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Sample records for belgian reactor 3

  1. Legal claims against Belgian reactors?; Rechtsmittel gegen belgische Reaktoren?

    Energy Technology Data Exchange (ETDEWEB)

    Raetzke, Christian [CONLAR Consulting on Nuclear Law and Regulation, Leipzig (Germany)

    2016-06-15

    The Belgian reactors Tihange 2 and Doel 3 have been restarted in November 2015 after the problem of hydrogen flakes in the reactor pressure vessels had been investigated. The permission to restart has been the object both of critical statements by the German Federal Ministry of the Environment (BMUB) and of lawsuits filed with Belgian law courts by a group of German municipalities led by the city of Aachen and by the Land North-Rhine-Westphalia. According to a general principle of the law of nations, a state is not permitted to operate installations near its border, which cause significant environmental damage in a neighbouring state. However, it is not quite clear how this principle applies to the issue of potential accidents of nuclear power plants. According to the author, a tangible threat of an accident is required; mere doubts and concerns about the extent of safety margins are not sufficient.

  2. Reactor pressure vessel steels[1997 Scientific Report of the Belgian Nuclear Research Centre

    Energy Technology Data Exchange (ETDEWEB)

    Van De Velde, J.; Fabry, A.; Van Walle, E.; Chaouuadi, R.

    1998-07-01

    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.

  3. Management and storage of nuclear fuel from Belgian research reactors

    International Nuclear Information System (INIS)

    Gubel, P.

    1996-01-01

    Experiences and problems with the storage of irradiated fuel at research reactors in Belgium are described. In particular, interim storage problems exist for spent fuel elements at the BR2 and the shut down BR3 reactors in Mol. (author). 1 ref

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

    International Nuclear Information System (INIS)

    Couplet, D.; Francoise, T.

    2001-01-01

    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. Belgian Contribution to the IAEA CRP-IV Programme on Assuring Structural Integrity of Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    Van Walle, E.; Chaouadi, R.; Scibetta, M.; Puzzolante, J.L.; Fabry, A.; Van de Velde, J.

    1997-10-01

    This report contains the actual status of the Belgian contribution to the IAEA CRP-IV program. Besides Charpy-V impact tests on as-received CRP-IV JRQ-specimens, fracture toughness tests were performed on two geometries: PCCV-specimens and CRB-specimens. The Charpy-V impact results correspond very well with the as-received CRP-III results. The fracture toughness data are also very consistent with identical tests recently performed on remaining as-received CRP-III material. Irradiated broken Charpy-V samples were reconstituted and tested in PCCV-mode. This was done in order to investigate the evolution of the ASME-curve versus the evolution of the mastercurve with irradiation. Initial results were reported. A new CHIVAS-irradiation in the CALLISTO-loop of the BR-2-reactor to support this investigation, is under preparation

  6. 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...

  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 Sprl 14. Groupe Hamon31. Thales Communication...

  8. 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...

  9. Monte Carlo modelling of the Belgian materials testing reactor BR2: present status

    International Nuclear Information System (INIS)

    Verboomen, B.; Aoust, Th.; Raedt, Ch. de; Beeckmans de West-Meerbeeck, A.

    2001-01-01

    A very detailed 3-D MCNP-4B model of the BR2 reactor was developed to perform all neutron and gamma calculations needed for the design of new experimental irradiation rigs. The Monte Carlo model of BR2 includes the nearly exact geometrical representation of fuel elements (now with their axially varying burn-up), of partially inserted control and regulating rods, of experimental devices and of radioisotope production rigs. The multiple level-geometry possibilities of MCNP-4B are fully exploited to obtain sufficiently flexible tools to cope with the very changing core loading. (orig.)

  10. Decommissioning of the BR3 pressurized-water reactor

    International Nuclear Information System (INIS)

    Massaut, V.

    1996-01-01

    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 programme, objectives, achievements in this research area at the Belgian Nuclear Research Centre SCK-CEN for 1995 are summarized

  11. Ageing Management of the reactor internals in Belgian nuclear units in view of Long Term Operation

    International Nuclear Information System (INIS)

    Gerard, R.; Bertolis, D.; Vissers, S.

    2012-01-01

    The reactor internals support the reactor core, distribute the coolant flow through the core, and guide and protect the rod control cluster assemblies and in-core instrumentation. Their integrity must be guaranteed in all operating and accident conditions. They are exposed to specific degradation mechanisms linked to the intense neutron irradiation, like Irradiation Assisted Stress Corrosion Cracking (IASCC) or potentially void swelling, in addition to more classical mechanisms like fatigue, wear and stress corrosion cracking. A rigorous follow-up of in-service degradation and an effective ageing management is therefore of crucial importance and contributes to the safe and economical operation of nuclear PWR units. (author)

  12. Assessing fundamental motor skills in Belgian children aged 3-8 years highlights differences to US reference sample.

    Science.gov (United States)

    Bardid, Farid; Huyben, Floris; Lenoir, Matthieu; Seghers, Jan; De Martelaer, Kristine; Goodway, Jacqueline D; Deconinck, Frederik J A

    2016-06-01

    This study aimed to understand the fundamental motor skills (FMS) of Belgian children using the process-oriented Test of Gross Motor Development, Second Edition (TGMD-2) and to investigate the suitability of using the United States (USA) test norms in Belgium. FMS were assessed using the TGMD-2. Gender, age and motor performance were examined in 1614 Belgian children aged 3-8 years (52.1% boys) and compared with the US reference sample. More proficient FMS performance was found with increasing age, from 3 to 6 years for locomotor skills and 3 to 7 years for object control skills. Gender differences were observed in object control skills, with boys performing better than girls. In general, Belgian children had lower levels of motor competence than the US reference sample, specifically for object control skills. The score distribution of the Belgian sample was skewed, with 37.4% scoring below average and only 6.9% scoring above average. This study supported the usefulness of the TGMD-2 as a process-oriented instrument to measure gross motor development in early childhood in Belgium. However, it also demonstrated that caution is warranted when using the US reference norms. ©2016 Foundation Acta Paediatrica. Published by John Wiley & Sons Ltd.

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

    International Nuclear Information System (INIS)

    Dadoumont, J.; Massaut, V.; Klein, M.; Demeulemeester, Y.

    2002-01-01

    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)

  14. The Belgian nuclear research centre

    International Nuclear Information System (INIS)

    Moons, F.

    2001-01-01

    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)

  15. 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.

  16. Introduction of mixed oxide fuel elements in the belgian cores

    International Nuclear Information System (INIS)

    Charlier, A.F.; Hollasky, N.A.

    1994-01-01

    The important amount of plutonium recovered from the reprocessing of spent fuel on the one hand, the national and international experience of the use of mixed oxide UO 2 -PuO 2 fuel in power reactors on the other hand, have led Belgian utilities to decide the introduction of Mixed-Oxide fuel in Doel unit 3 and Tihange unit 2 cores. The 'MOX' project has shown that it was possible without reducing safety or requiring modifications of the plant equipment. It has been approved by the Belgian 'Nuclear Safety Commission'. (authors). 1 tab., 2 figs

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

    International Nuclear Information System (INIS)

    Deen, J.R.; Snelgrove, J.L.

    1986-01-01

    Fuel cycle calculations have been performed with reference HEU fuel and LEU fuel using Cd wires or boron as burnable absorbers. The 235 U 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/m 3 . 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

  18. Annealing of the BR3 reactor pressure vessel

    International Nuclear Information System (INIS)

    Fabry, A.; Motte, F.; Stiennon, G.; Debrue, J.; Gubel, P.; Van de Velde, J.; Minsart, G.; Van Asbroeck, P.

    1985-01-01

    The pressure vessel of the Belgian BR-3 plant, a small (11 MWe) PWR presently used for fuel testing programs and operated since 1962, was annealed during March, 1984. The anneal was performed under wet conditions for 168 hours at 650 0 F with core removal and within plant design margins justification for the anneal, summary of plant characteristics, description of materials sampling, summary of reactor physics and dosimetry, development of embrittlement trend curves, hypothesized pressurized and overcooling thermal shock accidents, and conclusions are provided in detail

  19. Reactor Materials Research

    International Nuclear Information System (INIS)

    Van Walle, E.

    2001-01-01

    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)

  20. 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)

  1. Pressure Vessel Steel Research: Belgian Activities

    International Nuclear Information System (INIS)

    Van Walle, E.; Fabry, A.; Ait Abderrahim, H.; Chaouadi, R.; D'hondt, P.; Puzzolante, J.L.; Van de Velde, J.; Van Ransbeeck, T.; Gerard, R.

    1994-03-01

    A review of the Belgian research activities on Nuclear Reactor Pressure Vessel Steels (RPVS) and on related Neutron Dosimetry Aspects is presented. Born out of the surveillance programmes of the Belgian nuclear power plants, this research has lead to the development of material saving techniques, like reconstitution and miniaturization, and to improved neutron dosimetry techniques. A physically- justified RPVS fracture toughness indexation methodology, supported by micro-mechanistic modelling, is based on the elaborate use of the instrumented Charpy impact signal. Computational tools for neutron dosimetry allow to reduce the uncertainties on surveillance capsule fluences significantly

  2. Pressure Vessel Steel Research: Belgian Activities

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E; Fabry, A; Ait Abderrahim, H; Chaouadi, R; D` hondt, P; Puzzolante, J L; Van de Velde, J; Van Ransbeeck, T [Centre d` Etude de l` Energie Nucleaire, Mol (Belgium); Gerard, R [TRACTEBEL, Brussels (Belgium)

    1994-03-01

    A review of the Belgian research activities on Nuclear Reactor Pressure Vessel Steels (RPVS) and on related Neutron Dosimetry Aspects is presented. Born out of the surveillance programmes of the Belgian nuclear power plants, this research has lead to the development of material saving techniques, like reconstitution and miniaturization, and to improved neutron dosimetry techniques. A physically- justified RPVS fracture toughness indexation methodology, supported by micro-mechanistic modelling, is based on the elaborate use of the instrumented Charpy impact signal. Computational tools for neutron dosimetry allow to reduce the uncertainties on surveillance capsule fluences significantly.

  3. The utility of different reactor types for the research

    International Nuclear Information System (INIS)

    Stiennon, G.

    1983-01-01

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

    Science.gov (United States)

    Dewalque, Lucas; Pirard, Catherine; Charlier, Corinne

    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 metabolites of phthalates in samples collected from 261 participants living in and around Liege (Belgium). The analyses were carried out by liquid chromatography tandem mass spectrometry (LC-MS/MS) using isotopic dilution. To the best of our knowledge, this is the first time that the urinary levels of these 3 classes of chemicals are reported for the same general population in Belgium. Most of the parabens, the BP3, and all the phthalate metabolites were detected in 82.8 to 100.0% of the samples. For most of these chemicals, the exposure patterns significantly differ not only between children and adults, but also between males and females, especially with higher concentrations of parabens and phthalate metabolites in female and children subjects, respectively. PMID:24719881

  5. Two Belgian University Hospitals

    Directory of Open Access Journals (Sweden)

    M. Huylebrouck

    2012-01-01

    Full Text Available Background. Bevacizumab (BEV, a humanized immunoglobulin G1 monoclonal antibody that inhibits VEGF has demonstrated activity against recurrent high-grade gliomas (HGG in phase II clinical trials. Patients and Methods. Data were collected from patients with recurrent HGG who initiated treatment with BEV outside a clinical trial protocol at two Belgian university hospitals. Results. 19 patients (11 M/8 F were administered a total of 138 cycles of BEV (median 4, range 1–31. Tumor response assessment by MRI was available for 15 patients; 2 complete responses and 3 partial responses for an objective response rate of 26% for the intent to treat population were observed on gadolinium-enhanced T1-weighted images; significant regressions on T2/FLAIR were documented in 10 out of 15 patients (67%. A reduced uptake on PET was documented in 3 out of 4 evaluable patients. The six-month progression-free survival was 21% (95% CI 2.7–39.5. Two patients had an ongoing tumor response and remained free from progression after 12 months of BEV treatment. Conclusions. The activity and tolerability of BEV were comparable to results from previous prospective phase II trials. Reduced uptake on PET suggests a metabolic response in addition to an antiangiogenic effect in some cases with favorable clinical outcome.

  6. Dynamics of TRIGA-3 Salazar Reactor

    International Nuclear Information System (INIS)

    Gallardo S, L.F.

    1990-01-01

    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)

  7. Belgian national report

    International Nuclear Information System (INIS)

    Berthe, J.

    1998-01-01

    Status of Belgian nuclear power plants includes licensing, in-service inspection programs, state of electrical equipment and predictive maintenance. In terms of life management of NPPs degradation phenomena affect the design life of each component. Combination of in-service inspection, periodic testing, specific measurements, qualification test and overall experience supports maintenance programs and enable repairs and replacements in due time. These programs are part of continuous safety assessments performed. Managing NPP life encompasses technical aspects for safe and reliable operation and economical aspects. The approach of Belgian authorities resulted in high availability, competitive cost and reasonable long-term perspectives

  8. The Belgian Nuclear Higher Education Network

    International Nuclear Information System (INIS)

    Moons, F.; D'Haeseleer, W.; Giot, M.

    2004-01-01

    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)

  9. JENDL-3.3 thermal reactor benchmark test

    International Nuclear Information System (INIS)

    Akie, Hiroshi

    2001-01-01

    Integral tests of JENDL-3.2 nuclear data library have been carried out by Reactor Integral Test WG of Japanese Nuclear Data Committee. The most important problem in the thermal reactor benchmark testing was the overestimation of the multiplication factor of the U fueled cores. With several revisions of the data of 235 U and the other nuclides, JENDL-3.3 data library gives a good estimation of multiplication factors both for U and Pu fueled thermal reactors. (author)

  10. Reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Van De Velde, J.; Fabry, A.; Van Walle, E.; Chaouuadi, R.

    1998-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    1998-10-01

    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 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.

  12. 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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Decreton, M.

    1998-10-01

    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 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

  15. 3. Halden Reactor Project Workshop

    International Nuclear Information System (INIS)

    Louka, Michael N.

    2005-09-01

    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)

  16. EL-3 dismantling of an experimental reactor

    International Nuclear Information System (INIS)

    1989-01-01

    The EL3 experimental reactor has been definitively stopped in march 1979. Its decommissioning has been pronounced in the end of 1982. This article is consecrated at decontamination and dismantling works necessited by its passage at the dismantling level 2 [fr

  17. Integral test of JENDL-3.3 for fast reactors

    International Nuclear Information System (INIS)

    Chiba, Gou

    2003-01-01

    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)

  18. Belgian national report

    International Nuclear Information System (INIS)

    Berthe, J.

    1995-01-01

    At last IWG-LMNPP meeting, the approach on nuclear power plant life management in Belgium was presented. The present report focuses on results of in-service monitoring of major equipment, specifically reactor internals, reactor top-head penetrations and steam generators. Status of major backfitting on steam generators and balance of plant is developed as well as developments in the field of thermal stratification and qualification of ultrasonic inspection methods and personnel for in-service inspection. (author). (Abstract only)

  19. Reactor physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.

    1998-01-01

    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

  20. Systems analysis of the CANDU 3 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

    1993-07-01

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

  1. 3 Investment Scenarios for Fast Reactors

    International Nuclear Information System (INIS)

    Shoai Tehrani, Bianka; Da Costa, Pascal

    2013-01-01

    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

  2. Euthanasia: a "kit" sold in Belgian pharmacies.

    Science.gov (United States)

    2005-10-01

    (1) In France, legislation adopted in 2005 recognises the right of dying patients to refuse further treatment, and the right of physicians to ease their suffering with treatments that, due to adverse effects, may shorten their life. Measures deliberately aimed at hastening death are forbidden. (2) In Belgium, medical euthanasia was decriminalised in 2002, and can now be carried out either in hospital or at home. Nearly 20 cases of euthanasia are reported per month in Belgium. (3) A Belgian pharmacy chain now markets a "euthanasia kit".

  3. Experiment for search for sterile neutrino at SM-3 reactor

    Science.gov (United States)

    Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

    2016-11-01

    In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

  4. Completion of reconstruction for Japan Research Reactor No.3

    International Nuclear Information System (INIS)

    Kakefuda, K.; Tani, M.; Isshiki, M.

    1992-01-01

    The works of the reconstruction for the Japan Research Reactor No.3 (JRR-3) started in 1985 and initial criticality of the new reactor achieved in March, 1990. After commissioning test, the new JRR-3 has been operated some operational cycles since November, 1990. This paper presents outline of the removal work on the old JRR-3 and the new JRR-3. (author)

  5. Integral test of JENDL-3.3 for thermal reactors

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Mori, Takamasa

    2003-01-01

    Criticality benchmark testing was carried out for 59 experiments in various thermal reactors using a continues-energy Monte Carlo code MVP and its different libraries generated from JENDL-3.2, JENDL-3.3, JEF-2.2 and ENDF/B-VI (R8). From the benchmark results, we can say JENDL-3.3 generally gives better k eff values compared with other nuclear data libraries. However, further modification of JENDL-3.3 is expected to solve the following problems: 1) systematic underestimation of k eff depending on 235 U enrichment for the cores with low (less than 3wt.%) enriched uranium fueled cores, 2) dependence of C/E value of k eff on neutron spectrum and plutonium composition for MOX fueled cores. These are common problems for all of the nuclear data libraries used in this study. (author)

  6. Integral test of JENDL-3.3 on fast reactors

    International Nuclear Information System (INIS)

    Chiba, Gou; Hazama, Taira

    2003-05-01

    An integral test has been carried out to evaluate a performance of evaluated nuclear data library JENDL-3.3, which was newly released, in a view of applying neutronics analyses of fast reactors. Japan Nuclear Cycle Development Institute has a large amount of data of critical assembly experiments (ZPPR, BFS, MOZART and FCA) and power reactor tests (JOYO). The database was utilized in this test. In plutonium loaded cores, an improvement was observed about 0.3% ε k in criticality and 5% in the non-leakage term of sodium void reactivity by a revision form JENDL-3.2 to -3.3. These results shoed that the revision is valid in plutonium loaded cores. In uranium loaded cores, dependence of C/E values on control rod position became smaller in control rod worth in ZPPR cores. On the other hand, C/E values became worse both in criticality (0.6%εk) and in sodium void reactivity (30%) in BFS cores. The main cause was a revision of uranium-235 capture cross section, and it could not be concluded whether the revision is valid or not in uranium loaded cores. It is necessary to carry out a validation test at other independent critical experiments in which uranium fuel is used. (author)

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

    International Nuclear Information System (INIS)

    Kernbichler, W.; Miley, G.H.; Heindler, M.

    1989-01-01

    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

  8. 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…

  9. Reactor safety analysis

    International Nuclear Information System (INIS)

    Arien, B.

