currents was analyzed by a first-order reaction kinetics equation. The regression analysis implied that when the current density was less than 0.02 mA/cm2, ammonium removal was positively correlated to the current density. However, when the current density was more than 0.02 mA/cm2, the electric current became a limiting factor for the oxidation rate of ammonium and nitrogen removal efficiency. PMID:24649670

  19. Integral benchmark experiments of the Japanese Evaluated Nuclear Data Library (JENDL)-3.3 for the fusion reactor design

    International Nuclear Information System (INIS)

    JENDL-3.3 is a neutron cross section data library of 337 nuclei evaluated from the latest experimental data. JENDL-3.3 introduces double differential cross sections, which are energy- and angle-dependent ones of the scattered secondary neutrons, and are important for anisotropic neutron transport calculations for the fusion reactors design. This paper overviews benchmark experiments carried out for key fusion related nuclei such as Iron and Vanadium, and the results of analyses with JENDL-3.3, together with JENDL-3.2 and FENDL-2 for a comparison purpose. The experiments have been carried out at the Fusion Neutron Source (FNS) of JAERI. During the neutron injection into the assemblies, neutron and secondary gamma-ray spectra have been measured inside and outside the assemblies. For the test assemblies, we have used Iron, Copper, Vanadium and Tungsten as a single element material, and LiAlO2 and SiC as a compound material. From the integral benchmark experiments it was confirmed that the accuracy of JENDL-3.3 has been improved well compared with JENDL- 3.2 and FENDL-2 by the re-evaluation using latest experimental data, and JENDL-3.3 is suitable for the nuclear analysis of the fusion reactor. (author)

  20. Dry storage of MTR spent fuel from the Argentine radioisotope production reactor RA-3

    International Nuclear Information System (INIS)

    The nuclear fuel elements of the RA-3 reactor consist in 19 rectangular fuel plates held in position by two lateral structural plates. The whole assembly is coupled to the lower nozzles that fits in the reactor core grid. The inner plates are 1.5 mm thick, 70.5 mm wide and 655 mm long and the outer plates are 100 mm longer. The fuel plates are formed by a core of an AI-U alloy co-laminated between two plates of Al. Enrichment is 90% 235U. After being extracted from the reactor, the fuel elements have been let to cool down in the reactor storage pool and finally moved to the storage facility. This facility is a grid of vertical underground channels connected by a piping system. The system is filled with processed and controlled water. At the present the storage capacity of the facility is near to be depleted and some indications of deterioration of the fuel elements has been detected. Due to the present status of the facility and the spent fuel stored there, a decision has been taken to proceed to modify the present underwater storage to dry storage. The project consist in: a) Decontamination and conditioning of the storage channels to prepare them for dry storage. b) Disassembly of the fuel elements in hot cells in order to can only the active fuel plates in an adequate tight canister. c) The remnant structural pieces will be treated as low level waste. (author). 10 figs

  1. Safety and environmental aspects of the Apollo-L2 D-3He reactor

    International Nuclear Information System (INIS)

    This paper reports on Apollo-L2, a D-3He fueled tokamak reactor design that utilizes direct conversion. The reactor shield is made of steel and cooled with water. Three different austenitic steel alloys (PCA, 316 SS and Tenelon) were chosen to study the impact of material selection on the environmental and safety attractiveness of the reactor. The thermal response of the different shields following a loss of coolant accident (LOCA) was determined up to two weeks after an unscheduled shutdown of the reactor. The Tenelon structure encountered the highest temperature increase following the accident. The maximum temperature a Tenelon first wall reaches is 500degreeC. The nickel-stabilized steel structures (PCA and 316 SS) result in the highest off-site dose due to its high radioactive cobalt content. 58Co and 60Co are the major contributors to the calculated dose. The low temperature of the structure during a LOCA results in the release of a very small fraction of the radioactive inventory at the onset of an accident. Hence, the Apollo-L2 design achieves the inherent safety criteria with respect to activation products

  2. Proceedings of the fourth CSNI specialist meeting on fuel-coolant interaction in nuclear reactor safety - Volumes 2+3

    International Nuclear Information System (INIS)

    This document presents the volumes 2 and 3 of the proceedings of the fourth CSNI specialist meeting on fuel-coolant interaction in nuclear reactor safety that was held in Bournemouth, UK, 2-5 april 1979: seven papers for session IV (Specific and well-characterized integral FCI experiments), five papers for Session V (FCI studies directly related to reactor conditions), and seven papers for Session VI (Implications of FCI for reactor safety studies). Session VII presents two panel discussions, the first one related to the science of fuel coolant interactions, the second one to reactor safety implications

  3. Research on intelligent monitor for 3D power distribution of reactor core

    International Nuclear Information System (INIS)

    Highlights: • Core power distribution of ex-core measurement system has been reconstructed. • Building up an artificial intelligence model for 3-D core power distribution. • Error of the experiments has been reduced to 0.76%. • Methods for improving the accuracy of the model have been obtained. - Abstract: A real-time monitor for 3D reactor power distribution is critical for nuclear safety and high efficiency of NPP’s operation as well as for optimizing the control system, especially when the nuclear power plant (NPP) works at a certain power level or it works in load following operation. This paper was based on analyzing the monitor for 3D reactor power distribution technologies used in modern NPPs. Furthermore, considering the latest research outcomes, the paper proposed a method based on using an ex-core neutron detector system and a neural network to set up a real time monitor system for reactor’s 3D power distribution supervision. The results of the experiments performed on a reactor simulation machine illustrated that the new monitor system worked very well for a certain burn-up range during the fuel cycle. In addition, this new model could reduce the errors associated with the fitting of the distribution effectively, and several optimization methods were also obtained to improve the accuracy of the simulation model

  4. Study of heat transfer in 3D fuel rods of the EPRI-9R reactor modified

    International Nuclear Information System (INIS)

    This paper aims to conduct a case study of the fuel rods that have the highest and the lowest average power of the EPRI-9R 3D reactor modified , for various positions of the control rods banks. For this, will be addressed the verification of computer code, comparing the results obtained with analytical solutions. This check is important so that, subsequently, it is possible use the program to understand the behavior of the fuel rods and the coolant channel of the EPRI-9R 3D reactor modified. Thus, in view of the scope of this paper, first a brief introducing on the heat transfer is done, including the rod equations and the equation of energy in the channel to allow the analysis of the results

  5. Neutron Transport in Hexagonal Reactor Cores Modeled by Trigonal-Geometry Diffusion and Simplified P3 Nodal Methods

    OpenAIRE

    Duerigen, Susan

    2013-01-01

    The superior advantage of a nodal method for reactor cores with hexagonal fuel assemblies discretized as cells consisting of equilateral triangles is its mesh refinement capability. In this thesis, a diffusion and a simplified P3 (or SP3) neutron transport nodal method are developed based on trigonal geometry. Both models are implemented in the reactor dynamics code DYN3D. As yet, no other well-established nodal core analysis code comprises an SP3 transport theory model based on trigonal mesh...

  6. Efficiency Calibration of LaBr3(Ce) γ Spectroscopy in Analyzing Radionucles in Reactor Loop Water

    Institute of Scientific and Technical Information of China (English)

    CHEN; Xi-lin; QIN; Guo-xiu; GUO; Xiao-qing; CHEN; Yong-yong; MENG; Jun

    2013-01-01

    Monitoring the occurring and radioactivity concentration of fission products in nuclear reactor loop water is important for the nuclear reactor safe running evaluation,prevention of accidence and safe protection of working personnel.Study on the efficiency calibration for a LaBr3(Ce)detector experimental

  7. Heterogeneous computation tests of both substitution and reactivity worth experiments in the RB-3 reactor

    International Nuclear Information System (INIS)

    This report presents the results of several experiments carried out in the D2O-moderated RB-3 reactors at the CNEN's Laboratory of Montecuccolino, Bologna. The experiments referred to are either fuel-element substitution experiments or interstitial absorber experiments and were performed during the period 1972-1974. The results of measurements are compared with those obtained by means of computational procedure based on some ''cell'' codes coupled with heterogeneous codes. (authors)

  8. European contribution to Phase 3 of the benchmark core analysis for the BN-600 hybrid reactor

    International Nuclear Information System (INIS)

    This European participation in Phase 3 of the benchmark (BN-600) analysis consist of a joint contribution from France and the UK. Calculations were performed by ERANOS code and data system which has been developed in the framework of European cooperation on fast reactors. Results are presented for all the core neutronic parameters, both for homogeneous and heterogeneous core models and both for beginning and end of fuel cycle

  9. Risk assessment of a pressurized water reactor for Class 3-8 accidents

    Energy Technology Data Exchange (ETDEWEB)

    Hall, R.E.

    1979-10-01

    An assessment has been made of the impact on societal risk of Class 3-8 accident sequences as defined by Appendix D to 10 CFR50. The present analysis concentrates on a pressurized water reactor and utilizes realistic assumptions when practical. Conclusions are drawn as to the relative importance of the analyzed accidents and their impact on the development of a complete societal risk curve. 65 refs., 61 figs., 37 tabs.

  10. Contact allergy caused by methylisothiazolinone: the Belgian-French experience.

    Science.gov (United States)

    Aerts, Olivier; Goossens, An; Giordano-Labadie, Françoise

    2015-01-01

    The chemical Kathon CG(®), a mixture of the preservatives methylchloroisothiazolinone (MCI) and methylisothiazolinone (MI), was the leading cause of a worldwide epidemic of contact-allergic reactions in the eighties. From 2000 on, MI alone became allowed in industrial products and in 2005 authorities gave a green light for its use in leave-on and rinse-off cosmetics up to a maximum concentration of 100 ppm (0.01%). Following initial occupational cases, a continuously increasing number of consumers sensitized to MI have been reported and both Belgian and French allergy groups decided to routinely test MI in their baseline series from 2010 onwards. Two multicenter studies, comprising 8,680 and 7,874 patients in Belgium and France respectively, both clearly show the rise in contact allergy caused by MI, with a spectacular sensitization rate of ∼ 6.0% in 2012, even increasing to 7.0% in 2013. Mostly middle-aged women, presenting with facial-and/or hand dermatitis, were affected, although very young children were reported as well. Furthermore, the data confirmed that sensitization is primarily caused by cosmetics (mostly leave-on, but also rinse-off), household detergents and water-based paint. This unprecedented outbreak of contact sensitization to a preservative agent in Europe, and beyond, should have alerted the authorities much sooner and meanwhile the need for safer use concentrations of MI in cosmetics, detergents and industrial products is becoming more urgent every day. PMID:26412037

  11. New reactor concepts; Nieuwe rectorconcepten - nouveaux reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A

    1998-11-01

    The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost.

  12. Experimental and kinetic modeling study of 3-methylheptane in a jet-stirred reactor

    KAUST Repository

    Karsenty, Florent

    2012-08-16

    Improving the combustion of conventional and alternative fuels in practical applications requires the fundamental understanding of large hydrocarbon combustion chemistry. The focus of the present study is on a high-molecular-weight branched alkane, namely, 3-methylheptane, oxidized in a jet-stirred reactor. This fuel, along with 2-methylheptane, 2,5-dimethylhexane, and n-octane, are candidate surrogate components for conventional diesel fuels derived from petroleum, synthetic Fischer-Tropsch diesel and jet fuels derived from coal, natural gas, and/or biomass, and renewable diesel and jet fuels derived from the thermochemical treatment of bioderived fats and oils. This study presents new experimental results along with a low- and high-temperature chemical kinetic model for the oxidation of 3-methylheptane. The proposed model is validated against these new experimental data from a jet-stirred reactor operated at 10 atm, over the temperature range of 530-1220 K, and for equivalence ratios of 0.5, 1, and 2. Significant effort is placed on the understanding of the effects of methyl substitution on important combustion properties, such as fuel reactivity and species formation. It was found that 3-methylheptane reacts more slowly than 2-methylheptane at both low and high temperatures in the jet-stirred reactor. © 2012 American Chemical Society.

  13. The SPES3 Experimental Facility Design for the IRIS Reactor Simulation

    Directory of Open Access Journals (Sweden)

    Mario Carelli

    2009-01-01

    Full Text Available IRIS is an advanced integral pressurized water reactor, developed by an international consortium led by Westinghouse. The licensing process requires the execution of integral and separate effect tests on a properly scaled reactor simulator for reactor concept, safety system verification, and code assessment. Within the framework of an Italian R&D program on Nuclear Fission, managed by ENEA and supported by the Ministry of Economic Development, the SPES3 facility is under design and will be built and operated at SIET laboratories. SPES3 simulates the primary, secondary, and containment systems of IRIS with 1 : 100 volume scale, full elevation, and prototypical thermal-hydraulic conditions. The simulation of the facility with the RELAP5 code and the execution of the tests will provide a reliable tool for data extrapolation and safety analyses of the final IRIS design. This paper summarises the main design steps of the SPES3 integral test facility, underlying choices and phases that lead to the final design.

  14. A 3D transport-based core analysis code for research reactors with unstructured geometry

    International Nuclear Information System (INIS)

    Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal SN method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of keff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results

  15. DET3D - A CFD tool for simulating hydrogen combustion in nuclear reactor safety

    International Nuclear Information System (INIS)

    In this paper, the CFD code DET3D is described which has been used for simulating assumed accident scenarios in nuclear reactors (fission and fusion, see e.g. [4, 5, 11]) involving hydrogen detonations in complex 3-dimensional geometries. Two validation calculations against experiments are discussed in detail: (i) a hydrogen-air detonation in a 12 m long straight tube with a truly 3-dimensional inner obstacle, and (ii) a pure radiolytic gas detonation in a 5 m long U-shaped tube at 44 bar initial pressure. (author)

  16. Cost analysis and financial risk profile for severe reactor accidents at Waterford-3

    International Nuclear Information System (INIS)

    To support Louisiana Power and Light Company (LP and L) in determining an appropriate level of nuclear property insurance for Waterford Steam Electric Station, Unit 3 (Waterford-3), ABZ, Incorporated, performed a series of cost analyses and developed a financial risk profile. This five-month study, conducted in 1991, identified the potential Waterford-3 severe reactor accidents and described each from a cleanup perspective, estimated the cost and schedule to cleanup from each accident, developed a probability distribution of associated financial exposure, and developed a profile of financial risk as a function of insurance coverage

  17. Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program; Simulacion en 3 dimensiones de un ciclo de 18 meses para un reactor BWR usando el programa Nod3D

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez, N.; Alonso, G. [ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)]. E-mail: nhm@nuclear.inin.mx; Valle, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

    2004-07-01

    The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

  18. Photocatalytic reactors for treating water pollution with solar illumination, Part 3: a simplified analysis for recirculating reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sagawe, G.; Bahnemann, D. [Hannover Univ. (Germany). Inst. fuer Technische Chemie; Brandi, R.J.; Cassano, A.E. [Universidad Nacional de Litoral, Santa Fe (Argentina). Inst. de Desarrollo Tecnologico para la Imdustria Quimica

    2004-11-01

    A solar photoreactor operated in the batch, recirculating mode is analyzed in terms of very simple observable variables such as the impinging photon flux, the incident area, the initial concentration, the flow rate, the reactor volume and a property defined as the Observed Photonic Efficiency. The proposed equipment is made of a tubular reactor, a tank, a pump and the connecting pipes. The analysis is formulated in terms of the photon input corresponding to an equivalent batch system that is derived as a new reaction coordinate for photoreactions. Employing several plausible approximations, the pollutant concentration evolution in the tank is cast in terms of very simple analytical solutions. Process photonic efficiencies are defined for the system operation and calculated with respect to the maximum achievable yield corresponding to the differential operation of the solar recirculating reactor. (Author)

  19. A pilot application of the RELAP file to the steady state and transient analysis of a test section inside the BR2 reactor

    International Nuclear Information System (INIS)

    BR2 is a material test reactor sited in the Belgian Nuclear Research Centre in Mol. The main research programs carried out in BR2 are related to the safety of nuclear reactor structural materials and fuels, in normal and accidental conditions, plant lifetime evaluation and ageing of components. In this framework, a computer program that allows the performance of detailed, steady state analysis of several kinds of in-pile sections with an axisymmetrical geometry has been developed. Furthermore, comparing its results with those of the well known, extensively used, Relap5/Mod 3.2 code on a test problem has validated this program. This was performed in three steps: 1. modalisation development of a subsystem of a typical in-pile section. 2. steady state analysis and comparison with the above-mentioned program. 3. transient simulation of the same system; the considered transient consists of a loss of coolant flow. (author)

  20. A Small-Animal Irradiation Facility for Neutron Capture Therapy Research at the RA-3 Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Emiliano Pozzi; David W. Nigg; Marcelo Miller; Silvia I. Thorp; Amanda E. Schwint; Elisa M. Heber; Veronica A. Trivillin; Leandro Zarza; Guillermo Estryk

    2007-11-01

    The National Atomic Energy Commission of Argentina (CNEA) has constructed a thermal neutron source for use in Boron Neutron Capture Therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The Idaho National Laboratory (INL) and CNEA have jointly conducted some initial neutronic characterization measurements for one particular configuration of this source. The RA-3 reactor (Figure 1) is an open pool type reactor, with 20% enriched uranium plate-type fuel and light water coolant. A graphite thermal column is situated on one side of the reactor as shown. A tunnel penetrating the graphite structure enables the insertion of samples while the reactor is in normal operation. Samples up to 14 cm height and 15 cm width are accommodated.

  1. Accident consequence calculations and risk assessments for pressurized light water reactors with the computer code UFOMOD/B3

    International Nuclear Information System (INIS)

    With respect to the application of the accident consequence model of the German Risk Study (GRS) for light water reactors to risk assessments of other reactor types (high temperature reactor HTR-1160, fast breeder reactor SNR-300), the improved version UFOMOD/B3 was developed. The modifications mainly concern the deposition parameters, the resuspension process, the ingestion model and the dose factors. To make results comparable, recalculations for pressurized light water reactors were performed with the release categories of the GRS. The results show in contrast to the findings of the GRS a significant reduction of the acute fatality risk by a factor of 3.6. This essentially results from the smaller deposition parameters. The latent fatality risk was calculated nearly unchanged. (orig.)

  2. A Small-Animal Irradiation Facility for Neutron Capture Therapy Research at the RA-3 Research Reactor

    International Nuclear Information System (INIS)

    The National Atomic Energy Commission of Argentina (CNEA) has constructed a thermal neutron source for use in Boron Neutron Capture Therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The Idaho National Laboratory (INL) and CNEA have jointly conducted some initial neutronic characterization measurements for one particular configuration of this source. The RA-3 reactor (Figure 1) is an open pool type reactor, with 20% enriched uranium plate-type fuel and light water coolant. A graphite thermal column is situated on one side of the reactor as shown. A tunnel penetrating the graphite structure enables the insertion of samples while the reactor is in normal operation. Samples up to 14 cm height and 15 cm width are accommodated

  3. 3-D space time kinetics of compact high temperature reactor with fuel temperature feedback

    International Nuclear Information System (INIS)

    The Compact High Temperature Reactor (CHTR) is being developed as technology demonstrator for Indian High Temperature Reactor programme. Physics design of conceptual core of (Th-233U) fuelled CHTR is in advance stage and various core configurations have been proposed. Reactor core operation at high temperature necessitates sophisticated safety and anticipated transients analyses including postulated LORA, LOCA, and power set-back transients in CHTR. Recently, efficient IQS module in ARCH with adiabatic fuel temperature feedback capability has been developed. For accounting fuel and coolant temperature feedbacks in the simulation of 3D space time transients in CHTR, module for 1D (radial) heat conduction based module for heat transfer from fuel to coolant has been incorporated in 3D space-time analysis code ARCH. The AER benchmarking results of ARCH-IQS code with Doppler feedback and results of anticipated transient without scram (ATWS) of (Th-233U) fuelled CHTR with the present capability in ARCH-IQS code have been presented in this paper. (author)

  4. Simulation of nuclear reactor shielding experiment using the Sn code dot 3.5

    International Nuclear Information System (INIS)

    A large scale test facility representing the shielding arrangement of a fast breeder reactor has been simulated on the computer. The transport discrete discrete ordinate dot 3.5 has been used for this purpose. The shielding arrangement is a typical three dimensional geometry. A special strategy is developed to enable the simulation using two dimensional model. The neutron and gamma ray dose rates along the cavity, representing a reactor coolant channel, are compared with published experimental results and theoretical calculations using the Sn code twotran II. The calculation scheme adopted in the present work achieves better agreement with experiment than previous calculation- using the same code-with less computer time. 6 fig

  5. Revision of fast reactor group constant set JFS-3-J2

    International Nuclear Information System (INIS)

    To improve the fast reactor group constant set JFS-3-J2 to be applicable for high burnup reactor calculations, group constants for 155 fission product nuclides and the lumped group cross sections for four mother fission isotopes of U-235, U-238, Pu-239 and Pu-241 have been generated. Furthermore, the group constants for higher actinides such as Am and Cm have been produced on the basis of the JENDL-2 nuclear data, so as to be able to use for TRU-transmutation calculations. Benchmark test of this revised set has been performed by analysing the 21 fast critical experimental assemblies. Benchmark calculation system based on one-dimensional Sn-method has been developed to investigate the accuracy of one-dimensional diffusion calculations. Significant difference between the results obtained with the diffusion and transport calculations was observed for small cores and the assemblies with iron or nickel reflector. (author)

  6. Thermal hydraulic analysis for the Oregon State TRIGA reactor using RELAP5-3D

    International Nuclear Information System (INIS)

    Thermal hydraulic analyses have being conducted at Oregon State University (OSU) in support of the conversion of the OSU TRIGA reactor (OSTR) core from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel as part of the Reduced Enrichment for Research and Test Reactors program. The goals of the thermal hydraulic analyses were to calculate natural circulation flow rates, coolant temperatures and fuel temperatures as a function of core power for both the HEU and LEU cores; calculate peak values of fuel temperature, cladding temperature, surface heat flux as well as departure from nuclear boiling ratio (DNBR) for steady state and pulse operation; and perform accident analyses for the accident scenarios identified in the OSTR safety analysis report. RELAP5-3D Version 2.4.2 was implemented to develop a model for the thermal hydraulic study. The OSTR core conversion is planned to take place in late 2008. (author)

  7. Modifications of the PRONTO 3D finite element program tailored to fast burst nuclear reactor design

    International Nuclear Information System (INIS)

    This update discusses modifications of PRONTO 3D tailored to the design of fast burst nuclear reactors. A thermoelastic constitutive model and spatially variant thermal history load were added for this special application. Included are descriptions of the thermoelastic constitutive model and the thermal loading algorithm, two example problems used to benchmark the new capability, a user's guide, and PRONTO 3D input files for the example problems. The results from PRONTO 3D thermoelastic finite element analysis are benchmarked against measured data and finite difference calculations. PRONTO 3D is a three-dimensional transient solid dynamics code for analyzing large deformations of highly non-linear materials subjected to high strain rates. The code modifications are implemented in PRONTO 3D Version 5.3.3. 12 refs., 30 figs., 9 tabs

  8. SCALE-4 analysis of pressurized water reactor critical configurations. Volume 2: Sequoyah Unit 2 Cycle 3

    International Nuclear Information System (INIS)

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations reported herein is based on the codes and data provided in the SCALE-4 code system. This volume of the report documents the SCALE system analysis of three reactor critical configurations for the Sequoyah Unit 2 Cycle 3. This unit and cycle were chosen because of the relevance in spent fuel benchmark applications: (1) the unit had a significantly long downtime of 2.7 years during the middle of cycle (MOC) 3, and (2) the core consisted entirely of burned fuel at the MOC restart. The first benchmark critical calculation was the MOC restart at hot, full-power (HFP) critical conditions. The other two benchmark critical calculations were the beginning-of-cycle (BOC) startup at both hot, zero-power (HZP) and HFP critical conditions. These latter calculations were used to check for consistency in the calculated results for different burnups and downtimes. The keff results were in the range of 1.00014 to 1.00259 with a standard deviation of less than 0.001

  9. Neutron transport with the method of characteristics for 3-D full core boiling water reactor applications

    Science.gov (United States)

    Thomas, Justin W.

    2006-12-01

    The Numerical Nuclear Reactor (NNR) is a code suite that is being developed to provide high-fidelity multi-physics capability for the analysis of light water nuclear reactors. The focus of the work here is to extend the capability of the NNR by incorporation of the neutronics module, DeCART, for Boiling Water Reactor (BWR) applications. The DeCART code has been coupled to the NNR fluid mechanics and heat transfer module STAR-CD for light water reactor applications. The coupling has been accomplished via an interface program, which is responsible for mapping the STAR-CD and DeCART meshes, managing communication, and monitoring convergence. DeCART obtains the solution of the 3-D Boltzmann transport equation by performing a series of 2-D modular ray tracing-based method of characteristics problems that are coupled within the framework of 3-D coarse-mesh finite difference. The relatively complex geometry and increased axial heterogeneity found in BWRs are beyond the modeling capability of the original version of DeCART. In this work, DeCART is extended in three primary areas. First, the geometric capability is generalized by extending the modular ray tracing scheme and permitting an unstructured mesh in the global finite difference kernel. Second, numerical instabilities, which arose as a result of the severe axial heterogeneity found in BWR cores, have been resolved. Third, an advanced nodal method has been implemented to improve the accuracy of the axial flux distribution. In this semi-analytic nodal method, the analytic solution to the transverse-integrated neutron diffusion equation is obtained, where the nonhomogeneous neutron source was first approximated by a quartic polynomial. The successful completion of these three tasks has allowed the application of the coupled DeCART/STAR-CD code to practical BWR problems.