    1998-01-01

    Risk assessments of nuclear installations require accurate safety and reliability analyses to estimate the consequences of accidental events and their probability of occurrence. The objective of the work performed in this field at the Belgian Nuclear Research Centre SCK-CEN is to develop expertise in probabilistic and deterministic reactor safety analysis. The four main activities of the research project on reactor safety analysis are: (1) the development of software for the reliable analysis of large systems; (2) the development of an expert system for the aid to diagnosis; (3) the development and the application of a probabilistic reactor-dynamics method, and (4) to participate in the international PHEBUS-FP programme for severe accidents. Progress in research during 1997 is described

  10. History of the Belgian nuclear power controversy

    International Nuclear Information System (INIS)

    Laes, E.

    2009-01-01

    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

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

    International Nuclear Information System (INIS)

    1983-04-01

    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

  12. Reversed field pinch reactor study 3

    International Nuclear Information System (INIS)

    Hollis, A.A.; Mitchell, J.T.D.

    1977-12-01

    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)

  13. Removal in a lump of JRR-3 nuclear reactor

    International Nuclear Information System (INIS)

    Ohnishi, Nobuaki; Suzuki, Masanori; Nagase, Tetsuo; Watanabe, Morinari.

    1989-01-01

    The research reactor JRR-3 in Japan Atomic Energy Research Institute is called 'Home made No.1 reactor' as all except fuel and heavy water as the moderator and coolant were manufactured in Japan. The JRR-3 attained the criticality in 1962, and the cumulative time of operation reached 47135.5 hours, and the cumulative power output reached 419073.5 MWh. It was stopped in 1983. During the period, it was utilized for beam experiment, irradiation of fuel and materials, RI production and others. In order to cope with the expansion of utilization and the advance of utilizing technology of the research reactor, the reconstruction works are in progress, and the criticality of the reconstructed reactor is expected in 1990. On the site where the old reactor is removed, the reactor of different type is installed, and the first large cold neutron source is equipped. In this report, as to the removal of the old reactor proper, the method of working and the results are described. Considering the period of working, the cost and the management of the removed reactor, in the case of the JRR-3, the method of carrying it out in a lump was adopted as the optimum removal method. The plan, procedure and results of the removal working are reported. (K.I.)

  14. D-3He fueled FRC reactor 'ARTEMIS-L'

    International Nuclear Information System (INIS)

    Momota, Hiromu; Tomita, Yukihiro; Ishida, Akio; Kohzaki, Yasuji; Nakao, Yasuyuki; Nishikawa, Masabumi; Ohi, Shoichi; Ohnishi, Masami.

    1992-09-01

    A neutron-lean D- 3 He fueled field reversed configuration (FRC) fusion reactor is studied on the bases of former high-efficiency ARTEMIS design. Certain improvements such as effective axial contracting plasma heating and cusp-type direct energy converters as well as an empirical scale of the energy confinement are introduced. The resultant total neutron load onto the first wall of the plasma chamber is as low as 0.1 MW/m 2 , which enable the life of the first wall or the structural materials to be longer than the whole life of the reactor. The attractive characteristics of the neutron-lean reactor follow in the ARTEMIS design: it is socially acceptable in views of radioactivity and fuel resources, and the cost of electricity appears to be cheap compared with that from a light water reactor. Critical physics and engineering issues for performing the ARTEMIS-L reactor are clarified. (author)

  15. New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses

    Energy Technology Data Exchange (ETDEWEB)

    Nikitin, Evgeny [Ecole Polytechnique Federale de Lausanne (Switzerland); Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany); Fridman, Emil; Bilodid, Yuri; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

    2017-07-15

    The reactor dynamics code DYN3D being developed at the Helmholtz-Zentrum Dresden-Rossendorf is currently under extension for Sodium cooled Fast Reactor analyses. This paper provides an overview on the new version of DYN3D to be used for SFR core calculations. The current article shortly describes the newly implemented thermal mechanical models, which can account for thermal expansion effects of the reactor core. Furthermore, the methodology used in Sodium cooled Fast Reactor analyses to generate homogenized few-group cross sections is summarized. The conducted and planned verification and validation studies are briefly presented. Related publications containing more detailed descriptions are outlined for the completeness of this overview.

  16. Course of operators of the RA-3 reactor

    International Nuclear Information System (INIS)

    Caligiuri, G.A.

    1983-01-01

    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) [es

  17. The Flamanville 3 EPR reactor; Le reacteur EPR Flamanville 3

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    On April 10. 2007, the french government authorized EDF to create on the site of Flamanville ( La Manche) a nuclear base installation containing a pressurized water EPR type reactor. This nuclear reactor, conceived by AREVA NP and EDF, is the first copy of a generation susceptible to replace later, at least partly, the French nuclear reactors at present in operation.Within the framework of its mission of technical support of the Authority of Nuclear Safety ( A.S.N.), the I.R.S.N. widely contributed successively: to define the general objectives of safety assigned to this new generation of pressurized water nuclear reactors; to analyze the options of safety proposed by EDF for the EPR project; To deepen, upstream to the authorization of creation, the evaluation of the step of safety and the measures of conception retained by EDF that have to allow to respect the objectives of safety which were notified to it. (N.C.)

  18. Use of plate fuel elements for the RA3 reactor

    International Nuclear Information System (INIS)

    Parodi, C.; Parkanski, D.; Higa, M.; Marajofsky, A.

    1992-01-01

    The RA3 reactor is a pool reactor, redesigned for 5 MW dissipation. Nineteen plates are used in each fuel element. The utilization of 20% enriched U, gives the possibility of the development of rod type fuel with Al/U 3 O 8 cermets. The thermohydraulic and neutronic conditions are studied in this work in order to satisfy the stipulated power. In addition, the fabrication conditions of Al/U 3 O 8 and Al/U 3 O 8 /Zr H 2 cermets with densities within the limits imposed by the thermohydraulics and neutronics conditions are studied. (author)

  19. Dynamics of TRIGA-3 Salazar Reactor.; Dinamica del Reactor TRIGA Mark III del Centro Nuclear de Mexico.

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo S, L F

    1991-12-31

    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).

  20. From pralines to multinationals: The economic history of Belgian chocolate

    OpenAIRE

    Garrone, Maria; Pieters, Hannah; Swinnen, Johan F. M.

    2016-01-01

    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...

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

    International Nuclear Information System (INIS)

    Momota, H.; Ishida, A.; Kohzaki, Y.

    1991-07-01

    A comprehensive design study of the D- 3 He 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- 3 He 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- 3 He 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- 3 He FRC reactor are clarified. (author)

  2. 3D computer visualization and animation of CANDU reactor core

    International Nuclear Information System (INIS)

    Qian, T.; Echlin, M.; Tonner, P.; Sur, B.

    1999-01-01

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

  3. EL3 reactor description and safety analysis report

    International Nuclear Information System (INIS)

    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 14 neutrons/cm 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 [fr

  4. Ignition access in a D-3He helical reactor

    International Nuclear Information System (INIS)

    Mitarai, Osamu

    2003-01-01

    Ignition access in a D- 3 He helical reactor is studied based on 0-dimensional particle and power balance equations for deuterium, tritium, helium-3, alpha ash, proton ash, electron density and temperature. The calculations are based on the following experimental facts observed in LHD. (author)

  5. Modelling of MOCVD reactor: new 3D approach

    International Nuclear Information System (INIS)

    Raj, E; Lisik, Z; Niedzielski, P; Ruta, L; Turczynski, M; Wang, X; Waag, A

    2014-01-01

    The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

  6. Modelling of MOCVD Reactor: New 3D Approach

    Science.gov (United States)

    Raj, E.; Lisik, Z.; Niedzielski, P.; Ruta, L.; Turczynski, M.; Wang, X.; Waag, A.

    2014-04-01

    The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

  7. Nuclear research reactor 0.5 to 3 MW

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-05-15

    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 MW{sub TH}, with a minimum thermal neutron flux of approx, 10{sup 13} n/cm{sup 2}{center_dot}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

  8. Nuclear research reactor 0.5 to 3 MW

    International Nuclear Information System (INIS)

    1992-05-01

    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 MW TH , with a minimum thermal neutron flux of approx, 10 13 n/cm 2 ·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

  9. Advances in Reactor physics, mathematics and computation. Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume 3, are divided into sessions bearing on: - poster sessions on benchmark and codes: 35 conferences - review of status of assembly spectrum codes: 9 conferences - Numerical methods in fluid mechanics and thermal hydraulics: 16 conferences - stochastic transport and methods: 7 conferences.

  10. The World's Reactors no. 70 - Forsmark 3, BWR-75

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    A large pull-out wall chart is presented showing a coloured cut-away diagram of the Forsmark 3 station. It is accompanied by 2 small sketches one showing the layout of station buildings and the other the inside of the reactor vessel. Parameters are listed. (U.K.)

  11. 3D simulation of CANDU reactor regulating system

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  12. Economic evaluation of reprocessing - Indicative Belgian position

    International Nuclear Information System (INIS)

    1979-05-01

    This paper, which also appears as an Appendix to the final Working Group 4 report, forms part of the overall economic evaluation of reprocessing. The indicative national position and illustrative ''phase diagram'' for Belgium is presented. Other factors which influence the Belgian viewpoint and which are not included on the phase diagram are given

  13. Performance communication of the Belgian railway

    NARCIS (Netherlands)

    Gelders, Dave; Verckens, Jan Pieter; Galetzka, Mirjam; Seydel, E.R.

    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 of

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

    International Nuclear Information System (INIS)

    Goldschmidt, Adrian; Novara, Oscar; Lafuente, Jose

    2002-01-01

    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)

  15. Genetic parameters for chronic progressive lymphedema in Belgian Draught Horses.

    Science.gov (United States)

    De Keyser, K; Janssens, S; Peeters, L M; Foqué, N; Gasthuys, F; Oosterlinck, M; Buys, N

    2014-12-01

    Genetic parameters for chronic progressive lymphedema (CPL)-associated traits in Belgian Draught Horses were estimated, using a multitrait animal model. Clinical scores of CPL in the four limbs/horse (CPLclin ), skinfold thickness and hair samples (hair diameter) were studied. Due to CPLclin uncertainty in younger horses (progressive CPL character), a restricted data set (D_3+) was formed, excluding records from horses under 3 years from the complete data set (D_full). Age, gender, coat colour and limb hair pigmentation were included as fixed, permanent environment and date of recording as random effects. Higher CPLclin certainty (D_3+) increased heritability coefficients of, and genetic correlations between traits, with CPLclin heritabilities (SE) for the respective data sets: 0.11 (0.06) and 0.26 (0.05). A large proportion of the CPLclin variance was attributed to the permanent environmental effect in D_full, but less in D_3+. Date of recording explained a proportion of variance from 0.09 ± 0.03 to 0.61 ± 0.08. Additive genetic correlations between CPLclin and both skinfold thickness and hair diameter showed the latter two traits cannot be used as a direct diagnostic aid for CPL. Due to the relatively low heritability of CPLclin , selection should focus on estimated breeding values (from repeated clinical examinations) to reduce CPL occurrence in the Belgian Draught Horse. © 2014 Blackwell Verlag GmbH.

  16. 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)

  17. Reactor safety study applied to the Forsmark 3 Power Plant

    International Nuclear Information System (INIS)

    Ericsson, G.; Tiren, L.I.

    1978-01-01

    A reactor safety study of the Forsmark 3 BWR power plant has been carried out for the purpose of calculating core melt probabilities using WASH-1400 methods. A sensitivity analysis shows that the calculated core melt probability is changed by approximately a factor of 10 depending on assumptions made with respect to the probability of human error. The importance of the availability of off-site power and the influence of common cause failure is also discussed. (author)

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

    International Nuclear Information System (INIS)

    Wang, Jy-An John; Subramani, Ranjit

    2008-01-01

    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

  19. 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

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

    International Nuclear Information System (INIS)

    Liu Ping

    2001-01-01

    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)

  1. 3 D flow computations under a reactor vessel closure head

    International Nuclear Information System (INIS)

    Daubert, O.; Bonnin, O.; Hofmann, F.; Hecker, M.

    1995-12-01

    The flow under a vessel cover of a pressurised water reactor is investigated by using several computations and a physical model. The case presented here is turbulent, isothermal and incompressible. Computations are made with N3S code using a k-epsilon model. Comparisons between numerical and experimental results are on the whole satisfying. Some local improvements are expected either with more sophisticated turbulence models or with mesh refinements automatically computed by using the adaptive meshing technique which has been just implemented in N3S for 3D cases. (authors). 6 refs., 7 figs

  2. Belgian guidelines for economic evaluations: second edition.

    Science.gov (United States)

    Thiry, Nancy; Neyt, Mattias; Van De Sande, Stefaan; Cleemput, Irina

    2014-12-01

    The aim of this study was to present the updated methodological guidelines for economic evaluations of healthcare interventions (drugs, medical devices, and other interventions) in Belgium. The update of the guidelines was performed by three Belgian health economists following feedback from users of the former guidelines and personal experience. The updated guidelines were discussed with a multidisciplinary team consisting of other health economists, assessors of reimbursement request files, representatives of Belgian databases and representatives of the drugs and medical devices industry. The final document was validated by three external validators that were not involved in the previous discussions. The guidelines give methodological guidance for the following components of an economic evaluation: literature review, perspective of the evaluation, definition of the target population, choice of the comparator, analytic technique and study design, calculation of costs, valuation of outcomes, definition of the time horizon, modeling, handling uncertainty and discounting. We present a reference case that can be considered as the minimal requirement for Belgian economic evaluations of health interventions. These guidelines will improve the methodological quality, transparency and uniformity of the economic evaluations performed in Belgium. The guidelines will also provide support to the researchers and assessors performing or evaluating economic evaluations.

  3. Non-suicidal self-injury among Dutch and Belgian adolescents: Personality, stress and coping

    NARCIS (Netherlands)

    Kiekens, G.; Bruffaerts, R.; Nock, M.K.; Ven, M.O.M. van de; Witteman, C.L.M.; Mortier, P.; Demyttenaere, K.; Claes, L.

    2015-01-01

    Background This study examines: (1) the prevalence of Non-Suicidal Self-Injury (NSSI) among Dutch and Belgian adolescents, (2) the associations between Big Five personality traits and NSSI engagement/versatility (i.e., number of NSSI methods), and (3) whether these associations are mediated by

  4. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  5. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

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

    International Nuclear Information System (INIS)

    Ruan, Da

    2004-01-01

    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)

  7. The ARIES-III D-3He tokamak reactor

    International Nuclear Information System (INIS)

    Bathke, C.G.; Werley, K.A.; Miller, R.L.; Krakowski, R.A.; Santarius, J.F.

    1992-01-01

    The multi-institutional ARIES study has generated a conceptual design of another tokamak fusion reactor in a series that varies the assumed advances in technology and physics. The ARIES-III design uses a D- 3 He fuel cycle and requires advances in technology and physics for economical attractiveness. The optimal design was characterized through systems analyses for eventual conceptual engineering design. In this paper, results from the systems analysis are summarized, and a comparison with the high-field, D-T fueled ARIES-I is included

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

    International Nuclear Information System (INIS)

    Vicens, Hugo E.; Quintana, Jorge A.

    2001-01-01

    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

  9. Cultural differences in complex addition: efficient Chinese versus adaptive Belgians and Canadians.

    Science.gov (United States)

    Imbo, Ineke; LeFevre, Jo-Anne

    2009-11-01

    In the present study, the authors tested the effects of working-memory load on math problem solving in 3 different cultures: Flemish-speaking Belgians, English-speaking Canadians, and Chinese-speaking Chinese currently living in Canada. Participants solved complex addition problems (e.g., 58 + 76) in no-load and working-memory load conditions, in which either the central executive or the phonological loop was loaded. The authors used the choice/no-choice method to obtain unbiased measures of strategy selection and strategy efficiency. The Chinese participants were faster than the Belgians, who were faster and more accurate than the Canadians. The Chinese also required fewer working-memory resources than did the Belgians and Canadians. However, the Chinese chose less adaptively from the available strategies than did the Belgians and Canadians. These cultural differences in math problem solving are likely the result of different instructional approaches during elementary school (practice and training in Asian countries vs. exploration and flexibility in non-Asian countries), differences in the number language, and informal cultural norms and standards. The relevance of being adaptive is discussed as well as the implications of the results in regards to the strategy choice and discovery simulation model of strategy selection (J. Shrager & R. S. Siegler, 1998).

  10. New reactor concepts

    International Nuclear Information System (INIS)

    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

  11. Measurement of the natural radiation of the Belgian territory

    International Nuclear Information System (INIS)

    Gillard, J.; Flemal, J.M.; Deworm, J.P.; Slegers, W.

    1989-01-01

    A measurement campaign of natural occuring radionuclides was set up on the Belgian territory in order to assess the doses received by the Belgian population. The results of the measurements are published together with a map of natural occuring radionuclides and exposure rates. (L.D.C.)

  12. Physics analysis of the Apollo D-3He tokamak reactor

    International Nuclear Information System (INIS)

    Santarius, J.F.; Emmert, G.A.

    1990-01-01

    Recent developments in the analysis and conceptual design of Apollo, a D- 3 He Tokamak Reactor are presented. Encouraging experimental results on TEXT motivated a key change in the Apollo concept utilization of an ergodic magnetic limiter for impurity control instead of a divertor. Parameters for the updated Apollo design and an analysis of the ergoidc magnetic limiter are given. The Apollo reference case uses direct conversion of synchrotron radiation to electricity by rectifying antennas (rectennas) for its power conversion system. Previous analyses of this concept are expanded, including further details of the rectennas and of the loss of synchrotron power to the waveguides and walls. Although Apollo will burn D- 3 He fuel, a significant amount of unburned tritium will be generated by D4D reactions. The possibility of operating a short, dedicated, T+ 3 He burn phase to eliminate this tritium will be examined

  13. Atomization of U3Si2 for research reactor fuel

    International Nuclear Information System (INIS)

    Kim, C.K.; Kim, K.H.; Lee, C.T.; Kuk, I.H.

    1995-01-01

    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 U 3 Si and the mixture of U 3 Si 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 U 3 Si 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)

  14. Risk-return of Belgian SRI funds

    OpenAIRE

    Van Liedekerke, Luc; De Moor, Lieven; Vanwalleghem, Dieter

    2007-01-01

    We analyse the risk-return profile of Belgian SRI funds versus conventional investment funds. We apply a four-factor conditional Carhart model to establish whether there are significant differences in risk-return profile between an SRI portfolio and a conventional portfolio and test for learning effects in SRI funds. We show that there is no difference in risk-return profile between SRI and conventional funds. If return is not the problem, then what is it that limits the development of an SRI...

  15. Analyses for MARIA Research Reactor with RELAP/MOD3 code

    International Nuclear Information System (INIS)

    Szczurek, J.; Czerski, P.

    2004-01-01

    This paper deals with the application of the RELAP5/MOD3 code to the transient analyses for MARIA research reactor. Poland's MARIA Research Reactor is water and beryllium moderated, water-cooled reactor of a pool type with pressurized fuel channels containing concentric multi-tube assemblies of highly enriched uranium clad in aluminium. The RELAP5/MOD3 input data model includes the whole primary cooling circuit of the MARIA reactor. The model was qualified against the reactor data at steady state conditions and additionally against the existing reliable experimental data for a transient initiated by the reactor scram. The RELAP transient simulation was performed for loss of forced flow accidents including two scenarios with protected and unprotected (no scram) reactor core. Calculations allow estimating time margin for reactor scram initiation and reactivity feedbacks contribution to the results. (author)

  16. RA reactor exploitation, task 3.08/01

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-01-01

    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

  17. Direct harvesting of Helium-3 (3He) from heavy water nuclear reactors

    International Nuclear Information System (INIS)

    Bentoumi, G.; Didsbury, R.; Jonkmans, G.; Rodrigo, L.; Sur, B.

    2013-01-01

    The thermal neutron activation of deuterium inside a heavy-water-moderated or -cooled nuclear reactor produces a build-up of tritium in the heavy water. The in situ decay of tritium can, for certain reactor types and operating conditions, produce potentially useable amounts of 3 He, which can be directly extracted via the heavy-water cover gas without first separating, collecting and storing tritium outside the reactor. It is estimated that the amount of 3 He available for recovery from the moderator cover gas of a 700 MWe class Pressurized Heavy Water Reactor (PHWR) ranges from 0.1 to 0.7 m 3 (STP) per annum, varying with the tritium activity buildup in the moderator. The harvesting of 3 He would generate approximately 12.7 m 3 (STP) of 3 He, worth more than $30M at current market rates, over a typical 25-year operating cycle of the PHWR. This paper discusses the production of 3 He in the moderator of a PHWR and its extraction from the 4 He moderator cover gas system using conventional methods. (author)

  18. 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.

  19. Ageing management of the BR2 research reactor

    International Nuclear Information System (INIS)

    Verpoortem, J. R.; Van Dyck, S.

    2014-01-01

    At the Belgian nuclear research centre (SCK.CEN) several test reactors are operated. Among these, Belgian Reactor 2 (BR2) is the largest Material Test Reactor (MTR). This water-cooled, beryllium moderated reactor with a maximum thermal power of 100 MW became operational in 1962. Except for two major refurbishment campaigns of one year each, this reactor has been operated continuously over the past 50 years, with a frequency of 5-12 cycles per year. At present, BR2 is used for different research activities, the production of medical isotopes, the production of n-doped silicon and various training and education activities. (Author)

  20. Decommissioning of the BR3 PWR

    International Nuclear Information System (INIS)

    Massaut, V.

    1998-01-01

    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

  1. 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.

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

    International Nuclear Information System (INIS)

    Bencik, V.; Debrecin, N.; Foretic, D.

    2003-01-01

    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)

  3. 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.

  4. Irradiation of an uranium silicide prototype in RA-3 reactor

    International Nuclear Information System (INIS)

    Calabrese, R.; Estrik, G.; Notari, C.

    1996-01-01

    The factibility of irradiation of an uranium silicide (U 3 Si 2 ) prototype in the RA-3 reactor was studied. The standard RA-3 fuel element uses U 3 O 8 as fissible material. The enrichment of both standard and prototype is the same: 20% U 235 and also the frame geometry and number of plates is identical. The differences are in the plate dimensions and the fissile content which is higher in the prototype. The cooling conditions of the core allow the insertion of the prototype in any core position, even near the water trap, if the overall power is kept below 5Mw. Nevertheless, the recommendation was to begin irradiation near the periphery and later on move the prototype towards more central positions in order to increase the burnup rate. The prototype was effectively introduced in a peripheral position and the thermal fluxes were measured between plates with the foil activation technique. These were also evaluated with the fuel management codes and a reasonable agreement was found. (author). 5 refs., 3 figs., 3 tabs

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

    International Nuclear Information System (INIS)

    1983-07-01

    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

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

    International Nuclear Information System (INIS)

    2004-01-01

    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

  7. Construction of Research Reactors for Gen 3 and Gen 4 Reactors Development

    International Nuclear Information System (INIS)

    Behar, Christophe

    2014-01-01

    Christophe Behar, Director of the Nuclear Energy Division at CEA, detailed the different kind of research reactors and the issues in term of investment, use, side application such as the medical isotopes production

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

    International Nuclear Information System (INIS)

    Usui, Hozumi; Fujii, Yoshio; Shinohara, Yoshikuni

    1991-01-01

    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)

  9. Focal epilepsy in the Belgian shepherd

    DEFF Research Database (Denmark)

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

    2009-01-01

    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......, were included in the calculations of inheritance. results: Out of 199 family members, 66 dogs suffered from epilepsy. The prevalence of epilepsy in the family was 33%. Fifty-five dogs experienced focal seizures with or without secondary generalisation, while four dogs experienced primary generalised...... 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...

  10. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  11. Study on the reactivity behavior partially loaded reactor cores using SIMULATE-3

    International Nuclear Information System (INIS)

    Holzer, Robert; Zeitz, Andreas; Grimminger, Werner; Lubczyk, Tobias

    2009-01-01

    The reactor core design for the NPP Gundremmingen unit B and C is performed since several years using the validated 3D reactor core calculation program SIMULATE-3. The authors describe a special application of the program to study the reactivity for different partial core loadings. Based on the comparison with results of the program CASMO-4 the program SIMULATE-3 was validated for the calculation of partially loaded reactor cores. For the planned reactor operation in NPP Gundremmingen using new MOX fuel elements the reactivity behavior was studied with respect to the KTA-Code requirements.

  12. 3D CAD model of the subcritical nuclear reactor of IPN; Modelo CAD 3D del reactor nuclear subcritico del IPN

    Energy Technology Data Exchange (ETDEWEB)

    Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ibarra R, G.; Del Valle G, E.; Sanchez R, A., E-mail: narehc@hotmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN, Edif. 9, Unidad Profesional Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico)

    2016-09-15

    The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

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

    International Nuclear Information System (INIS)

    Brognon, T.

    1993-01-01

    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.)

  14. 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

  15. The Health Risks of Belgian Illicit Indoor Cannabis Plantations.

    Science.gov (United States)

    Vanhove, Wouter; Cuypers, Eva; Bonneure, Arne-Jan; Gotink, Joachim; Stassen, Mirna; Tytgat, Jan; Van Damme, Patrick

    2018-04-10

    We assessed the prevalence of potential health hazards to intervention staff and cannabis growers in Belgian indoor cannabis plantations. Surface mold swab samples were taken at 16 Belgian indoor plantations contained mostly Penicillium sp. and Aspergillus sp. However, their precise health impact on intervention staff and illicit growers is unclear as no molds spore concentrations were measured. Atmospheric gas monitoring in the studied cannabis plantations did not reveal dangerous toxic substances. Health symptoms were reported by 60% of 221 surveyed police, but could not be linked to specific plantation characteristics. We conclude that Belgian indoor cannabis plantations pose a potential health threat to growers and intervention staff. AS there are currently no clear safety guidelines for seizure and dismantling of Belgian indoor cannabis plantations, we recommend first responders to follow strict safety rules when entering the growth rooms, which include wearing appropriate personal protective equipment. © 2018 American Academy of Forensic Sciences.

  16. Fusion reactor control study. Volume 3. Tandem mirror reactors. Final report

    International Nuclear Information System (INIS)

    Chang, F.R.; DeCanio, F.; Fisher, J.L.; Madden, P.A.

    1982-03-01

    A study of the control requirements of the Tandem Mirror Reactor concept is reported. The study describes the development of a control simulator that is based upon a spatially averaged physics code of the reactor concept. The simulator portrays the evolution of the plasma through the complete reactor operating cycle; it includes models of the control and measurement system, thus allowing the exploration of various strategies for reactor control. Startup, shutdown, and control during the quasi-steady-state power producing phase were explored. Configurations are described which use a variety of control effectors including modulation of the refueling rate, beam current, and electron cyclotron resonance heating. Multivariable design techniques were used to design the control laws and compensators for the feedback controllers and presume the practical measurement of only a subset of the plasma and machine variables. Performance of the various controllers is explored using the nonlinear control simulator. Derivative control strategies using new or developed sensors and effectors appropriate to a power reactor environment are postulated, based upon the results of the control configurations tested. Research and development requirements for these controls are delineated

  17. Comparison of 2D and 3D Neutron Transport Analyses on Yonggwang Unit 3 Reactor

    International Nuclear Information System (INIS)

    Maeng, Aoung Jae; Kim, Byoung Chul; Lim, Mi Joung; Kim, Kyung Sik; Jeon, Young Kyou; Yoo, Choon Sung

    2012-01-01

    10 CFR Part 50 Appendix H requires periodical surveillance program in the reactor vessel (RV) belt line region of light water nuclear power plant to check vessel integrity resulting from the exposure to neutron irradiation and thermal environment. Exact exposure analysis of the neutron fluence based on right modeling and simulations is the most important in the evaluation. Traditional 2 dimensional (D) and 1D synthesis methodologies have been widely applied to evaluate the fast neutron (E > 1.0 MeV) fluence exposure to RV. However, 2D and 1D methodologies have not provided accurate fast neutron fluence evaluation at elevations far above or below the active core region. RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries) program for 3D geometries calculation was therefore developed both by Westinghouse Electronic Company, USA and Korea Reactor Integrity Surveillance Technology (KRIST) for the analysis of In-Vessel Surveillance Test and Ex-Vessel Neutron Dosimetry (EVND). Especially EVND which is installed at active core height between biological shielding material and concrete also evaluates axial neutron fluence by placing three dosimetries each at Top, Middle and Bottom part of the angle representing maximum neutron fluence. The EVND programs have been applied to the Korea Nuclear Plants. The objective of this study is therefore to compare the 3D and the 2D Neutron Transport Calculations and Analyses on the Yonggwang unit 3 Reactor as an example

  18. Development of 3D CFD simulation method in nuclear reactor safety analysis

    International Nuclear Information System (INIS)

    Rosli Darmawan; Mariah Adam

    2012-01-01

    One of the most prevailing issues in the operation of nuclear reactor is the safety of the system. Worldwide publicity on a few nuclear accidents as well as the notorious Hiroshima and Nagasaki bombing have always brought about public fear on anything related to nuclear. Most findings on the nuclear reactor accidents are closely related to the reactor cooling system. Thus, the understanding of the behaviour of reactor cooling system is very important to ensure the development and improvement on safety can be continuously done. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last three decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. This paper discusses the development of 3D CFD usage in nuclear reactor safety analysis worldwide. A brief review on the usage of CFD at Malaysia's Reactor TRIGA PUSPATI is also presented. (author)

  19. Shaba II: The French and Belgian Intervention in Zaire in 1978

    Science.gov (United States)

    1993-04-01

    morning. By 0500, the sun was climbing. Some of the foreigners were awake and preparing breakfast. Some looking out their doors found the streets...project and was equipped with its own aircraft. With two DC-3s, one Piper Aztec , and two Bell 206 helicopters, the company was already considering an...flew in to compliment the French and Belgians on the operation. Taken to visit the massacre site in the new town, the shocked dictator gasped, "My God

  20. A quality control exercise of radionuclide calibrators among Belgian hospitals

    International Nuclear Information System (INIS)

    Reher, D.F.G.; Merlo, P.

    1990-01-01

    On the initiative of the Belgian Association of Hospital Physicists, eleven Belgian hospitals participated in a quality control of radionuclide calibrators conducted in collaboration with the Central Bureau for Nuclear Measurements of the Commission of the European Communities. For practical reasons the nuclide 57 Co was chosen. The results from 20 different radionuclide calibrators show a fair agreement with a similar comparison carried out in 1980 in the UK. (orig.)

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

    International Nuclear Information System (INIS)

    Viais J, J.

    1994-01-01

    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)

  2. Estimation of reactor pool water temperature after shutdown in JRR-3M

    International Nuclear Information System (INIS)

    Yagi, Masahiro; Sato, Mitsugu; Kakefuda, Kazuhiro

    1999-01-01

    The reactor pool water temperature increasing by the decay heat was estimated by calculation. The reactor pool water temperature was calculated by increased enthalpy that was estimated by the reactor decay heat, the heat released from the reactor biological shielding concrete, reactor pool water surface, the heat conduction from the canal and the core inlet piping. These results of calculation were compared with the past measured data. As the results of estimation, after the JRR-3M shutdown, the calculated reactor pool temperature first increased sharply. This is because the decay heat was the major contribution. And then, rate of increased reactor pool temperature decreased. This is because the ratio of heat released from reactor biological shielding concrete and core inlet piping to the decay heat increased. Besides, the calculated reactor pool water temperature agreed with the past measured data in consequence of correcting the decay heat and the released heat. The corrected coefficient k 1 of decay heat was 0.74 - 0.80. And the corrected coefficient k 2 of heat released from the reactor biological shielding concrete was 3.5 - 4.5. (author)

  3. Study for improvement of light water reactor technology, (3)

    International Nuclear Information System (INIS)

    Suzuki, Hideaki; Morita, Terumichi; Igarashi, Hiroshi; Tabata, Hiroaki

    1991-01-01

    The Japan Atomic Power Company has performed some studies, which are referred to as 'some feasibility studies of LWR technology', in order to help improve and up-grade the light water reactor technology. We would like to show the key results of the above studies in an orderly fashion in this document. As the third issue, this paper describes the study of the feasibility of applying a suppression pool system in a 4-loop PWR plant in order to reduce containment volume and evaluates the merits of such a system. The results confirmed the feasibility of such a plant consisting of a 4-loop plant with a suppression pool system. The expected merits of a suppression pool type PWR are as follows: (1) The volume within the containment boundary is half of that for the conventional plant. This reduces the material quantity substantially. (2) A wider layout space is obtained since the operating floor is located outside the containment are. And this improves the maneuverability of plant outage. (3) Low center of gravity of the plant contributes to improving the ability to withstand seismic activity. Although there are some open items left that should be confirmed, we consider that PWR with small CV is an appealing plant in the light of further sales points such as relaxing siting conditions, extending the use of robotics and so on. (author)

  4. Level 3 decommissioning of Triton - Nereide research reactor

    International Nuclear Information System (INIS)

    Lopes, E.; Pillette-Cousin, L.

    2002-01-01

    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)

  5. Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

    International Nuclear Information System (INIS)

    Edwards, Robert; Huang, Zhengyu

    2001-01-01

    The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been expanded for boiling water reactor (BWR) out-of-phase behavior. During BWR out-of-phase oscillation half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. A description of the new HRS is given; three computers are employed to handle all the computations required, including real-time data processing and graph generation. BWR out-of-phase oscillation was successfully simulated. By adjusting the reactivity feedback gains from boiling channels to the TRIGA reactor and to the first harmonic mode power simulation, limit cycle can be generated with both reactor power and the simulated first harmonic power. A 3-D display of spatial power distributions of fundamental mode, first harmonic, and total powers over the reactor cross section is shown

  6. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

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

    International Nuclear Information System (INIS)

    Chambaud, B.

    1974-01-01

    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. Antibodies to elastin peptides in sera of Belgian Draught horses with chronic progressive lymphoedema.

    Science.gov (United States)

    van Brantegem, L; de Cock, H E V; Affolter, V K; Duchateau, L; Hoogewijs, M K; Govaere, J; Ferraro, G L; Ducatelle, R

    2007-09-01

    Chronic progressive lymphoedema (CPL) is a recently recognised disease of the lymphatic system characterised by lesions in the skin of the lower legs in several draught horse breeds, including the Belgian Draught hourse. Clinical signs slowly progress and result in severe disfigurement of the limbs. Ideally, supportive treatment should be started early in the disease process. However early diagnosis and monitoring progression of CPL is still a challenge. Elastin changes, characterised by morphological alterations as well as increased desmosine levels, in the skin of the distal limbs of horses affected with CPL are probably associated with a marked release of elastin degradation products, which elicit production of circulating anti-elastin antibodies (AEAbs) in the serum. An enzyme-linked immunosorbent assay (ELISA) for detection of serum AEAbs may document elastin breakdown. An ELISA technique was used to evaluate levels of AEAbs in sera of 97 affected Belgian Draught horses that were clinically healthy except for possible skin lesions, associated with CPL in their distal limbs. The horses were divided into 5 groups according to the severity of these skin lesions: normal horses (Group 1, n = 36), horses with mild lesions (Group 2, n = 43), horses with moderate lesions (Group 3, n = 8), horses with severe lesions (Group 4, n = 10) and, as a control, healthy Warmblood horses, unaffected by the disease (Group 5, n = 83). Horses with clinical signs of CPL had significantly higher AEAb levels compared to clinically normal Belgian Draught horses and to healthy Warmblood horses. These levels correlated with severity of lesions. CPL in draught horses is associated with an increase of serum AEAbs. Evaluation of serum levels of AEAbs by ELISA might be a useful diagnostic aid for CPL. Pathological degradation of elastic fibres, resulting in deficient support of the distal lymphatics, is proposed as a contributing factor for CPL in Belgian Draught horses.

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

    International Nuclear Information System (INIS)

    1966-01-01

    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

  10. AMNT 2014. Key topic: Reactor operation, safety - report. Pt. 3

    Energy Technology Data Exchange (ETDEWEB)

    Bohnstedt, Angelika [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany). Programm Nukleare Sicherheitsforschung (NUKLEAR); Mull, Thomas [AREVA GmbH, Erlangen (Germany). Nuclear Fusion, HTR and Transverse Issues (PTDH-G); Starflinger, Joerg [Stuttgart Univ. (Germany). Inst. fuer Kernenergetik und Energiesysteme (IKE)

    2015-01-15

    Summary report on the following sessions of the Annual Conference on Nuclear Technology held in Frankfurt, 6 to 8 May 2014: - Reactor Operation, Safety: Radiation Protection (Angelika Bohnstedt); - Competence, Innovation, Regulation: Fusion Technology - Optimisation Steps in the ITER Design (Thomas Mull); - Competence, Innovation, Regulation: Education, Expert Knowledge, Knowledge Transfer (Joerg Starflinger). The other Sessions of the Key Topics 'Reactor Operation, Safety', 'Competence, Innovation, Regulation' and 'Fuel, Decommissioning and Disposal' have been covered in atw 10 and 12 (2015) and will be covered in further issues of atw.