  10. Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program

    International Nuclear Information System (INIS)

    The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

  11. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR). Version 3.5, Quick Reference Guide

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert, B.G.; Richards, R.E.; Reece, W.J.; Gertman, D.I.

    1992-10-01

    This Reference Guide contains instructions on how to install and use Version 3.5 of the NRC-sponsored Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR). The NUCLARR data management system is contained in compressed files on the floppy diskettes that accompany this Reference Guide. NUCLARR is comprised of hardware component failure data (HCFD) and human error probability (HEP) data, both of which are available via a user-friendly, menu driven retrieval system. The data may be saved to a file in a format compatible with IRRAS 3.0 and commercially available statistical packages, or used to formulate log-plots and reports of data retrieval and aggregation findings.

  12. Moderators for the design of a cold neutron source for the RA 3 reactor

    International Nuclear Information System (INIS)

    The cold neutron production of hydrogenous materials was studied, taking into account their radiation resistance, for the conceptual design of a cold neutron source for the RA-3 reactor.Low spontaneous release of chemical energy was found in mesitylene.Libraries for hidrogen in mesitylene were generated using the NJOY nuclear processing system and the resulting cross sections were compared with experimental data.Good agreement between measurements and calculations was found in those cases where data are available.New calculations using the RA-3 geometry and these validated libraries will be performed

  13. Neutron fluence determination at reactor filters by 3He proportional counters: Comparison of unfolding algorithms

    International Nuclear Information System (INIS)

    Multichannel pulse height measurements with a cylindrical 3He proportional counter obtained at a reactor filter of natural iron are taken to investigate the properties of three algorithms for neutron spectrum unfolding. For a systematic application of uncertainty propagation the covariance matrix of previously determined 3He response functions is evaluated. The calculated filter transmission function together with a covariance matrix estimated from cross-section uncertainties of the filter material is used as fluence pre-information. The results obtained from algorithms with and without pre-information differ in shape and uncertainties for single group fluence values, but there is sufficient agreement when evaluating integrals over neutron energy intervals

  14. Neutrino-4 experiment on the search for a sterile neutrino at the SM-3 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Serebrov, A. P., E-mail: serebrov@pnpi.spb.ru; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Chernyi, A. V.; Zherebtsov, O. M. [National Research Centre “Kurchatov Institute,”, Konstantinov Petersburg Nuclear Physics Institute (Russian Federation); Martemyanov, V. P.; Tsinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I. [National Research Centre “Kurchatov Institute,” (Russian Federation); Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K. [State Scientific Centre Research Institute of Atomic Reactors (Russian Federation); and others

    2015-10-15

    In view of the possibility of the existence of a sterile neutrino, test measurements of the dependence of the reactor antineutrino flux on the distance from the reactor core has been performed on SM-2 reactor with the Neutrino-2 detector model in the range of 6–11 m. Prospects of the search for reactor antineutrinos at short distances have been discussed.

  15. Power Distribution Analysis for the ORNL High Flux Isotope Reactor Critical Experiment 3

    International Nuclear Information System (INIS)

    The mission of the Reduced Enrichment for Research and Test Reactors Program is to minimize and, to the extent possible, eliminate the use of highly enriched uranium (HEU) in civilian nuclear applications by working to convert research and test reactors, as well as radioisotope production processes, to low-enriched uranium (LEU) fuel and targets. Oak Ridge National Laboratory (ORNL) is currently reviewing the design bases and key operating criteria including fuel operating parameters, enrichment-related safety analyses, fuel performance, and fuel fabrication in regard to converting the fuel of the High Flux Isotope Reactor (HFIR) from HEU to LEU. The purpose of this study is to validate Monte Carlo methods currently in use for conversion analyses. The methods have been validated for the prediction offlux values in the reactor target, reflector, and beam tubes, but this study focuses on the prediction of the power density profile in the core. Power distributions were calculated in the fuel elements of the HFIR, a research reactor at ORNL, via MCNP and were compared to experimentally obtained data. This study was performed to validate Monte Carlo methods for power density calculations and to observe biases. A current three-dimensional MCNP model was modified to replicate the 1965 HFIR Critical Experiment 3 (HFIRCE-3). In this experiment, the power profile was determined by counting the gamma activity at selected locations in the core. 'Foils' (chunks of fuel meat and clad) were punched out of the fuel elements in HFIRCE-3 following irradiation, and experimental relative power densities were obtained by measuring the activity of these foils and comparing each foil's activity to the activity of a normalizing foil. This analysis consisted of calculating corresponding activities by inserting volume tallies into the modified MCNP model to represent the punchings. The average fission density was calculated for each foil location and then normalized to the reference foil

  16. Pedometer-Determined Physical Activity and Its Comparison with the International Physical Activity Questionnaire in a Sample of Belgian Adults

    Science.gov (United States)

    De Cocker, Katrien; Cardon, Greet; De Bourdeaudhuij, Ilse

    2007-01-01

    Pedometer-determined physical activity (PA) levels in Belgian adults were provided and compared to PA scores reported in the International Physical Activity Questionnaire (IPAQ). The representative sample (N = 1,239) of the Belgian population took on average 9,655 (4,526) steps/day. According to pedometer indices 58.4% were insufficiently active.…

  17. Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira, E-mail: ecfoliveira@hotmail.com, E-mail: lazara.castrillo@upe.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica de Pernambuco

    2015-07-01

    Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

  18. Semi-catalyzed deuterium reactors for co-generation of 3He and synfuels (the CoSCD concept)

    International Nuclear Information System (INIS)

    The potential of developing semi-catalyzed deuterium reactors for co-generation of 3He and synthetic fuels is discussed. Such factors as environmental impact, siting, energy basics, and engineering technology are also discussed

  19. CASMO-4/SIMULATE-3/MCNPX analysis of a reactor pressure vessel scraping test

    International Nuclear Information System (INIS)

    To respond to current-day increased requirements on the quality of neutron fluence estimations, an up-to-date methodology for 3D fast fluence evaluation for light water reactors is under development at the Paul Scherrer Inst., centered around the MCNP(X) code using continuous-energy neutron data libraries. An essential part of the verification and validation of the entire calculation route covering the core-follow and neutron transport simulation tools - CASMO/SIMULATE/ MCNPX - is the analysis of available experimental data on the fast neutron fluence at the inside of the reactor pressure vessel. The fluence estimations are based on actual reactor operation history data. The reference measurement-based fluence estimates were obtained from evaluations of pressure vessel scraping data that have been performed previously at the Paul Scherrer Inst.. The paper describes the general features of the developed and applied methodology and particular modeling options which were selected through accompanying comprehensive sensitivity and optimization studies. Also presented are the results of the calculations and their comparison with measurements. Very good agreement is found between the calculated results and the evaluated experimental data, indicating a high level of accuracy in the fixed-source, ex-core neutron transport calculations. (authors)

  20. Thermal Hydraulics Analysis for the 3MW TRIGA MARK-II Research Reactor Under Transient Condition

    International Nuclear Information System (INIS)

    Some important thermal hydraulic parameters of the 3 MW TRIGA MARK-II research reactor operating under transient condition were investigated using two computer codes PULTRI and TEMPUL. Major transient parameters, such as, peak power and prompt energy released after pulse, maximum fuel and coolant temperature, surface heat flux, time and radial distribution of temperature within fuel element after pulse, fuel, fuel-cladding gap width variation, etc. were computer and compared with the experimental and operational values as reported in the safety Analysis Report (SAR). It was observed that pulsing of the reactor inserting an excess reactivity of $2.00 shoots the reactor power level to 854.353 MW compared to an experimental value of 852 MW; the maximum fuel temperature corresponding to this peak power was found to be 846.76o C which is much less than the limiting maximum value of fuel temperature of 11500 C as reported in SAR. During a pulse if the film boiling occurs for a peak adiabatic fuel temperature of 1000o C, the calculated outer cladding wall temperature was observed to be 702.390 C compared to a value of 760o C reported in SAR under the same condition. The investigated other results were also found to be in good agreement with the values reported in the SAR. 16 refs., 22 figs. (author)

  1. Feasibility Study of Using PAGAT Polymer Gel Dosimeter for 3D Dosimetry Around the Reactor Core

    International Nuclear Information System (INIS)

    An important problem for samples irradiation in research reactors is determination of three dimensional dose distributions in the vicinity of reactor core. Polymer gel dosimeters can be used to measure complex three dimensional dose distributions as well as the integrated dose accurately with no dependency on the dose rate. Furthermore, as they are tissue-equivalent, they may be used as a phantom. So far, polymer gel dosimeters have been used for photon, electron, proton, neutron and heavy ions, but there is a lack of application of polymer gel dosimeters for dosimetry of the mixed field of radiation of different linear ionization concentration. In this research, PAGAT polymer gel dosimeters are fabricated in the laboratory and then were irradiated with the mixed neutron gamma field from the fission process of the Tehran Research Reactor. The gel response was determined by the nuclear magnetic resonance imaging technique as a change in the relaxation rate (R2) of the gel dosimeters. The gel response as a function of normalized dose was investigated and a bi-exponential fitting was adjusted to the dose-R2 data. The region with a linear response, is called dynamic range. The slope of the region as the sensitivity of PAGAT gels to the normalized dose resulted from the neutron-gamma mixed field, was estimated to be 1.695 s-1. The results of this research showed that PAGAT polymer gel dosimeter is a useful tool for 3D dose distribution to determine the neutron gamma mixed field.

  2. Compliance of Companies with Corporate Governance Codes: Case Study on Listed Belgian

    Directory of Open Access Journals (Sweden)

    Sven H. De Cleyn

    2014-09-01

    Full Text Available Listed and large companies become increasingly subject to internal and external pressure to comply with ethical and social standards. This article focuses on one aspect of this matter, namely the corporate governance issue. Within the framework of recent corporate scandals, this paper investigates whether and to which extent Belgian publicly listed SMEs comply with the Belgian Code on Corporate Governance after its first year of introduction, which has been constituted in the framework of the European Action Plan on Corporate Governance.In a sample of 78 Belgian listed SMEs, the compliance with the Code is analysed. After its first year of introduction, companies comply with on average 70% of the Code’s provisions. The most problematic topics in terms of disclosure of information seem to relate to (individual remuneration, private information and content of shareholders’ meetings.

  3. PCBs and OCPs in marine species from the Belgian North Sea and the Western Scheldt Estuary

    Energy Technology Data Exchange (ETDEWEB)

    Voorspoels, S.; Covaci, A.; Maervoet, J.; Schepens, P. [Antwerp Univ., Wilrijk (Belgium). Toxicological Centre

    2004-09-15

    The use and/or production of polychlorinated biphenyls (PCBs) and organochlorine pesticides (OCPs), such as 2,2-bis-(4-chlorophenyl)-1,1,1-trichloroethane (DDT), hexachlorobenzene (HCB) and lindane ({gamma}-HCH) have been banned in most developed countries since the 1970's. Despite this measure, these compounds are among the most prevalent environmental pollutants and they can be found in various environmental compartments, both biotic and abiotic. Their widespread presence is due to their extremely persistant and lipophilic nature, resulting in enrichment throughout the food chain. Prolonged exposure to these pollutants can interfere with normal physiology and biochemistry3, resulting in adverse effects in various organisms, including starfish, shrimp, crabs, and fish4. Because humans readily consume seafood, such as shrimp, crab and various fish species, these organisms are of great scientific value to estimate the possible exposure to PCBs and OCPs through marine food sources. The area studied in this investigation covered both commercial fishing grounds (Belgian North Sea - BNS) and a recreational fishing area (Western Scheldt Estuary - SE). The drainage basin of the SE covers a very densely populated and highly industrialised region, causing a high level of pollution in the SE. In this work, PCBs and OCPs were determined in benthic invertebrates and different fish species from both BNS and SE in order to evaluate trends in levels, congener distribution, and geographical variation.

  4. Radiation dose to premature new-borns in the belgian neonatal intensive care units

    International Nuclear Information System (INIS)

    In the neonatal intensive care units (NICU), premature new-borns may be exposed to important doses. Because of their increased radiosensitivity and longer life expectancy, dose optimisation is of importance. The present study aimed at evaluating the dose of the most common radiographs in the Belgian NICU. Entrance surface kerma (ESK) and kerma area product (KAP) were collected in 17 NICU (among 19 in Belgium). Median ESK ranged from 13 to 172 μGy and from 8 to 117 μGy for chest and combined chest-abdomen radiographs, respectively; median KAP ranged from 1.4 to 14.2 mGy cm2 and from 3.8 to 28.1 mGy cm2 for chest and combined chest-abdomen radiographs, respectively. Those differences were due to large variations in the examination settings. Diagnostic reference levels (DRL) were set for chest and combined chest-abdomen radiographs. Though the radiograph dose was usually low, the cumulative dose per stay could be high. The wide, intercentre differences indicate that there is scope for dose reduction. The use of DRL should contribute to achieve this object. (authors)

  5. Floating seaweed in the neustonic environment: A case study from Belgian coastal waters

    Science.gov (United States)

    Vandendriessche, Sofie; Vincx, Magda; Degraer, Steven

    2006-02-01

    Floating seaweeds form the most important natural component of all floating material found on the surface of oceans and seas. Notwithstanding the absence of natural rocky shores, ephemeral floating seaweed clumps are frequently encountered along the Belgian coast. From October 2002 to April 2003, seaweed samples and control samples (i.e. surface water samples from a seaweed-free area) were collected every other week. Multivariate analysis on neustonic macrofaunal abundances showed significant differences between seaweed and control samples in the fraction > 1 mm. Differences were less conspicuous in the 0.5-1 mm fraction. Seaweed samples were characterised by the presence of seaweed fauna e.g. Acari, Idotea baltica, Gammarus sp ., while control samples mainly contained Calanoida, Larvacea, Chaetognatha, and planktonic larvae of crustaceans and polychaetes. Seaweed samples (1 mm fraction) harboured considerably higher diversities (× 3), densities (× 18) and biomasses (× 49) compared to the surrounding water column (control samples). The impact of floating seaweeds on the neustonic environment was quantified by the calculation of the added values of seaweed samples considering biomass and density. These calculations resulted in mean added values of 311 ind m - 2 in density and 305 mg ADW m - 2 in biomass. The association degree per species was expressed as the mean percentage of individuals found in seaweed samples in proportion to the total density and biomass of that species (seaweed samples + control samples). Thirteen species showed an association percentage > 95%, and can therefore be considered members of the floating seaweed fauna.

  6. Energy conversion options for ARIES-III - A conceptual D-3He tokamak reactor

    International Nuclear Information System (INIS)

    The potential for highly efficient conversion of fusion power to electricity provides one motivation for investigating D-3He fusion reactors. This stems from: (1) the large fraction of D-3He power produced in the forms of charged particles and synchrotron radiation which are amenable to direct conversion, and (2) the low neutron fluence and lack of tritium breeding constraints, which increase design flexibility. The design team for a conceptual D-3He tokamak reactor, ARIES-III, has investigated numerous energy conversion options at a scoping level in attempting to realize high efficiency. The energy conversion systems have been studied in the context of their use on one or more of three versions of a D-3He tokamak: a first stability regime device, a second stability regime device, and a spherical torus. The set of energy conversion options investigated includes bootstrap current conversion, compression-expansion cycles, direct electrodynamic conversion, electrostatic direct conversion, internal electric generator, liquid metal heat engine blanket, liquid metal MHD, plasma MHD, radiation boiler, scrape-off layer thermoelectric, synchrotron radiation conversion by rectennas, synchrotron radiation conversion by thermal cycles, thermionic/AMTEC/thermal systems, and traveling wave conversion. The original set of options is briefly discussed, and those selected for further study are described in more detail. The four selected are liquid metal MHD, plasma MHD, rectenna conversion, and direct electrodynamic conversion. Thermionic energy conversion is being considered, and some options may require a thermal cycle in parallel or series. 17 refs., 3 figs., 1 tab

  7. Application of the 3-D Nodal Equivalence Theory to the CANDU Reactor

    International Nuclear Information System (INIS)

    The RFSP code is found to be subject to inconsistency issue mainly due to the lack of nodal equivalence. In Ref. 2, it has been shown that nodal equivalence theory can be effective for the 2-D CANDU core analysis. In this work, the 3-D nodal equivalence theory was applied to see its effectiveness in a 3-dimensional CANDU reactor analysis. The 3-D nodal equivalence is applied to the whole core analysis of a clean CANDU6 core. Both the radial and the axial DFs are quite different for different reactivity devices inside the fuel lattice. It has been demonstrated that the application of the conventional 2-D nodal equivalence theory gives better accuracy in terms of the k-eff and power profiles, while the 3-D equivalence theory only results in marginal improvements. The relative ineffectiveness of the axial discontinuity factor may be ascribed to simplifications of the very complicated core geometry and some assumptions in modeling both radial and axial reflectors of the CANDU reactor. For a more accurate evaluation of the 3-D equivalence theory, more realistic reflector models are currently under development

  8. Combined phenol and acetate degradation by O3/UV in two different reactor configurations

    OpenAIRE

    Van de Moortel, Wim; Van Eyck, Kwinten; Liers, Sven; Degrève, Jan; Luyten, Jan

    2010-01-01

    Aromatic compounds, such as phenol, are widely used in various industries and can cause serial environmental damage when present in natural waters. As several intermediates are formed during the mineralisation pathway of phenol, the reactor design may influence the product distribution. This study aims to examine the sensitivity of the phenol degradation products upon varying the reactor configuration. The two reactor configurations examined are a 10 liter batch reactor and a 50 liter reactor...

  9. Remodeling and Characterization of Pneumatic Transfer System (PTS no.3) of HANARO Research Reactor for Neutron Activation Analysis

    International Nuclear Information System (INIS)

    A pneumatic transfer system (PTS) is one of the facilities to be used in an irradiation of a target material for an instrumental neutron activation analysis (INAA) in a research reactor. This technical report describes for the irradiation test and the characteristics and removing, remodelling of the reinstalled PTS no.3 in NAA no.3 irradiation hole at HANARO for user information and the reactor management

  10. Spectrum and density of neutron flux in the irradiation beam line no. 3 of the IBR-2 reactor

    Science.gov (United States)

    Shabalin, E. P.; Verkhoglyadov, A. E.; Bulavin, M. V.; Rogov, A. D.; Kulagin, E. N.; Kulikov, S. A.

    2015-03-01

    Methodology and results of measuring the differential density of the neutron flux in irradiation beam line no. 3 of the IBR-2 reactor using neutron activation analysis (NAA) are presented in the paper. The results are compared to the calculation performed on the basis of the 3D MCNP model. The data that are obtained are required to determine the integrated radiation dose of the studied samples at various distances from the reactor.

  11. High resolution mapping of the tropospheric NO2 distribution in three Belgian cities based on airborne APEX remote sensing

    Science.gov (United States)

    Tack, Frederik; Merlaud, Alexis; Fayt, Caroline; Danckaert, Thomas; Iordache, Daniel; Meuleman, Koen; Deutsch, Felix; Adriaenssens, Sandy; Fierens, Frans; Van Roozendael, Michel

    2015-04-01

    An approach is presented to retrieve tropospheric nitrogen dioxide (NO2) vertical column densities (VCDs) and to map the NO2 two dimensional distribution at high resolution, based on Airborne Prism EXperiment (APEX) observations. APEX, developed by a Swiss-Belgian consortium on behalf of ESA (European Space Agency), is a pushbroom hyperspectral imager with a high spatial (approximately 3 m at 5000 m ASL), spectral (413 to 2421 nm in 533 narrow, contiguous spectral bands) and radiometric (14-bit) resolution. VCDs are derived, following a similar approach as described in the pioneering work of Popp et al. (2012), based on (1) spectral calibration and spatial binning of the observed radiance spectra in order to improve the spectral resolution and signal-to-noise ratio, (2) Differential Optical Absorption Spectroscopy (DOAS) analysis of the pre-processed spectra in the visible wavelength region, with a reference spectrum containing low NO2 absorption, in order to quantify the abundance of NO2 along the light path, based on its molecular absorption structures and (3) radiative transfer modeling for air mass factor calculation in order to convert slant to vertical columns. This study will be done in the framework of the BUMBA (Belgian Urban NO2 Monitoring Based on APEX hyperspectral data) project. Dedicated flights with APEX mounted in a Dornier DO-228 airplane, operated by Deutsches Zentrum für Luft- und Raumfahrt (DLR), are planned to be performed in Spring 2015 above the three largest and most heavily polluted Belgian cities, i.e. Brussels, Antwerp and Liège. The retrieved VCDs will be validated by comparison with correlative ground-based and car-based DOAS observations. Main objectives are (1) to assess the operational capabilities of APEX to map the NO2 field over an urban area at high spatial and spectral resolution in a relatively short time and cost-effective way, and to characterise all aspects of the retrieval approach; (2) to use the APEX NO2 measurements

  12. RA Reactor operation and maintenance (I-IX), part VIII, Task 3.08/05, Decontamination of the reactor

    International Nuclear Information System (INIS)

    Permanent increase of radiation in the heavy water system was noticed during first three year of the RA reactor operation, even when the reactor was shutdown. It was found that there was no failure of the fuel element cladding. Radioactive cobalt was found in the heavy water which was rather strange. During repair of the heavy water system, it has been found that stellite was used for coating the heavy water pumps. Since stellite is a cobalt alloy, this could have been the source of radioactive cobalt in the heavy water. The stellite coating was damaged due to friction and particle of cobalt appeared in the coolant, they were activated since they were in the core. decontamination of the heavy water and the heavy water coolant loop was a must . Beside the detailed report on the contamination and decontamination of the heavy water system this volume includes 14 annexes describing the investigation of the event and the whole procedure of decontamination

  13. Non-destructive method for internal quality determination of belgian endive (cichorium intybus l.)

    OpenAIRE

    De Baerdemaeker J.; Quenon V.

    2000-01-01

    A method and process were developed to nondestructively measure the length of the floral stalk in Belgian endive Cichorium intybus L. Current X-ray technology proved to be a feasible method. A detection algorithm was developed based on the minimal transmitted intensities along the length. The method is very accurate with an absolute precision of 4.9 mm and allows the study of the influence of storage conditions and time on the Belgian endive internal quality. The growth of the floral stalk is...

  14. Core and fuel feasibility study for improved flexibility on the Belgian Nuclear Power Plants

    International Nuclear Information System (INIS)

    A feasibility study has been performed for extended power modulations on Belgian NPPs. The goal is to make the existing nuclear power units in Belgium more flexible without implementing hardware modifications and guaranteeing safety at all times. As the critical part of the feasibility study, the impacts on the core behaviour and fuel performance have been studied in detail. It is concluded that all existing fuels loaded in the Belgian plants allow up to 30 power modulations per fuel cycle without changing the currently applied fuel cycle management. This is also supported by the extensive experience feedback of the fuel products for flexible operations in European countries. (author)

  15. Sand dynamics along the Belgian coast based on airborne hyperspectral data and lidar data

    OpenAIRE

    Deronde, B; Houthuys, R.; Sterckx, S.; Fransaer, D.

    2005-01-01

    The goal of this project was to explore the possibilities of airborne hyperspectral data and airborne lidar data to study sand dynamics on the Belgian backshore and foreshore. The Belgian coast is formed by a sandy strip at the southern edge of the North Sea Basin which is commonly known as the Southern Bight. Since the beach is prone to structural and occasional erosion, it is very important to obtain a better understanding of the processes controlling it. The combination of multi-temporal h...

  16. A faunistic survey of the Belgian marine molluscs: a first progress report

    OpenAIRE

    Backeljau, T.

    1988-01-01

    A first step towards the compilation of a critical faunistic inventory of the Belgian marine mollusks, was the publication of a preliminary nomenclatural list in 1986. This list mentioned 136 species which until then bad been recorded alive in Belgium. Yet, such records do not necessarily mean that the species involved actually belong to our fauna. Hence, the next thing to do, was to start a more profound survey of the Belgian marine malacofauna in order to determine which species form well-e...

  17. Study of the U3O8-Al thermite reaction and strength of reactor fuel tubes

    International Nuclear Information System (INIS)

    Research and test reactors are presently operated with aluminum-clad fuel elements containing highly enriched uranium-aluminum alloy cores. To lower the enrichment and still maintain reactivity, the uranium content of the fuel element will need to be higher than currently achievable with alloy fuels. This will necessitate conversion to other forms such as U3O8-aluminum cermets. Above the aluminum melting point, U3O8 and aluminum undergo an exothermic thermite reaction and cermet fuel cores tend to keep their original shape. Both factors could affect the course and consequences of a reactor accident, and prompted an investigation of the behavior of cermet fuels at elevated temperatures. Tests were carried out using pellets and extruded tube-sections with 53 wt % U3O8 in aluminum. This content corresponds to a theoretical uranium density of 1.9 g/cc. Results indicate that the thermite reaction occurs at about 9000C in air without a violent effect. The heat of reaction was approximately 123 cal/g of U3O8-aluminum fuel. Tensile and compressive strength of the fuel tube section is low above 6600C. In tension, sections failed at about the aluminum melting point. In compression with 2-psi average axial stress, failure occurred at 9170C, while 7 psi average axial stress produced failure at 6690C

  18. Handbook of nuclear engineering: vol 1: nuclear engineering fundamentals; vol 2: reactor design; vol 3: reactor analysis; vol 4: reactors of waste disposal and safeguards

    CERN Document Server

    2013-01-01

    The Handbook of Nuclear Engineering is an authoritative compilation of information regarding methods and data used in all phases of nuclear engineering. Addressing nuclear engineers and scientists at all academic levels, this five volume set provides the latest findings in nuclear data and experimental techniques, reactor physics, kinetics, dynamics and control. Readers will also find a detailed description of data assimilation, model validation and calibration, sensitivity and uncertainty analysis, fuel management and cycles, nuclear reactor types and radiation shielding. A discussion of radioactive waste disposal, safeguards and non-proliferation, and fuel processing with partitioning and transmutation is also included. As nuclear technology becomes an important resource of non-polluting sustainable energy in the future, The Handbook of Nuclear Engineering is an excellent reference for practicing engineers, researchers and professionals.