  11. 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

  12. 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). Elle est

  13. Mechanical degradation processes: The Belgian experience

    International Nuclear Information System (INIS)

    Lafaille, J.P.; Hennart, J.C.

    1998-01-01

    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)

  14. Antimicrobial use in Belgian broiler production.

    Science.gov (United States)

    Persoons, Davy; Dewulf, Jeroen; Smet, Annemieke; Herman, Lieve; Heyndrickx, Marc; Martel, An; Catry, Boudewijn; Butaye, Patrick; Haesebrouck, Freddy

    2012-08-01

    The use of antimicrobials in production animals has become a worldwide concern in the face of rising resistance levels in commensal, pathogenic and zoonotic bacteria. In the years 2007 and 2008 antimicrobial consumption records were collected during two non consecutive production cycles in 32 randomly selected Belgian broiler farms. Antimicrobials were used in 48 of the 64 monitored production cycles, 7 farms did not use any antimicrobials in both production cycles, 2 farms only administered antimicrobials in one of the two production cycles, the other 23 farms applied antimicrobial treatment in both production cycles. For the quantification of antimicrobial drug use, the treatment incidences (TI) based on the defined daily doses (the dose as it should be applied: DDD) and used daily doses (the actual dose applied: UDD) were calculated. A mean antimicrobial treatment incidence per 1000 animals of 131.8 (standard deviation 126.8) animals treated daily with one DDD and 121.4 (SD 106.7) animals treated daily with one UDD was found. The most frequently used compounds were amoxicillin, tylosin and trimethoprim-sulphonamide with a mean TI(UDD) of 37.9, 34.8, and 21.7, respectively. The ratio of the UDD/DDD gives an estimate on correctness of dosing. Tylosin was underdosed in most of the administrations whereas amoxicillin and trimethoprim-sulphonamide were slightly overdosed in the average flock. Copyright © 2012 Elsevier B.V. All rights reserved.

  15. Report on the Survey of the Design Review of New Reactor Applications. Volume 3: Reactor

    International Nuclear Information System (INIS)

    Downey, Steven; Monninger, John; Nevalainen, Janne; Lorin, Aurelie; ); Webster, Philip; Joyer, Philippe; Kawamura, Tomonori; Lankin, Mikhail; Kubanyi, Jozef; Haluska, Ladislav; Persic, Andreja; Reierson, Craig; Kang, Kyungmin; Kim, Walter

    2016-01-01

    At the tenth meeting of the CNRA Working Group on the Regulation of New Reactors (WGRNR) in March 2013, the Working Group agreed to present the responses to the Second Phase, or Design Phase, of the Licensing Process Survey as a multi-volume text. As such, each report will focus on one of the eleven general technical categories covered in the survey. The general technical categories were selected to conform to the topics covered in the International Atomic Energy Agency (IAEA) Safety Guide GS-G-4.1. This document, which is the third report on the results of the Design Phase Survey, focuses on the Reactor. The Reactor category includes the following technical topics: fuel system design, reactor internals and core support, nuclear design and core nuclear performance, thermal and hydraulic design, reactor materials, and functional design of reactivity control system. For each technical topic, the member countries described the information provided by the applicant, the scope and level of detail of the technical review, the technical basis for granting regulatory authorisation, the skill sets required and the level of effort needed to perform the review. Based on a comparison of the information provided by the member countries in response to the survey, the following observations were made: - Although the description of the information provided by the applicant differs in scope and level of detail among the member countries that provided responses, there are similarities in the information that is required. - All of the technical topics covered in the survey are reviewed in some manner by all of the regulatory authorities that provided responses. - Design review strategies most commonly used to confirm that the regulatory requirements have been met include document review and independent verification of calculations, computer codes, or models used to describe the design and performance of the core and the fuel. - It is common to consider operating experience and

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

    International Nuclear Information System (INIS)

    Shi, Dunfu

    2015-01-01

    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

  17. Terminal patients in Belgian nursing homes: a cost analysis.

    Science.gov (United States)

    Simoens, Steven; Kutten, Betty; Keirse, Emmanuel; Vanden Berghe, Paul; Beguin, Claire; Desmedt, Marianne; Deveugele, Myriam; Léonard, Christian; Paulus, Dominique; Menten, Johan

    2013-06-01

    Policy makers and health care payers are concerned about the costs of treating terminal patients. This study was done to measure the costs of treating terminal patients during the final month of life in a sample of Belgian nursing homes from the health care payer perspective. Also, this study compares the costs of palliative care with those of usual care. This multicenter, retrospective cohort study enrolled terminal patients from a representative sample of nursing homes. Health care costs included fixed nursing home costs, medical fees, pharmacy charges, other charges, and eventual hospitalization costs. Data sources consisted of accountancy and invoice data. The analysis calculated costs per patient during the final month of life at 2007/2008 prices. Nineteen nursing homes participated in the study, generating a total of 181 patients. Total mean nursing home costs amounted to 3,243 € per patient during the final month of life. Total mean nursing home costs per patient of 3,822 € for patients receiving usual care were higher than costs of 2,456 € for patients receiving palliative care (p = 0.068). Higher costs of usual care were driven by higher hospitalization costs (p < 0.001). This study suggests that palliative care models in nursing homes need to be supported because such care models appear to be less expensive than usual care and because such care models are likely to better reflect the needs of terminal patients.

  18. Severe polysaccharide storage myopathy in Belgian and Percheron draught horses.

    Science.gov (United States)

    Valentine, B A; Credille, K M; Lavoie, J P; Fatone, S; Guard, C; Cummings, J F; Cooper, B J

    1997-05-01

    A severe myopathy leading to death or euthanasia was identified in 4 Belgian and 4 Percheron draught horses age 2-21 years. Clinical signs ranged from overt weakness and muscle atrophy in 2 horses age 2 and 3 years, to recumbency with inability to rise in 6 horses age 4-21 years. In 5 horses there was mild to severe increases in muscle enzyme levels. Clinical diagnoses included equine motor neuron disease (2 horses), post anaesthetic myopathy (2 horses), exertional myopathy (2 horses), myopathy due to unknown (one horse), and equine protozoal myelitis (one horse). Characteristic histopathology of muscle from affected horses was the presence of excessive complex polysaccharide and/or glycogen, revealed by periodic acid-Schiff staining in all cases and by electron microscopy in one case. Evaluation of frozen section histochemistry performed on 2 cases indicated that affected fibres were Type 2 glycolytic fibres. Subsarcolemmal and intracytoplasmic vacuoles were most prominent in 3 horses age 2-4 years, and excessive glycogen, with little or no complex polysaccharide, was the primary compound stored in affected muscle in these young horses. Myopathic changes, including fibre size variation, fibre hypertrophy, internal nuclei, and interstitial fat infiltration, were most prominent in 5 horses age 6-21 years, and the accumulation of complex polysaccharide appeared to increase with age. Mild to moderate segmental myofibre necrosis was present in all cases.

  19. The Chernobyl reactor accident and its consequences. 3. rev. ed.

    International Nuclear Information System (INIS)

    1987-01-01

    The report presents a comprehensive survey of measured data explaining the radiation exposure in the Land Hessen, and a chronological survey of the decisions and measures taken by the Hessian regional government in response to the Chernobyl reactor accident. The measures for instance included selection of appropriate measuring methods and sites, checking of various environmental material, waste disposal surveillance, and dose assessments, and a range of monitoring programmes. (PW) [de

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

    International Nuclear Information System (INIS)

    Revankar, S. T.

    2004-01-01

    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

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

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-12-01

    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

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

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi

    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.)

  3. 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.)

  4. Inventory of nuclear liabilities - The Belgian perspective

    International Nuclear Information System (INIS)

    Minon, Jean-Paul

    2003-01-01

    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

  5. Security and Privacy Improvements for the Belgian eID Technology

    Science.gov (United States)

    Verhaeghe, Pieter; Lapon, Jorn; de Decker, Bart; Naessens, Vincent; Verslype, Kristof

    The Belgian Electronic Identity Card enables Belgian citizens to prove their identity digitally and to sign electronic documents. At the end of 2009, every Belgian citizen older than 12 years will have such an eID card. In the future, usage of the eID card may be mandatory. However, irresponsible use of the card may cause harm to individuals.

  6. The Belgian Nuclear Higher Education Network: Your way to the European Master in Nuclear Engineering

    International Nuclear Information System (INIS)

    Moons, F.; D'haeseleer, W.; Giot, M.

    2004-01-01

    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. More information: http://www.sckcen.be/BNEN. (author)

  7. 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.

  8. Benchmark tests of JENDL-3.2 for thermal and fast reactors

    International Nuclear Information System (INIS)

    Takano, Hideki

    1995-01-01

    Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

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

    International Nuclear Information System (INIS)

    Gonzalez M, J.L.

    1994-01-01

    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)

  10. The simplified P3 approach on a trigonal geometry in the nodal reactor code DYN3D

    International Nuclear Information System (INIS)

    Duerigen, S.; Fridman, E.

    2011-01-01

    DYN3D is a three-dimensional nodal diffusion code for steady-state and transient analyses of Light-Water Reactors with square and hexagonal fuel assembly geometries. Currently, several versions of the DYN3D code are available including a multi-group diffusion and a simplified P 3 (SP 3 ) neutron transport option. In this work, the multi-group SP 3 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 WWER-type Pressurized Water Reactors as well as for innovative reactor concepts including block type High-Temperature Reactors and Sodium Fast Reactors. In this paper, the theoretical background for the trigonal SP 3 methodology is outlined and the results of a preliminary verification analysis are presented by means of a simplified WWER-440 core test example. The accordant cross sections and reference solutions were produced by the Monte Carlo code SERPENT. The DYN3D results are in good agreement with the reference solutions. The average deviation in the nodal power distribution is about 1%. (Authors)

  11. Report on generation IV technical working group 3 : liquid metal reactors

    International Nuclear Information System (INIS)

    Lineberry, M. J.; Rosen, S. L.; Sagayama, Y.

    2002-01-01

    This paper reports on the first round of R and D roadmap activities of the Generation IV (Gen IV) Technical Working Group (TWG) 3, on liquid metal-cooled reactors. Liquid metal coolants give rise to fast spectrum systems, and thus the reactor systems considered in this TWG are all fast reactors. Gas-cooled fast reactors are considered in the context of TWG 2. As is noted in other Gen IV papers, this first round activity is termed ''screening for potential'', and includes collecting the most complete set of liquid metal reactor/fuel cycle system concepts possible and evaluating the concepts against the Gen IV principles and goals. Those concepts or concept groups that meet the Gen IV principles and which are deemed to have reasonable potential to meet the Gen IV goals will pass to the next round of evaluation. Although we sometimes use the terms ''reactor'' or ''reactor system'' by themselves, the scope of the investigation by TWG 3 includes not only the reactor systems, but very importantly the closed fuel recycle system inevitably required by fast reactors. The response to the DOE Request for Information (RFI) on liquid metal reactor/fuel cycle systems from principal investigators, laboratories, corporations, and other institutions, was robust and gratifying. Thirty three liquid metal concept descriptions, from eight different countries, were ultimately received. The variation in the scope, depth, and completeness of the responses created a significant challenge for the group, but the TWG made a very significant effort not to screen out concepts early in the process

  12. Ageing management of baffle former bolts in Belgian nuclear power plants

    International Nuclear Information System (INIS)

    Somville, F.; Gerard, R.; Bosch, R.W.; Bertolis, D.; Vissers, S.

    2015-01-01

    The Pressurized Water Reactors internals support the reactor core, distribute the coolant flow through the core, and guide and protect the rod control cluster assemblies and in-core instrumentation. The baffle-to-former bolts are used in Pressurized Water Reactors to attach the baffle plates to the former plates in the reactor vessel lower internals. The resulting structure forms a boundary for the flow of coolant and provides lateral support to the fuel assemblies. Some edge bolts are also present, assembling together the baffle plates. After an operating time of the order of 120.000 hours, some bolts exhibit cracking at the junction of the head and the shank of the bolt. Examinations of failed bolts have made it possible to identify the cause of cracking as irradiation assisted stress corrosion cracking (IASCC). Up to now, baffle bolt cracking has been detected in units older than 15 years, where the baffle bolts are not cooled (no holes in the former to allow a water flow on the bolt shank). In Belgium, the concerned units are Tihange 1 and Doel 1-2. The paper summarizes the experience with baffle bolts cracking in Belgian units and the strategy implemented to mitigate this problem, consisting of structural integrity analyses, baffle bolts inspections and replacement, and research programs in the field of IASCC, including examinations of highly irradiated replaced bolts. (authors)

  13. 3D CAD model of the subcritical nuclear reactor of IPN

    International Nuclear Information System (INIS)

    Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A.; Ibarra R, G.; Del Valle G, E.; Sanchez R, A.

    2016-09-01

    The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

  14. RELAP5-3D code validation of RBMK-1500 reactor reactivity measurement transients

    International Nuclear Information System (INIS)

    Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

    2003-01-01

    This paper deals with the modeling of transients taking place during the measurements of the void and fast power reactivity coefficients performed at Ignalina NPP. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Following the simulation of the two above mentioned transients with RELAP5-3D code, a conclusion was made that the obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data. Despite of the small differences, RELAP5-3D code predicts reactivity and the total reactor core power behavior during the transients in a reasonable manner. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core

  15. 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.

  16. G2 and G3 reactors design; Description des reacteurs G2 et G3

    Energy Technology Data Exchange (ETDEWEB)

    Herreng,; Ertaud,; Pasquet, [Societe Alsacienne de Constructions Mecaniques (France)

    1958-07-01

    '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 CO{sub 2} 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 CO{sub 2} 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/cm{sup 2}). Steam can be condensed in the event of a group turbo-generator stopping, with no modifion for the normal operating conditions of the reactor. Auxiliary circuits have to assure the continuous purifying of cooling CO{sub 2}, its storage and drain. 49 boron carbide rods are used to control the

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

    International Nuclear Information System (INIS)

    Fucks, I.; Hardeman, F.

    2002-01-01

    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)

  18. RA reactor exploitation, task 3.08/01; Zadatak 3.08/01 - Eksploatacija reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    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.

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

    International Nuclear Information System (INIS)

    1979-11-01

    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

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

    International Nuclear Information System (INIS)

    Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

    2005-01-01

    G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) 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 CO 2 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 ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

  1. Artificial intelligence in nuclear reactor operation

    International Nuclear Information System (INIS)

    Da Ruan; Benitez-Read, J.S.

    2005-01-01

    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)

  2. Large-signal, dynamic simulation of the slowpoke-3 nuclear heating reactor

    International Nuclear Information System (INIS)

    Tseng, C.M.; Lepp, R.M.

    1983-07-01

    A 2 MWt nuclear reactor, called SLOWPOKE-3, is being developed at the Chalk River Nuclear Laboratories (CRNL). This reactor, which is cooled by natural circulation, is designed to produce hot water for commercial space heating and perhaps generate some electricity in remote locations where the costs of alternate forms of energy are high. A large-signal, dynamic simulation of this reactor, without closed-loop control, was developed and implemented on a hybrid computer, using the basic equations of conservation of mass, energy and momentum. The natural circulation of downcomer flow in the pool was simulated using a special filter, capable of modelling various flow conditions. The simulation was then used to study the intermediate and long-term transient response of SLOWPOKE-3 to large disturbances, such as loss of heat sink, loss of regulation, daily load following, and overcooling of the reactor coolant. Results of the simulation show that none of these disturbances produce hazardous transients

  3. Prestressed concrete nuclear reactor containment structures. Revision 3

    International Nuclear Information System (INIS)

    Reuter, H.R.; Chang-Lo, P.L.C.; Pfeifer, B.W.; Shah, G.H.; Whitcraft, J.S.

    1975-02-01

    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.)

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

    International Nuclear Information System (INIS)

    Fraix, J.

    1992-01-01

    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)

  5. Open Access to the Belgian Nuclear higher Education Network

    International Nuclear Information System (INIS)

    Simons, S.

    2005-01-01

    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

  6. Decomposing dynamic profit inefficiency of Belgian dairy farms

    NARCIS (Netherlands)

    Ang, Frederic; Lansink, Alfons Oude

    2018-01-01

    This paper introduces a nonparametric framework for analysing dynamic profit inefficiency and applies this to a sample of Belgian, specialised dairy farms from 1996 to 2008. Profit inefficiency is decomposed into technical and allocative inefficiency. The paper also decomposes profit inefficiency

  7. Positioning and role of public relations in large Belgian organizations

    NARCIS (Netherlands)

    Gorp, B. van; Pauwels, L.M.A.

    2007-01-01

    This paper studies the position and role of public relations in the hierarchical structure of Belgian organizations of at least 50 employees. Empirical data was collected from a web survey (n = 750) to find out to what extent principles of excellence in public relations are applied in Belgium. The

  8. Will Dutch Become Flemish? Autonomous Developments in Belgian Dutch

    Science.gov (United States)

    Van de Velde, Hans; Kissine, Mikhail; Tops, Evie; van der Harst, Sander; van Hout, Roeland

    2010-01-01

    In this paper a series of studies of standard Dutch pronunciation in Belgium and the Netherlands is presented. The research is based on two speech corpora: a diachronic corpus of radio speech (1935-1995) and a synchronic corpus of Belgian and Netherlandic standard Dutch from different regions at the turn of the millennium. It is shown that two…

  9. Physical fitness of elite Belgian soccer players by player position.

    Science.gov (United States)

    Boone, Jan; Vaeyens, Roel; Steyaert, Adelheid; Vanden Bossche, Luc; Bourgois, Jan

    2012-08-01

    The purpose of this study was to gain an insight into the physical and physiological profile of elite Belgian soccer players with specific regard to the player's position on the field. The sample consisted of 289 adult players from 6 different first division teams. The players were divided into 5 subgroups (goalkeepers, center backs, full backs, midfielders, and strikers) according to their self-reported best position on the field. The subjects performed anaerobic (10-m sprint, 5 × 10-m shuttle run [SR], squat jump [SJ], and countermovement jump [CMJ]) and aerobic (incremental running protocol) laboratory tests. The strikers had significantly shorter sprinting times (5-, 5- to 10-m time, and SR) compared with the midfielders, center backs, and goalkeepers, whereas the full backs were also significantly faster compared with the goalkeepers and the center backs. The goalkeepers and the center backs displayed higher jumping heights (total mean SJ = 40.7 ± 4.6 cm and CMJ = 43.1 ± 4.9 cm) compared with the other 3 positions, whereas the strikers also jumped higher than the full backs and the midfielders did. Regarding the aerobic performance, both full backs and the midfielders (61.2 ± 2.7 and 60.4 ± 2.8 ml · min(-1) · kg(-1), respectively) had a higher VO2max compared with the strikers, center backs, and goalkeepers (56.8 ± 3.1, 55.6 ± 3.5, and 52.1 ± 5.0 ml · min(-1) · kg(-1), respectively). From this study, it could be concluded that players in different positions have different physiological characteristics. The results of this study might provide useful insights for individualized conditional training programs for soccer players. Aside from the predominant technical and tactical skills, a physical profile that is well adjusted to the position on the field might enhance game performance.