  19. A nodal expansion method for solving the multigroup SP3 equations in the reactor code DYN3D

    International Nuclear Information System (INIS)

    The core model DYN3D which has been developed for three-dimensional analyses of steady states and transients in thermal reactors with quadratic or hexagonal fuel assemblies is based on nodal methods for the solution of the two-group neutron diffusion equation. Loading cores with higher content of MOX fuel, the increase of the fuel cycle length and new types of reactors are challenging for these standard methods. A nodal expansion method for solving the equations of the simplified P3 approximation (SP3) of the multigroup transport equation was developed to improve the accuracy of the DYN3D code. In this paper, the method used in DYN3D-SP3 is described. It is applied for the pin-wise calculation of a steady state of the OECD/NEA and U.S. NRC PWR MOX/UO2 Core Transient Benchmark. The eigenvalue keff, assembly powers and the pin powers are computed. The results calculated with different approaches including diffusion theory are compared with the reference solution obtained from a heterogeneous transport calculation with the code DeCART. Different approaches of the diffusion coefficient used in the SP3 equations are investigated. The SP3 results obtained with the transport cross section of multigroup diffusion theory show the smallest deviations from the reference solution. These deviations are in the same order as the results of the code DORT, whereas the DORT and DYN3D calculations were carried out with the same library of group constants for homogenized pin cells. (authors)

  20. Unstructured 3D core calculations with the descartes system application to the JHR research reactor

    International Nuclear Information System (INIS)

    Recent developments in the DESCARTES system enable neutronics calculations dealing with very complex unstructured geometrical configurations. The discretization can be made either by using a very fine Cartesian mesh and the fast simplified transport (SPN) solver MINOS, or a discretization based on triangles and the SP1 solver MINARET. In order to perform parallel calculations dealing with a very fine mesh in 3D, a domain decomposition with non overlapping domains has been implemented. To illustrate these capabilities, we present an application on the future European research reactor JHR dedicated to technological irradiations. (authors)

  1. Simple computational modeling for human extracorporeal irradiation using the BNCT facility of the RA-3 Reactor

    International Nuclear Information System (INIS)

    We present a simple computational model of the reactor RA-3 developed using Monte Carlo transport code MCNP. The model parameters are adjusted in order to reproduce experimental measured points in air and the source validation is performed in an acrylic phantom. Performance analysis is carried out using computational models of animal extracorporeal irradiation in liver and lung. Analysis is also performed inside a neutron shielded receptacle use for the irradiation of rats with a model of hepatic metastases.The computational model reproduces the experimental behavior in all the analyzed cases with a maximum difference of 10 percent. (author)

  2. New conception of the radioisotope producer reactor core using U3 Si2-Al fuel

    International Nuclear Information System (INIS)

    In this work the neutronic studies developed for a new RPR reactor core are presented, utilizing a U3 Si2-Al fuel with high density in Uranium. It is first shown the results of a optimization study in order to verify the neutronic parameters in function of the fuel element geometry. Next some core configurations are tested through the calculations system and refueling scheme until the equilibrium fuel cycle is obtained. Lastly, it is presented a configuration most adequate to the objectives, restrictions and safety criteria. (author)

  3. Fabrication of zero power reactor fuel elements containing 233U3O8 powder

    International Nuclear Information System (INIS)

    Oak Ridge National Laboratory, under contract with Argonne National Laboratory, completed the fabrication of 1743 fuel elements for use in their Zero Power Reactor. The contract also included recovery of 20 kg of 233U from rejected elements. This report describes the steps associated with conversion of purified uranyl nitrate (as solution) to U3O8 powder (suitable for fuel) and subsequent charging, sealing, decontamination, and testing of the fuel elements (packets) preparatory to shipment. The nuclear safety, radiation exposures, and quality assurance aspects of the program are discussed

  4. Reaction balance and efficiency analysis of a D-D fusion/fission hybrid with satellite D-3He reactors

    International Nuclear Information System (INIS)

    Selected reactor physics and isotope balance characteristics of a fusion hybrid supported D-3He satellite nuclear energy system are formulated and investigated. The system consists of two types of reactors: a parent D-fueled fusion device and a number of smaller reactors optimized for D-3He fusion. The parent hybrid station breeds the helium-3 for the satellites and also breeds fissile fuel for an existing fission reactor economy. Various hybrid operational regimes are examined in order to determine favorable reactor Q values and effective fusion and fission efficiencies. A number of analytical correlations between power output, plasma energetics, blanket neutronics, breeding capacity, and energy conversion cycles are established and evaluated. Numerical examples of performance parameters such as fission-tofusion power, overall conversion efficiency, and the ratio of satellite to parent fusion power are presented. The range of reactor efficiencies is elucidated as affected by the internal plasma power balances. As an upper bound based on optimistic injection and direct conversion efficiencies, we find the D-3He satellite system power output attaining at best 1/3 of the parent fusion power

  5. Maintenance activities based on condition monitoring at research reactor 'JRR-3'

    International Nuclear Information System (INIS)

    The construction of the research reactor 'JRR-3' started for the purpose of development of nuclear technology and first critical was achieved in 1961. A large scale modification was done to meet more demand from researchers in 1990. So far about operation of JRR-3, there have been no big troubles and JRR-3 has been under stable operation for more than 15 years from the modification. But it is necessary to investigate a maintenance methods considering high aging. On the contrary, it is difficult to keep present maintenance because of the reduction of the budget and staff. In such situation, we should find an effective maintenance method which does not spoil safety and reliability. In this presentation, our investigation of the effective maintenance and actual maintenance activities on JRR-3 are shown. (author)

  6. Ignition analysis for D plasma with non-Maxwellian 3He minority in fusion reactors

    International Nuclear Information System (INIS)

    Possible fusion reactivity enhancement due to 3He minority ICRF heating in D-3He toroidal plasma is demonstrated in present numerical simulations. On this purpose the particle code based on test-particle approach is developed. This code solves guiding center equations for 3He ions in toroidal magnetic field including Coulomb collisions of these ions with the background deuterons and electrons. A simple Monte Carlo model for ICRF heating is implemented in this code as well. The transformation of 3He distribution function from Maxwellian to non-Maxwellian due to heating plays the key role for reactivity enhancement. The formation of significant energetic tail gives rise to the reactivity enhancement. This is an important issue for the performance of fusion reactors with minority heating of ICRF. (author)

  7. Status of spent fuel in the 3MW BAEC MK-II research reactor facility of Bangladesh

    International Nuclear Information System (INIS)

    Bangladesh has been operating a 3 MW TRIGA MARK II research reactor since 1986. The reactor is installed in the campus of the Atomic Energy Research Establishment (AERE) at Savar, which is located about 40 km northwest of Dhaka. It is one of the main nuclear research facilities in the country. The reactor uses TRIGA LEU fuel with uranium content of 20% by weight. The enrichment level of the fuel is 19.7%. So far the reactor has been operated for 5624 hours with a total cumulative burnup (BU) of 10 690 MWh (445 MWd). The main areas of use are: training of man-power for research reactor operation and applications, radioisotope (RI) production, neutron activation analysis (NAA), neutron radiography (NR) and neutron scattering. Radioisotopes produced to date are: I-131, Sc-46 and Tc-99m. Bangladesh is a peace loving country with a strong commitment towards nuclear nonproliferation. Accordingly, it has signed several multilateral and bilateral agreements, protocols, treaties, etc. prevailing in the International Nuclear Non-proliferation regime. Bangladesh has also signed a Nuclear Cooperation Agreement with the USA on 17 September 1981, which facilitated export of nuclear technology from the USA to Bangladesh. The research reactor was procured under the provisions of this agreement. In 2003, the tenure of the Agreement was extended up to 2012. At present, there does not exist any spent fuel element in the reactor facility. However, with the recently undertaken RI production enhancement program, it is expected that the reactor will start generating spent fuels from the year 2012. It is to be mentioned that Bangladesh is aware of the US DOE's 'Take Back Program' in connection with the research reactor spent fuel of US origin, and is very much interested to take part in this program. The paper presents the current status of handling and storage facilities available for spent fuel and strategy for the safe management spent fuel to be generated from the research reactor in

  8. Attitude of a group of Belgian stakeholders towards proposed agricultural countermeasures after a radioactive contamination: synthesis of the discussions within the Belgian EC-FARMING group

    International Nuclear Information System (INIS)

    In the case of radioactive contamination of the environment with an impact on the food chain, the remediation strategy will not only be based on scientific knowledge and technical experience, but will also be dictated by peculiarities of the country. These characteristics include the agro-industrial structure, the local and international economical contexts and the political configuration including the distribution of responsibilities and competencies. This paper identifies and illustrates the most relevant characteristics of the Belgian agricultural system and political environment; it also describes the past experience with food chain contamination, which is expected to influence the attitude of Belgian stakeholders, who would be involved in the setting up of countermeasure strategies for maintaining agricultural production and food safety. The picture drawn explains why several countermeasures aiming to reduce the contamination in food products, although scientifically sound and technically feasible, are hardly acceptable or even not acceptable at all, to the stakeholders

  9. A new DYN3D library for the WWER-1000 reactors

    International Nuclear Information System (INIS)

    A new library of two-group diffusion and kinetics parameters has been generated for the neutron kinetics code DYN3D, intended for analysis of reactivity initiated accidents for the WWER-1000 reactors. All assembly types for the 3-year fuel cycle are included. The Helios-1.5 code and its 190-group library have been used at the stage of lattice calculations. The approximation methodology is based on a combination between interpolation over the moderator temperature and density, and approximation over the rest of the independent state parameters. High accuracy is achieved by applying square interpolation over the moderator temperature and cubical interpolation over the moderator density. The axial/radial reflectors are described by effective diffusion parameters, including reference discontinuity factors, calculated on the base of one/two-dimensional heterogeneous 23-group transport theory solutions by Mariko. DYN3D has been modified in order to use the new library. Results by BIPR-7, used for operational neutronics calculations of the WWER-1000 reactors at the Kozloduy NPP, have been used to validate the new library (Authors)

  10. 3 MW TRIGA Research Reactor facility of BAEC and its Utilization

    Energy Technology Data Exchange (ETDEWEB)

    Molla, N.I.; Bhuiyan, S.I.; Wadud Mondal, M.A.; Ahmed, F.U.; Islam, M.N.; Hossain, S.M.; Ahmed, K.; Zulquarnain, A.; Abedin, Z. [Bangladesh Atomic Energy Commission, Atomic Energy Research Establishment, Dhaka (Bangladesh)

    1999-08-01

    The paper briefly describes the Utilisation of 3 MW TRIGA Research Reactor of BAEC for neutron beam research, neutron activation analysis are isotope production. It includes the installation of the triple axis neutron spectrometer at the radial piercing beam port and a neutron radiography set-up at the tangential beam port and their uses for material analysis and condensed matter research and material testing. Nuclear and magnetic structures of some ferrites have been studied in powder diffraction method in the double axis mode. SANS technique with double crystal diffraction known as Bonse and Hart's method has been adopted in an experiment with alumina sample. The neutron radiography set-up and its use in the detection of corrosion in alumina have been reported. Determination of arsenic concentration in drinking water from tube well via Instrumental Neutron Activation Analysis and production of radioiodine-131 by dry distillation method are presented. Our experience on the removal of N-16 decay tank because of the leakage of coolant and bringing the research reactor back to operational by-passing the decay tank have been focussed. A possible reconfiguration of the existing TRIGA core, without exceeding the safety margins, providing additional irradiation channel and upgrading the neutron flux for increased radioisotope production has been attempted. Cross section library ENDF/B-VI and JENDL3.2, code NJOY94.10, WIMSD package, 3-D code CITATION, PARET and Monte Carlo code MCNP4B2 have been employed to achieve the objective. (author)

  11. Conceptual analysis of the fuel management strategy for the RA-3 research reactor at 10 MW

    International Nuclear Information System (INIS)

    The Argentine Research Reactor RA-3 was designed to produce radioisotopes and it operates with LEU (U3O8) fuel since 1990. Its initial power was 5 MW and it has recently been upgraded to 10 MW. The National Atomic Energy Commission (CNEA) is both its owner and operator. At the beginning of this year, the Nuclear Regulatory Authority extended its operation license to an authorised power of 10 MW after a series of modifications and tests carried out by the installation during 2002 and 2003. As a consequence of this power increase, the installation introduced some non-systematic modifications in its fuel management strategy with the purpose of preserving the operation period in 20 days approximately. The main change was to load 2 fuel elements per cycle in some cycles (instead of 1 as it used to be at 5 MW). The purpose of this work is to perform a conceptual analysis of possible fuel management strategies for the RA-3 reactor, that could provide quantitative elements for a safety assessment, as well as to evaluate the fuel management flexibility at 10 MW in compliance with standards in force. It is concluded that operation at 10 MW with a 2 FE/cycle strategy leads to a significant excess reactivity at the beginning of cycle, but still in compliance with the margins established by the standards of application. (author)

  12. 3 MW TRIGA Research Reactor facility of BAEC and its Utilization

    International Nuclear Information System (INIS)

    The paper briefly describes the Utilisation of 3 MW TRIGA Research Reactor of BAEC for neutron beam research, neutron activation analysis are isotope production. It includes the installation of the triple axis neutron spectrometer at the radial piercing beam port and a neutron radiography set-up at the tangential beam port and their uses for material analysis and condensed matter research and material testing. Nuclear and magnetic structures of some ferrites have been studied in powder diffraction method in the double axis mode. SANS technique with double crystal diffraction known as Bonse and Hart's method has been adopted in an experiment with alumina sample. The neutron radiography set-up and its use in the detection of corrosion in alumina have been reported. Determination of arsenic concentration in drinking water from tube well via Instrumental Neutron Activation Analysis and production of radioiodine-131 by dry distillation method are presented. Our experience on the removal of N-16 decay tank because of the leakage of coolant and bringing the research reactor back to operational by-passing the decay tank have been focussed. A possible reconfiguration of the existing TRIGA core, without exceeding the safety margins, providing additional irradiation channel and upgrading the neutron flux for increased radioisotope production has been attempted. Cross section library ENDF/B-VI and JENDL3.2, code NJOY94.10, WIMSD package, 3-D code CITATION, PARET and Monte Carlo code MCNP4B2 have been employed to achieve the objective. (author)

  13. Development of a 3-dimensional calculation model of the Danish research reactor DR3 to analyse a proposal to a new core design called ring-core

    International Nuclear Information System (INIS)

    A 3-dimensional calculation model of the Danish research reactor DR3 has been developed. Demands of a more effective utilization of the reactor and its facilities has required a more detailed calculation tool than applied so far. A great deal of attention has been devoted to the treatment of the coarse control arms. The model has been tested against measurements with satisfying results. Furthermore the model has been used to analyse a proposal to a new core design called ring-core where 4 central fuel elements are replaced by 4 dummy elements to increase the thermal flux in the center of the reactor. (author)

  14. Annual report of Department of Research Reactor and Tandem Accelerator, JFY2007. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and tandem accelerator

    International Nuclear Information System (INIS)

    The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor-3), JRR-4(Japan Research Reactor-4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2007 and March 31, 2008. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator. (2) Utilization of research reactors and tandem accelerator. (3) Upgrading of utilization techniques of research reactors and tandem accelerator. (4) Safety administration for research reactors and tandem accelerator. (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, commendation, plans and outcomes in service and technical developments and so on. (author)

  15. Annual report of Department of Research Reactor and Tandem Accelerator, JFY2009. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and Tandem Accelerator

    International Nuclear Information System (INIS)

    The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor-3), JRR-4(Japan Research Reactor-4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2009 and March 31, 2010. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator. (2) Utilization of research reactors and tandem accelerator. (3) Upgrading of utilization techniques of research reactors and tandem accelerator. (4) Safety administration for research reactors and tandem accelerator. (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, commendation, outcomes in service and technical developments and so on. (author)

  16. Annual report of Department of Research Reactor and Tandem Accelerator, JFY2008. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and Tandem Accelerator

    International Nuclear Information System (INIS)

    The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor-3), JRR-4(Japan Research Reactor-4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2008 and March 31, 2009. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator. (2) Utilization of research reactors and tandem accelerator. (3) Upgrading of utilization techniques of research reactors and tandem accelerator. (4) Safety administration for research reactors and tandem accelerator. (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, commendation, outcomes in service and technical developments and so on. (author)

  17. Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2005

    International Nuclear Information System (INIS)

    In the fiscal year 2005, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 37 cycles (daily operation : 137 days). JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography. Irradiation for activation analyses, radioisotope (RI) productions, fission tracks. Irradiation test of reactor materials etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT). Prompt gamma-ray analyses. Sensitivity measurement of radiation detectors. Experiment in the nuclear reactor training. Practice of Reactor operation. Irradiation for activation analyses, RI productions, fission tracks etc. The volume contains 100 activity reports, which are categorized into the fields of neutron scattering (9 subcategories), neutron radiography, neutron activation analyses, RI productions, prompt gamma-ray analyses, and others submitted by the users in JAEA and from other organizations. (author)

  18. Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2009

    International Nuclear Information System (INIS)

    JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. In the fiscal year 2009, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 6 cycles (daily operation : 24 days). The volume contains 138 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, prompt gamma-ray analyses, neutron activation analyses, RI productions, and others submitted by the users in JAEA and from other organizations. (author)

  19. Detailed modeling of KALININ-3 NPP VVER-1000 reactor pressure vessel by the coupled system code ATHLET/BIPR-VVER

    International Nuclear Information System (INIS)

    The paper gives an overview of the recent developments of a new reactor pressure vessel (RPV) model of VVER-1000 for the coupled system code ATHLET/BIPR-VVER. Based on the previous experience a methodology is worked out for modeling the RPV in a pseudo-3D way with the help of a multiple parallel thermal-hydraulic channel scheme that follows the hexagonal fuel assembly structure from the bottom to the top of the reactor. The results of the first application of the new modeling are discussed on the base of the OECD/NEA coupled code benchmark for Kalinin-3 NPP transient. Coolant mass flow distributions in reactor volume of VVER 1000 reactor are presented and discussed. It is shown that along the core height a mass flow re-distribution of the coolant takes place starting approximately at an axial layer located 1 meter below the core outlet. (author)

  20. Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2006

    International Nuclear Information System (INIS)

    In the fiscal year 2006, the research reactor JRR-3 was operated 7 cycles (cycle operation: 26 days/cycle) for utilization sharing of the facility. And JRR-4 was operated 37 cycles (daily operation: 151 days). JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. The volume contains 294 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, neutron activation analyses, RI productions, prompt gamma-ray analyses, and others submitted by the users in JAEA and from other organizations. (author)

  1. Nuclear criticality safety of fuel element storage for upgraded research reactor JRR-3

    International Nuclear Information System (INIS)

    Criticality aspects of storing 20 % enriched U Alsub(x) - Al fuel elements were evaluated for the upgraded research reactor JRR-3. Criticality calculations were carried out as a function of the number of fuel elements, lattice pitch, and water density in the moderator. The effects of neutron absorbers on neutron multiplication were also examined for storage arrays of the fuel element. Results show that the arrays in the storage racks proposed for the JRR-3 are subcritical enough. The fuel elements can be safely stored against any possible storage circumstances. The obtained data are presented in a form in which interpolation may be made to estimate the neutron multiplication factor of any element storage configulations of the fuel elements of the JRR-3. (author)

  2. Research of 3-D hexagonal nodal transport method for fast reactor

    International Nuclear Information System (INIS)

    The 3-D hexagonal nodal transport theory calculation method for fast reactor core was studied. Based on this method, 3-D hexagonal nodal transport code NAST was developed. The surface average angular fluxes were approximated by an azimuthally symmetric double Pn-expansion DP1 and DP3, and 1-D discrete ordinates equations were solved on a fine spatial mesh within the node. Considering the characteristics of the nodal method, the response matrix method was used in the iterations. Therefore, the calculation within the node was simplified and time was saved. The code was tested for the keff, calculation of CEFR and BN-600. A good agreement with the reference results was achieved. (authors)

  3. Eu2O3 and B4C worth calculations in fast reactor spectra

    International Nuclear Information System (INIS)

    Work is now in progress to design and fabricate europia (Eu2O3) control rods for irradiation testing in one of the early operating cycles of the Fast Test Reactor (FTR). These tests will provide data for evaluating europia as a possible control rod absorber material in fast reactors. Prediction of the reactivity worth of the europia rods is a necessary part of the design process. To improve the accuracy of these predictions, an experiment was performed in the FTR Engineering Mockup Critical (EMC) to determine the relative reactivity worth of Eu2O3 and boron carbide (B4C) in control rod size quantities. Upon completion of the experiment, calculations were performed to obtain reactivity worths for comparison with the measured worths. From this comparison, calculation-vs.-experiment (C/E) bias factors were obtained for use in correcting the computed reactivity worth of europia in the FTR. A brief description of the experiment is presented together with the experimental results, a description of the analytical methods, the calculated results, and a comparison of the calculated and experimental results. Also included are discussions of the effects of resonance self-shielding and mesh spacing on computed absorber rod worths

  4. The Oak Ridge Research Reactor: Safety analysis: Volume 2, Supplement 3

    International Nuclear Information System (INIS)

    The Oak Ridge Research Reactor (ORR) was constructed in the mid 1950s. Since it is an older facility, the issue of life-limiting conditions or material deterioration resulting from prolonged exposure to the normal operating environment is an item that should be addressed in the safety analysis for the ORR. Life-limiting conditions were considered in the original design of ORR; but due to the limited data that were available at that time on material performance in research reactors, various studies were completed during the first 10 years of operation at ORR to verify the applicable life-limiting parameters. Based on today's knowledge of life limiting conditions and the previous 30 years of operating experience at the ORR facility, the three specific areas of concern are addressed in this supplement: (1) embrittlement of the structures due to radiation damage, which is described in Section 2; (2) fatigue due to the effects of both thermal cycling and vibration, which is addressed in Section 3; and (3) the effects of corrosion on the integrity of the primary system, which is described in Section 4. The purpose of this document is to provide a review of the applicable safety studies which have been performed, and to state the status of the ORR with regard to embrittlement, fatigue (due to thermal cycling and vibration), and corrosion

  5. Usual control of RA-3 fuel assemblies for the upgrading the reactor power and fuel density

    International Nuclear Information System (INIS)

    In recent years, the Argentine Atomic Energy Commission has been working actively on the upgrading of its RA-3 research reactor and the development of new fuel compounds. The last and most important milestones were the irradiation of fuel prototypes with higher U mass and higher density that standard, and the upgrading of the reactor power from 5 to 10 MW. The RA-3 is a typical Material Testing and Research Reactor (MTR), built and operated by CNEA, and located at Ezeiza Atomic Center in Buenos Aires, Argentina. It is a pool type reactor, moderated and cooled with light water. Cooling is provided by down going forced convection. Since its commissioning, the reactor had mainly operated at a thermal power of 5 MW until a power increase program was initiated. Recently, the target of 10 MW was reached successfully. This involved mainly the improvement of primary and secondary cooling and water purification systems. The primary coolant flow was increased from 750 to 1350 m3/h, this was achieved through a third pipeline with its corresponding pump and heat exchanger. The current core consists of 21 standard fuel elements (Fe) and 4 control fuel elements. These fuel assemblies are Mtr-type box plates of low enrichment Uranium (Leu), with an U3O8-Al fuel meat matrix with a meat density of 3.2 gu/cm3 and Aluminum-6061 cladding, having 19 and 14 fuel plates respectively. Since 1997, three full scale prototypes of U3Si2-Al matrix with a meat density of 4.8 gU/cm3 and having up to 30% higher U mass have been irradiated. The irradiation of FE with even higher U mass and increased density compounds is about to start. All of these modifications brought about the need of a more demanding control of the core components, the fuel elements, and the radiological and chemical quality of water of the primary cooling system (PWS). Before the power increase program was initiated, all the graphite reflector boxes were inspected in the annex built-in hot cell of the reactor in order to

  6. Development of a 3D multigroup program for Dancoff factor calculation in pebble bed reactors

    International Nuclear Information System (INIS)

    Highlights: • Development of a 3D Monte Carlo based code for pebble bed reactors. • Dancoff sensitivity to clad, moderator and fuel cross sections is considered. • Sensitivity of Dancoff to number of energy groups is considered. • Sensitivity of Dancoff to number of fuel and their arrangement is considered. • Excellent agreements vs. MCNP code. - Abstract: The evaluation of multigroup constants in reactor calculations depends on several parameters. One of these parameters is the Dancoff factor which is used for calculating the resonance integral and flux depression in the resonance region in heterogeneous systems. In the current paper, a computer program (MCDAN-3D) is developed for calculating three dimensional black and gray Dancoff coefficients, based on Monte Carlo, escape probability and neutron free flight methods. The developed program is capable to calculate the Dancoff factor for an arbitrary arrangement of fuel and moderator pebbles. Moreover this program can simulate fuels with homogeneous and heterogeneous compositions. It might generate the position of Triso particles in fuel pebbles randomly as well. It could calculate the black and gray Dancoff coefficients since fuel region might have different cross sections. Finally, the effects of clad and moderator are considered and the sensitivity of Dancoff factor with fuels arrangement variation, number of TRISO particles and neutron energy has been studied

  7. The advanced 3D method for activation analysis of fusion reactor materials

    International Nuclear Information System (INIS)

    The method allows analyzing the complex objects activated by neutrons (e.g. fusion reactors) combining advantages of the 3D radiation transport by MCNP program with calculations of multiple activation and radioactive decay chains by FISPACT program. The problem of preparing the gamma-ray sources in cells of 3D geometry was solved by creation of an interface between the MCNP and FISPACT programs. The interface allows optimizing the process of activation analysis by revealing dominant sources of radiation. The developed interface essentially reduces the time needed for calculations. The main advantage of the method is realization of so-called 'multibox' procedure for decay gamma source sampling during decay gamma transport in very large and complex fusion reactor models. Shutdown dose rate calculations are faster (up to 600 times in ITER cryostat) in comparison with applied MCNP standard source definition by using an external user-supplied source subroutine of the 'multibox' procedure. The offered method is intended for solution of the activation tasks with deep penetration of radiation. The method was used in the engineering design of ITER-FEAT and RF DEMO-S

  8. ORPHEE research reactor: 3D core depletion calculation using Monte-Carlo code TRIPOLI-4®

    Science.gov (United States)

    Damian, F.; Brun, E.