  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. Belgian experience in applying the open-quotes leak-before-breakclose quotes concept to the primary loop piping

    International Nuclear Information System (INIS)

    Gerard, R.; Malekian, C.; Meessen, O.

    1997-01-01

    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 cutsclose quotes technique. The influence of such residual loads on the LBB margins is discussed in details and typical results are presented

  12. 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.

  13. Review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Tetsuya; Araki, Masaaki; Ohba, Toshinobu; Torii, Yoshiya [Japan Atomic Energy Agency, Tokai, Ibaraki (Japan); Takeuchi, Masaki [Nuclear Safety Commission (Japan)

    2012-03-15

    JRR-3(Japan Research Reactor No.3) with the thermal power of 20MW is a light water moderated and cooled, swimming pool type research reactor. JRR-3 has been operated without major troubles. This paper presents about review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors. In addition, some topics concerning damages in JRR-3 due to the Great East Japan Earthquake are presented. (author)

  14. Pilot plant production at Riso of LEU silicide fuel for the Danish reactor DR3

    International Nuclear Information System (INIS)

    Toft, P.; Borring, J.; Adolph, E.

    1988-01-01

    A pilot plant for fabricating LEU silicide fuel elements has been established at Riso National Laboratory. Three test elements for the Danish reactor DR3 have been fabricated, based on 19.88% enriched U 3 Si 2 powder that has been purchased elsewhere. The pilot plant has been set up and 3 test elements fabricated without any major difficulties

  15. The qualification of U3O8 as research reactor fuel

    International Nuclear Information System (INIS)

    Krull, W.

    1983-01-01

    This report summarizes the today knowledge of the qualification status of U 3 O 8 as low enriched ( 3 O 8 is so far qualified to start testing of ten (10) fuel elements with an U-density of 3.1 g U/cc in the FRG-2 research reactor. (orig.) [de

  16. Plutonium-burn high temperature gas-cooled reactor for 3E+3S

    International Nuclear Information System (INIS)

    Okamoto, Koji

    2015-01-01

    The Nuclear Energy Development in Japan is facing a very difficult conditions after Fukushima-Daiichi NPP Accident. Nuclear Energy has strong advantages on 3E, i.e., Energy security, Economical efficiency and Environment. However, people does not believe the Safety 'S' of Nuclear Energy, now. The disadvantage of 'S' overrides the advantages of '3E'. In Nuclear Energy, 'S' is expanded into 3S, i.e., Safety, Security and Safeguards. Especially, the management of Plutonium inventory in Spent Fuel generated by the NPP operation is very important in the viewpoints of non-proliferation. The high-temperature gas cooled reactor (HTGR) is the solution of these disadvantages of '3S' in Nuclear Energy. The fuel of HTGR is composed by 1 mm spherical fuel particle, i.e., TRISO made by fuel, graphite and silicon-carbide. The silicon-carbide can confine the fission products in any conditions of fuel life cycle, i.e., during operation, accidents and disposal for 1 million years. The confinement of the radioactive materials can be confirmed by the TRISO. The HTGR core has strong negative feedback for temperature. So, the fission automatically stopped at the accidental conditions, such as loss of flow and LOCA. Also, the residual heat can be cooled by the radiation heat transfer to reactor vessel wall. The HTGR system usually has passive vessel wall cooling system. When the passive cooling system had been failed, the heat can be transferred to the land by heat conductions, and fuel does not reach the SiC broken temperature. The fission chain reaction has been stopped automatically by negative feedback, i.e., physics. The residual heat had been cooled automatically by radiation. The radioactive materials had been confined automatically by silicon-carbide. The HTGR is superior for 'S' safety. Plutonium can be burned by the HTGR. In the viewpoints of non-proliferation, the fuel should be made by YSZ-PuO 2 , stabilized buffer

  17. Research reactor core conversion guidebook. V. 3: Analytical verification (Appendices G and H)

    International Nuclear Information System (INIS)

    1992-04-01

    Volume 3 consists of Appendix G which contains detailed results of a safety-related benchmark problem for an idealized reactor and Appendix H which contains detailed comparisons of calculated and measured data for actual cores with moderately enriched uranium and low enriched uranium fuels. The results of the benchmark calculations in Appendix G are summarized in Chapter 7 of Volume 1 and the results of the comparisons between calculations and measurements are summarized in Chapter 8 of Volume 1. Both the approaches described in these appendices are very useful in ensuring that the calculational methods employed in the preparation of a Safety Report are accurate. As a first step, it is recommended that reactor operators/physicists use their own methods and codes to first calculate the benchmark problem, and then compare the results of calculations with measurements in their own reactor or in one of the reactors for which measured data is available in Appendix H. (author). Refs, figs and tabs

  18. On the major DYN3D developments for fast reactor design and transient analysis

    International Nuclear Information System (INIS)

    Merk, B.; Kliem, S.

    2013-01-01

    Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

  19. On the major DYN3D developments for fast reactor design and transient analysis

    Energy Technology Data Exchange (ETDEWEB)

    Merk, B.; Kliem, S. [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety Div.

    2013-07-01

    Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

  20. Application of RELAP5-3D code for thermal analysis of the ADS reactor core

    International Nuclear Information System (INIS)

    Fernandes, Gustavo Henrique Nazareno

    2018-01-01

    Nuclear power is essential to supply global energy demand. Therefore, in order to use nuclear fuel more efficiently, more efficient nuclear reactors technologies researches have been intensified, such as hybrid systems, composed of particle accelerators coupled into nuclear reactors. In order to add knowledge to such studies, an innovative reactor design was considered where the RELAP5-3D thermal-hydraulic analysis code was used to perform a thermal analysis of the core, either in stationary operation or in situations transitory. The addition of new kind of coolants, such as, liquid salts, among them Flibe, lead, lead-bismuth, sodium, lithium-bismuth and lithium-lead was an important advance in this version of the code, making possible to do the thermal simulation of reactors that use these types of coolants. The reactor, object of study in this work, is an innovative reactor, due to its ability to operate in association with an Accelerator Driven System (ADS), considered a predecessor system of the next generation of nuclear reactors (GEN IV). The reactor selected was the MYRRHA (Multi-purpose Hybrid Research Reactor for High tech Applications) due to the availability of data to perform the simulation. In the modeling of the reactor with the code RELAP5-3D, the core was simulated using nodules with 1, 7, 15 and 51 thermohydraulic channels and eutectic lead-bismuth (LBE) as coolant. The parameters, such as, pressure, mass flow and coolant and heat structure temperature were analyzed. In addition, the thermal behavior of the core was evaluated by varying the type of coolant (sodium) in substitution for the LBE of the original design using the model with 7 thermohydraulic channels. The results of the steady-state calculations were compared with data from the literature and the proposed models were verified certifying the ability of the RELAP5-3D code to simulate this innovative reactor. After this step, it was analysed cases of transients with loss of coolant flow

  1. Biodiversity of a wreck from the Belgian Continental Shelf: monitoring using scientific diving: preliminary results

    OpenAIRE

    Massin, Cl.; Norro, A.; Mallefet, J.

    2002-01-01

    Scientific diving from aboard the r/v Belgica has been employed to carry out a preliminary study of the macrofauna living on a wreck (the Birkenfels) located on the Belgian Continental Shelf (BCS). The study revealed an extremely rich sessile and slow moving fauna (at least 40 species), 3 jellyfish species and 8 fish species. The presence of at least 51 species represents a biodiversity of macrofauna on the wreck that is much higher than that found in nearly all known surrounding soft bottorn...

  2. EPR, a GEN 3 Reactor providing a competitive electricity cost

    International Nuclear Information System (INIS)

    Salhi, Othman

    2006-01-01

    Since the very beginning of the development of what was to become the EPR, several European entities were involved. The French and German safety authorities expressed that reinforced safety was compulsory. Additional measures were then included to prevent the occurrence of events likely to damage the core, and reduce the possibility of exposure of operating and maintenance personnel. However, not with standing these safety related features resulting from the requirements of the safety authorities, we will focus today on another group of entities that were key players in EPR development: the Utilities. The Utilities voiced their need for a competitive electricity produced and a competitive nuclear reactor. The tradeoff was then to reach both targets in a unique product: a safer and more competitive NPP. Today, the EPR presents features that enable our clients to compete with the cheapest fossil-based electricity production plants. Increased thermal efficiency is obtained both through a higher steam pressure and through careful optimization of the secondary system thermal cycle

  3. 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

    Energy Technology Data Exchange (ETDEWEB)

    Fabry, A.; Chaouadi, T.; Puzzolante, J.L.; Van de Velde, J. [Centre de l``Etude de l``Energie Nucleaire, Mol (Belgium); Biemiller, E.C. [Yankee Atomic Electric Company, Bolton (United States); Rossinski, S.T.; Carter, R.G. [Electric Power Research Institute, Charlotte (United States)

    1996-07-01

    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

  4. 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)

    Fabry, A.; Chaouadi, T.; Puzzolante, J.L.; Van de Velde, J.; Biemiller, E.C.; Rossinski, S.T.; Carter, R.G.

    1996-07-01

    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

  5. Studies on the liquid fluoride thorium reactor: Comparative neutronics analysis of MCNP6 code with SRAC95 reactor analysis code based on FUJI-U3-(0)

    Energy Technology Data Exchange (ETDEWEB)

    Jaradat, S.Q., E-mail: sqjxv3@mst.edu; Alajo, A.B., E-mail: alajoa@mst.edu

    2017-04-01

    Highlights: • The verification for FUJI-U3-(0)—a molten salt reactor—was performed. • The MCNP6 was used to study the reactor physics characteristics for FUJI-U3 type. • The results from the MCNP6 were comparable with the ones obtained from literature. - Abstract: The verification for FUJI-U3-(0)—a molten salt reactor—was performed. The reactor used LiF-BeF2-ThF4-UF4 as the mixed liquid fuel salt, and the core was graphite moderated. The MCNP6 code was used to study the reactor physics characteristics for the FUJI-U3-(0) reactor. Results for reactor physics characteristic of the FUJI-U3-(0) exist in literature, which were used as reference. The reference results were obtained using SRAC95 (a reactor analysis code) coupled with ORIGEN2 (a depletion code). Some modifications were made in the reconstruction of the FUJI-U3-(0) reactor in MCNP due to unavailability of more detailed description of the reactor core. The assumptions resulted in two representative models of the reactor. The results from the MCNP6 models were compared with the reference results obtained from literature. The results were comparable with each other, but with some notable differences. The differences are because of the approximations that were done on the SRAC95 model of the FUJI-U3 to simplify the simulation. Based on the results, it is concluded that MCNP6 code predicts well the overall simulation of neutronics analysis to the previous simulation works using SRAC95 code.

  6. Emergence of bovine ehrlichiosis in Belgian cattle herds.

    Science.gov (United States)

    Guyot, Hugues; Ramery, Eve; O'Grady, Luke; Sandersen, Charlotte; Rollin, Frédéric

    2011-06-01

    Bovine ehrlichiosis is a tick-borne rickettsial disease caused by Anaplasma phagocytophilum. The disease can also be transmitted to humans. Outbreaks in cattle have been described in many European countries. In Belgium, infections caused by A. phagocytophilum have been reported in humans and dogs; however, this paper details the first report of ehrlichiosis in cattle herds in Belgium. The first case described was in a dairy herd located in eastern Belgium. Clinical signs included hyperthermia, polypnea, and swelling of the limbs. The other case was diagnosed in a second, mixed purpose herd in western Belgium. Within the second herd, all of the affected animals came from the same pasture. All animals in that pasture showed recurrent hyperthermia, and some also showed signs of mastitis and late-term abortions. Blood smears and serology revealed the presence of A. phagocytophilum in the majority of animals with pyrexia. Furthermore, the presence of leptospirosis, Neospora caninum, and Q fever antibodies was tested by serological analysis, but all results were negative. Paired serology for Adenovirus, BHV-4, BHV-1, BVD, PI3, and RSV-B did not show any significant seroconversion. Milk samples from cows affected by mastitis revealed minor pathogens. Fecal testing for the presence of Dictyocaulus viviparus in the first herd was negative. Recurrent pyrexia in pastured cattle is a non-specific sign, and can be related to several different pathogens. Bovine ehrlichiosis is transmitted by the tick species Ixodes ricinus which is known to be present throughout Belgium. Belgian practitioners should include ehrlichiosis in their differential diagnosis when confronted with pastured cattle suffering from recurrent pyrexia. Copyright © 2011 Elsevier GmbH. All rights reserved.

  7. Antimicrobial prescribing behaviour in dogs and cats by Belgian veterinarians.

    Science.gov (United States)

    Van Cleven, Alexia; Sarrazin, Steven; de Rooster, Hilde; Paepe, Dominique; Van der Meeren, Sofie; Dewulf, Jeroen

    2018-03-17

    The objective of this study is to survey general prescribing behaviour by Belgian companion animal veterinarians and to assess agreement of these practices with current treatment guidelines. Therefore an online survey was administered with five realistic and frequently occurring first-line cases to primary-care veterinary practitioners. For each case a predefined pattern of questions were asked about whether or not they would prescribe antimicrobials, if they would prescribe a non-antimicrobial treatment and if they would perform additional diagnostic steps. The responses were compared with recommendations in national guidelines and recent literature. The overall most prescribed antimicrobials were potentiated amoxicillin (43.0 per cent), fluoroquinolones (14.7 per cent), third-generation and fourth-generation cephalosporins (10.9 per cent) and tetracyclines (10.9 per cent). Only 48.3 per cent of the veterinarians complied with the guidelines in nearly all of the clinical scenarios (ie, prescribing antimicrobials when indicated, not prescribing antimicrobials when it is not indicated). Moreover, when prescribing highest priority critically important antimicrobials, susceptibility testing on bacterial cultures was performed in only 12.4 per cent of the prescriptions. The results showed that the prescribing behaviour of antimicrobial compounds by primary-care veterinary practitioners in dogs and cats is often not in agreement with national guidelines. Focus in improvement of this prescribing behaviour should be on performing the appropriate diagnostic steps and decreasing the use of highest priority critically important antimicrobials. © British Veterinary Association (unless otherwise stated in the text of the article) 2018. All rights reserved. No commercial use is permitted unless otherwise expressly granted.

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

    International Nuclear Information System (INIS)

    Leclere, Robert; Van Landeghem, Yves

    2010-01-01

    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 emotions: bring reason into a mainly emotional

  9. Optimization of reactor power by taking into consideration temperature increase in a reactor pumped 3He-Xe laser

    International Nuclear Information System (INIS)

    Cetin, Fuesun

    2009-01-01

    In nuclear pumped lasers, gas parameters are optimized in a manner such that output power is increased for constructing a high power laser. Since output power increases with the increase of energy deposited in the gas, high output power requires high pumping power. However, the high energy loading results in elevated gas temperature. Temperature increase of this magnitude can detrimentally influence the laser gain and efficiency, since it negatively impacts several important laser kinetic.processes. This fact may cause laser output to abruptly terminate before the peak of the pump pulse [1-3]. A nuclear pumped laser using a volumetric energy source through the 3 He(n, p) 3 H reaction has here been considered. It is assumed that TRIGA Mark II Reactor at Istanbul Technical University is used for nuclear pumping as the neutron source. In the previous papers, the optimal parameters for improving both output power and optical homogeneity were determined [4-5]. Spatial and temporal variations of gas temperature during pumping pulse for maximum peak power (1200 MW) were determined for various operating pressures in Ref. [6]. It was seen that gas temperature reaches up to 1000 0 K near the peak of the pumping pulse for the initial pressures of 1-4 atm. This means that laser output may terminate before the peak of the pump pulse due to overheating of laser gas. Under these conditions, a question arises about a further optimisation taking into consideration gas temperature. This question has been examined in this study. Experimental results (Batyrbekov et al, 1989) showed that temperature rise up to 650 C had no influence on Xe laser characteristics [ 7]. Therefore, It has here been assumed that the lasing will terminate when gas temperature reaches 1000 0 K for a Xe-laser with 3 He buffer gas. Under these conditions optimum reactor power is investigated by taking into consideration lasing duration also. (orig.)

  10. 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.

  11. Level of zinc in the Belgian calamine-bearing lichens: Stereocaulon nanodes Tuck f tyroliense (nyl. ) M. Lamb

    Energy Technology Data Exchange (ETDEWEB)

    Maquinay, A; Lamb, I M; Lambinon, J; Ramaut, J L

    1961-01-01

    It has been shown that Stereocaulon nanodes, a lichen species native to Belgian calamine-bearing soils takes up zinc and in some cases accumulates it in the thallus. This lichen, growing on a substrate slag with 700 p.p.m. of zinc may contain up to 3300 p.p.m. This abnormally high content as compared with the usual plant tolerance for this element does not seem to disturb its metabolism. 3 references, 4 tables.

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

    International Nuclear Information System (INIS)

    Islam, M.S.; Haque, M.M.; Salam, M.A.; Rahman, M.M.; Khandokar, M.R.I.; Sardar, M.A.; Saha, P.K.; Haque, A.; Malek Sonar, M.A.; Uddin, M.M.; Hossain, S.M.S.; Zulquarnain, M.A.

    2004-01-01

    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 ( 131 I, 99m Tc, 46 Sc), 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 TeO 2 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 ( 131 I). 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

  13. Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

    International Nuclear Information System (INIS)

    Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

    2003-01-01

    The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

  14. Characteristics of D-3He fueled frc reactor: ARTEMIS-L

    International Nuclear Information System (INIS)

    Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

    1993-11-01

    The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author)

  15. Characteristics of D(-3)He fueled FRC reactor: ARTEMIS-L

    Science.gov (United States)

    Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

    1993-11-01

    The characteristics of D(-3)He fueled commercial fusion reactor ARTEMIS-L are discussed. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L becomes compact and its veta-value is extremely high. Consequently, it is possible to construct an economical fusion power plant based on this concept. The life of the structural materials is found during the full reactor life (30 years) and the safety of the reactor is intrinsic to D(-3)He fuels. The amount of disposed materials is rather small and the level of the intruder dose is so low that the plant appears to be acceptable in regards to the environment.

  16. Characteristics of D-{sup 3}He fueled frc reactor: ARTEMIS-L

    Energy Technology Data Exchange (ETDEWEB)

    Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

    1993-11-01

    The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author).

  17. Verification of using SABINE-3.1 code for calculations of radioactive inventory in reactor shield

    International Nuclear Information System (INIS)

    Moukhamadeev, R.; Suvorov, A.

    2000-01-01

    This report presents the results of calculations of radioactive inventory and doses of activation radiation for the International Benchmark Calculations of Radioactive Inventory for Fission Reactor Decommissioning, IAEA, and measurements of activation doses in shield of WWER-440 (Armenian NPP), using one-dimension modified code SABINE-3.1. For decommissioning of NPP it is very important to evaluate in correct manner radioactive inventory in reactor construction and shield materials. One-dimension code SABINE-3.1 (removing-diffusion method for neutron calculation) was modified to perform calculation of radioactive inventory in reactor shield materials and dose from activation photons behind them. These calculations are carried out on the base of nuclear constant system ABBN-78 and new library of activation data for a number of long-lived isotopes, prepared by authors on the base of [9], which present at shield materials as microimpurities and manage radiation situation under the decay more than 1 year. (Authors)

  18. Station Blackout Analysis for a 3-Loop Westinghouse PWR Reactor Using Trace

    International Nuclear Information System (INIS)

    El-Sahlamy, N.M.

    2017-01-01

    One of the main concerns in the area of severe accidents in nuclear reactors is that of station blackout (SBO). The loss of offsite electrical power concurrent with the unavailability of the onsite emergency alternating current (AC) power system can result in loss of decay heat removal capability, leading to a potential core damage which may lead to undesirable consequences to the public and the environment. To cope with an SBO, nuclear reactors are provided with protection systems that automatically shut down the reactor, and with safety systems to remove the core residual heat. This paper provides a best estimate assessment of the SBO scenario in a 3-loop Westinghouse PWR reactor. The evaluation is performed using TRACE, a best estimate computer code for thermal-hydraulic calculations. Two sets of scenarios for SBO analyses are discussed in the current work. The first scenario is the short term SBO where it is assumed that in addition to the loss of AC power, there is no DC power; i.e., no batteries are available. In the second scenario, a long term SBO is considered. For this scenario, DC batteries are available for four hours. The aim of the current SBO analyses for the 3-loop pressurized water reactor presented in this paper is to focus on the effect of the availability of a DC power source to delay the time to core uncovers and heatup

  19. 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.

  20. Iron microbial communities in Belgian Frasnian carbonate mounds

    OpenAIRE

    Boulvain, F.; De Ridder, C.; Mamet, B.; Preat, A.; Gillan, D.

    2001-01-01

    The Belgian Frasnian carbonate mounds occur in three stratigraphic levels in an overall backstepping succession. Petit-Mont and Arche Members form the famous red and grey “marble” exploited for ornamental stone since Roman times. The evolution and distribution of the facies in the mounds is thought to be associated with ecologic evolution and relative sea-level fluctuations. Iron oxides exist in five forms in the Frasnian mounds; four are undoubtedly endobiotic organized structures: (1) micro...

  1. Dynamic Profit Inefficiency: A DEA Application to Belgian Dairy Farms

    OpenAIRE

    Ang, Frederic; Oude Lansink, Alfons

    2014-01-01

    Using a nonparametric framework, we analyze dynamic profit inefficiency for a sample of Belgian, specialized dairy farms from 1996–2008. Profit inefficiency is decomposed into contributions of output, input, and investment. Moreover, we identify the contributions of technical and allocative inefficiency in each input and output. The results suggest substantial profit inefficiency under the current dairy-quota system, mainly driven by an average underproduction of approximately 50 percent and ...

  2. Estimation of the furan contamination across the Belgian food chain

    OpenAIRE

    Scholl , Georges; Scippo , Marie-Louise; Eppe , Gauthier; De Pauw , Edwin; Saegerman , Claude

    2011-01-01

    Abstract The current paper provides the estimate of the furan content in Belgian foods. The objective of the study was to achieve the best food chain coverage with a restrictive number of samples (n=496). The geographic distribution, the different market chains and labels, but also the consumption frequencies were taken into account for the sampling plan construction. Weighting factors on contamination levels, consumption frequency and diversity of food items were applied to set ...

  3. 3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers

    International Nuclear Information System (INIS)

    Chvetsov, I.; Volkov, A.

    2000-01-01

    For advanced fast reactors (EFR, BN-600M, BN-1600, CEFR) the special complementary loop is envisaged in order to ensure the decay heat removal from the core in the case of LOF accidents. This complementary loop includes immersion coolers that are located in the hot reactor plenum. To analyze the transient process in the reactor when immersion coolers come into operation one needs to involve 3-D thermal hydraulics code. Furthermore sometimes the problem becomes more complicated due to necessity of simulation of the thermal hydraulics processes into the core interwrapper space. For example on BN-600M and CEFR reactors it is supposed to ensure the effective removal of decay heat from core subassemblies by specially arranged internal circulation circuit: 'inter-wrapper space'. For thermal hydraulics analysis of the transients in the core and in the whole reactor including hot plenum with immersion coolers and considering heat and mass exchange between the main sodium flow and sodium that moves in the inter-wrapper space the code GRIFIC (the version of GRIF code family) was developed in IPPE. GRIFIC code was tested on experimental data obtained on RAMONA rig under conditions simulating decay heat removal of a reactor with the use of immersion coolers. Comparison has been made of calculated and experimental result, such as integral characteristics (flow rate through the core and water temperature at the core inlet and outlet) and the local temperatures (at thermocouple location) as well. In order to show the capabilities of the code some results of the transient analysis of heat removal from the core of BN-600M - type reactor under loss-of-flow accident are presented. (author)

  4. The Odd One Out? Revisiting the Belgian Welfare State

    Directory of Open Access Journals (Sweden)

    Cor Wagenaar

    2014-08-01

    Full Text Available Michael Ryckewaert publication Building the Economic Backbone of the Belgian Welfare State. Infrastructure, planning and architecture 1945-1973 describes the evolution of the welfare state and Belgium, more specifically its spatial characteristics. This by now historical socio-political model had decidedly collectivist traits, culminating in the provision of social security networks and a vast expansion of the public domain. If collectivism was one of the key elements of the welfare state, the absence of centralized planning appears to make the Belgian variant somewhat problematic.Whereas in countries like the Netherlands, Germany and France, modernism became the house style of the welfare state, thanks to the massive investments in public housing, this did not happen in Belgium. Here, the De Taeye Act of 1948 sponsored the construction of individual, detached houses; not surprisingly, most clients preferred traditional architecture and refrained from modern experiments. Industrial parks, office buildings and shops, on the other hand, developed into the cornerstones of Belgian modern architecture after 1945. Both the low-density sprawl and the industrial parks depend heavily on the use of the car, which was accommodated by the construction of a network of highways.

  5. Comparison of composition and quality traits of meat from young finishing bulls from Belgian Blue, Limousin and Aberdeen Angus breeds.

    Science.gov (United States)

    Cuvelier, C; Clinquart, A; Hocquette, J F; Cabaraux, J F; Dufrasne, I; Istasse, L; Hornick, J L

    2006-11-01

    Thirty-six young finishing bulls from three breeds (Belgian Blue, Limousin and Aberdeen Angus) were fattened over five months with finishing diets based either on sugar-beet pulp or on cereals. Nutritional quality traits of meat - fat content and fatty acid composition with emphasis on the n-6 and n-3 polyunsaturated fatty acids - along with some organoleptic quality traits were measured. The Belgian Blue bulls had the lowest intramuscular fat content associated with lower saturated and monounsaturated fatty acid contents. The polyunsaturated fatty acid content did not differ to a large extent between the breeds, the Aberdeen Angus bulls showing slightly higher values. Relative to energy intake, the overall contribution of meat to the n-3 fatty acid recommended intake was small, whatever the breed. By contrast, the contribution of meat to daily fat intake was of greater importance, especially for the Aberdeen Angus bulls. The quality traits of meat varied also according to the breed: compared to the Aberdeen Angus, the Belgian Blue bull meat had the stablest colour, the highest drip and the lowest cooking losses. The meat of Limousin bulls had intermediate characteristics for all the parameters.

  6. Effect of fuel assembly when changing from AFA 2G to AFA 3G on seismic loads of reactor internal

    International Nuclear Information System (INIS)

    Liu Wenjin; Zeng Zhongxiu; Ye Xianhui; Wu Wanjun

    2013-01-01

    Nonlinear seismic model for reactor with fuel assemblies of AFA 2G and AFA 3G is established. Using ANSYS software, seismic nonlinear time -history analysis is completed and the effects on seismic loads of reactor system are obtained. The result shows that when the fuel assembly changing from AFA 2G to AFA 3G, it is necessary to reevaluate the fuel assembly itself, but not the reactor internal. (authors)

  7. Neutron flux calculations for the Rossendorf research reactor in (hex)- and (hex,z)-geometry using SNAP-3D

    International Nuclear Information System (INIS)

    Koch, R.; Findeisen, A.

    1986-04-01

    The multigroup neutron diffusion theory code SNAP-3D has been used to perform time independent neutron flux and power calculations of the 10 MW Rossendorf research reactor of the type WWR-SM. The report describes these calculations, as well as the actual reactor configuration, some details of the code SNAP-3D, and two- and three-dimensional reactor models. For evaluating the calculations some flux values and control rod worths have been compared with those of measurements. (author)

  8. ASAMPSA2 best-practices guidelines for L2 PSA development and applications. Volume 3 - Extension to Gen IV reactors

    International Nuclear Information System (INIS)

    Bassi, C.; Bonneville, H.; Brinkman, H.; Burgazzi, L.; Polidoro, F.; Vincon, L.; Jouve, S.

    2010-01-01

    The main objective assigned to the Work Package 4 (WP4) of the 'ASAMPSA2' project (EC 7. FPRD) consist in the verification of the potential compliance of L2PSA guidelines based on PWR/BWR reactors (which are specific tasks of WP2 and WP3) with Generation IV representative concepts. Therefore, in order to exhibit potential discrepancies between LWRs and new reactor types, the following work was based on the up-to-date designs of: - The European Fast Reactor (EFR) which will be considered as prototypical of a pool-type Sodium-cooled Fast Reactor (SFR); - The ELSY design for the Lead-cooled Fast Reactor (LFR) technology; - The ANTARES project which could be representative of a Very-High Temperature Reactor (VHTR); - The CEA 2400 MWth Gas-cooled Fast Reactor (GFR). (authors)

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

    International Nuclear Information System (INIS)

    Loika, E.F.

    1994-01-01

    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

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

    International Nuclear Information System (INIS)

    Gomez, Angel

    2003-01-01

    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 16 N 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 16 N measuring system, an estimator, that is a function of the ratio of this currents, is proposed in order to detect flow changes. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

    2003-04-01

    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.

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

    International Nuclear Information System (INIS)

    Greene, S.R.; Fisher, S.E.; Bevard, B.B.

    1996-09-01

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

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

    International Nuclear Information System (INIS)

    Blum, A.S.; Moir, R.W.

    1977-01-01

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

  14. GRIMH3: A new reactor calculation code at Savannah River Site

    International Nuclear Information System (INIS)

    Le, T.T.; Pevey, R.E.

    1993-01-01

    The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex. The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex

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

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Fisher, S.E.; Bevard, B.B. [and others

    1996-09-01

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

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

    International Nuclear Information System (INIS)

    2010-04-01

    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

  17. Intake of bisphenol A from canned beverages and foods on the Belgian market.

    Science.gov (United States)

    Geens, Tinne; Apelbaum, Tali Zipora; Goeyens, Leo; Neels, Hugo; Covaci, Adrian

    2010-11-01

    Bisphenol A (BPA), a contaminant which may be present in the coating of cans, was determined in 45 canned beverages and 21 canned food items from the Belgian market. Beverages had an average BPA concentration of 1.0 ng/ml, while canned foods had a higher average concentration of 40.3 ng/g. The amount of BPA present in food items was dependent on the type of can and sterilisation conditions rather than the type of food. For example, BPA was not detected in non-canned beverages (canned food items had a very low average concentration of 0.46 ng/g. Using detailed information from the Belgian food consumption survey, the BPA intake of adults through canned foods and beverages was estimated to be 1.05 µg/day or 0.015 µg/kg body weight/day (assuming an average adult weight of 70 kg). Intake assessments, based on urinary metabolite concentrations from the literature, resulted in slightly higher BPA intakes (range 0.028-0.059 µg/kg body weight/day). This suggests that sources other than canned foods and beverages contribute to BPA exposure in humans.

  18. Contribution of the Belgian hospital physicists association to quality assurance in radiotherapy

    International Nuclear Information System (INIS)

    Hoornaert, M.Th.; Vynckier, S.; Dam, J. van; Bouiller, A.

    1997-01-01

    In 1987, the Belgian Hospital Physicists Association (BHPA) has started a program in order to uniformize the dosimetry in the Belgian radiotherapy centres. Several initiatives were taken: a) Dosimetry, of photon beams: Endorsement of the Dutch dosimetry, code of practice (NCS) (1), calibration of ionisation chambers in a common laboratory (Laboratory for standard dosimetry, RUG), on site visits where, besides mechanical checks of simulators and radiation units, absorbed dose was measured at different locations in a water phantom. Since 1987, a total of 23 centres were visited involving 18 simulators, 17 cobalt units and 22 linear accelerators with 33 photon beams. The energy of those photon beams ranged from 4 to 25 MeV (2). b) Dosimetry of electron beams: Endorsement of the Dutch dosimetry code of practice (3), calibration of several parallel plate chambers following the recommendations of the IAEA (4) and the NCS, on site visits for local measurements in electron beams. This program started last year. three centres were visited with a total of 23 energies ranging from 4.5 to 21 MeV. c) Elaboration of procedures and common reporting form for daily quality control will be published. (author)

  19. Development of covariance date for fast reactor cores. 3

    International Nuclear Information System (INIS)

    Shibata, Keiichi; Hasegawa, Akira

    1999-03-01

    Covariances have been estimated for nuclear data contained in JENDL-3.2. As for Cr and Ni, the physical quantities for which covariances are deduced are cross sections and the first order Legendre-polynomial coefficient for the angular distribution of elastically scattered neutrons. The covariances were estimated by using the same methodology that had been used in the JENDL-3.2 evaluation in order to keep a consistency between mean values and their covariances. In a case where evaluated data were based on experimental data, the covariances were estimated from the same experimental data. For cross section that had been evaluated by nuclear model calculations, the same model was applied to generate the covariances. The covariances obtained were compiled into ENDF-6 format files. The covariances, which had been prepared by the previous fiscal year, were re-examined, and some improvements were performed. Parts of Fe and 235 U covariances were updated. Covariances of nu-p and nu-d for 241 Pu and of fission neutron spectra for 233,235,238 U and 239,240 Pu were newly added to data files. (author)

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

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-12-01

    This volume covers the following reports concerned with the maintenance and repair work of the RA reactor: repair of the technical water system; maintenance of the transportation equipment; vacuuming and drying during refurbishment; repair and decontamination of the distillation device; and the report on participation of the operational dosimetry division in the RA reactor refurbishment activities

  1. 3D CFD for chemical transport profiles in a rotating disk CVD reactor

    Science.gov (United States)

    Han, Jong-Hyun; Yoon, Do-Young

    2010-06-01

    The RDCVD (Rotating Disk Chemical Vapor Deposition) technique is an appropriate method for uniform deposition of grains, such as compound semiconductior materials. The substrate temperature and rotation speed are the major factors, which determine the thickness uniformity of the deposited films. This paper investigates 3D CFD (3 Dimensional Computational Fluid Dynamics) simulation results of flow and heat transfer in a reactor of RDCVD using Fluent. In order to establish the reducibility of buoyancy effect on deposition quality, the chemical transport profile upon the disk heated is examined successfully in 3D domain for different rotating speeds. The resulting vortex flows due the simultaneous buoyance and centrifuge are discussed qualitatively in the 3D virtual system of a RDCVD reactor. 3D CFD is even more effective to describe the internal vortex flows due to the competitive inlet, buoyancy and centrifuge flows, which cannot be realized in the general 2D (2 Dimensional) CFD.[Figure not available: see fulltext.

  2. Tritium production, management and its impact on safety for a D-3He fusion reactor

    International Nuclear Information System (INIS)

    Sze, D.K.; Herring, S.; Sawan, M.

    1991-11-01

    About three percent of the fusion energy produced by a D- 3 He 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, 3 He 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- 3 He 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. 3 refs., 2 figs., 3 tabs

  3. 3-D analysis of reactor loop isolation valves

    International Nuclear Information System (INIS)

    Dietrich, D.E.

    1975-01-01

    A full three-dimensional analysis for the design and operational loading conditions was performed on a 29 inch loop isolation valve using the Westinghouse finite element computer code. The 3-D analysis was employed for the valve design in place of utilizing the standard ASME valve design criteria. The valve design employs the design by analysis concept allowed for nuclear class valve. The valve design was evaluated for a set of independent load including pipe reactions and internal pressure. The design pipe reaction loads were based upon maximum fiber pipe stresses at yield for the bending moments, pipe membrane stresses at half yield for the axial load, and pipe maximum shear stress at half yield for the torsional moment. The valve design pressure was the system loop design pressure. The operating and accident condition evaluation included pipe reactions, extended structure forces, system pressure, and system thermal transients. The valve was analyzed for the normal operating, upset, emergency, and faulted loading conditions. These operating and accident conditions used various specified combinations of the supplied generic system pressure, deadweight, thermal, seismic, and LOCA pipe load components. The generic pipe loads are the worst possible postulated loads for any system design. These generic pipe load components were supplied as maximums and minimums so a simplified nozzle analysis was performed to determine the worst case combination for each loading condition. The valve design was shown to meet the design, operating, and accident condition requirements of the ASME code. The design by analysis concept for nuclear class 1 valves gave a significant reduction in required minimum wall thickness, 3.75 inches vs. 5.4 inches. These translate into significant material savings

  4. Decision, Annex 3[Organizational structure of the Division for reactor maintenance]; Prilog br. 3, Odluka

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [Reaktor RA, Odelenje odrzavanja, Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1965-12-15

    The objective of the decision about the change in the organizational scheme of the Division for RA reactor maintenance is to achieve efficiency. The tasks are precisely defined as to divide the maintenance and repair tasks from special ones related to improvement of reactor operation its applicability. [Serbo-Croat] U cilju poboljsanja organizacije rada Odelenja odrzavanja reaktora RA, kao i efikasnijeg koriscenja raspolozivog kadra izvrsice se razgranicenje poslova odrzavanja i remonta od posebnih zadataka koji se odnose na poboljsanje rada reaktora i povecanje njegovih mogucnosti.