    2014-06-01

    ORPHEE is a research reactor located at CEA Saclay. It aims at producing neutron beams for experiments. This is a pool-type reactor (heavy water), and the core is cooled by light water. Its thermal power is 14 MW. ORPHEE core is 90 cm height and has a cross section of 27x27 cm2. It is loaded with eight fuel assemblies characterized by a various number of fuel plates. The fuel plate is composed of aluminium and High Enriched Uranium (HEU). It is a once through core with a fuel cycle length of approximately 100 Equivalent Full Power Days (EFPD) and with a maximum burnup of 40%. Various analyses under progress at CEA concern the determination of the core neutronic parameters during irradiation. Taking into consideration the geometrical complexity of the core and the quasi absence of thermal feedback for nominal operation, the 3D core depletion calculations are performed using the Monte-Carlo code TRIPOLI-4® [1,2,3]. A preliminary validation of the depletion calculation was performed on a 2D core configuration by comparison with the deterministic transport code APOLLO2 [4]. The analysis showed the reliability of TRIPOLI-4® to calculate a complex core configuration using a large number of depleting regions with a high level of confidence.

  9. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 2-Sequoyah Unit 2 Cycle 3

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, S.M.

    1995-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations reported herein is based on the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of three reactor critical configurations for the Sequoyah Unit 2 Cycle 3. This unit and cycle were chosen because of the relevance in spent fuel benchmark applications: (1) the unit had a significantly long downtime of 2.7 years during the middle of cycle (MOC) 3, and (2) the core consisted entirely of burned fuel at the MOC restart. The first benchmark critical calculation was the MOC restart at hot, full-power (HFP) critical conditions. The

  10. Can metacognition compensate for intelligence in the first year of Belgian higher education?

    NARCIS (Netherlands)

    Minnaert, A

    1996-01-01

    This study reports on the effects of metacognitive knowledge and skills compensating for intelligence in relation to academic performance in the first year of Belgian higher education. About 600 freshmen of educational sciences, medicine and psychology participated in this project. Tasks and questio

  11. Comparing Compositional Effects in Two Education Systems: The Case of the Belgian Communities

    Science.gov (United States)

    Danhier, Julien; Martin, Émilie

    2014-01-01

    The Belgian educational field includes separate educational systems reflecting the division of the country into linguistic communities. Even if the French-speaking and the Dutch-speaking communities keep sharing important similarities in terms of funding rules and structures, they present a huge gap between their respective pupils'…

  12. Performance Measurement in Belgian Hospitals : a state-of-the-art

    OpenAIRE

    Van Caillie, Didier; Rouhana, Rima; Santin, Sarah

    2007-01-01

    This communication proposes a global state-of-the-art around the central question : "How is performance measured and controlled in Belgian hospitals. As a first step in a global research project dedicated to the use of Balanced ScoreCard in publics hospitals around the world, it is essentially focused on global economic aspects and on major macroeconomic statistics.

  13. Clonal Expansion of the Belgian Phytophthora ramorum Populations Based on New Microsatellite Markers

    Science.gov (United States)

    Coexistence of both mating types A1 and A2 within the EU1 lineage of Phytophthora ramorum has only been observed in Belgium, begging the question whether sexual reproduction is occurring. A collection of 411 Belgian P. ramorum isolates was established during a seven year survey. Our main objective w...

  14. Osteochondrosis of the occipital condyles and atlanto-occipital dysplasia in a Belgian horse

    OpenAIRE

    Muirhead, Tammy; McClure, J.T.; Bourque, Andrea; Pack, LeeAnn

    2003-01-01

    A lesion in the cervical region of a 14-month-old Belgian gelding with severe ataxia was suspected. Necropsy revealed symmetric focal cartilage defects compatible with osteochondrosis of the occipital condyles and atlanto-occipital dysplasia. To our knowledge this is the first equine report of symmetrical osteochondrosis of the occipital condyles causing neurologic signs.

  15. The role of the sickness funds in the Belgian health care market

    NARCIS (Netherlands)

    W. Nonneman (Nonneman); E.K.A. van Doorslaer (Eddy)

    1994-01-01

    textabstractThis article reviews some of the salient features of the Belgian health care finance and delivery system. Special attention is paid to the role played by the third-party payers, i.e. the Health Insurance Associations (HIAs) in administering the compulsory national health insurance progra

  16. Chikungunya infection confirmed in a Belgian traveller returning from Phuket (Thailand)

    OpenAIRE

    Bottieau, E; Van Esbroeck, M.; Cnops, L.; Clerinx, J; van Gompel, A.

    2009-01-01

    Chikungunya infection has been increasingly reported in international travellers following its epidemic re-emergence in the Indian Ocean islands in 2006 and its spread to southern Asia thereafter. We describe the first case of chikungunya in a Belgian traveller returning from Phuket, Thailand and discuss the potential implications of chikungunya cases imported to European countries for patient management and public health.

  17. Speaking Turkish in Belgian primary schools: teacher beliefs versus effective consequences

    NARCIS (Netherlands)

    O. Ağırdağ; K. Jordens; M. Van Houtte

    2014-01-01

    In this mixed-method study, we explore teachers’ beliefs concerning the use of the Turkish language by Turkish children in Belgian primary schools, and we compare these findings with the effective consequences of language maintenance. The qualitative analyses revealed that teachers have very negativ

  18. Multiyear Serological Surveillance of Notifiable Influenza A Viruses in Belgian Poultry: A Retrospective Analysis.

    Science.gov (United States)

    Marché, Sylvie; Houdart, Philippe; van den Berg, Thierry; Lambrecht, Bénédicte

    2016-05-01

    Surveillance of notifiable avian influenza (NAI) virus is mandatory in European member states, and each year a serological survey is performed to detect H5 and H7 circulation in poultry holdings. In Belgium, this serological monitoring is a combination of a stratified and a risk-based approach and is applied to commercial holdings with more than 200 birds. Moreover, a competitive nucleoprotein (NP) ELISA has been used as first screening method since 2010. A retrospective analysis of the serological monitoring performed from 2007 through 2013 showed sporadic circulation of notifiable low-pathogenicity avian influenza (LPAI) viruses in Belgian holdings with a fluctuating apparent flock seroprevalence according to years and species. Overall, the highest apparent flock seroprevalence was detected for the H5 subtype in domestic Anatidae, with 20%-50% for breeding geese and 4%-9% for fattening ducks. Positive serology against non-H5/H7 viruses was also observed in the same species with the use of the IDScreen influenza A antibody competition ELISA kit (ID-vet NP ELISA), and confirmed by isolation of H2, H3, H6, and H9 LPAI viruses. Among Galliformes, the apparent flock seroprevalence was lower, ranging between 0.3% and 1.3%. Circulation of notifiable LPAI viruses was only observed in laying hens with a similar seroprevalence for H5 and H7. Based on ID-vet NP ELISA results, no circulation of LPAI viruses, regardless the subtype, was observed in breeding chickens and fattening turkeys. Retrospectively, the use of an ELISA as first-line test not only reduced the number of hemagglutination inhibition tests to be performed, but also gave a broader evaluation of the prevalence of LPAI viruses in general, and might help to identify the most at-risk farms. PMID:27309088

  19. Multiyear Serological Surveillance of Notifiable Influenza A Viruses in Belgian Poultry: A Retrospective Analysis.

    Science.gov (United States)

    Marché, Sylvie; Houdart, Philippe; van den Berg, Thierry; Lambrecht, Bénédicte

    2015-12-01

    Surveillance of notifiable avian influenza (NAI) virus is mandatory in European member states, and each year a serological survey is performed to detect H5 and H7 circulation in poultry holdings. In Belgium, this serological monitoring is a combination of a stratified and a risk-based approach and is applied to commercial holdings with more than 200 birds. Moreover, a competitive nucleoprotein (NP) ELISA has been used as first screening method since 2010. A retrospective analysis of the serological monitoring performed from 2007 through 2013 showed sporadic circulation of notifiable low-pathogenicity avian influenza (LPAI) viruses in Belgian holdings with a fluctuating apparent flock seroprevalence according to years and species. Overall, the highest apparent flock seroprevalence was detected for the H5 subtype in domestic Anatidae, with 20%-50% for breeding geese and 4%-9% for fattening ducks. Positive serology against non-H5/H7 viruses was also observed in the same species with the use of the IDScreen influenza A antibody competition ELISA kit (ID-vet NP ELISA), and confirmed by isolation of H2, H3, H6, and H9 LPAI viruses. Among Galliformes, the apparent flock seroprevalence was lower, ranging between 0.3% and 1.3%. Circulation of notifiable LPAI viruses was only observed in laying hens with a similar seroprevalence for H5 and H7. Based on ID-vet NP ELISA results, no circulation of LPAI viruses, regardless the subtype, was observed in breeding chickens and fattening turkeys. Retrospectively, the use of an ELISA as first-line test not only reduced the number of hemagglutination inhibition tests to be performed, but also gave a broader evaluation of the prevalence of LPAI viruses in general, and might help to identify the most at-risk farms. PMID:26629630

  20. Chemical-looping gasification of biomass in a 10k Wth interconnected fluidized bed reactor using Fe2 O3/Al2 O3 oxygen carrier

    Institute of Scientific and Technical Information of China (English)

    HUSEYIN Sozen; WEI Guo-qiang; LI Hai-bin; HE Fang; HUANG Zhen

    2014-01-01

    The aim of this research is to design and operate a 10 kW hot chemical-looping gasification ( CLG) unit using Fe2 O3/Al2 O3 as an oxygen carrier and saw dust as a fuel. The effect of the operation temperature on gas composition in the air reactor and the fuel reactor, and the carbon conversion of biomass to CO2 and CO in the fuel reactor have been experimentally studied. A total 60 h run has been obtained with the same batch of oxygen carrier of iron oxide supported with alumina. The results show that CO and H2 concentrations are increased with increasing temperature in the fuel reactor. It is also found that with increasing fuel reactor temperature, both the amount of residual char in the fuel reactor and CO2 concentration of the exit gas from the air reactor are degreased. Carbon conversion rate and gasification efficiency are increased by increasing temperature and H2 production at 870 ℃reaches the highest rate. Scanning electron microscopy (SEM), X-ray diffraction (XRD) and BET-surface area tests have been used to characterize fresh and reacted oxygen carrier particles. The results display that the oxygen carrier activity is not declined and the specific surface area of the oxygen carrier particles is not decreased significantly.

  1. Development of Nb3Sn strands for the International Thermonuclear Experimental Reactor (ITER) in Japan

    International Nuclear Information System (INIS)

    Development programs focusing on Nb3Sn strands for the International Thermonuclear Experimental Reactor (ITER) are being performed by each party of the ITER group. The object of development is not only to achieve high superconductor performance such as high current density and low hysteresis loss but also enable reliable low-cost mass production, one way of which is to reduce strand breaking during the manufacturing process, and long-length strand fabrication. In Japan, the Japan Atomic Energy Research Institute (JAERI) is involved in this program backed by strong collaboration with many domestic industries. Totally, 11.1 tons (2,400 km-length) of strands that meet ITER specifications have been fabricated successfully. Through this development work, the Nb3Sn strand mass-production technique in Japan has been substantially improved. (author)

  2. Ignition curves for deuterium/helium-3 fuel in spherical tokamak reactor

    Indian Academy of Sciences (India)

    Motevalli S M; Fadaei F

    2016-04-01

    In this paper, ignition curve for deuterium/helium-3 fusion reaction is studied. Four fusion reactions are considered. Zero-dimensional model for the power balance equation has been used. The closed ignition curves for $\\rho$ = constant (ratio of particle to energy confinement time) have been derived. The results of our calculations show that ignited equilibria for deuterium/helium-3 fuel in a spherical tokamak is only possible for $\\rho$ = 5.5 and 6. Then, by using the energy confinement scaling and parameters of the spherical tokamak reactor, the plasma stability limits have been obtained in $n_e, T$ plane and, to determine the thermal instability of plasma, the time dependent transport equations have been solved.

  3. Central Reactivity Measurements on Assemblies 1 and 3 of the Fast Reactor FR0

    International Nuclear Information System (INIS)

    The reactivity effects of small samples of various materials have been measured, by the period method at the core centre of Assemblies 1 and 3 of the fast zero power reactor FR0. For some materials the reactivity change as a function of sample size has also been determined experimentally. The core of Assembly 1 consisted only of uranium enriched to 20 % whereas the core of Assembly 3 was diluted with 30 % graphite. The results have been compared with calculated values obtained with a second-order transport-theoretical perturbation model and using differently shielded cross sections depending upon sample size. Qualitative agreement has generally been found, although discrepancies still exist. The spectrum perturbation caused by the experimental arrangement has been analyzed and found to be rather important

  4. Prevalence and origin of HIV-1 group M subtypes among patients attending a Belgian hospital in 1999.

    Science.gov (United States)

    Snoeck, Joke; Van Dooren, Sonia; Van Laethem, Kristel; Derdelinckx, Inge; Van Wijngaerden, Eric; De Clercq, Erik; Vandamme, Anne-Mieke

    2002-04-23

    HIV-1 group M strains are usually subtyped based on gag and/or env gene sequences. In our lab, part of the pol gene sequence was available in order to determine the genotypic anti-HIV drug resistance profile. To estimate the prevalence of the different HIV-1 subtypes in patients visiting the University Hospitals in Leuven in 1999 and for whom a genotypic drug resistance test was needed, we tried to use the pol sequence for subtyping. Recombination was investigated by similarity plots and bootscanning and subtyping was performed by phylogenetic analysis. The overall region spanning the entire protease and 747 nucleotides of the reverse transcriptase proved very suitable for subtyping, although there was a low phylogenetic signal at the beginning of the reverse transcriptase (nucleotides 0-250), as we demonstrated by likelihood mapping. Of the 41 samples analyzed, 21 belonged to subtype B. Of the other 20 non-B strains, 9 belonged to subtype C, 2 to subtype D and 1 to subtype A, G, H and J, respectively, 3 were CRF_02 (Circulating Recombinant Form), 1 was recombinant with a novel breakpoint and 1 sample was untypable. Although subtype B is still the most prevalent subtype in Belgium, it seems to be responsible for only half of the infections in this study. We could also document that the prevalence of subtype C is high in the Belgian native patients, especially among the heterosexually infected population. This could possibly be an indication for an epidemic spread of HIV-1 subtype C in Belgium, as for one third of these patients, no link to an endemic region could be found. The other non-B subtypes and the recombinants are mainly introduced by immigrants or by Belgian citizens traveling abroad. PMID:11955642

  5. Delayed neutron fraction in low enrichment nuclear fuel. Case of RA3 reactor, CNEA

    International Nuclear Information System (INIS)

    Between 2000 and 2002 several experiments were carried out to evaluate RA-3 Reactor power by two independent experimental methods: neutron noise and thermohydraulics. Because the neutron noise technique needs βeff value, it was recalculated following the lines applied to RA-4 Reactor of CNEA. The 'up to this work' accepted value of βeff was 8.14E-03. For the reevaluation the neutron diffusion code PUMA and transport code WIMS was used. A five groups neutron spectrum, several percentages of burn-up, was considered. The νd's from various authors were evaluated to choose the best fit. Two step calculation was carried out: a) the nuclear β (βn, and the effective β (βeff). The adopted value of βn was 6.84E-03. Effective β was obtained using PUMA diffusion code using a five macrogroups option. The six spectra of delayed neutron were condensed from Rudstam tables. So βeff obtained was 12% smaller than the previously obtained value 8.14E-03

  6. Spectral history modeling in the reactor dynamics code DYN3D

    International Nuclear Information System (INIS)

    A new method of treating spectral history effects in reactor core calculations was developed and verified in this dissertation. The nature of history effects is a dependence of fuel properties not only on the burnup, but also on the local spectral conditions during burnup. The basic idea of the proposed method is the use of the plutonium-239 concentration as the spectral history indicator. The method was implemented in the reactor dynamics code DYN3D and provides a correction for nodal cross sections according to the local spectral history. A verification of the new method was performed by single-assembly calculations in comparison with results of the lattice code HELIOS. The application of plutonium-based history correction significantly improves the cross section estimation accuracy both for UOX and MOX fuel, with quadratic and hexagonal geometry. The new method was applied to evaluate the influence of history effects on full-core calculation results. Analysis of a PWR equilibrium fuel cycle has shown a significant effect on the axial power distribution during a whole cycle, which causes axial temperature and burnup redistributions. The observed neutron flux redistribution improves neutron economy, so the fuel cycle is longer than in calculations without history corrections. Analyses of hypothetical control rod ejection accidents have shown a minor influence of history effects on the transient course and safety relevant parameters.

  7. Neutronic Analysis of the 3 MW TRIGA MARK II Research Reactor, Part I: Monte Carlo Simulation

    International Nuclear Information System (INIS)

    This study deals with the neutronic analysis of the current core configuration of a 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh and validation of the results by benchmarking with the experimental, operational and available Final Safety Analysis Report (FSAR) values. The three-dimensional continuous-energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the TRIGA core. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Continuous energy cross-section data from ENDF/B-VI and S(α, β) scattering functions from the ENDF/B-V library were used. The validation of the model against benchmark experimental results is presented. The MCNP predictions and the experimentally determined values are found to be in very good agreement, which indicates that the Monte Carlo model is correctly simulating the TRIGA reactor. (author)

  8. The management routes for materials produced by the dismantling of the BR3-PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Klein, M.; Demeulemeester, Y.; Ponnet, M.; Emond, M.; Emond, O.; Dadoumont, J.; Massaut, V. [SCK. CEN, Belgian Nuclear Research Centre, Mol (Belgium)

    2000-07-01

    The dismantling of the BR3 reactor produces quite large masses of contaminated materials, mainly metals or concrete. The main management routes are: conditioning of the radioactive wastes and disposal, recycling of radioactive materials in the nuclear sector and the recycling of free released materials in the industrial sector or their evacuation as industrial waste. The conditioning of the radioactive wastes is essentially performed in the installations of Belgoprocess and must follow the specifications imposed by the national radwaste management agency ONDRAF/NIRAS. The conditioning of the pieces produced during the cutting of the reactor pressure vessel is given as example. The recycling of radioactive materials in the nuclear sector is possible for metals and for concrete. For metals, SCK.CEN has an agreement with a nuclear foundry which reuses these materials for the fabrication of shieldings. For concrete, an R and D programme is going on with the objective to demonstrate the possible reuse of baryte concrete as raw materials for the production of mortar used in the conditioning of radioactive wastes. The free release of radioactive materials and their reuse or evacuation as radioactive wastes requires the strict respect of procedures and the use of low level measurement techniques. Various decontamination techniques are used at SCK.CEN to reach this objective. For the metals, we use mainly simple washing, abrasive decontamination and hard chemical decontamination. For concrete, we use mainly scabbling or shaving techniques. (authors)

  9. The management routes for materials produced by the dismantling of the BR3-PWR reactor

    International Nuclear Information System (INIS)

    The dismantling of the BR3 reactor produces quite large masses of contaminated materials, mainly metals or concrete. The main management routes are: conditioning of the radioactive wastes and disposal, recycling of radioactive materials in the nuclear sector and the recycling of free released materials in the industrial sector or their evacuation as industrial waste. The conditioning of the radioactive wastes is essentially performed in the installations of Belgoprocess and must follow the specifications imposed by the national radwaste management agency ONDRAF/NIRAS. The conditioning of the pieces produced during the cutting of the reactor pressure vessel is given as example. The recycling of radioactive materials in the nuclear sector is possible for metals and for concrete. For metals, SCK.CEN has an agreement with a nuclear foundry which reuses these materials for the fabrication of shieldings. For concrete, an R and D programme is going on with the objective to demonstrate the possible reuse of baryte concrete as raw materials for the production of mortar used in the conditioning of radioactive wastes. The free release of radioactive materials and their reuse or evacuation as radioactive wastes requires the strict respect of procedures and the use of low level measurement techniques. Various decontamination techniques are used at SCK.CEN to reach this objective. For the metals, we use mainly simple washing, abrasive decontamination and hard chemical decontamination. For concrete, we use mainly scabbling or shaving techniques. (authors)

  10. Deliberation of Post-3.11 Fast Reactor R&D Strategy in Japan

    International Nuclear Information System (INIS)

    The severe accident on 11 March 2011 at the Fukuhsima Daiichi nuclear power plant in Japan has changed the public perception towards nuclear power and it is now predicted that nuclear power will decrease its contribution to energy supply in the long term. Taking this prediction into consideration, Japan is in the process of deliberating the future R&D strategy for the SFR and its fuel cycle technology. Currently, the following are under discussion as major R&D objectives for the SFR and its fuel cycles technology: (1) to accomplish the full power operation of MONJU, improving its safety feature in accordance with new safety standards to be set by the NRA; (2) to explore innovative fast reactor technology for reducing the amount and toxic level of radioactive waste, (3) to explore innovative technologies for strengthening the safety of SNRs and (4) to support innovation in nuclear energy technology, safety, waste management, non-proliferation and security relevant to fast reactors and their fuel cycle technology. As these are major objectives of R&D for a sustainable nuclear energy system, Japan will carry out activities to achieve these objectives in close cooperation with the international community. (author)

  11. Detection and identification of xerophilic fungi in Belgian chocolate confectionery factories.

    Science.gov (United States)

    De Clercq, Nikki; Van Coillie, Els; Van Pamel, Els; De Meulenaer, Bruno; Devlieghere, Frank; Vlaemynck, Geertrui

    2015-04-01

    Chocolate confectionery fillings are generally regarded as microbiologically stable. The stability of these fillings is largely due to the general practice of adding either alcohol or preservatives. Consumer demands are now stimulating producers to move away from adding alcohol or other preservatives to their confectionery fillings and instead to search for innovative formulations. Such changes in composition can influence the shelf life of the product and may lead to spoilage by xerophilic fungi. The aim of this study was to test whether the production environment of Belgian chocolate confectionery factories and common ingredients of chocolate confectioneries could be potential sources of contamination with xerophilic fungal species. In the factory environment, the general and strictly xerophilic fungal spore load was determined using an RCS Air Sampler device in combination with DG18 and MY50G medium, respectively. Four basic ingredients of chocolate confectionery fillings were also examined for fungal spore levels using a direct plating technique. Detected fungi were identified to species level by a combination of morphological characterization and sequence analysis. Results indicated a general fungal spore load in the range of 50-250 colony forming units per cubic meter of air (CFU/m(3) air) and a more strict xerophilic spore load below 50 CFU/m(3) air. These results indicate rather low levels of fungal spores present in the factory environment. The most prevalent fungi in the factory environment were identified as Penicillium spp., particularly Penicillium brevicompactum. Examination of the basic ingredients of confectionery fillings revealed nuts to be the most likely potential source of direct contamination. In nuts, the most prevalent fungal species identified were Eurotium, particularly Eurotium repens. PMID:25475302

  12. The BR2 materials testing reactor. Past, ongoing and under-study upgradings

    International Nuclear Information System (INIS)

    The BR2 reactor (Mol, Belgium) is a high-flux materials testing reactor. The fuel is 93% 235U enriched uranium. The nominal power ranges from 60 to 100 MW. The main features of the design are the following: 1) maximum neutron flux, thermal: 1.2 x 1015 n/cm2 s; fast (E > 0.1 MeV) : 8.4 x 1014 n /cm2 s; 2) great flexibility of utilization: the core configuration and operation mode can be adapted to the experimental loading; 3) neutron spectrum tailoring; 4) availability of five 200 mm diameter channels besides the standard channels (84 mm diameter); 5) access to the top and bottom covers of the reactor authorizing the irradiation of loops. The reactor is used to study the behaviour of fuel elements and structural materials intended for future nuclear power stations of several types (fission and fusion). Irradiations are carried out in connection with performance tests up to very high burn-up or neutron fluence as well as for safety experiments, power cycling experiments, and generally speaking, tests under off-normal conditions. Irradiations for nuclear transmutation (production of high specific activity radio-isotopes and transplutonium elements), neutron-radiography, use of beam tubes for physics studies, and gamma irradiations are also carried out. The BR2 is used in support of Belgian programs, at the request of utilities, industry and universities and in the framework of international agreements. The paper reviews the past and ongoing upgrading and enhancement of reactor capabilities as well as those under study or consideration, namely with regard to: reactor equipment, fuel elements, irradiation facilities, reactor operation conditions and long-term strategy. (author)