  5. 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)

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

    International Nuclear Information System (INIS)

    Longoni, G.

    2010-01-01

    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 InfiniBand R 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)

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

    International Nuclear Information System (INIS)

    Jasiulevicius, A.

    2001-01-01

    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

  8. The use of pesticides in Belgian illicit indoor cannabis plantations.

    Science.gov (United States)

    Cuypers, Eva; Vanhove, Wouter; Gotink, Joachim; Bonneure, Arne; Van Damme, Patrick; Tytgat, Jan

    2017-08-01

    Cannabis (Cannabis spp.) use and cultivation continue to increase in many (European) countries. The illicit indoor cannabis plantations that supply Belgian and European cannabis markets create problems and concerns about health and safety of intervention staff, dismantling companies, the direct environment of cannabis plantations and, eventually, of cannabis users. Main risks may come from pesticide residues on plants, cultivation infrastructure and materials; left-over plant growth-promoting substances; mycotoxins from fungal pathogens on harvested plants; and/or high levels of cannabinoids in cannabis plant parts for consumption. In the present research, we report on pesticides found in illicit indoor cannabis plantations in Belgium. EN15662 QuEChERS extraction method and LC-MS/MS analysis were used to identify pesticides in indoor cannabis plantations and thus to evaluate the hazards associated with the use, cultivation and removal of cannabis plants in plantations as well as with dismantling activities in the cultivation rooms. We found pesticides in 64.3% of 72 cannabis plant samples and in 65.2% of 46 carbon filter cloth samples. Overall, 19 pesticides belonging to different chemical classes were identified. We found o-phenylphenol, bifenazate, cypermethrin, imidacloprid, propamocarb, propiconazole and tebuconazole, which is consistent with the commonly reported pesticides from literature. In only a few cases, pesticides found in bottles with a commercial label, were also identified in plant or stagnant water samples collected from the growth rooms where the bottles had been collected. We further revealed that, even though most pesticides have a low volatility, they could be detected from the carbon filters hanging at the ceiling of cultivation rooms. As a result, it is likely that pesticides also prevail in the plantation atmosphere during and after cultivation. The risk of inhaling the latter pesticides increases when plants sprayed with pesticides are

  9. Safety analysis of loss of flow transients in a typical research reactor by RELAP5/MOD3.3

    International Nuclear Information System (INIS)

    Di Maro, B.; Pierro, F.; Adorni, M.; Bousbia Salah, A.; D'Auria, F.

    2003-01-01

    The main aim of the following study is to assess the RELAP5/MOD3.3 code capability in simulating transient dynamic behaviour in nuclear research reactors. For this purpose typical loss of flow transient in a representative MTR (Metal Test Reactor) fuel type Research Reactor is considered. The transient herein considered is a sudden pump trip followed by the opening of a safety valve in order to allow passive decay heat removal by natural convection. During such transient the coolant flow decay, originally downward, leads to a flow reversal and the cooling process of the core passes from forced, mixed and finally to natural circulation. This fact makes it suitable for evaluating the new features of RELAP5 to simulate such specific operating conditions. The instantaneous reactor power is derived through the point kinetic calculation, both protected and unprotected cases are considered (with and without Scram). The results obtained from this analysis were also compared with previous results obtained by old version RELAP5/MOD2 code. (author)

  10. 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

  11. Fusion blankets for catalyzed D--D and D--He3 reactors

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.R.

    1977-01-01

    Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β noncircular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphynyl coolant

  12. Effect of 3-D moderator flow configurations on the reactivity of CANDU nuclear reactors

    International Nuclear Information System (INIS)

    Zadeh, Foad Mehdi; Etienne, Stephane; Chambon, Richard; Marleau, Guy; Teyssedou, Alberto

    2017-01-01

    Highlights: • 3-D CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • The interaction between moderator temperatures with reactivity is determined. - Abstract: The reactivity of nuclear reactors can be affected by thermal conditions prevailing within the moderator. In CANDU reactors, the moderator and the coolant are mechanically separated but not necessarily thermally isolated. Hence, any variation of moderator flow properties may change the reactivity. Until now, nuclear reactor calculations have been performed by assuming uniform moderator flow temperature distribution. However, CFD simulations have predicted large time dependent flow fluctuations taking place inside the calandria, which can bring about local temperature variations that can exceed 50 °C. This paper presents robust CANDU 3-D CFD moderator simulations coupled to neutronic calculations. The proposed methodology makes it possible to study not only different moderator flow configurations but also their effects on the reactor reactivity coefficient.

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

    International Nuclear Information System (INIS)

    Herring, J.S.; Dolan, T.J.

    1992-01-01

    This paper reports on the ARIES-III reactor study, an extensive examination of the viability of a D- 3 He-fueled commercial tokamak powder reactor. Because neutrons are produced only through side reactions (D+D- 3 HE+N; and D+D-T+p followed by D+T- 4 He+n), 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 authors 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. The authors also modeled a loss-of-cooling accident (LOCA) in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. Because the maximum temperature is low, degree C, release fractions are small. The authors analyzed the disposition of the 20 g/day of tritium that is produced by D-D reactions and removed by the vacuum pumps

  14. Fusion blankets for catalyzed D--D and D--3He reactors

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.R.

    1977-01-01

    Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β non-circular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphenyl coolant

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

    International Nuclear Information System (INIS)

    Raynal, A.

    1987-01-01

    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 [fr

  16. Master-3.0: multi-purpose analyzer for static and transient effects of reactors

    International Nuclear Information System (INIS)

    Cho, Byung Oh; Joo, Han Gyu; Cho, Jin Young; Song, Jae Seung; Zee, Sung Quun

    2002-03-01

    MASTER-3.0 (Multi-purpose Analyzer for Static and Transient Effects of Reactors) is a nuclear design code based on the multi-group diffusion theory to calculate the steady-state and transient pressurized water reactor core in a 3-dimensional Cartesian or hexagonal geometry. Its neutronics model solves the space-time dependent neutron diffusion equations with NIM (Nodal Integration Method), NEM (Nodal Expansion Method), AFEN (Analytic Function Expansion Nodal Method)/NEM Hybrid Method, NNEM (Non-linear Nodal Expansion Method) or NANM (Non-linear Analytic Nodal Method) for a Cartesian geometry and with NTPEN (Non-linear Triangle-based Polynomial Expansion Nodal Method), AFEN (Analytic Function Expansion Nodal)/NEM Hybrid Method or NLFM (Non-linear Local Fine-Mesh Method) for a hexagonal one. Coarse mesh rebalancing, Krylov Subspace method, energy group restriction/prolongation method and asymptotic extrapolation method are implemented to accelerate the convergence of iteration process. MASTER-3.0 performs microscopic depletion calculations using microscopic cross sections provided by CASMO-3 or HELIOS and also has the reconstruction capability of pin information by use of MSS-IAS (Method of Successive Smoothing with Improved Analytic Solution). For the thermal-hydraulic calculation, fuel temperature table or COBRA3-C/P or MATRA model can be used selectively. In addition, MASTER-3.0 is designed to cover various PWRs including SMART as well as WH- and CE-type reactors, providing all data required in their design procedures

  17. 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

  18. 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.

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

    International Nuclear Information System (INIS)

    Kim, Yeun Seung; Eom, Young Sam; Lee, Suk Hee; Ryu, Seung Hyun

    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

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

    International Nuclear Information System (INIS)

    Agueda, Horacio C.; Estevez, Esteban; Gerding, Jose R.; Markiewicz, Mario E.

    2003-01-01

    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 U 3 O 8 or U 3 Si 2 as carrying phase of the fissile material with an enrichment of 19.70% of 235 U. 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)

  1. Hydrogen production using Rhodopseudomonas palustris WP 3-5 with hydrogen fermentation reactor effluent

    International Nuclear Information System (INIS)

    Chi-Mei Lee; Kuo-Tsang Hung

    2006-01-01

    The possibility of utilizing the dark hydrogen fermentation stage effluents for photo hydrogen production using purple non-sulfur bacteria should be elucidated. In the previous experiments, Rhodopseudomonas palustris WP3-5 was proven to efficiently produce hydrogen from the effluent of hydrogen fermentation reactors. The highest hydrogen production rate was obtained at a HRT value of 48 h when feeding a 5 fold effluent dilution from anaerobic hydrogen fermentation. Besides, hydrogen production occurred only when the NH 4 + concentration was below 17 mg-NH 4 + /l. Therefore, for successful fermentation effluent utilization, the most important things were to decrease the optimal HRT, increase the optimal substrate concentration and increase the tolerable ammonia concentration. In this study, a lab-scale serial photo-bioreactor was constructed. The reactor overall hydrogen production efficiency with synthetic wastewater exhibiting an organic acid profile identical to that of anaerobic hydrogen fermentation reactor effluent and with effluent from two anaerobic hydrogen fermentation reactors was evaluated. (authors)

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

    International Nuclear Information System (INIS)

    Di Marco, A.; Gillaume, E.J.; Ruggirello, G.; Zaweruchi, A.

    1996-01-01

    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% 235 U. 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

  3. BNAIC 2008 : Proceedings of BNAIC 2008, the twentieth Belgian-Dutch Artificial Intelligence Conference

    NARCIS (Netherlands)

    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

  4. Improvements in performance of Doel-3 and -4

    International Nuclear Information System (INIS)

    Smissaert, G.; Miert, J. Van; Meire, D.

    1988-01-01

    At the end of 1987 the seven Belgian nuclear power units (all PWRs) arrived at an average lifetime load factor of 78.8 %, which puts Belgium among the leading nuclear countries. This article examines the reasons for this performance, paying particular attention to the maintenance and output management approach in Doel-3 and -4. Good performance indicators are the results of several factors, such as good design, efficient maintenance and short refuelling outages. Optimum performance, quick intervention and low dose capability is incorporated at the design state. As a result of experience gained with the time consuming repair of radioactive NSSS components of the SENA reactor, Belgian nuclear power plants have the possibility of dividing the reactor pool into separate pits. It is possible to store the upper and lower internal parts independently under water and to lower the water level in the reactor pit to vessel flange level to reduce repair time when applicable. All reactors and usually all the steam generators' manholes are equipped with automatic stud tensioners leading to shorter opening and closure times and reduced dose for the personnel. 1 fig

  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. Impact of a VAP bundle in Belgian intensive care units.

    Science.gov (United States)

    Jadot, Laurent; Huyghens, Luc; De Jaeger, Annick; Bourgeois, Marc; Biarent, Dominique; Higuet, Adeline; de Decker, Koen; Vander Laenen, Margot; Oosterlynck, Baudewijn; Ferdinande, Patrick; Reper, Pascal; Brimioulle, Serge; Van Cromphaut, Sophie; De Clety, Stéphane Clement; Sottiaux, Thierry; Damas, Pierre

    2018-05-21

    In order to decrease the incidence of ventilator-associated pneumonia (VAP) in Belgium, a national campaign for implementing a VAP bundle involving assessment of sedation, cuff pressure control, oral care with chlorhexidine and semirecumbent position, was launched in 2011-2012. This report will document the impact of this campaign. On 1 day, once a year from 2010 till 2016, except in 2012, Belgian ICUs were questioned about their ventilated patients. For each of these, data about the application of the bundle and the possible treatment for VAP were recorded. Between 36.6 and 54.8% of the 120 Belgian ICUs participated in the successive surveys. While the characteristics of ventilated patients remained similar throughout the years, the percentage of ventilated patients and especially the duration of ventilation significantly decreased before and after the national VAP bundle campaign. Ventilator care also profoundly changed: Controlling cuff pressure, head positioning above 30° were obtained in more than 90% of cases. Oral care was more frequently performed within a day, using more concentrated solutions of chlorhexidine. Subglottic suctioning also was used but in only 24.7% of the cases in the last years. Regarding the prevalence of VAP, it significantly decreased from 28% of ventilated patients in 2010 to 10.1% in 2016 (p ≤ 0.0001). Although a causal relationship cannot be inferred from these data, the successive surveys revealed a potential impact of the VAP bundle campaign on both the respiratory care of ventilated patients and the prevalence of VAP in Belgian ICUs encouraging them to follow the guidelines.

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

    International Nuclear Information System (INIS)

    Xia, Hong; Li, Bin; Liu, Jianxin

    2014-01-01

    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

  8. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

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

    Energy Technology Data Exchange (ETDEWEB)

    Merny, R. [Association Vincotte, Avenue du Roi 157, B-1060 Bruxelles/Brussels (Belgium)

    1986-02-15

    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)

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

    International Nuclear Information System (INIS)

    Merny, R.

    1986-01-01

    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)

  11. The actual practice of air cleaning in Belgian nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Goossens, W.R. [PEGO, Mol (Belgium)

    1995-02-01

    With 60% of its power generation from nuclear stations Belgium has 7 nuclear power stations in operation with a total capacity of 5.4 MWe. Enriched uranium is imported and converted to fuel assemblies. The actinides of reprocessed fuel are recycled as MOX fuel. A main waste conditioning operation has been performed in the PAMELA vitrifier. The actual practice of nuclear air cleaning in the Belgian PWR station DOEL-4 and in the PAMELA -vitrification plant for high level liquid waste is reviewed.

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

    International Nuclear Information System (INIS)

    Affonso, Renato Raoni Werneck; Lava, Deise Diana; Borges, Diogo da Silva; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes

    2014-01-01

    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

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

    International Nuclear Information System (INIS)

    Broccoli, U.; Cambi, G.; Vanossi, A.; Zapellini, G.

    1977-01-01

    This report presents the results of several experiments carried out in the D 2 O-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)

  14. Examination of U3Si2-Al fuel elements from the Oak Ridge Research Reactor

    International Nuclear Information System (INIS)

    Copeland, G.L.; Snelgrove, J.L.; Hofman, G.L.

    1986-01-01

    The results of postirradiation examination of low-enriched U 3 Si 2 fuel elements from the Oak Ridge Research Reactor are presented. The elements replaced standard high-enriched elements and were handled routinely except that the burnup of half the elements was extended beyond normal limits up to about 98% peak. The elements were manufactured by commercial fuel suppliers. The performance was completely satisfactory for all the elements

  15. A solution to level 3 dismantling of gas-cooled reactors: Graphite incineration

    International Nuclear Information System (INIS)

    Dubourg, M.

    1993-01-01

    This paper presents an approach developed to solve the specific decommissioning problems of the G2 and G3 gas cooled reactors at Marcoule and the strategy applied with emphasis in incinerating the graphite core components, using a fluidized-bed incinerator developed jointly between the CEA and FRAMATOME. The incineration option was selected over subsurface storage for technical and economic reasons. Studies have shown that gaseous incineration releases are environmentally acceptable

  16. 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.

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

    International Nuclear Information System (INIS)

    Zhang, Tengfei; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi; Li, Yunzhao

    2013-01-01

    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 S N 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 k eff 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

  18. 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.

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

    International Nuclear Information System (INIS)

    Houtte, Paul van; Bourgois, Nicolas; Renard, Francoise; Huget, Philippe; D'hoore, William; Scalliet, Pierre

    2007-01-01

    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

  20. 550 tests to support Tihange 2 and Doel 3 integrity assessment. Extensive research based on SCK-CEN's long-term expertise

    International Nuclear Information System (INIS)

    2014-01-01

    The article discusses the analyses that were performed in order to assess of the impact of the hydrogen flakes on the mechanical properties of the vessels of the Doel 3 and Tihange 2 nuclear power plants . Results from 550 tests shows no significant influence on the mechanical properties. Both reactors were given the green light by the regulator AFCN/FANC to restart operation. On specific request of the Belgian safety authorities, an interim research programme was set up in parallel to study the effects of irradiation on the material. The BR2 reactor is engaged for this research.

  1. Reactor Dosimetry Applications Using RAPTOR-M3G:. a New Parallel 3-D Radiation Transport Code

    Science.gov (United States)

    Longoni, Gianluca; Anderson, Stanwood L.

    2009-08-01

    The numerical solution of the Linearized Boltzmann Equation (LBE) via the Discrete Ordinates method (SN) requires extensive computational resources for large 3-D neutron and gamma transport applications due to the concurrent discretization of the angular, spatial, and energy domains. This paper will discuss the development RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries), a new 3-D parallel radiation transport code, and its application to the calculation of ex-vessel neutron dosimetry responses in the cavity of a commercial 2-loop Pressurized Water Reactor (PWR). RAPTOR-M3G is based domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architectures. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor, yielding an efficient solution methodology for large 3-D problems. Measured neutron dosimetry responses in the reactor cavity air gap will be compared to the RAPTOR-M3G predictions. This paper is organized as follows: Section 1 discusses the RAPTOR-M3G methodology; Section 2 describes the 2-loop PWR model and the numerical results obtained. Section 3 addresses the parallel performance of the code, and Section 4 concludes this paper with final remarks and future work.

  2. 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)

  3. 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)

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

    International Nuclear Information System (INIS)

    Cutbush, J.D.; Abbott, E.C.; Carpenter, W.L. Jr.

    1992-01-01

    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

  5. 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)

  6. 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.

  7. 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.

  8. Development of multi-physics code systems based on the reactor dynamics code DYN3D

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Schuetze, Jochen [ANSYS Germany GmbH, Darmstadt (Germany); Frank, Thomas [ANSYS Germany GmbH, Otterfing (Germany); Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany)

    2011-07-15

    The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

  9. Development of multi-physics code systems based on the reactor dynamics code DYN3D

    International Nuclear Information System (INIS)

    Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich; Schuetze, Jochen; Frank, Thomas; Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo

    2011-01-01

    The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

  10. Horizontal bioreactor for ethanol production by immobilized cells. Pt. 3. Reactor modeling and experimental verification

    Energy Technology Data Exchange (ETDEWEB)

    Woehrer, W

    1989-04-05

    A mathematical model which describes ethanol formation in a horizontal tank reactor containing Saccharomyces cerevisiae immobilized in small beads of calcium alignate has been developed. The design equations combine flow dynamics of the reactor as well as product formation kinetics. The model was verified for 11 continuous experiments, where dilution rate, feed glucose concentration and bead volume fraction were varied. The model predicts effluent ethanol concentration and CO/sub 2/ production rate within the experimental error. A simplification of the model is possible, when the feed glucose concentration does not exceed 150 kg/m/sup 3/. The simplification results in an analytical solution of the design equation and hence can easily be applied for design purposes as well as for optimization studies.

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

    International Nuclear Information System (INIS)

    Takano, Hideki; Kaneko, Kunio.

    1989-10-01

    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)

  12. 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.