  13. The BR1 research facilities to calibrate fuzzy logic technology for nuclear reactor control

    International Nuclear Information System (INIS)

    During the last three decades SCK-CEN has participated in various international programmes using the BR1 (Belgian Reactor 1) facilities for various research and calibration purposes. The BR1 has proved to be an excellent for calibration and validation of techniques, integral nuclear data validation, activation analysis, characterisation of materials by neutron transmission, and physics experiments. Moreover, the knowledge, built up at BR1 has lead to the best calibration conditions for applying fuzzy logic control (FLC) for nuclear reactor control

  14. JEFF-3.1.1 nuclear data validation for sodium fast reactors

    International Nuclear Information System (INIS)

    The JEFF-3.1.1 Nuclear Data Library is the latest version of the Joint Evaluated Fission and Fusion Library. The complete suite of data was released in 2008, and contains general purpose nuclear data evaluations compiled at the NEA Data Bank in co-operation with several laboratories in NEA Data Bank member countries. JEFF-3.1.1 contains also radioactive decay data, activation data and fission yields data. It combines the efforts of the JEFF and EFF Working Groups who have contributed to this combined fission and fusion file. The library contains neutron reaction data, incident proton data and thermal neutron scattering law data in the ENDF-6 format. The aim of this paper is to present the status of the validation of this library using the Monte Carlo Code TRIPOLI4.5 for fast reactor calculations. To reach that goal, we reanalyse a selected set of integral experiments performed in MASURCA Mock-up at CEA/CADARACHE, in ZPPR mock-up at INL USA and in SUPERPHENIX power Reactor. These experiments are: - The CIRANO program in MASURCA (1994-1997) was meant to extend the validation of ERANOS (code, schemes, data libraries) to Pu-burning fast reactors (CAPRA project) via the progressive substitution of fertile blankets by steel reflectors; - The ZPPR10A experiment proposed in IRPhE of the NEA Data Bank. This experiment is complementary of the first one because sodium void effects have been measured and are available; - Several experiments made during de commissioning of SUPERPHENIX that give information on different types of critical states of the core. All these experiments are modelled with the TRIPOLI code to avoid most of the errors due to deterministic models and to focus only on the nuclear data biases. An example of the SUPERPHENIX core modelling is given on the figure 1. Ongoing analysis shows the capability of the new JEFF3.1.1 nuclear library to predict the SFR neutronic behaviours.From this work and from the qualification work performed with ERANOS2, some required

  15. DRAGON 3.05D, Reactor Cell Calculation System with Burnup

    International Nuclear Information System (INIS)

    1 - Description of program or function: The computer code DRAGON contains a collection of models that can simulate the neutron behavior of a unit cell or a fuel assembly in a nuclear reactor. It includes all of the functions that characterize a lattice cell code, namely: the interpolation of microscopic cross sections supplied by means of standard libraries; resonance self-shielding calculations in multidimensional geometries; multigroup and multidimensional neutron flux calculations that can take into account neutron leakage; transport-transport or transport-diffusion equivalence calculations as well as editing of condensed and homogenized nuclear properties for reactor calculations; and finally isotopic depletion calculations. 2 - Methods: The code DRAGON contains a multigroup flux solver conceived that can use a various algorithms to solve the neutron transport equation for the spatial and angular distribution of the flux. Each of these algorithms is presented in the form of a one-group solution procedure where the contributions from other energy groups are considered as sources. The current release of DRAGON contains five such algorithms. The JPM option that solves the integral transport equation using the J+- method, (interface current method applied to homogeneous blocks); the SYBIL option that solves the integral transport equation using the collision probability method for simple one dimensional (1-D) or two dimensional (2-D) geometries and the interface current method for 2-D Cartesian or hexagonal assemblies; the EXCELL/NXT option to solve the integral transport equation using the collision probability method for more general 2-D geometries and for three dimensional (3-D) assemblies; the MOCC option to solve the transport equation using the method of cyclic characteristics in 2-D Cartesian, and finally the MCU option to solve the transport equation using the method of characteristics (non cyclic) for 3-D Cartesian geometries. The execution of DRAGON is

  16. Uranium transport around the reactor zone at Okelobondo (Oklo). Data evaluation with M3 and HYTEC

    International Nuclear Information System (INIS)

    The Swedish Nuclear Fuel and Waste Management Company (SKB) is conducting and participating in Natural Analogue activities as part of various studies regarding the final disposal of high level nuclear waste (HLW). The aim of this study is to use the hydrogeological and hydrochemical data from Okelobondo (Oklo Natural Analogue) to compare the outcome of two independent modelling approaches (HYTEC and M3). The modelling helps to evaluate the processes associated with nuclear natural reactors such as redox, adsorption/desorption and dissolution/precipitation of the uranium and to develop more realistic codes which can be used for site investigations and data evaluation. HYTEC (1D and 2D) represents a deterministic, transport and multi-solutes reactive coupled code developed at Ecole des Mines de Paris. M3 (Multivariate Mixing and Mass balance calculations) is a mathematical-statistical concept code developed for SKB. M3 can relatively easily be used to calculate mixing portions and to identify sinks or sources of element concentrations that may exist in a geochemical system. M3 helped to address the reactions in the coupled code HYTEC. Thus, the major flow-paths and reaction paths were identified and used for transport evaluation. The reactive transport results (one-dimensional and two-dimensional simulations) are in good agreement with the statistical approach using the M3 model. M3 and HYTEC show a dissolution of the uranium layer in contact with upwardly oxidising waters. M3 and HYTEC show a gain of manganese rich minerals downstream the reactor. A comparison of the U and Mn plots for M3 deviation and HYTEC results showed an almost mirror behaviour. The U transport stops when the Mn gain increases. Thus, HYTEC and M3 modelling predict that a possible reason for not having U transport up to the surface in Okelobondo is due to an inorganic trap which may hinder the uranium transport. The two independent modelling approaches can be used to complement each other and to

  17. Benchmark tests of JENDL-3.3 and ENDF/B-VI data files using Monte Carlo simulation of the 3 MW TRIGA MARK II research reactor

    International Nuclear Information System (INIS)

    The three-dimensional continuous-energy Monte Carlo code MNCP4C was used to develop a versatile and accurate full-core model of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Validation of the newly generated continuous energy cross section data from JENDL-3.3 was performed against some well-known benchmark lattices using MCNP4C and the results were found to be in very good agreement with the experiment and other evaluations. For TRIGA analysis continuous energy section data from JENDL-3.3 and ENDF/B-VI in combination with the JENDL-3.2 and ENDF/B-V data files (for natZr, natMo, natCr, natFe, natNi, natSi, and natMg) at 300K evaluations were used. Full S(α, β) scattering functions from ENDF/B-V for Zr in ZrH, H in ZrH and water molecule, and for graphite were used in both cases. The validation of the model was performed against the criticality and reactivity benchmark experiments of the reactor. The MNCP calculated values for effective multiplication factor keff underestimated 0.0250%Δk/k and 0.2510%Δk/k for control rods critical positions and overestimated 0.2098%Δk/k and 0.0966%Δk/k for all control rods withdrawn positions using JENDL-3.3 and ENDF/B-VI, respectively. The core multiplication factor differs appreciably (∼3.3%) between the no S(α, β) (when temperature representation for free gas treatment is about 300K) and 300K S(α, β) case. However, there is ∼20.0% decrease of thermal neutron flux occurs when the thermal library is removed. Effect of erbium isotope that is present in the TRIGA fuel over the criticality analysis of the reactor was also studied. In addition to the keff values, the well known integral parameters: δ28, δ25, ρ25, and C were calculated and compared for both JENDL3.3

  18. Opinion of Belgian Egg Farmers on Hen Welfare and Its Relationship with Housing Type

    Science.gov (United States)

    Stadig, Lisanne M.; Ampe, Bart A.; Van Gansbeke, Suzy; Van den Bogaert, Tom; D’Haenens, Evelien; Heerkens, Jasper L.T.; Tuyttens, Frank A.M.

    2015-01-01

    Simple Summary Until 2012, laying hens in the EU were often housed in conventional cages that offered limited space and few opportunities to perform highly motivated behaviors. Conventional cages are now banned in the EU in order to improve animal welfare. In this study, egg farmers were surveyed (winter 2013–2014) to assess whether they perceived any changes in animal welfare since changing housing systems, what role hen welfare played in choosing a new housing system, and which aspects of hen welfare they find most important. The data show that the answers differ depending on which housing system the farmers currently use and whether they had used conventional cages in the past. Abstract As of 2012, the EU has banned the use of conventional cages (CC) for laying hens, causing a shift in housing systems. This study’s aim was to gain insight into farmers’ opinions on hen health and welfare in their current housing systems. A survey was sent to 218 Belgian egg farmers, of which 127 (58.3%) responded, with 84 still active as egg farmer. Hen welfare tended to be less important in choosing the housing system for farmers with cage than with non-cage systems. Respondents currently using cage systems were more satisfied with hen health than respondents with non-cage systems. Reported mortality increased with farm size and was higher in furnished cages than in floor housing. Feather pecking, cannibalism, smothering and mortality were perceived to be higher in current housing systems than in CC, but only by respondents who shifted to non-cage systems from previously having had CC. Health- and production-related parameters were scored to be more important for hen welfare as compared to behavior-related parameters. Those without CC in the past rated factors relating to natural behavior to be more important for welfare than those with CC. This difference in opinion based on farmer backgrounds should be taken into account in future research. PMID:26703742

  19. Incorporating higher order WINKLER springs with 3-D finite element model of a reactor building for seismic SSI analysis

    International Nuclear Information System (INIS)

    In order to fulfill the seismic safety requirements, in the frame of seismic requalification activities for NPP Muehleberg, Switzerland, detailed seismic analysis performed on the Reactor Building and the results are presented previously. The primary objective of the present investigation is to assess the seismic safety of the reinforced concrete structures of reactor building. To achieve this objective requires a rather detailed 3-D finite element modeling for the outer shell structures, the drywell, the reactor pools, the floor decks and finally, the basemat. This already is a complicated task, which enforces need for simplifications in modelling the reactor internals and the foundation soil. Accordingly, all internal parts are modelled by vertical sticks and the Soil Structure Interaction (SSI) effects are represented by sets of transitional and higher order rotational WINKLER springs, i.e. avoiding complicated finite element SSI analysis. As a matter of fact, the availability of the results of recent investigations carried out on the reactor building using diversive finite element SSI analysis methods allow to calibrate the WINKLER springs, ensuring that the overall SSI behaviour of the reactor building is maintained

  20. Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

    Energy Technology Data Exchange (ETDEWEB)

    M. Chen; CM Regan; D. Noe

    2006-01-09

    Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

  1. 3-D seismic response of a base-isolated fast reactor

    International Nuclear Information System (INIS)

    This paper describes a 3-D response analysis methodology development and its application to a base-isolated fast breeder reactor (FBR) plant. At first, studies on application of a base-isolation system to an FBR plant were performed to identify a range of appropriate characteristics of the system. A response analysis method was developed based on mathematical models for the restoring force characteristics of several types of the systems. A series of shaking table tests using a small scale model was carried out to verify the analysis method. A good agreement was seen between the test and analysis results in terms of the horizontal and vertical responses. Parametric studies were then made to assess the effects of various factors which might be influential to the seismic response of the system. Moreover, the method was applied to evaluate three-dimensional response of the base-isolated FBR. (author)

  2. Alize 3 - first critical experiment for the franco-german high flux reactor - calculations

    International Nuclear Information System (INIS)

    The results of experiments in the light water cooled D2O reflected critical assembly ALIZE III have been compared to calculations. A diffusion model was used with 3 fast and epithermal groups and two overlapping thermal groups, which leads to good agreement of calculated and measured power maps, even in the case of strong variations of the neutron spectrum in the core. The difference of calculated and measured keff was smaller than 0.5 per cent δk/k. Calculations of void and structure material coefficients of the reactivity of 'black' rods in the reflector, of spectrum variations (Cd-ratio, Pu-U-ratio) and to the delayed photoneutron fraction in the D2O reflector were made. Measurements of the influence of beam tubes on reactivity and flux distribution in the reflector were interpreted with regard to an optimum beam tube arrangement for the Franco- German High Flux Reactor. (author)

  3. Once-through thorium fuel cycle evaluation for TVA's Browns Ferry-3 Boiling Water Reactor

    International Nuclear Information System (INIS)

    This report documents benchmark evaluations to test thorium lattice predictive methods and neutron cross sections against available data and summarizes specific evaluations of the once-through thorium cycle when applied to the Browns Ferry-3 BWR. It was concluded that appreciable uncertainties in thorium cycle nuclear data cloud the ability to reliably predict the fuel cycle performance and that power reactor irradiations of ThO2 rods in BWRs are desirable to resolve uncertainties. Benchmark evaluations indicated that the ENDF/B-IV data used in the evaluations should cause an underprediction of U-233/ThO2 fuel reactivity, and, therefore, the results of the preliminary evaluations completed under the program should be conservative

  4. Ordering under heating of cold tempered and irradiated in reactor Fe3Al alloy

    International Nuclear Information System (INIS)

    Here we are presenting results of structure changes processes investigations taking place in preliminary ordered intermetallic compound Fe3Al at cold tempered (deformation degree > 80%), cold worked samples irradiation in reactor WWR-K (fluences from 1 x 1018 up to 5 x 1019 n cm-2, E ≥ 0,1 MeV, Tirr 18 n/cm-2 samples have completely disordered structure. Under the heating process the ordering takes place in both set of samples but it begins at different temperatures: at 300 C for cold worked samples and at 350C for cold worked and irradiated ones. The results are interpreted on the base of the model of atoms separation in the region of displacement cascades. (A.A.D.)

  5. Single-line-to-ground fault test on a 3-phase superconducting fault current limiting reactor

    International Nuclear Information System (INIS)

    The current limiting behavior of a 3-phase superconducting fault current limiting reactor (SCFCLR) in a model power system at a single-line-to-ground fault is experimentally confirmed. A small arc gap is attached on one phase of the model transmission line. And, a single-line-to-ground fault is activated. The behavior of the grounding current and the arc are observed. This paper reports that the experimental result show that, the fault current is limited to very small value by the large zero-phase-sequence reactance of the SCFCLR, and the self-extinction of the fault arc is observed, the power flow is not disturbed by a single-line-to-ground fault at all, and any power system disturbances are not observed, the windings do not quench for a single-line-tosingle-line-to-ground fault

  6. STAR 3D nodal kinetics and thermal-hydraulic model for the Pennsylvania State TRIGA reactor

    International Nuclear Information System (INIS)

    A detailed three-dimensional (3D) time-dependent STAR nodal kinetics model coupled to a one-dimensional (1 D) thermal-hydraulics WIGL model has been developed to describe conservatively the peak power and pulse behavior of the Penn State University (PSU) Breazeale TRIGA reactor. This paper describes how the STAR model and its cross section data input was developed and benchmarked against actual TRIGA pulse experiments. Different core configurations (i.e., different core loading patterns, and with/without the TRIGA core next to the D20 tank) were used for several TRIGA pulse tests with different reactivity insertion worths (1.5$, 2.0$ , 2.5$). This paper shows that the STAR nodal kinetics code adequately simulates TRIGA pulses when group constants are generated from physics codes (i.e., WIMS-D4) that can accurately model the TRIGA uranium-zirconium-hydride fuel. (author)

  7. Study of the U3O8-Al thermite reaction and strength of reactor fuel tubes

    International Nuclear Information System (INIS)

    Heating tests using 53 wt % U3O8-Al pellets show that an exothermic reaction occurs between 875 and 10000C and takes 10 to 20 seconds to reach maximum temperature. The maximum temperature is a function of particle size of the U3O8 with large particles exhibiting lower peak temperatures. The calculated energy release was 123 cal/g of U3O8-aluminum fuel. Tests using aluminum clad outer fuel tube sections gave lower peak temperatures than for pellets. No violent reactions occurred. The results are reasonably consistent with recent reported data indicating that the exothermic U3O8-Al reaction is not an important energy source. The compressive and tensile strengths of U3O8 tubes above 6600C are low. In compression, sections with 2 psi average axial stress failed at 9170C, while sections with 7 psi failed at 6690C. Tubes with U-Al alloy cores failed at about 6700C with no applied load. The stresses in fuel tubes during a reactor transient may range up to several hundred psi and are less than 7 psi only in the upper part of the fuel tube

  8. Finnish EPR Olkiluoto 3. The world's first third-generation reactor now under construction

    International Nuclear Information System (INIS)

    The EPR was developed by Framatome and Siemens KWU (the nuclear division of Siemens), whose nuclear activities were combined in January 2001 to form Framatome ANP, now AREVA NP. The French electricity utility EDF (Electricite de France), together with the major German utilities, played an active role in the project. The safety authorities of the two countries joined forces to bring their respective safety standards into line and draw up joint design rules for the new reactor. On December 18, 2003, the consortium formed by AREVA and Siemens - and led by AREVA - signed a contract with TVO for the turnkey construction of the EPR. The overall Olkiluoto 3 project cost has been estimated by TVO at around euros 3 Billion. TVO is responsible for the overall project management and licensing process with the Finnish Safety Authority STUK. In the pre-qualification phase, STUK concluded that the EPR can meet the Finnish licensing requirements. All specific comments will be taken into account for the realization of the project. In January 2005, STUK emphasized in its safety assessment that the evolutionary EPR design compared to predecessor product lines has been further enhanced by AREVA. This paper presents first, The Finnish energy situation (Electricity consumption and supply, Finland's Kyoto CO2 cutback, Competitiveness of nuclear power), and then the EPR in Olkiluoto (General schedule of responsibilities, Important milestones of the project). Finally, the EPR third-generation and advanced reactor is presented with its position in the international competition (Targeted design objectives, Main characteristics, competitiveness, safety, Additional measures to prevent the occurrence of events likely to damage the core, Increased protection against the consequences of core melt)

  9. Refurbishment, Modernization and Ageing Management Program of The 3MW TRIGA Mark-II Research Reactor of Bangladesh

    International Nuclear Information System (INIS)

    The 3 MW TRIGA MK-II research reactor of Bangladesh Atomic Energy Commission (BAEC) achieved its first criticality on 14 September 1986. The reactor has been used for manpower training, radioisotope production and various R and D activities in the field of neutron activation analysis, neutron radiography and neutron scattering. Reactor Operation and Maintenance Unit (ROMU) is responsible for operation and maintenance of the research reactor. During the past twenty seven years ROMU carried out several refurbishments, replacement, modification and modernization activities in the reactor facility. The major tasks carried out under refurbishment program were replacement of the corrosion damaged N-16 decay tank by a new one, replacement of the fouled shell and tube type heat exchanger by a plate type one, modification of the shielding arrangements around the N-16 decay tank and ECCS system and solving the radial beam port-1 leakage problem. All of these refurbishment activities were performed under an annual development project (ADP) funded by Bangladesh government. BAEC research reactor (RR) was operated by analogue console system from its commissioning to July, 2011. Old analog based console has been replaced by digital console on June, 2012. Modernization program for the reactor control console due to obsolescence and unavailability of spare parts of I and C system was vital to restore the safe operation of the reactor. Considering these facts, installation of a digital control console and I and C system based on the state-of-the-art digital technology became necessary. Reactor digital console system installation tasks were performed under another ADP funded project by Bangladesh government. Now the reactor is operating with the digital control system. Besides this, the Neutron Radiography (NR) facility has been modernized by the addition of a digital neutron radiography set-up at the tangential beam port. The Neutron Scattering (NS) facility also has been upgraded

  10. Refurbishment, Modernization and Ageing Management Program of The 3MW TRIGA Mark-II Research Reactor of Bangladesh

    Energy Technology Data Exchange (ETDEWEB)

    Salam, M. A. [Atomic Energy Research Establishment, Dhaka (Bangladesh)

    2013-07-01

    The 3 MW TRIGA MK-II research reactor of Bangladesh Atomic Energy Commission (BAEC) achieved its first criticality on 14 September 1986. The reactor has been used for manpower training, radioisotope production and various R and D activities in the field of neutron activation analysis, neutron radiography and neutron scattering. Reactor Operation and Maintenance Unit (ROMU) is responsible for operation and maintenance of the research reactor. During the past twenty seven years ROMU carried out several refurbishments, replacement, modification and modernization activities in the reactor facility. The major tasks carried out under refurbishment program were replacement of the corrosion damaged N-16 decay tank by a new one, replacement of the fouled shell and tube type heat exchanger by a plate type one, modification of the shielding arrangements around the N-16 decay tank and ECCS system and solving the radial beam port-1 leakage problem. All of these refurbishment activities were performed under an annual development project (ADP) funded by Bangladesh government. BAEC research reactor (RR) was operated by analogue console system from its commissioning to July, 2011. Old analog based console has been replaced by digital console on June, 2012. Modernization program for the reactor control console due to obsolescence and unavailability of spare parts of I and C system was vital to restore the safe operation of the reactor. Considering these facts, installation of a digital control console and I and C system based on the state-of-the-art digital technology became necessary. Reactor digital console system installation tasks were performed under another ADP funded project by Bangladesh government. Now the reactor is operating with the digital control system. Besides this, the Neutron Radiography (NR) facility has been modernized by the addition of a digital neutron radiography set-up at the tangential beam port. The Neutron Scattering (NS) facility also has been upgraded

  11. Tritium production in He-3 gas cells immersed in the Tokamak Fusion Test Reactor neutron field

    International Nuclear Information System (INIS)

    Tritium generated in an external cell by the reaction 3He(n,p)T can be used as a gauge of long-term fusion neutron production, because of the 12-year half-life of T and the relative ease of measuring the T content either by sampling or from the saturation current of the cell when operated as an ionization chamber. Two high-pressure 3He gas cells enclosed in polyethylene neutron moderators were exposed to Tokamak Fusion Test Reactor (TFTR) neutrons during high-power D - T operation. The tritium produced in the cells was assayed by the Princeton Differential Atmospheric Tritium Sampler. The measured tritium generated per 1019 fusion neutrons was 510 pCi/cc at 2.3 m from the TFTR vessel and 1.3 m below the midplane, and 2020 pCi/cc at 1.0 m from the TFTR vessel in the midplane. Combining these results with previous measurements at a third location, we found 0.11 to 0.23 triton produced per neutron incident on the projected cell cross section, with an asymptotic local tritium breeding ratio of 0.32. copyright 1999 American Institute of Physics

  12. 3D modeling of the primary circuit in the reactor pressure vessel of a PHWR

    International Nuclear Information System (INIS)

    A computational fluid dynamics (CFD) simulation of the reactor pressure vessel (RPV) of the pressurized heavy water reactor (PHWR) of 745 electrical MW Atucha II nuclear power plant was carried out. A three dimensional (3D) detailed model was employed to simulate coolant circuit considering the upper and lower plenums, the downcomer and the hot and cold legs. Control rods and coolant channel tubes at the upper plenum were included to quantify the mixing flow with more realism. The whole set of 451 coolant channels were modeled by means of a zero dimensional methodology. That is, the effect of each coolant channel was modeled through the introduction of a source point at the upper plenum and a sink point at the lower plenum. For each coupled sink/source points (SSP) the mass, momentum and energy balance were solved considering the local pressure difference and the temperature between the corresponding points where sinks and sources were placed. Based on this strategy, three models with increasingly level of approximation were implemented. For the first model the 451 coolant channels were reduced to only 57 pairs of SSP to represent all the coolant channels, concentrating the effect of several coolant channels in a unique pair of sink and source while taking into account geometric design details. For the second model, 225 pairs of SSP were introduced. Finally, for the third model each one of the 451 coolant channels were modeled by means of one pair of SSP. Depending on the coolant channel location, the radial power distribution and the pressure loss caused by the corresponding flow restrictor present by design were considered. Simulations carried out give insight in the complexity of the flow. As expected, the greater the details of the model the better the accuracy reached in the representation of the RPV behavior. In addition, the flow distributor located at the lower plenum showed to be very efficient since, the mass flow at each channel was found to be fairly

  13. Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2007

    International Nuclear Information System (INIS)

    In the fiscal year 2007, the research reactor JRR-3 was operated for 7 cycles (cycle operation : 26days/cycle) and the JRR-4 was operated for 92 days. JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment and practice in the nuclear reactor training, Irradiation for activation analyses, RI productions, fission tracks, etc. The volume contains 262 activity reports, which are categorized into the fields of neutron scattering (10 subcategories), neutron radiography, neutron activation analyses, prompt gamma-ray analyses, and others submitted by the users in JAEA and other Organizations. (author)

  14. Suspended Particulate Matter (SPM) mapping from MERIS imagery. Calibration of a regional algorithm for the Belgian coastal waters

    OpenAIRE

    B. Nechad; De Cauwer, V.; PARK, Y.; Ruddick, K. G.

    2003-01-01

    A hydro-optical algorithm based on reflectance at 555nm has been used in the past for suspended particulate matter concentrations (SPM) retrieval from SeaWiFS over the Belgian coastal (case II) waters in Southern North Sea. The extra spectral resolution of MERIS offers the possibility of improvements, though necessitates algorithm recalibration. This study presents the calibration of the hydro-optical model used to derive SPM from MERIS reflectance for Belgian coastal waters. The model is bas...

  15. Adapting to the system or the student? Exploring teacher adaptations to disadvantaged students in an English and Belgian secondary school

    OpenAIRE

    Stevens, Peter; Van Houtte , Mieke

    2011-01-01

    This article builds on research on teacher adaptations to students by exploring how a Belgian and English national context influence teachers’ definitions of educational success, their explanations of educational failure and allocation of scarce educational resources to disadvantaged students. Ethnographic data from one Flemish (Belgian) and one English secondary, multicultural school suggest that teachers in both schools adapt their expectations to students in line with the perceived ability...

  16. Transportable nuclear power plant T3C-M with two reactor plants of improved safety

    International Nuclear Information System (INIS)

    Development and cultivating of districts in Siberia, North, Far East, Kamchatka and other remote or almost inaccessible district of the country depends to a large degree on their providing with power. The specific character of these districts imposes in turn a wide variety of special requirements upon the power sources. In particular, it is essential to provide the following; maximum manufacture availability of the whole equipment at the minimum volume of construction and installation work on operation site, high safety, longterm service life, ecologically, minimum scope of work on equipment in-service maintenance and inspection, etc. Taking into account the well-known difficulties connected with the delivery of conventional energy carriers to the above-mentioned districts and the situation with the alternative power sources, the application of the low-power nuclear plants (NPP) for these purposes looks definitely promising. Among the probable trends in creating the NPPs of this type as very promising is considered the possibility to apply the two-circuit reactor plant of the vessel type with the liquid lead as a primary coolant and free air as a secondary coolant and working medium in the open gas-turbine cycle. The nuclear plant T3C-M of improved safety with two of this type reactor plants with total electric power of 8 MW is developed by CDB of Machine Building with participation of several enterprises of St. Petersburg under the scientific leadership and is intended for generation of electric power and up to 4 Gcal/h of heat for populated areas and installations placed at long distance from the main electric power supply sources where it is difficult or non-efficient economically to deliver the conventional kinds of fuel. The main principles being laid as a basis when developing the proposed NPP will allow one to create mobile power sources which possess a high degree of safety and inherent self-protection

  17. Deployment of Smart 3D Subsurface Contaminant Characterization at the Brookhaven Graphite Research Reactor

    International Nuclear Information System (INIS)

    The Brookhaven Graphite Research Reactor (BGRR) Historical Site Assessment (BNL 1999) identified contamination inside the Below Grade Ducts (BGD) resulting from the deposition of fission and activation products from the pile on the inner carbon steel liner during reactor operations. Due to partial flooding of the BGD since shutdown, some of this contamination may have leaked out of the ducts into the surrounding soils. The baseline remediation plan for cleanup of contaminated soils beneath the BGD involves complete removal of the ducts, followed by surveying the underlying and surrounding soils, then removing soil that has been contaminated above cleanup goals. Alternatively, if soil contamination around and beneath the BGD is either non-existent/minimal (below cleanup goals) or is very localized and can be ''surgically removed'' at a reasonable cost, the BGD can be decontaminated and left in place. The focus of this Department of Energy Accelerated Site Technology Deployment (DOE ASTD) project was to determine the extent (location, type, and level) of soil contamination surrounding the BGD and to present this data to the stakeholders as part of the Engineering Evaluation/Cost Analysis (EE/CA) process. A suite of innovative characterization tools was used to complete the characterization of the soil surrounding the BGD in a cost-effective and timely fashion and in a manner acceptable to the stakeholders. The tools consisted of a tracer gas leak detection system that was used to define the gaseous leak paths out of the BGD and guide soil characterization studies, a small-footprint Geoprobe to reach areas surrounding the BGD that were difficult to access, two novel, field-deployed, radiological analysis systems (ISOCS and BetaScint) and a three-dimensional (3D) visualization system to facilitate data analysis/interpretation. All of the technologies performed as well or better than expected and the characterization could not have been completed in the same time or at

  18. Evaluation on activation activity of reactor in JRR-2 applied 3 dimensional model to neutron flux calculation

    International Nuclear Information System (INIS)

    Revaluation to activation activity of reactor evaluated at the notification of dismantling submitted in 1997 was carried out in JRR-2 where decommissioning was advanced now. In the revaluation, estimation accuracy on neutron streaming at various horizontal experimental tubes was improved by applying 3 dimensional model to neutron transport calculation that had been carried out by 2 dimensional model, and calculating with TORT. As the result, excessive overestimations on horizontal experimental tubes and biological shield that had greatly contributed to total activation activity in evaluation at the notification of dismantling was revised, sum of their activation activities in the revaluation decreased to 1/18 (case after 1 year from the permanent shutdown of reactor) of evaluation at the notification of dismantling, and the structural materials that had large activation activity were changed. By the above, it was shown that introducing 3 dimensional model was effective in evaluation on activation activity of the research reactor that had a lot of various experimental tubes. Total activation activity of reactor by the revaluation depended on control rods, thermal shield plates and horizontal experimental tubes, and the value after 1 year from the permanent shutdown of reactor was 1.9x1014 Bq. (author)

  19. Analysis of a possible experimental assessment of a prototype fuel element containing burnable poison in the RA-3 reactor

    International Nuclear Information System (INIS)

    The Argentine RA-3 research reactor (5 MW) is presently operated with LEU fuel by the National Atomic Energy Commission (CNEA). It belongs to the group of nuclear installations controlled, from the radiological and nuclear safety point of view, by the Nuclear Regulatory Authority (ARN). A new type of fuel elements containing burnable absorbers, with similar enrichment as the standard fuel elements but greater fissile contents, has recently been proposed for a new Argentine reactor design (RRR). In this framework the ARN considers interesting, if technically possible, the performance of an experiment in the RA-3 reactor. The experiment might enable, for such fuel element containing burnable poison, the verification of its neutronic behaviour under irradiation as well as a validation of the calculation line by comparison to measured values. It should be desirable that such experiment could reproduce as much as possible those conditions estimated for the RRR reactor, still under design in Argentina, having Silicide fuel elements with burnable poison, in the shape of cadmium wires in their structure. We here analyse a possible experiment consisting in the loading of a prototype fuel element with burnable poison in a normally loaded RA-3 core configuration. It would essentially be a standard RA-3 fuel element, having cadmium wires in its frame. This experiment would enable the verification of the prototype behaviour under irradiation, its operation limits and conditions, and particularly, the reactivity safety margins established in Argentine Standards, both calculated and measured. The main part of the experiment would imply some 200 full power days of operation at 5 MW, which would be drastically reduced if the reactor power is increased to 10 MW, as foreseen. We also show that under the proposed conditions, the experiment would not represent a significant penalty to the reactor normal operation. (author)

  20. Simplified 3D model of a PWR reactor vessel using fluid dynamics code ANSYS CFX computational; Modelo simplificado 3D de la vasija de un reactor PWR mediante el codigo de dinamica de fluidos computacional ANSYS CFX

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, M.; Miro, R.; Barrachina, T.; Verdu, G.

    2011-07-01

    This paper presents the results from the calculation of the steady state simulation with model of CFD (computational fluid dynamic) operating under conditions of operation at full power (Hot Full Power). Development and the CFD model results show the usefulness of these codes for calculating 3D of the variable thermohydraulics of these reactors.

  1. Modelling activities of experimental facilities related to advanced reactors. Considerations on 1D/3D issues

    International Nuclear Information System (INIS)

    The state of art of modelling activities related to integral experimental facilities of advanced passive reactors show to date important open items. The main advantage of using 1D plant codes is the capability of simulating the full interaction between components traditionally correctly modelled (condensers, heat exchangers, pipes and vessels) and other components for which codes are not 100% suitable (pools and containments). Polytechnical University of Catalonia (UPC) and Polytechnical University of Valencia (UPV) cooperated with other European research organizations in the 'Technology Enhancement for Passive Safety Systems' (TEPSS) project, within the European Fourth Framework Programme. It was a task of both Universities to supply analytical support of PANDA tests. The paper deals with the 1D/3D discussion in the framework of modelling activities related to integral passive facilities like PANDA. It starts choosing reference tests among those corresponding to our participation in TEPSS project. The discrepancies observed in a 1D simulation of the selected tests will be shown and analyzed. An evaluation of how the 3D version can lead to a better agreement with data will be included. Disadvantages of 3D codes will be shown too. Combining the use of different codes, and considering analyst criteria, will make possible to establish suitable recommendations from both engineering and scientific point of view. (author)

  2. 3 Investments Scenarios for Fast Reactors in Europe. 2012, November 30 - Preliminary Version

    International Nuclear Information System (INIS)

    The article aims at widening the scope of MIT report « The future of nuclear power after Fukushima » (2012) to the European electric supply by studying the particular conditions of industrial development of Fast Reactors. We specially focus on France, Germany, United Kingdom, Spain and Italy, from now up to 2040, for the technology is supposed to be available by then. These conditions can be either favorable or not to FRs, according to 3 main dynamically quantified drivers: ''technical change'', i.e. relative evolutions of efficiency and costs of available technologies (gas, coal, wind…); ''policy'', i.e. incentive framework given by European energy policies (nuclear, climate...); ''economic, i.e. structure of electricity markets (level of centralization...). A total of 24 scenarios are developed using an imaginative approach, i.e. assuming different possibilities for the future change according the latter 3 main drivers. 3 of them proving to be favorable to the FRs are then discussed in view of the quantitative drivers mentioned above and of an additional driver about public acceptance. (author)

  3. Safety Design Strategy for the Advanced Test Reactor Diesel Bus (E-3) and Switchgear Replacement Project

    Energy Technology Data Exchange (ETDEWEB)

    Noel Duckwitz

    2011-06-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  4. CAST3M/ARCTURUS: a coupled heat transfer / CFD code for thermal-hydraulic analyses of gas cooled reactors

    International Nuclear Information System (INIS)

    , heat exchangers...) are highly recommended. Nevertheless, in case of a total loss of station service power, the safety demonstration of the concept should be guaranteed by natural circulation decay heat removal. This could be performed by keeping a relatively high back-up pressure for pure He natural convection and also by heavy gas injection. So, it also necessary to compute the mixing of different gases and the on-set of natural convection. In this paper, we will report on the developments of the CAST3M/ARCTURUS thermal-hydraulics code developed at CEA, including its coupling to the neutronics code CRONOS2 and the system code CATHARE. Elementary validation cases will be detailed, as well as application of the code to benchmark problems such as the GT-MHR or the HTTR. Examples of thermal-hydraulic calculations decay heat removal of fast reactors designs (GCFR) will also be described. (authors)

  5. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

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

  6. Production and benchmark tests of fast reactor group constant set JFS-3-J2

    International Nuclear Information System (INIS)

    The JFS-3-J2 set is an advanced set for JENDL-2B-70, which was produced as the group constants of the JAERI-Fast set type. The JFS-3-J2 has been produced with the use of a processing code system TIMS-PGG. The characteristics of JFS-3-J2 are as follows. The group averaged cross sections are calculated by weighting with the collision density spectrum for core composition in a typical fast reactor, to correct the overestimate of elastic removal cross sections caused by using ''1/E-spectrum''. The weighting method of the collision density spectrum is called ''REMO-correction'', and produces harder neutron spectra than those calculated from ''1/E-spectrum'' method, especially for low energy range below about 1 keV. Hence, the nuclear characteristics such as Doppler reactivity coefficients sensitive to the flux shape in the low energy region are considerably affected by the harder spectra. The temperature dependent self-shielding factors for structural materials, Fe, Cr and Ni are calculated. The Doppler effect of structural materials is 0.24% δk/k for temperature change from 0 to 3000K in the ZPR-6 assembly 7. The scheme of self-shielding factor tables is corrected to obtain high accuracy of interpolation for f-tables by using cubic apline function. The self-shielding effect of inelastic scattering cross sections is considered. The neutron group to group transfer materices of elastic scattering are expanded from P0 to P3. In this report, the effects of ''REMO-correction'' considered in the generation of JFS-3-J2 on the nuclear characteristics are studied. The benchmark tests of JFS-3-J2 are performed and the results are discussed by being compared with those calculated with the JENDL-2B-70 set. (author)

  7. Analysis of a total flow blockage of a Fuel Assembly in a typical MTR Research Reactor by RELAP5/MOD3.3

    International Nuclear Information System (INIS)

    The lack of full understanding of complex mechanisms connected with the interaction between thermal-hydraulics and neutronics still challenge the design and the operation of nuclear reactors by the adoption of conservative safety limits. The recent availability of powerful computer and computational techniques together with the continuing increase in operational experience imposes the revisiting of those areas and eventually the identification of design/safety requirements that can be relaxed [1]. Currently, the enlarged commercial exploitation of nuclear Research Reactors (RR) has increased the consideration to their corresponding safety issues. Almost all of the safety analyses have so far been performed using conservative computational tools [2]. Nowadays, the application of Best-Estimate (BE) methods constitutes a real necessity in order to increase their commercial productivity. In this framework, an attempt is made to apply the BE technique to perform a safety evaluation under research reactors operational conditions. In fact, this technique has been largely verified and validated for power reactors using coupled system thermal-hydraulic and three-dimensional neutron kinetics [1]. For this purpose, as typical representative of research reactors, the IAEA 10 MW MTR Research Reactors problem [3] is considered. The system thermal-hydraulic RELAP5 [4] code was developed to simulate transient scenarios in Power reactors such PWR, BWR, VVER, etc. However, only limited work was performed to access the applicability of the code to Research Reactors operating conditions (low pressure, mass flow rates, power, etc) [5]. Previous works performed in this field are reported in [5], [6] and [7]. In this framework, total and partial blockage of a single Fuel Assembly cooling channel are investigated. As a first attempt the calculations are performed by applying the BE thermal-hydraulic system code RELAP5 alone using its point kinetic model to derive the instantaneous core

  8. Modeling the thermal–hydraulic behavior of the reactor cavity cooling system using RELAP5-3D

    International Nuclear Information System (INIS)

    Highlights: • The RCCS complex geometry and heat transfer mechanisms were modeled with RELAP5-3D. • Code limitations were overcome by applying special heat structures modeling techniques. • The simulation results were found to be in good agreement with the experimental data. • RELAP5-3D was found to be an adequate tool for analysis of HTGR components. - Abstract: The Very High Temperature Gas-Cooled Reactor (VHTR), one of the six proposed designs for the next generation nuclear reactor, was conceived to achieve high temperatures to support industrial applications and power generation. Due to the high temperature reached during normal operation, the design included new passive safety systems. The Reactor Cavity Cooling System (RCCS) is a new passive safety system designed to remove the heat from the reactor cavity during normal operation (steady-state) and accident scenarios. Computational tools such as system codes have been selected to simulate the reactor system and, in particular, the new safety components. The capabilities of these codes are being investigated to verify their ability in predicting the phenomena involved in the RCCS operation during steady-state and accident conditions. A RELAP5-3D input model of a small scale water-cooled experimental facility was prepared to simulate steady-state. The simulation results were compared with data produced during the experimental steady-state run. The results obtained and presented in the paper showed a good agreement of the code prediction with the experimental data. The paper also provides a set of modeling techniques to overcome some of the limitations of the current version of the computer code in simulating complex geometries with combined heat transfer mechanisms in the reactor cavity of the VHTR

  9. SAFIR-2 and the Belgian methodological R and D programme on deep disposal

    International Nuclear Information System (INIS)

    Peter de Preter (NIRAS/ONDRAF, Belgium) provided an overview of the Belgian programme of research and development on deep disposal. The planned submission in December 2001 of SAFIR 2 (the second Safety Assessment and Feasibility Interim Report) would mark an important milestone, as the report would inform a decision by the Belgian government on the nature of future research. An independent committee of scientists established by NIRAS/ONDRAF had reviewed a draft version of the report. The committee was generally in agreement with the technical R and D priorities proposed in the report but suggested that there should be more integration of technical and societal aspects. The committee also recommended that a future research programme should compare the option of deep disposal with other strategies for long-term management of radioactive wastes. It was suggested that a strategic environmental assessment might provide an appropriate mechanism for comparing alternative management strategies, and would enable societal dimensions also to be addressed

  10. TRIO a general computer code for reactor 3-D flows analysis. Application to a LMFBR hot plenum

    International Nuclear Information System (INIS)

    TRIO is a code developed at CEA to investigate general incompressible 2D and 3D viscous flows. Two calculations are presented: the lid driven cubic cavity at Re=400; steady state (velocity and temperature field) of a LMFBR hot plenum, carried out in order to prepare the calculation of a cold shock consecutive to a reactor scram. 8 refs., 26 figs.

  11. Production of Carbon Nanotubes over Pre-reduced LaCoO3 Using Fluidized-bed Reactor

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    Carbon nanotubes were synthesized on a large scale by the catalytic decomposition of hydrocarbons over pre-reduced LaCoO3 using a fluidized-bed reactor. Reaction parameters such as reduction temperature, reduction time and reaction temperature were discussed.

  12. Comparison and performance analysis of the W-3 and EPRI correlations for critical heat flux in PWR reactors

    International Nuclear Information System (INIS)

    The present work presents a comparison between the W 3 and EPRI correlations for critical heat flux. Experimental data were used in order to verify the behavior of the above-mentioned correlations. The conclusions presented in this work allow a better definition of the correlations for the operating safety limits calculations of a PWR type reactor. (author)

  13. Sensitivity analyses of thermal bridges: confrontation with the new Belgian EPB-methodology

    OpenAIRE

    Delghust, Marc; Huyghe, Willem; Janssens, Arnold

    2011-01-01

    As governments continue to impose more and higher energetic requirements for buildings, they also need better assessment-tools to take into account as many parameters as possible. This results in continuous developments of new calculation methods and softwares, where a balance has to be found between practicality and accuracy. To answer this problem, specifically with regard to the thermal bridges, the three Belgian regions developed a new and common pragmatic approach for assessing therma...

  14. The Gay Men Sex Studies: prevalence of sexual dysfunctions in Belgian HIV+ gay men

    OpenAIRE

    Vansintejan J; Janssen J; Van De Vijver E; Vandevoorde J; Devroey D

    2013-01-01

    Johan Vansintejan, Joris Janssen, Erwin Van De Vijver, Jan Vandevoorde, Dirk Devroey Department of Family Medicine, Vrije Universiteit Brussel (VUB), Brussels, Belgium Abstract: The aim of this Internet-based survey was to investigate the prevalence and associated predictors of sexual dysfunctions in Belgian self-reported HIV-positive men who have sex with other men. Of the 72 participants, 56% had a mild-to-severe erectile dysfunction, and 15% reported a hypoactive sexual desire disorder. Th...

  15. The Gay Men Sex Studies: prevalence of sexual dysfunctions in Belgian HIV+ gay men

    OpenAIRE

    Vansintejan, Johan

    2013-01-01

    Johan Vansintejan, Joris Janssen, Erwin Van De Vijver, Jan Vandevoorde, Dirk Devroey Department of Family Medicine, Vrije Universiteit Brussel (VUB), Brussels, Belgium Abstract: The aim of this Internet-based survey was to investigate the prevalence and associated predictors of sexual dysfunctions in Belgian self-reported HIV-positive men who have sex with other men. Of the 72 participants, 56% had a mild-to-severe erectile dysfunction, and 15% reported a hypoactive sexual desire disorder. T...

  16. Average daily nitrate and nitrite intake in the Belgian population older than 15 years

    OpenAIRE

    Temme, Liesbeth; Vandevijvere, Stefanie Marie; Vinkx, Christine; Huybrechts, Inge; Goeyens, Leo; Van Oyen, Herman

    2011-01-01

    Abstract The aim of this study was to assess the dietary intake of nitrate and nitrite in Belgium. The nitrate content of processed vegetables, cheeses and meat products was analyzed. These data were completed by data from non-targeted official control and from literature. In addition, the nitrite content of meat products was measured. Concentration data for nitrate and nitrite were linked to food consumption data of the Belgian Food Consumption Survey. This study included 3245 res...

  17. Estimating an Ex Ante Cost Function for Belgian Arable Crop Farms

    OpenAIRE

    Hansen, Kristiana; Baudry, Alexandre; De Blander, Rembert; Frahan, Bruno Henry de; Polome, Philippe

    2009-01-01

    We estimate a farm-level cost function for Belgian crop farms using FADN data over the study period 1996-2006. We rely on an estimation of farmers' expected yields at the time cropping decisions are made rather than actual yields observed in the FADN data. The use of an ex ante cost function improves the cost function estimation. We subsequently suggest how our cost function can be used in simulations to analyze farmer response to changes in output price risk.

  18. Waste disposal[1997 Scientific Report of the Belgian Nuclear Research Centre

    Energy Technology Data Exchange (ETDEWEB)

    Neerdael, B.; Marivoet, J.; Put, M.; Verstricht, J.; Van Iseghem, P.; Buyens, M.

    1998-07-01

    The primary mission of the Waste Disposal programme at the Belgian Nuclear Research Centre SCK/CEN is to propose, develop, and assess solutions for the safe disposal of radioactive waste. In Belgium, deep geological burial in clay is the primary option for the disposal of High-Level Waste and spent nuclear fuel. The main achievements during 1997 in the following domains are described: performance assessment, characterization of the geosphere, characterization of the waste, migration processes, underground infrastructure.

  19. Use of Information, Product Innovation and Financial Performance on Belgian Glasshouse Holdings

    OpenAIRE

    Taragola, Nicole; Huylenbroeck, Guido Van; Van Lierde, Dirk

    2002-01-01

    In order to meet the changing needs and preferences of consumers it will be important for Belgian glasshouse growers to change from a production-driven to a customer-driven strategy. More than ever, use of information and product innovation become critical factors in the changing competitive environment. The aim of the research is to analyse the relationship between business and managerial characteristics, use of information sources, product innovation and financial performance of the firm. T...

  20. Mafia Practices and Italian Entrepreneurial Activities in the Belgian Food Sector. Research Objectives

    OpenAIRE

    De Biase, Marco

    2014-01-01

    This research aims to study the social and economic processes which encourage the spread of mafia practices among some Italian entrepreneurs in the Belgian food sector. My working hypothesis is that mafia practices are violent and predatory strategies to monopolize the economic market, which rely on the use of heterogeneous and cross-class social networks. In this view, these activities of old and new Italian entrepreneurs in Belgium are not the result of the exportation of mafia methods thro...

  1. Geochemical behavior of trace elements in sub-tidal marine sediments of the Belgian coast

    OpenAIRE

    Gao, Y.; Lesven, L.; Gillan, D.; K Sabbe; Billon, G.; De Galan, S.; Elskens, M.; Baeyens, W.; Leermakers, M.

    2009-01-01

    High resolution profiles of trace elements (Fe, Mn, Co, As, Cu, Cr, Ni and Pb) were assessed using the DET (Diffusive Equilibrium in Thin films) and DGT (Diffusive Gradients in Thin films) techniques in silty, organically enriched, sub-tidal sediments of the Belgian coast during late winter and spring 2008. The general chemical properties of the sediments such as dissolved oxygen, pH, Eh and sulfide profiles, controlling precipitation/ mobilization reactions, were determined with electrodes (...

  2. New Belgian Law on Research on Human Embryos: Trust in Progress Through Medical Science

    OpenAIRE

    Pennings, G

    2003-01-01

    The new Belgian law on research on embryos in vitro accepts all types of research directed at therapeutic purposes and at increased medical knowledge. This includes research for germline and somatic gene therapy, therapeutic cloning, and the development of embryonic stem cell lines. As this presupposes the creation of embryos for research, this too is allowed. Other goals like sex selection for nonmedical reasons, eugenic practices and reproductive cloning are prohibited. In general, the law ...

  3. Recent changes in saving behaviour by Belgian households : the impact of uncertainty

    OpenAIRE

    Raïsa Basselier; Geert Langenus

    2014-01-01

    Belgian households save a lot in comparison to the average for the euro area. Recently, the household saving rate has shown large swings. During the great recession, it temporarily jumped to a record high but since then it has gradually fallen to record lows. Different factors of a structural or cyclical nature may account for these swings. The objective of the article is to gauge to what extent uncertainty related to the general economic outlook or households’ income prospects has contribute...

  4. Perceived work ability and turnover intentions: a prospective study among Belgian healthcare workers

    OpenAIRE

    Derycke, Hanne; Clays, Els; Vlerick, Peter; D'hoore, William; Hasselhorn, Hans Martin; Braeckman, Lutgart

    2012-01-01

    AIM: To report a study exploring prospective relations between nurses' perceived work ability and three forms of turnover intentions, respectively, intent to leave the ward, organization and profession. BACKGROUND: Turnover of nursing staff is a major challenge for healthcare settings and for healthcare in general, urging the need to improve retention. DESIGN: Survey. METHODS: Based on the longitudinal data of the Belgian sample from the European Nurses' Early Exit study, a total of 1531 heal...

  5. Signs of neurotoxicity in a Belgian Blue herd after ingestion of moulded silage

    OpenAIRE

    Guyot, Hugues; Sandersen, Charlotte; Brihoum, Mounir; Vandeputte, Sébastien; Rollin, Frédéric

    2011-01-01

    After ingestion of moulded beet pulp silage, cases of cerebro-cortical necrosis (CCN) and mortalities were observed in a dual purpose Belgian Blue (BB) herd. Contamination with Paecilomyces spp., a mould that produces byssochlamic acid, malformins and patulin, was proven. Twenty-five days after progressive introduction of beet pulp silage into the ration, most of the animals showed diminished appetite, excessive salivation and decreased milk production. Some of them showed a...

  6. Adaptation of the European crop growth monitoring system to the Belgian conditions.

    OpenAIRE

    Buffet, D.; Dehem, Didier; Wouters, K.; Tychon, Bernard; Oger, Robert; Veroustraete, F.

    1999-01-01

    The aim of the Belgian Crop Growth Monitoring System (B-CGMS) is the elaboration of an integrated information system predicting reliable, timely and objective estimates of crop yields and monitoring calamity sites at regional scales. Seven major crops are concerned by the project : winter wheat, winter barley, fodder maize, winter rape seed, potatoes, sugar beet and permanent meadow. The main tasks in the adaptation of the European model come down to the completion and the improvement of the ...

  7. An empirical study on human resource planning in Belgian production companies

    OpenAIRE

    Van den Bergh, Jorne; Belien, Jeroen; Hoskens, Brent

    2013-01-01

    This paper investigates human resource planning in Belgian production companies. First, a literature study is developed to serve as a basis for the results of the empirical research. The literature study is mainly based on papers in the field of operations research that provide interesting insights such as the research-application gap, which is the lack of implementation of models provided by literature. The most important part of this paper is the empirical research. The empirical research i...

  8. Salt water infiltration in two artificial sea inlets in the Belgian dune area

    OpenAIRE

    Vandenbohede, A.; Lebbe, L.; Gysens, S.; Delecluyse, K.; DeWolf, P.

    2008-01-01

    In the dune area of the Westhoek Nature Reserve, situated in the western Belgian coastal plain, two artificial tidal inlets were made aiming to enhance biodiversity. The infiltration of salt water in these tidal inlets was carefully monitored because a fresh water lens is present in the phreatic dune aquifer. This forms an important source of fresh water which is for instance exploited by a water company. The infiltration was monitored over a period of two years by means of electromagnetic bo...

  9. Psychological dimensions of unemployment: a gender comparison between Belgian and South African unemployed.

    OpenAIRE

    Yannick Griep; Sebastiaan Rothmann; Wouter Vleugels; Hans De Witte

    2012-01-01

    This study sought to compare South African and Belgian unemployed in their subjective experience of unemployment, committed towards employment and job search behaviour. We also considered gender differences regarding the psychological dimensions of unemployment between Belgium and South Africa. A cross-sectional survey design was used. Unemployed people were sampled from the Potchefstroom area in South Africa (N = 381) and the Brussels area in Belgium (N = 305). The Experiences of Unemploymen...

  10. Information Availability, Information Quality and the Financial Structure of Belgian SME's.

    OpenAIRE

    Van Campenhout, Geert; Van Caneghem, Tom

    2009-01-01

    In this paper we test whether the amount and/or quality of financial statement information affect the financial structure of Small and Medium Enterprises (SMEs). We explore this issue for Belgian SMEs because there are important differences in disclosure and audit requirements among them. Consistent with the traditional view that asymmetric or incomplete information restricts access to external funds, our results indicate that both the amount and the quality of financial statement information...

  11. Long-term changes in oil pollution off the Belgian coast: evidence from beached bird monitoring

    OpenAIRE

    Seys, J.; Offringa, H; P. Meire; Van Waeyenberge, J.; Kuijken, E.

    2002-01-01

    Trends in oil pollution in the southernmost (Belgian) part of the North Sea were analysed using a dataset of 37 years (1962-99) of annual national beached bird surveys conducted in February each year. The most abundant seabird groups represented in the beached birds were auks (31 %), gulls (28%), scoters (17%) and Kittiwake (9%). Oil rates of most bird species/taxa indicate a decline in oil pollution, though only Larus-gulls, Common Guillemot and Razorbill show significant reductions. The slo...

  12. Software support for manufacturing operations in Belgian SMEs: one size fits all?

    OpenAIRE

    Desmarey, Thierry; Degryse, Kris; Cottyn, Johannes

    2011-01-01

    Manufacturing companies face a big challenge to bridge the gap between their business and manufacturing processes. The urge to increase efficiency makes it necessary to align the business and manufacturing processes. Small and Medium-sized Enterprises (SMEs) experience several barriers to adopt software support for manufacturing operations. This paper gives an overview of a research study conducted in Belgian SMEs. The research studied the current adoption of software support for manufacturin...

  13. Sustainability of Global and Local Food Value Chains: An Empirical Comparison of Peruvian and Belgian Asparagus

    OpenAIRE

    Jana Schwarz; Monica Schuster; Bernd Annaert; Miet Maertens; Erik Mathijs

    2016-01-01

    The sustainability of food value chains is an increasing concern for consumers, food companies and policy-makers. Global food chains are often perceived to be less sustainable than local food chains. Yet, thorough food chain analyses and comparisons of different food chains across sustainability dimensions are rare. In this article we analyze the local Belgian and global Peruvian asparagus value chains and explore their sustainability performance. A range of indicators linked to environmental...

  14. Floating seaweed in the neustonic environment: a case study from Belgian coastal waters

    OpenAIRE

    Vandendriessche, S; Vincx, M.; Degraer, S.

    2007-01-01

    Floating seaweeds form the most important natural component of all floating material found on the surface of oceans and seas. Notwithstanding the absence of natural rocky shores, ephemeral floating seaweed clumps are frequently encountered along the Belgian coast. From October 2002 to April 2003, seaweed samples and control samples (i.e. surface water samples from a seaweed-free area) were collected every other week. Multivariate analysis on neustonic macrofaunal abundances showed significant...

  15. Using Firm-Level Data to Assess Gender Wage Discrimination in the Belgian Labour Market

    OpenAIRE

    Borowczyk Martins, Daniel; Vandenberghe, Vincent

    2010-01-01

    In this paper we explore a matched employer-employee data set to investigate the presence of gender wage discrimination in the Belgian private economy labour market. We identify and measure gender wage discrimination from firm-level data using a labour index decomposition pioneered by Hellerstein and Neumark (1995), which allows us to compare direct estimates of a gender productivity differential with those of a gender labour costs differential. We take advantage of the panel structure of the...

  16. Firm-level Evidence on Gender Wage Discrimination in the Belgian Private Economy

    OpenAIRE

    Vandenberghe, Vincent

    2011-01-01

    In this paper we explore a matched employer-employee data set to investigate the presence of gender wage discrimination in the Belgian private economy labour market. Contrary to many existing papers, we analyse gender wage discrimination using an independent productivity measure. Using firm-level data, we are able to compare direct estimates of a gender productivity differential with those of a gender wage differential. We take advantage of the panel structure to identify gender-related diffe...

  17. Grid Integration of Large-scale Wind Turbines Equipped With Full Converters: Belgian Case Study

    OpenAIRE

    De Rijcke, Simon; Ergun, Hakan; De Vos, Kristof; Driesen, Johan

    2011-01-01

    This paper describes the results of a study regarding the integration of wind power into the Belgian electricity system. The main focus of this study is the contribution of wind power plants to reduce voltage deviations due to load and wind power variations. It is shown, how wind power plants can maintain and improve the voltage in the existing electricity network using their reactive power operation characteristics. To cover a broad variety of study cases, two different locations with differ...

  18. Development of Nb3Sn based multi-filamentary superconductor wires for fusion reactor magnets

    International Nuclear Information System (INIS)

    Nb3Sn is a proposed type II superconductor material to be used as superconducting magnet in fusion reactor for its superior superconducting properties. Fabrication of long single length wire containing Nb3Sn filaments is a challenge. The usual manufacturing philosophy involves deforming an assembly of tin and niobium in copper matrix to the final size, followed by the heat treatment to produce superconducting phase at Nb-Cu interface. Multi-filamentary wires were fabricated by hot extrusion of superconductor billet followed by several stages of cold drawing. Heat treatments at various temperature and time were carried out on as formed wire containing multiple filaments in order to see the growth of superconducting intermetallic phase during subsequent characterization. Post heat treatment characterization through SEM, EBSD and EDS revealed the presence of intermetallic phase of Nb and Sn, hypo stoichiometric in Sn, at the Cu-Nb interface growing towards the center of Nb filament. The manufacturing process till the desired final size of the wire happened to be a challenge, mainly because it required extraordinary co-deformability between various materials in such an assembly. Post-trial failure analysis through destructive testing using optical and scanning electron micrographs revealed the propensity of internal radial cracks at Cu-Sn interfaces, while the Nb-Cu interfaces were found to be relatively unaffected. This paper will discuss the details of the fabrication process. (author)

  19. Reactor safety issues resolved by the 2D/3D program

    International Nuclear Information System (INIS)

    The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated. This report was prepared in a coordination among US, Germany and Japan. US and Germany have published the report as NUREG/IA-0127 and GRS-101 respectively. (author)

  20. Recommendations of the Reactor Safety Commission (RSK) 1975-1977. Vol. 3

    International Nuclear Information System (INIS)

    After the recommendations are published in the 'Bundesanzeiger' the Office of the RSK publishes them as a closed report, by order of the Federal Minister of the Interior. The reports are devided in two parts: Part I contains the recommendations which are given by the RSK; part II contains the official notices concerning the RSK. There also is a subject index. Three volumes are published by now: Volume 1: IRS-A-9 (December 1975), Recommendations of the RSK 1971 to 1974 (68th to 96th meeting), pages: I-1 to I-116 and II-1 to II-23 appendix a: RSK-guide lines for PWR, edition 04.74; Volume 2: IRS-A-11 (August 1976), Recommendations of the RSK 1974 to 1975, (97th to 105th meeting), pages: I-117 to I-145; Volume 3: GRS-12 (August 1978), Recommendations of the RSK 1975 to 1977 (106th to 129th meeting), pages: I-146 to I-276 and II-24 to II-25. Volume 3 also contains the recommendation on German Fuel Cycle Center which was given by the Reactor Safety Commision (RSK) in community with the Radiological Protection Commission (SSK), in autumn 1977. (orig./HP) 891 HP

  1. Detection of the contamination of air by tritiated water vapour around the reactor EL3

    International Nuclear Information System (INIS)

    The authors describe the apparatus used for the detection of the tritiated water vapour contamination in the air around the reactor EL 3. The apparatus consists of two air-circulation ionisation chambers; the air in one of these is dried by passage through a silica-gel column. By carrying out a differential measurement of the ionization currents, it is possible to measure the tritiated water vapour concentration. A theoretical study of the response of the chambers is carried out for two types of emission of the tritiated water vapour: continuous, or in bursts. The experimental work comprises: calibration in the measurement range employed; study of the selectivity for other active gases; study of typical accidents; the interpretation of the results in the case of discontinuous emission, taking into account the desorption from the walls of the measurement chamber, a phenomenon which is observed during the emptying process. The authors give finally actual examples of how to use the results. The apparatus built makes it possible to detect, in less than ten minutes, contamination by tritiated water vapour in the presence of other active gases, in a measurement range of between 3 and 2200 MPC, and with an accuracy of about 25 per cent. A transposition to calculations of the risk to workers should be made with the utmost caution; an envelope of this risk can be drawn up more or less accurately depending on particular cases. (authors)

  2. Boiling Water Reactor Turbine Trip (TT) Benchmark - Vol. IV Volume IV: Summary Results of Exercise 3

    International Nuclear Information System (INIS)

    In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts, as well as for current applications. Recently developed 'best-estimate' computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for coupling of the core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for that purpose. The present volume is the last in a series of four and summarises the results of the third benchmark exercise, which analyses a turbine trip (TT) in a BWR in its entirety, involving pressurization events in which the coupling between core phenomena and system dynamics plays an important role. Exercise 3 also analyses four extreme scenarios which allowed participants to test the capabilities of their code(s) in terms of coupling and feedback modelling. The data made available from experiments carried out at the plant make the present benchmark particularly valuable. The data used are from events at the Peach Bottom 2 reactor (a GE-designed BWR/4). (authors)

  3. Reactor safety issues resolved by the 2D/3D Program

    International Nuclear Information System (INIS)

    The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated

  4. Simulation of a channel blockage transient in the Angra 2 Nuclear Reactor using a RELAP5-3D model

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez-Mantecon, Javier; Costa, Antonella L.; Veloso, Maria Auxiliadora F.; Pereira, Claubia; Reis, Patricia A.L.; Scari, Maria E., E-mail: mantecon1987@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: patricialire@yahoo.com.br, E-mail: melizabethscari@yahoo.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Escola de Engenharia. Departamento de Engenharia Nuclear

    2015-07-01

    The Angra 2 Nuclear Power Plant (NPP) is a Pressurized Water Reactor (PWR) type with electrical output of about 1350 MW. The RELAP5-3D code was used to develop a detailed thermal hydraulic model of such reactor using reference data from the Angra 2 Final Safety Analysis Report (FSAR). In this work, a blockage transient has been investigated at full power operation. The transient herein considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)

  5. RV Belgica II: The new Belgian research vessel to replace the existing RV A962 Belgica

    OpenAIRE

    Naudts, L.; Cox, D.; Roose, P.; Monteny, F.

    2014-01-01

    For 30 years RV Belgica has been the marine infrastructure for Belgian marine scientists working on the Belgian Part of the North Sea and in the marine realm stretching from Norway, over Ireland to Morocco. Even if the ship is still performing research activities for more than 180 days per year, the increasing number of technical breakdowns and the rising maintenance cost associated with her age makes the replacement by a new RV Belgica II a necessity for the Belgian marine science community.

  6. Microbial diversity and metabolite composition of Belgian red-brown acidic ales.

    Science.gov (United States)

    Snauwaert, Isabel; Roels, Sanne P; Van Nieuwerburg, Filip; Van Landschoot, Anita; De Vuyst, Luc; Vandamme, Peter

    2016-03-16

    Belgian red-brown acidic ales are sour and alcoholic fermented beers, which are produced by mixed-culture fermentation and blending. The brews are aged in oak barrels for about two years, after which mature beer is blended with young, non-aged beer to obtain the end-products. The present study evaluated the microbial community diversity of Belgian red-brown acidic ales at the end of the maturation phase of three subsequent brews of three different breweries. The microbial diversity was compared with the metabolite composition of the brews at the end of the maturation phase. Therefore, mature brew samples were subjected to 454 pyrosequencing of the 16S rRNA gene (bacteria) and the internal transcribed spacer region (yeasts) and a broad range of metabolites was quantified. The most important microbial species present in the Belgian red-brown acidic ales investigated were Pediococcus damnosus, Dekkera bruxellensis, and Acetobacter pasteurianus. In addition, this culture-independent analysis revealed operational taxonomic units that were assigned to an unclassified fungal community member, Candida, and Lactobacillus. The main metabolites present in the brew samples were L-lactic acid, D-lactic acid, and ethanol, whereas acetic acid was produced in lower quantities. The most prevailing aroma compounds were ethyl acetate, isoamyl acetate, ethyl hexanoate, and ethyl octanoate, which might be of impact on the aroma of the end-products. PMID:26802571

  7. Last developments in the Belgian disposal programme for low and intermediate short-lived waste

    International Nuclear Information System (INIS)

    After an historical reminder of the several phases of the Belgian program for the disposal of low and medium level short-lived waste since the creation of ONDRAF/NIRAS and the bad results obtained in the 90's by using a pure technical approach, the presentation will explain the main lines of the new methodology developed, as a consequence of the government decision of 16 January 1998 in ONDRAF/NIRAS to improve local acceptance for the disposal project. The way local partnerships were created with four nuclear municipalities under the form of a non-profit organization with a clear mission, the functioning, on a voluntary base, of the different partnerships during four to six years and the concrete results obtained until now using this very innovative method will be addressed. The last developments of the Belgian program for the disposal of low and medium level and short-lived waste will be presented, including the recent and very important decision of the Belgian government of 23 June 2006 to dispose of the low and medium active short-lived waste in a surface disposal installation on the territory of the municipality Dessel. (author)

  8. Air crew exposure on board of long-haul flights of the Belgian airlines

    International Nuclear Information System (INIS)

    New European radiation protection recommendations state that measures need to be taken for flight crew members whose annual radiation exposure exceeds 1 mSv. This will be the case for flight crew members who accumulate most of their flying hours on long-haul flights. The Recommendations for the Implementation of the Basic Safety Standards Directive states that for annual exposure levels between 1 and 6 mSv individual dose estimates should be obtained, whereas for annual exposures exceeding 6 mSv, which might rarely occur, record keeping with appropriate medical surveillance is recommended. To establish the exposure level of Belgian air crews, radiation measurements were performed on board of a total of 44 long-haul flights of the Belgian airlines. The contribution of low linear energy transfer (LET) radiation (photons, electrons, protons) was assessed by using TLD-700H detectors. The exposure to high-LET radiation (mostly neutrons) was measured with bubble detectors. Results were compared to calculations with an adapted version of the computer code CARI. For the low-LET radiation the calculations were found to be in good agreement with the measurements. The measurements of the neutron dose were consistently lower than the calculations. With the current flight schedules used by the Belgian airlines, air crew members are unlikely to receive annual doses exceeding 4 mSv. (author)

  9. Distribution of doses resulting from cosmic rays exposure for Belgian airlines

    International Nuclear Information System (INIS)

    The Belgian Radiation Protection Act of 2001 requires air line companies registered in Belgium to evaluate the doses resulting from the exposure of their crew to cosmic radiation. If the annual dose of 1 mSv is exceeded, the crew must be informed about their individual doses and pregnant women have to be protected, among other measures. The Federal Agency for Nuclear Control (FANC, the competent Belgian radiation protection authority) has issued guidelines in order to help air line companies to fulfil their duties. Following the publication of these guidelines, all commercial Belgian air line companies have sent data on the exposure of their personnel. These data and, in particular, the distribution of the doses are presented. Except in the cases of small 'air taxi' companies, which fly only on very short distances and low altitudes, a significant number of air crew members gets doses of more than 1 mSv/y. The maximum value amounts to ∼ 4 mSv/y. The computer codes used by the companies in order to evaluate the individual doses are PCAIRE, CARI and IASON-FREE. The FANC imposed the concerned companies to reassess yearly the individual doses. (author)

  10. Predicting the environmental risks of radioactive discharges from Belgian nuclear power plants

    International Nuclear Information System (INIS)

    An environmental risk assessment (ERA) was performed to evaluate the impact on non-human biota from liquid and atmospheric radioactive discharges by the Belgian Nuclear Power Plants (NPP) of Doel and Tihange. For both sites, characterisation of the source term and wildlife population around the NPPs was provided, whereupon the selection of reference organisms and the general approach taken for the environmental risk assessment was established. A deterministic risk assessment for aquatic and terrestrial ecosystems was performed using the ERICA assessment tool and applying the ERICA screening value of 10 μGy h−1. The study was performed for the radioactive discharge limits and for the actual releases (maxima and averages over the period 1999–2008 or 2000–2009). It is concluded that the current discharge limits for the Belgian NPPs considered do not result in significant risks to the aquatic and terrestrial environment and that the actual discharges, which are a fraction of the release limits, are unlikely to harm the environment. -- Highlights: • Impact of radioactive discharges by the Belgian NPPs of Doel and Tihange on wildlife was evaluated. • Deterministic risk assessment for aquatic and terrestrial ecosystems performed with the ERICA tool. • NPP discharge limits do not result in significant risks to the aquatic and terrestrial environment. • Actual discharges, a fraction of the release limits, are unlikely to harm the environment

  11. Belgian canine population and purebred study for forensics by improved mitochondrial DNA sequencing.

    Science.gov (United States)

    Desmyter, Stijn; Gijsbers, Leonie

    2012-01-01

    In canine population studies for forensics, the mitochondrial DNA is profiled by sequencing the two hyper variable regions, HV1 and HV2 of the control region. In a first effort to create a Belgian population database some samples showed partially poor sequence quality. We demonstrated that a nuclear pseudogene was co-amplified with the mtDNA control region. Using a new combination of primers this adverse result was no longer observed and sequencing quality was improved. All former samples with poor sequence data were reanalyzed. Furthermore, the forensic canine population study was extended to 208 breed and mixed dogs. In total, 58 haplotypes were identified, resulting in an exclusion capacity of 0.92. The profile distribution of the Belgian population sample was not significantly different from those observed in population studies of three other countries. In addition to the total population study 107 Belgian registered pedigree dogs of six breeds were profiled. Per breed, the obtained haplotypes were supplemented with those from population and purebred studies. The combined data revealed that some haplotypes were more or less prominent present in particular dog breeds. The statistically significant differences in haplotype distribution between breeds and population sample can have consequences on mtDNA databasing and matching probabilities in forensics. PMID:21489897

  12. A combined 1D/3D fuel burnup analysis of generation IV light water reactor IRIS

    International Nuclear Information System (INIS)

    A combined 1D/3D methodology for the fuel burnup analysis of generation IV light water reactors with thin boron coating that covers the fuel rods is described in this paper. This methodology is founded on three approximations. The first approximation assumes that the problem of fuel depletion in the entire 3D core can be resolved into two independent problems. One is a 3D Monte Carlo evolution of power distribution in large volumes (nodes) with the KENO-V.a code, and the other is a transport method evolution of burnup dependent fuel composition in 1D Wigner-Seitz cell for each node independently. With the second approximation, the time-dependent fuel composition in the node (e.g., in the fuel assembly) is calculated by using a 1D fuel depletion analysis with the SAS2H control module from the SCALE-4.4a code system. The third approximation involves smearing the boron coating with the clad (by volume homogenization). The proposed SAS2H/KENO-V.a methodology is verified for the case of 2D x-y model of IRIS 15x15 fuel assembly (with a reflective boundary condition) by using two well benchmarked code systems. The first one is MOCUP, a coupled MCNP-4C and ORIGEN2.1 utility code, and the second is KENO-V.a/ORIGEN2.1 code system recently developed by authors of this paper. It has been found that the proposed SAS2H/KENO-V.a methodology gives a satisfactory accuracy for keff and nuclide composition. Finally, this methodology was applied for 3D burnup analysis of IRIS-1000 benchmark≠44 core. Detailed keff and power density evolution with burnup are reported. (author)

  13. A 3D model of a reverse vortex flow gliding arc reactor

    Science.gov (United States)

    Trenchev, G.; Kolev, St.; Bogaerts, A.

    2016-06-01

    In this computational study, a gliding arc plasma reactor with a reverse-vortex flow stabilization is modelled for the first time by a fluid plasma description. The plasma reactor operates with argon gas at atmospheric pressure. The gas flow is simulated using the k-ε Reynolds-averaged Navier–Stokes turbulent model. A quasi-neutral fluid plasma model is used for computing the plasma properties. The plasma arc movement in the reactor is observed, and the results for the gas flow, electrical characteristics, plasma density, electron temperature, and gas temperature are analyzed.

  14. Generation of a library for reactor calculations and some applications in core and safety parameter studies of the 3-MW TRIGA MARK-II research reactor

    International Nuclear Information System (INIS)

    This paper reports on a data base of the TRIGAP code that is generated for the 3-MW TRIGA MARK-II research reactor in Bangladesh. The library is created using the WIMS-D/4 code. Cross sections are calculated from zero burnup to 37% of initial 235U in 20 burnup steps. The created TRIGAP library is tested through practical calculations and is compared with experimental values or with values in the safety analysis report (SAR). Excess reactivity of the fresh core configuration is measured and determined to be 10.27 dollars, while a value of 10.267 dollars is obtained using the generated library. By choosing burnup steps of 0, 50, 350, and 750, WM · h, the whole operating history is covered. The calculated temperature defect at 1 and 3 MW is 1.15 and 3.59 dollars compared with the experimental value of 1.02 and 3.64 dollars, respectively. The xenon value obtained at 1 and 3 MW is 2.21 and 3.20 dollars, respectively, compared with 3.57 dollars at 3 MW in the SAR. The TRIGAP code with its new library is used for calculating fast and thermal flux distributions close to values from the SAR

  15. Characterization of mandibular molar root and canal morphology using cone beam computed tomography and its variability in Belgian and Chilean population samples

    Energy Technology Data Exchange (ETDEWEB)

    Torres, Andres; Jacobs, Reinhilde; Lambrechts, Paul [Katholieke Universiteit Leuven, Leuven (Belgium); Brizuela, Claudia; Cabrera, Carolina; Concha, Guillermo; Pedemonte, Maria Eugenia [Universidad de los Andes, Santiago (Chile)

    2015-06-15

    This study used cone-beam computed tomography (CBCT) to characterize mandibular molar root and canal morphology and its variability in Belgian and Chilean population samples. We analyzed the CBCT images of 515 mandibular molars (257 from Belgium and 258 from Chile). Molars meeting the inclusion criteria were analyzed to determine (1) the number of roots; (2) the root canal configuration; (3) the presence of a curved canal in the cross-sectional image of the distal root in the mandibular first molar and (4) the presence of a C-shaped canal in the second mandibular molar. A descriptive analysis was performed. The association between national origin and the presence of a curved or C-shaped canal was evaluated using the chi-squared test. The most common configurations in the mesial root of both molars were type V and type III. In the distal root, type I canal configuration was the most common. Curvature in the cross-sectional image was found in 25% of the distal canals of the mandibular first molars in the Belgian population, compared to 11% in the Chilean population. The prevalence of C-shaped canals was 10% or less in both populations. In cases of unclear or complex root and canal morphology in the mandibular molars, CBCT imaging might assist endodontic specialists in making an accurate diagnosis and in treatment planning.

  16. Characterization of mandibular molar root and canal morphology using cone beam computed tomography and its variability in Belgian and Chilean population samples

    International Nuclear Information System (INIS)

    This study used cone-beam computed tomography (CBCT) to characterize mandibular molar root and canal morphology and its variability in Belgian and Chilean population samples. We analyzed the CBCT images of 515 mandibular molars (257 from Belgium and 258 from Chile). Molars meeting the inclusion criteria were analyzed to determine (1) the number of roots; (2) the root canal configuration; (3) the presence of a curved canal in the cross-sectional image of the distal root in the mandibular first molar and (4) the presence of a C-shaped canal in the second mandibular molar. A descriptive analysis was performed. The association between national origin and the presence of a curved or C-shaped canal was evaluated using the chi-squared test. The most common configurations in the mesial root of both molars were type V and type III. In the distal root, type I canal configuration was the most common. Curvature in the cross-sectional image was found in 25% of the distal canals of the mandibular first molars in the Belgian population, compared to 11% in the Chilean population. The prevalence of C-shaped canals was 10% or less in both populations. In cases of unclear or complex root and canal morphology in the mandibular molars, CBCT imaging might assist endodontic specialists in making an accurate diagnosis and in treatment planning

  17. Effect of Co3O4 and CeO2 Infiltration on the Activity of a LSM15/GDC10 Highly Porous Electrochemical Reactor

    DEFF Research Database (Denmark)

    Ippolito, Davide; Kammer Hansen, Kent

    2014-01-01

    VOC component of Diesel engine exhausts, over a wide range of temperatures. The entire reactor was thought as a highly porous catalytic filter for a possible application in a Diesel exhausts purification system. The porous reactor was used as a backbone for the infiltration of Co3O4 and Co3O4/CeO2...

  18. 3-D kinetics simulations of the NRU reactor using the DONJON code

    International Nuclear Information System (INIS)

    The NRU reactor is highly heterogeneous, heavy-water cooled and moderated, with online refuelling capability. It is licensed to operate at a maximum power of 135 MW, with a peak thermal flux of approximately 4.0 x 1018 n.m-2 . s-1. In support of the safe operation of NRU, three-dimensional kinetics calculations for reactor transients have been performed using the DONJON code. The code was initially designed to perform space-time kinetics calculations for the CANDUR power reactors. This paper describes how the DONJON code can be applied to perform neutronic simulations for the analysis of reactor transients in NRU, and presents calculation results for some transients. (authors)

  19. FLOW3D model for below-core thermal mixing in the Oconee pressurised water reactor

    International Nuclear Information System (INIS)

    The computational fluid dynamics code FLOW3D is being used to develop a model for calculating the mixing of cold leg flows inside the vessel of a pressurised water reactor. To assess the capabilities of the model, a simulation was made of a thermal mixing test at the Oconee-1 Nuclear Station. The test measured temperature deviations at the core inlet produced by an imposed temperature difference between cold legs. Both the tests results and the simulation showed that most of the cold leg flows arrive unmixed at the core inlet. However, the simulation was unable to reproduce the asymmetric irregularities observed in the core inlet temperature distribution, and consequently the degree of mixing was under-predicted. Various sensitivity studies were carried out on the model, but these did not reveal the source of the asymmetry. It was therefore concluded that the asymmetry source was outside the scope of the model, but the model was nevertheless able to make plausible but pessimistic estimates of mixing. (author)

  20. Design and development of the high performance shutdown rods for the CANDU 3 reactor

    International Nuclear Information System (INIS)

    The high performance Shutdown Rod (SDR) unit has been successfully developed to improve insertion performance significantly. Because this unit is the reactor's primary emergency shutdown device, reliability is very important. It was therefore prudent to use an evolutionary approach, based on the well-established Shutoff Rod (SOR) unit, rather than attempt radical new concepts. Based on analyses of the key performance influences, the absorber rod's weight was reduced 55% by thinning its absorber and sheath tubes. The release clutch was uprated by 60% by adding extra friction plates and utilizing its full rated voltage. Its release time was shortened simply by working it harder. The proven drive mechanism was otherwise unchanged. The rod and clutch change permitted increasing the accelerating spring force and stroke both 250%. These changes yielded a reduction of insertion time from 0.56 sec (old type SOR) to 0.33 sec, and is 0.11 sec faster than the original CANDU 3 design target. Full scale rig testing has verified it can reliably perform partial drop tests and also endure 3000 full drops, 5 times its rated service demand

  1. Overview of European Community (Activity 3) work on materials properties of fast reactor structural materials

    International Nuclear Information System (INIS)

    The Fast Reactor Coordinating Committee set up in 1974 the Working Group Codes and Standards, and organized its work into four main activities: Manufacturing standards, Structural analysis, Materials and Classification of components. The main purpose of materials activity is to compare and contrast existing national specifications and associated properties relevant to structural materials in fast reactors. Funds are available on a yearly basis for tasks to be carried out through Study Contracts. At present about four Study Contract Reports are prepared each year

  2. Licensed operating reactors. Status summary report, data as of February 28, 1986. Volume 10, No. 3

    International Nuclear Information System (INIS)

    This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  3. Technology assessment HTR. Part 3. Economics of new concept of the modular High Temperature Reactor

    International Nuclear Information System (INIS)

    In this study the economic feasibility of new concepts of the High Temperature Reactor were investigated. These new concepts are characterized as inherently safe. The different concepts were used as industrial heat/power reactors and compared with a gas fired Steam and Gas turbine installation. The best economic advantages are offered by a HTR with a Thorium/Uranium cycle as compared with a gas fired steam- and gas turbine. 6 figs, 9 tabs, 21 refs

  4. The third generation pressurized water reactor Olkiluoto 3 in Finland becomes reality

    International Nuclear Information System (INIS)

    The European Pressurized Water Reactor (EPR) is the world's first third-generation pressurized water reactor (PWR). The first steps towards realization of an EPR nuclear power plant were taken in Olkiluoto, Finland, in 2004 by preliminary works on the construction site. The first concrete will be poured after the construction licence has been granted by the Finnish government which is anticipated to happen in the beginning of 2005. Commerical operation is scheduled to start in 2009. (orig.)

  5. Thermal Hydraulic Analysis of 3 MW TRIGA Research Reactor of Bangladesh Considering Different Cycles of Burnup

    OpenAIRE

    M. H. Altaf; N.H. Badrun

    2014-01-01

    Burnup dependent steady state thermal hydraulic analysis of TRIGA Mark-II research reactor has been carried out utilizing coupled point kinetics, neutronics and thermal hydraulics code EUREKA-2/RR. From the previous calculations of neutronics parameters including percentage burnup of individual fuel elements performed so far for 700 MWD burnt core of TRIGA reactor showed that the fuel rod predicted as hottest at the beginning of cycle (fresh core) was found to remain as the hottest until 200 ...

  6. OECD MCCI project enhancing instrumentation for reactor materials experiments, Rev. 0 September 3, 2002.

    Energy Technology Data Exchange (ETDEWEB)

    Lomperski, S.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    Reactor safety experiments for studying the reactions of a molten core (corium) with water and/or concrete involve materials at extremely high temperature. Such high temperature severely restricts the types of sensors that can be employed to measure characteristics of the corium itself. Yet there is great interest in improving instrumentation so that the state of the melt can be established with more precision. In particular, it would be beneficial to increase both the upper range limit and accuracy of temperature measurements. The poor durability of thermocouples at high temperature is also an important issue. For experiments involving a water-quenched melt, direct measurements of the growth rate of the crust separating the melt and water would be of great interest. This is a key element in determining the nature of heat transfer between the melt and coolant. Despite its importance, no one has been able to directly measure the crust thickness during such tests. This paper considers three specialized sensors that could be introduced to enhance melt characterization: (1) A commercially fabricated, single point infrared temperature measurement with the footprint of a thermowell. A lens assembly and fiber optic cable linked to a receiver and amplifier measures the temperature at the base of a tungsten thermowell. The upper range limit is 3000 C and accuracy is {+-}0.25% of the reading. (2) In-house development of an ultrasonic temperature sensor that would provide multipoint measurements at temperatures up to {approx}3000 C. The sensors are constructed from tungsten rods and have a high temperature durability that is superior to that of thermocouples. (3) In-house development of an ultrasonic probe to measure the growth rate of the corium crust. This ultrasonic sensor would include a tungsten waveguide that transmits ultrasonic pulses up through the corium melt towards the crust and detects reflections from the melt/crust interface. A measurement of the echo time

  7. Electrical properties of nano Li2TiO3 for fusion reactors

    International Nuclear Information System (INIS)

    In the development of tritium breeding blankets for fusion reactor, lithium based ceramic such as lithium orthosilicate (Li4SiO4), lithium titanate (Li2TiO3), lithium zirconate (Li2ZrO3), and lithium oxide (Li2O) for breeding blankets. Among them Lithium titanate (LT) is one of the most promising tritium breeding materials due to their reasonable lithium atom density, low activation, good compatibility with structural materials, excellent tritium release performance and chemical stability. Electrical properties may reflect some characteristic features, hence analysis of electrical charge transport in small grained Li2TiO3 ceramics, as envisaged for tritium breeding, may contribute to gain information of certain high energy ball-milling process. The main attribute of current study analyzes the electrical conductivity behavior of Li2TiO3 ceramics. The 10h milled powder of Li2TiO3 by High energy ball milling (HEBM) calcined at 700 °C for 2h. The calcinied powder was pressed uniaxially with 3wt. % PVA (polyvinyl alcohol) solution added as binder. The rectangular disk samples of diameter 12.7mm and thickness 12mm was made by hydraulic press with 400 Mpa pressure. The samples were sintered at 700 °C, 800 °C, 900 °C and 1000 °C 2h in conventional sintering. Silver contacts were made on the opposite disc faces and heated at 700 °C for 15 minutes with a heating 5 °C per minute for electrical measurement. It was found that microwave sintered samples shows higher thermal conductivity then conventional sintered one. The Ea value decreases with increase in frequency, due to the increase in ionic conductivity. The ionic conductivity is a combination of both macroscopic and microscopic conduction, which indirectly depend on the bulk Rb and grain boundary Rgb resistance. At high temperature only single semicircle could be found, using high frequency data, indicate dominant behavior of grain. The value of activation energy (0.238eV) and conductivity range (10-3 to 10-4 S

  8. Design of a New Research Reactor: Preliminary Conceptual Design (3rd Year)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, B. C.; Chae, H. T. and others

    2006-01-15

    A research reactor design is a kind of integral engineering project and a process to obtain a concrete shape through several years of concept development, conceptual design, basic design and detail design. So it requires close cooperation in various areas as well as lots of manpower and cost. The overall process at each stage may be said to be similar except for some stage-specific works. In 2005 as last year of a concept development stage, investigations on the various concepts of the fuel, reactor structure and systems which can meet the requirements established. The requirements for the process systems and I and C systems have also been embodied. The major tasks planned at the early of 2005 have been performed for each area of reactor design as follows: Establishment of the fuel and reactor core concept, and the core analysis, Preliminary thermal-hydraulic and safety analyses for the conceptual cores, Establishment and improvement of analysis system, Concept developments of the reactor structures and major systems, Test and test plan to verify the developed concepts, International cooperation to establish the foundations for exporting a research reactor.

  9. CAST3M/ARCTURUS: A coupled heat transfer CFD code for thermal-hydraulic analyzes of gas cooled reactors

    International Nuclear Information System (INIS)

    The safety of gas cooled reactors (High Temperature Reactors (HTR), Very High Temperature Reactors (VHTR) or Gas Cooled Fast Reactors (GFR)) must be ensured by systems (active or passive) which maintain loads on component (fuel) and structures (vessel, containment) within acceptable limits under accidental conditions. To achieve this objective, thermal-hydraulics computer codes are necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Some key safety questions are related to the evaluation of decay heat removal and containment pressure and thermal loads. This requires accurate simulations of conduction, convection, thermal radiation transfers and energy storage. Coupling with neutronics is also an important modeling aspect for the determination of representative parameters such as neutronics coefficient (Doppler coefficient, Moderator coefficient, ...), critical position of control rods, reactivity insertion aspects, .... For GFR, the high power density of the core and its necessary reduced dimension cannot rely only on passive systems for decay heat removal. Therefore, forced convection using active safety systems (gas blowers, heat exchangers, ...) are highly recommended. Nevertheless, in case of station black-out, the safety demonstration of the concept should be guaranteed by natural circulation heat removal. This could be performed by keeping a relatively high back-up pressure for pure helium convection and also by heavy gas injection. So, it is also necessary to model mixing of different gases, the on-set of natural convection and the pressure and thermal loads onto the proximate or guard containment. In this paper, we report on the developments of the CAST3M/ARCTURUS thermal-hydraulics (Lumped Parameter and CFD) code developed at CEA, including its coupling to the neutronics code CRONOS2 and the system code CATHARE. Elementary validation cases are detailed, as well as application of the code to benchmark

  10. Simulations of RUTA-70 reactor with CERMET fuel using DYN3D/ATHLET and DYN3D/RELAP5 coupled codes

    International Nuclear Information System (INIS)

    RUTA-70 model for simulations with the internally coupled codes DYN3D/ATHLET and DYN3D/RELAP5 was developed. A 3-D power distribution in the core is calculated by DYN3D with thermal-hydraulic feedback from the system codes. A steady-state corresponding to the full reactor power and an accident scenario initiated by failure of all primary coolant pumps were simulated with the DYN3D/ATHLET and DYN3D/RELAP5 coupled code systems to verify these codes. The compared coupled codes give close predictions for the initial and final states of the simulated accident but not for the transition between them. The observed deviations are explained by differences in the subcooled boiling models of the employed versions of ATHLET and RELAP5. Nevertheless, both simulations confirm a high level of the reactor inherent safety. The allowed safety margins were not reached. (orig.)

  11. Simulations of RUTA-70 reactor with CERMET fuel using DYN3D/ATHLET and DYN3D/RELAP5 coupled codes

    Energy Technology Data Exchange (ETDEWEB)

    Kozmenkov, Y.; Rohde, U. [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany); Baranaev, Y.; Glebov, A. [State Scientific Center of the Russian Federation, Obninsk, Kaluga Region (Russian Federation). Inst. for Physics and Power Engineering

    2012-08-15

    RUTA-70 model for simulations with the internally coupled codes DYN3D/ATHLET and DYN3D/RELAP5 was developed. A 3-D power distribution in the core is calculated by DYN3D with thermal-hydraulic feedback from the system codes. A steady-state corresponding to the full reactor power and an accident scenario initiated by failure of all primary coolant pumps were simulated with the DYN3D/ATHLET and DYN3D/RELAP5 coupled code systems to verify these codes. The compared coupled codes give close predictions for the initial and final states of the simulated accident but not for the transition between them. The observed deviations are explained by differences in the subcooled boiling models of the employed versions of ATHLET and RELAP5. Nevertheless, both simulations confirm a high level of the reactor inherent safety. The allowed safety margins were not reached. (orig.)

  12. Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2008

    International Nuclear Information System (INIS)

    JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment and practice in the nuclear reactor training, Irradiation for activation analyses, RI productions, fission tracks etc. In the fiscal year 2008, the research reactor JRR-3 was operated for 7 cycles (cycle operation : 26days/cycle) for utilization sharing of facility. The research reactor JRR-4 was not operated in 2008. Because a crack was found on the weld of the aluminum cladding of a graphite reflector element. JRR-4 has remained shutdown until the reflector elements were replaced. The volume contains 250activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, neutron activation analyses, and others submitted by the users in JAEA and other Organizations. (author)

  13. Gamma-ray spectrum measurement in Japan research reactor no. 3 using a portable Ge(Li) spectrometer

    International Nuclear Information System (INIS)

    A portable Ge(Li) gamma-ray spectrometer having 2.6% peak detection efficiency and 3.5 keV energy resolution was made using a 7.5 liter liquid nitrogen dewar. The total weight of the spectrometer including the detector, cryostat, preamplifier, high voltage filter, and 7.5 liter liquid nitrogen was 11 kg. Gamma-ray spectra were measured at various places in Japan Research Reactor No. 3 using the spectrometer. Gamma-rays from natural radioactive nuclides such as 40K, 208Tl, 214Bi, and from 60Co, which was an induced radioactive nuclide of the reactor constructing components, were observed at all the places. During the reactor in operation, gamma-rays from 41Ar, the induced radioactive nuclide of argon in air, were observed also at all the places. High-energy gamma-rays from the neutron capture reaction in iron and from 16N induced by 16O(n,p)16N reaction in the oxygen in heavy-water coolant were found in the first floor of the reactor room; the former seemed to originate from the monochromator crystals of the neutron diffractometers. Noble gas fission product gamma-rays were observed in helium cover gas in the FFD system. Pulse height distributions and counting rates of these gamma-rays were shown. (author)

  14. Reactor inlet header critical break identification and analysis for KAPP-3 and 4 using computer code RELAP-5/Mod.3.2

    International Nuclear Information System (INIS)

    Kakrapar Atomic Power Project units-3 and 4 (KAPP-3 and 4) are 700 MWe Pressurized Heavy Water Reactors (PHWR) are presently under construction. This paper presents the identification of critical break in Reactor Inlet Header and its analysis performed, for KAPP-3 and 4 as a part of safety studies to investigate the plant behavior. The limiting/critical break size at Reactor Inlet Header is identified by considering the peak sheath temperature during the Loss of coolant accident. System thermal hydraulics code RELAP-5/MOD3.2 has been used for the analysis. Here the overall thermal hydraulics of the plant along with various control systems, trip and actuation logics have been simulated. High pressure accumulators and low pressure recirculation system of emergency core cooling system are modeled. The modeling of secondary system includes modeling of Atmospheric Steam Discharge Valves (ASDVs), Safety Relief Valves (SRVs), Condensate Steam Discharge Valves (CSDVs), and Governor Valves, the U-tubes of the steam generator, the riser, the separator and the steam drum. Using this model, critical break size in the Reactor Outlet Header was identified and consequence of the event on maximum peak clad temperature and core parameters were evaluated. Following postulated accidents, the event progression and the variations of different parameters like different Header pressures, mass flow rate in the core, fuel clad temperature and rate of discharge from break etc have been studied. (author)

  15. Simulation of the operational monitoring of a BWR with Simulate-3; Simulacion del seguimiento operacional de un reactor BWR con Simulate-3

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez F, J. O.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: ace.jo.cu@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    This work was developed in order to describe the methodology for calculating the fuel burned of nuclear power reactors throughout the duration of their operating cycle and for each fuel reload. In other words, simulate and give monitoring to the main operation parameters of sequential way along its operation cycles. For this particular case, the operational monitoring of five consecutive cycles of a reactor was realized using the information reported by their processes computer. The simulation was performed with the Simulate-3 software and the results were compared with those of the process computer. The goal is to get the fuel burned, cycle after cycle for obtain the state conditions of the reactor needed for the fuel reload analyses, stability studies and transients analysis, and the development of a methodology that allows to manage and resolve similar cases for future fuel cycles of the nuclear power plant and explore the various options offered by the simulator. (Author)

  16. Microbiota characterization of a Belgian protected designation of origin cheese, Herve cheese, using metagenomic analysis.

    Science.gov (United States)

    Delcenserie, V; Taminiau, B; Delhalle, L; Nezer, C; Doyen, P; Crevecoeur, S; Roussey, D; Korsak, N; Daube, G

    2014-10-01

    Herve cheese is a Belgian soft cheese with a washed rind, and is made from raw or pasteurized milk. The specific microbiota of this cheese has never previously been fully explored and the use of raw or pasteurized milk in addition to starters is assumed to affect the microbiota of the rind and the heart. The aim of the study was to analyze the bacterial microbiota of Herve cheese using classical microbiology and a metagenomic approach based on 16S ribosomal DNA pyrosequencing. Using classical microbiology, the total counts of bacteria were comparable for the 11 samples of tested raw and pasteurized milk cheeses, reaching almost 8 log cfu/g. Using the metagenomic approach, 207 different phylotypes were identified. The rind of both the raw and pasteurized milk cheeses was found to be highly diversified. However, 96.3 and 97.9% of the total microbiota of the raw milk and pasteurized cheese rind, respectively, were composed of species present in both types of cheese, such as Corynebacterium casei, Psychrobacter spp., Lactococcus lactis ssp. cremoris, Staphylococcus equorum, Vagococcus salmoninarum, and other species present at levels below 5%. Brevibacterium linens were present at low levels (0.5 and 1.6%, respectively) on the rind of both the raw and the pasteurized milk cheeses, even though this bacterium had been inoculated during the manufacturing process. Interestingly, Psychroflexus casei, also described as giving a red smear to Raclette-type cheese, was identified in small proportions in the composition of the rind of both the raw and pasteurized milk cheeses (0.17 and 0.5%, respectively). In the heart of the cheeses, the common species of bacteria reached more than 99%. The main species identified were Lactococcus lactis ssp. cremoris, Psychrobacter spp., and Staphylococcus equorum ssp. equorum. Interestingly, 93 phylotypes were present only in the raw milk cheeses and 29 only in the pasteurized milk cheeses, showing the high diversity of the microbiota

  17. 2D fluid model analysis for the effect of 3D gas flow on a capacitively coupled plasma deposition reactor

    Science.gov (United States)

    Kim, Ho Jun; Lee, Hae June

    2016-06-01

    The wide applicability of capacitively coupled plasma (CCP) deposition has increased the interest in developing comprehensive numerical models, but CCP imposes a tremendous computational cost when conducting a transient analysis in a three-dimensional (3D) model which reflects the real geometry of reactors. In particular, the detailed flow features of reactive gases induced by 3D geometric effects need to be considered for the precise calculation of radical distribution of reactive species. Thus, an alternative inclusive method for the numerical simulation of CCP deposition is proposed to simulate a two-dimensional (2D) CCP model based on the 3D gas flow results by simulating flow, temperature, and species fields in a 3D space at first without calculating the plasma chemistry. A numerical study of a cylindrical showerhead-electrode CCP reactor was conducted for particular cases of SiH4/NH3/N2/He gas mixture to deposit a hydrogenated silicon nitride (SiN x H y ) film. The proposed methodology produces numerical results for a 300 mm wafer deposition reactor which agree very well with the deposition rate profile measured experimentally along the wafer radius.

  18. Computational Analysis of Nuclear Safety Parameters of 3 MW TRIGA Mark-II Research Reactor Based on Evaluated Nuclear Data Libraries JENDL-3.3 and ENDF/B-VII.0

    Energy Technology Data Exchange (ETDEWEB)

    Khan, Jahirul Haque [Bangladesh Atomic Energy Commission, Dhaka (Bangladesh)

    2013-07-01

    The objective of this study is to explain the main nuclear safety parameters of 3 MW TRIGA Mark-II Research Reactor at AERE, Savar, Dhaka, Bangladesh from the viewpoint of reactor safety and also reactor operator. The most important nuclear reactor physics safety parameters are power distribution, power peaking factors, shutdown margin, control rod worth, excess reactivity and fuel temperature reactivity coefficient. These parameters are calculated using the chain of the computer codes the SRAC-PIJ for cell calculation based on neutron transport theory and the SRAC-CITATION for core calculation based on neutron diffusion equation. To achieve this objective the TRIGA model is developed by the 3-D diffusion code SRAC-CITATION based on the group constants that come from the collision probability transport code SRAC-PIJ. In this study the evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 are used. The calculated most important reactor physics parameters are compared to the safety analysis report (SAR) values as well as earlier published MCNP results (numerically benchmark). It was found that the calculated results show a good agreement between the said libraries. Besides, in most cases the calculated results reveal a reasonable agreement with the SAR values (by General Atomic) as well as the MCNP results. In addition, this analysis can be used as the inputs for thermal-hydraulic calculations of the TRIGA fresh core in the steady state and pulse mode operation. Because of power peaking factors, power distributions and temperature reactivity coefficients are the most important reactor safety parameters for normal operation and transient safety analysis in research as well as in power reactors. They form the basis for technical specifications and limitations for reactor operation such as loading pattern limitations for pulse operation (in TRIGA). Therefore, this analysis will be very important to develop the nuclear safety parameters data of 3 MW TRIGA Mark

  19. Computational Analysis of Nuclear Safety Parameters of 3 MW TRIGA Mark-II Research Reactor Based on Evaluated Nuclear Data Libraries JENDL-3.3 and ENDF/B-VII.0

    International Nuclear Information System (INIS)

    The objective of this study is to explain the main nuclear safety parameters of 3 MW TRIGA Mark-II Research Reactor at AERE, Savar, Dhaka, Bangladesh from the viewpoint of reactor safety and also reactor operator. The most important nuclear reactor physics safety parameters are power distribution, power peaking factors, shutdown margin, control rod worth, excess reactivity and fuel temperature reactivity coefficient. These parameters are calculated using the chain of the computer codes the SRAC-PIJ for cell calculation based on neutron transport theory and the SRAC-CITATION for core calculation based on neutron diffusion equation. To achieve this objective the TRIGA model is developed by the 3-D diffusion code SRAC-CITATION based on the group constants that come from the collision probability transport code SRAC-PIJ. In this study the evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 are used. The calculated most important reactor physics parameters are compared to the safety analysis report (SAR) values as well as earlier published MCNP results (numerically benchmark). It was found that the calculated results show a good agreement between the said libraries. Besides, in most cases the calculated results reveal a reasonable agreement with the SAR values (by General Atomic) as well as the MCNP results. In addition, this analysis can be used as the inputs for thermal-hydraulic calculations of the TRIGA fresh core in the steady state and pulse mode operation. Because of power peaking factors, power distributions and temperature reactivity coefficients are the most important reactor safety parameters for normal operation and transient safety analysis in research as well as in power reactors. They form the basis for technical specifications and limitations for reactor operation such as loading pattern limitations for pulse operation (in TRIGA). Therefore, this analysis will be very important to develop the nuclear safety parameters data of 3 MW TRIGA Mark

  20. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

    Energy Technology Data Exchange (ETDEWEB)

    Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)