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

    International Nuclear Information System (INIS)

    Bombaerts, Gunter; Bovy, Michel; Laes, Erik

    2006-01-01

    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

  14. 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

  15. Predicting mortality and incident immobility in older Belgian men by characteristics related to sarcopenia and frailty

    DEFF Research Database (Denmark)

    Kruse, C; Goemaere, S; De Buyser, S

    2018-01-01

    and bone mineral density scores were the most important predictors. INTRODUCTION: Machine learning principles were used to predict 5-year mortality and 3-year incident severe immobility in a population of older men by frailty and sarcopenia characteristics. METHODS: Using prospective data from 1997 on 264......There is an increasing awareness of sarcopenia in older people. We applied machine learning principles to predict mortality and incident immobility in older Belgian men through sarcopenia and frailty characteristics. Mortality could be predicted with good accuracy. Serum 25-hydroxyvitamin D...... the most important predictors of immobility. Sarcopenia assessed by lean mass estimates was relevant to mortality prediction but not immobility prediction. CONCLUSIONS: Using advanced statistical models and a machine learning approach 5-year mortality can be predicted with good accuracy using a Bayesian...

  16. Thermal fluid dynamics study of nuclear advanced reactors of high temperature using RELAP5-3D

    International Nuclear Information System (INIS)

    Scari, Maria Elizabeth

    2017-01-01

    Fourth Generation nuclear reactors (GEN-IV) are being designed with special features such as intrinsic safety, reduction of isotopic inventory and use of fuel in proliferation-resistant cycles. Therefore, the investigation and evaluation of operational and safety aspects of the GEN-IV reactors have been the subject of numerous studies by the international community and also in Brazil. In 2008, in Brazil, was created the National Institute of Science and Technology of Innovative Nuclear Reactors, focusing on studies of projects and systems of new generation reactors, which included GEN-IV reactors as well as advanced PWR (Pressurized Water Reactor) concepts. The Department of Nuclear Engineering of the Federal University of Minas Gerais (DEN-UFMG) is a partner of this Institute, having started studies on the GEN-IV reactors in the year 2007. Therefore, in order to add knowledge to these studies, in this work, three projects of advanced reactors were considered to verify the simulation capability of the thermo-hydraulic RELAP5-3D code for these systems, either in stationary operation or in transient situations. The addition of new working fluids such as ammonia, carbon dioxide, helium, hydrogen, various types of liquid salts, among them Flibe, lead, lithium-bismuth, lithium-lead, was a major breakthrough in this version of the code, allowing also the simulation of GEN-IV reactors. The modeling of the respective core of an HTTR (High Temperature Engineering Test Reactor), HTR-10 (High Temperature Test Module Reactor) and LS-VHTR (Liquid-Salt-Cooled Very-High-Temperature Reactor) were developed and verified in steady state comparing the values found through the calculations with reference data from other simulations, when it is possible. The first two reactors use helium gas as coolant and the LS-VHTR uses a mixture of 66% LiF and 34% of BeF 2 , the LiF-BeF 2 , also know as Flibe. All the studied reactors use enriched uranium as fuel, in form of TRISO (Tristructural

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

    International Nuclear Information System (INIS)

    Hernandez, N.; Alonso, G.; Valle, E. del

    2004-01-01

    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. Belgian experience with radiation technologies for sterilization

    International Nuclear Information System (INIS)

    Bonet, H.; Dardenne, P.; Herier, A.

    1998-01-01

    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 Co 60 , source having a capacity for sterilization of more than 100 m 3 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)

  19. 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...

  20. 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.

  1. 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.

  2. Verification of the CASMO-3/SIMULATE-3 pin power accuracy by comparison with operating boiling water reactor measurements

    International Nuclear Information System (INIS)

    Uegata, T.; Saji, E.; Tanaka, H.

    1993-01-01

    Intranodal pin power distributions calculated by the CASMO-3/SIMULATE-3 code have been compared with pin gamma scan measurements. These data were obtained from the depleted core of an operating boiling water reactor (BWR), which is more complicated than a pressurized water reactor to calculate because of the existence of coolant void distributions and cruciform control blades. Furthermore, measured bundles include mixed-oxide (MOX) bundles in which steep thermal flux gradients occur. Both UO 2 and MOX bundles have been calculated in the same manner based on the standard CASMO-3/SIMULATE-3 methods. The total pin power root-mean-square (rms) error is 2.7%, which includes measurement error, from an 896-point comparison. There is no obvious dependency on axial elevations (void fractions) and no significant difference between fuel types (UO 2 or MOX), although the errors in a peripheral bundle, which is less important from the standpoint of core design, are somewhat larger than those in the internal bundles. If the peripheral bundle is excluded, the total rms error is reduced to 2.2%. From these results, it is concluded that excellent agreement has been obtained between the calculations and measurements and that the calculational capability of CASMO-3/SIMULATE-3 for the intranodal pin power distribution is quite satisfactory and useful for BWR core design

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

    International Nuclear Information System (INIS)

    Cantargi, F; Sbaffoni, M; Granada, R

    2004-01-01

    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 [es

  4. The ARIES-III D-3He tokamak reactor: Design-point determination and parametric studies

    International Nuclear Information System (INIS)

    Bathke, C.G.; Werley, K.A.; Miller, R.L.; Krakowski, R.A.; Santarius, J.F.

    1991-01-01

    The multi-institutional ARIES study has generated a conceptual design of another tokamak fusion reactor in a series that varies the assumed advances in technology and physics. The ARIES-3 design uses a D- 3 He fuel cycle and requires advances in technology and physics for economical attractiveness. The optimal design was characterized through systems analyses for eventual conceptual engineering design. Results from the systems analysis are summarized, and a comparison with the high-field, D-T fueled ARIES-1 is included. 11 refs., 5 figs

  5. 3-dimensional finite element modelling of reactor building internal structure for static analysis

    International Nuclear Information System (INIS)

    Joshi, M.H.; Reddy, V.J.; Kushwaha, H.S.; Reddy, G.R.; Karandikar, G.V.

    1991-01-01

    a) Thin shell element gives fairly accurate results when compared to 3-D Brick element for the type of structure and loading in Reactor Building. b) The maximum element size is fixed from model 3(c) i.e. 2.0 m. c) Openings with size smaller than 0.5 m can be neglected without affecting the results very much. d) For any such problem, the methodology described in this paper can be used to take rational decisions which will ensure reasonable accuracy. (author)

  6. Extension of the GeN-Foam neutronic solver to SP3 analysis and application to the CROCUS experimental reactor

    International Nuclear Information System (INIS)

    Fiorina, Carlo; Hursin, Mathieu; Pautz, Andreas

    2017-01-01

    Highlights: • Development and verification of an SP 3 solver based on OpenFOAM. • Integration into the GeN-Foam multi-physics platform. • Application of the new GeN-Foam SP 3 solver to the CROCUS reactor. - Abstract: The Laboratory for Reactor Physics and Systems Behaviour at the PSI and at the EPFL has been developing since 2013 a multi-physics platform for coupled reactor analysis named GeN-Foam. The developed tool includes a solver for the eigenvalue and transient solution of multi-group neutron diffusion equations. Although frequently used in reactor analysis, the diffusion theory shows some limitations for core configurations involving strong anisotropies, which is the case for the CROCUS research reactor at the EPFL. The use of an SP 3 approximation to neutron transport can often lead to visible improvements in a code predictive capabilities, especially for one-directional anisotropies, with acceptable added computational cost vs diffusion. Following some modelling issues for the CROCUS reactor, and in order to improve the GeN-Foam modelling capabilities, the GeN-Foam diffusion solver has been extended to allow for SP 3 analyses. The present paper describes such extension and a preliminary verification using a mini-core PWR benchmark. The newly developed solver is then applied to the analysis of the CROCUS experimental reactor and results are compared to Monte Carlo calculations, as well as to the results of the diffusion solver.

  7. Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code

    Science.gov (United States)

    Pinem, Surian; Malem Sembiring, Tagor; Tukiran; Deswandri; Sunaryo, Geni Rina

    2018-02-01

    The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/°C, -1.00518 pcm/°C to 1.00649 pcm/°C and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 °C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.

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

    International Nuclear Information System (INIS)

    Huda, M.Q.; Bhuiyan, S.I.; Mondal, M.A.W.

    1996-12-01

    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 846.76 o C which is much less than the limiting maximum value of fuel temperature of 1150 0 C as reported in SAR. During a pulse if the film boiling occurs for a peak adiabatic fuel temperature of 1000 o C, the calculated outer cladding wall temperature was observed to be 702.39 0 C compared to a value of 760 o 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)

  9. MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

    Energy Technology Data Exchange (ETDEWEB)

    Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A. [AREVA - Tour AREVA, 1 Place Jean Millier, 92084 Paris La Defense (France)

    2013-07-01

    In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO{sub 2} fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory.

  10. MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

    International Nuclear Information System (INIS)

    Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A.

    2013-01-01

    In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO 2 fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory

  11. Development of M3C code for Monte Carlo reactor physics criticality calculations

    International Nuclear Information System (INIS)

    Kumar, Anek; Kannan, Umasankari; Krishanani, P.D.

    2015-06-01

    The development of Monte Carlo code (M3C) for reactor design entails use of continuous energy nuclear data and Monte Carlo simulations for each of the neutron interaction processes. BARC has started a concentrated effort for developing a new general geometry continuous energy Monte Carlo code for reactor physics calculation indigenously. The code development required a comprehensive understanding of the basic continuous energy cross section sets. The important features of this code are treatment of heterogeneous lattices by general geometry, use of point cross sections along with unionized energy grid approach, thermal scattering model for low energy treatment, capability of handling the microscopic fuel particles dispersed randomly. The capability of handling the randomly dispersed microscopic fuel particles which is very useful for the modeling of High-Temperature Gas-Cooled reactor fuels which are composed of thousands of microscopic fuel particle (TRISO fuel particle), randomly dispersed in a graphite matrix. The Monte Carlo code for criticality calculation is a pioneering effort and has been used to study several types of lattices including cluster geometries. The code has been verified for its accuracy against more than 60 sample problems covering a wide range from simple (like spherical) to complex geometry (like PHWR lattice). Benchmark results show that the code performs quite well for the criticality calculation of the system. In this report, the current status of the code, features of the code, some of the benchmark results for the testing of the code and input preparation etc. are discussed. (author)

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

    International Nuclear Information System (INIS)

    Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira

    2015-01-01

    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)

  13. 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)

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

    International Nuclear Information System (INIS)

    1980-01-01

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

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

    International Nuclear Information System (INIS)

    Santarius, J.F.; Blanchard, J.P.; Emmert, G.A.; Sviatoslavsky, I.N.; Wittenberg, L.J.; Ghoneim, N.M.; Hasan, M.Z.; Mau, T.K.; Greenspan, E.; Herring, J.S.; Kernbichler, W.; Klein, A.C.; Miley, G.H.; Miller, R.L.; Peng, Y.K.M.

    1989-01-01

    The potential for highly efficient conversion of fusion power to electricity provides one motivation for investigating D- 3 He fusion reactors. This stems from: (1) the large fraction of D- 3 He 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- 3 He 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- 3 He 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

  16. Study of seismic responses of Candu-3 reactor building using isolator bearings

    International Nuclear Information System (INIS)

    Biswas, J.K.

    1992-01-01

    Seismic isolator bearings are known to increase reliability, reduce cost and increase the potential sitings for nuclear power plants located in regions of high seismicity. High seismic activities in Canada occur mainly in the western coast, the Grand Banks and regions of Quebec along the St. Lawrence river. In Canada, nuclear power plants are located in Ontario, Quebec and New Brunswick where the seismicity levels are low to moderate. Consequently, seismic isolator bearings have not been used in the existing nuclear power plants in Canada. The present paper examines the effect of using seismic isolator bearings in the design for the new CANDU3 which would be suitable for regions having high seismicity. The CANDU3 Nuclear Power Plant is rated at 450 MW of net output power and is a smaller version of its predecessor CANDU6 successfully operating in Canada and abroad. The design of CANDU3 is being developed by AECL CANDU. Advanced technologies for design, construction and plant operation have been utilized. During the conceptual development of the CANDU3 design, various design options including the use of isolator bearings were considered. The present paper presents an overview of seismic isolation technology and summarizes the analytical work for predicting the seismic behavior of the CANDU3 reactor building. A lumped-parameter dynamic model for the reactor building is used for the analysis. The characteristics of the bearings are utilized in the analysis work. The time-history modal analysis has been used to compute the seismic responses. Seismic responses of the reactor building with and without isolator bearings are compared. The isolator bearings are found to reduce the accelerations of the reactor building. As a result, a lower level of seismic qualification for components and systems would be required. The use of these bearings however increases rigid body seismic displacements of the structure requiring special considerations in the layout and interfaces for

  17. Processing requirements for property optimization of Eu2O3-W cermets for fast reactor neutron absorber applications

    International Nuclear Information System (INIS)

    Pasto, A.E.; Tennery, V.J.

    1977-01-01

    Europium sesquioxide is a candidate fast reactor neutron absorber material. It possesses several desirable characteristics for this application, but has a low thermal conductivity. This gives rise to pellet cracking during reactor operation. To increase the thermal conductivity without great sacrifice in nuclear worth, addition of tungsten to Eu 2 O 3 has been evaluated. Synthesis and fabrication techniques described allow preparation of high density compacts of Eu 2 O 3 -15 vol. percent tungsten, possessing favorable thermal conductivity and thermal expansion characteristics

  18. 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).

  19. Belgian citizens' and broiler producers' perceptions of broiler chicken welfare in Belgium versus Brazil.

    Science.gov (United States)

    Vanhonacker, F; Tuyttens, F A M; Verbeke, Wim

    2016-07-01

    New EU regulations require more stringent country-of-origin labeling, while imports of broiler meat from non-EU countries are increasing. In light of these trends, we have studied citizens' and producers' perceptions of broiler meat originating from Belgium versus Brazil and their perception of broiler production in Belgium versus Brazil. A particular focus was the association between country of origin and perceived level of animal welfare. We also investigated the perception of scaling-up and outdoor access in terms of perceived level of animal welfare. Cross-sectional survey data was collected among Flemish citizens (n = 541) and broiler producers (n = 114). In accordance with literature on general farm animal welfare, both stakeholder types claimed to allocate great importance to broiler welfare and generally agreed with the Welfare Quality model of broiler welfare. Citizens disagreed with the producers that 1) consumers are not willing to pay more for higher welfare products, 2) that broilers suffer little, 3) that broiler welfare in current Belgian production units is generally non-problematic, 4) that scaling-up production units would not have a positive impact on profitability nor a profoundly negative impact on broiler welfare, and 5) that the impact of providing broilers with outdoor access is negative for consumers, farmers, and broilers. Country of origin had a strong influence on the perception of both broiler production and broiler meat. Belgian citizens, and producers (much more than citizens) considered nearly all aspects related to broiler production and broiler meat to be significantly superior for chicken produced in Belgium compared to Brazil. Further research should focus on how these perceptions influence purchase intentions and production decisions. Future avenues for research are to quantify market opportunities for country-of-origin labeling and to investigate to which extent stakeholders' perceptions correspond with reality. © 2016 Poultry

  20. Proceedings of 2. Yugoslav symposium on reactor physics, Part 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-07-01

    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.

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

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-12-01

    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

  2. Comparison and analysis of 1D/2D/3D neutronics modeling for a fusion reactor

    International Nuclear Information System (INIS)

    Li, J.; Zeng, Q.; Chen, M.; Jiang, J.; Wu, Y.

    2007-01-01

    During the course of analyzing the characteristics for fusion reactors, the refined calculations are needed to confirm that the nuclear design requirements are met. Since the long computational time is consumed, the refined three-dimensional (3D) representation has been used primarily for establishing the baseline reference values, analyzing problems which cannot be reduced by symmetry considerations to lower dimensions, or where a high level of accuracy is desired locally. The two-dimensional (2D) or one-dimensional (1D) description leads itself readily to resolve many problems, such as the studies for the material fraction optimization, or for the blanket size optimization. The purpose of this paper is to find out the differences among different geometric descriptions, which can guide the way to approximate and simplify the computational model. The fusion power reactor named FDS-II was designed as an advanced fusion power reactor to demonstrate and validate the commercialization of fusion power by Institute of Plasma Physics, Chinese Academy of Science. In this contribution, the dual-cooled lithium lead (DLL) blanket of FDS-II was used as a reference for neutronics comparisons and analyses. The geometric descriptions include 1D concentric sphere model, 1D, 2D and 3D cylinder models. The home-developed multi-functional neutronics analysis code system VisualBUS, the Monte Carlo transport code MCNP and nuclear data library HENDL have been used for these analyses. The neutron wall loading distribution, tritium breeding ratio (TBR) and nuclear heat were calculated to evaluate the nuclear performance. The 3D calculation has been used as a comparison reference because it has the least errors in the treatment of geometry. It is suggested that the value of TBR calculated by the 1D approach should be greater than 1.3 to satisfy the practical need of tritium self-sufficiency. The distribution of nuclear heat based on the 2D and 3D models were similar since they all consider

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

    International Nuclear Information System (INIS)

    Peacock, H.B.

    1984-01-01

    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 U 3 O 8 -aluminum cermets. Above the aluminum melting point, U 3 O 8 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 therefore 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 % U 3 O 8 in aluminum. This content corresponds to a theoretical uranium density of 1.9 g/cc. Results indicate that the thermite reaction occurs at about 900 0 C in air without a violent effect. The heat of reaction was approximately 123 cal/g of U 3 O 8 -aluminum fuel. Tensile and compressive strength of the fuel tube section is low above 660 0 C. In tension, sections failed at about the aluminum melting point. In compression with 2 psi average axial stress, failure occurred at 917 0 C, while 7 psi average axial stress produced failure at 669 0 C. (author)

  4. Development of a 3-D flow analysis computer program for integral reactor

    International Nuclear Information System (INIS)

    Youn, H. Y.; Lee, K. H.; Kim, H. K.; Whang, Y. D.; Kim, H. C.

    2003-01-01

    A 3-D computational fluid dynamics program TASS-3D is being developed for the flow analysis of primary coolant system consists of complex geometries such as SMART. A pre/post processor also is being developed to reduce the pre/post processing works such as a computational grid generation, set-up the analysis conditions and analysis of the calculated results. TASS-3D solver employs a non-orthogonal coordinate system and FVM based on the non-staggered grid system. The program includes the various models to simulate the physical phenomena expected to be occurred in the integral reactor and will be coupled with core dynamics code, core T/H code and the secondary system code modules. Currently, the application of TASS-3D is limited to the single phase of liquid, but the code will be further developed including 2-phase phenomena expected for the normal operation and the various transients of the integrator reactor in the next stage

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

    International Nuclear Information System (INIS)

    Peacock, H.B.

    1983-01-01

    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 U 3 O 8 -aluminum cermets. Above the aluminum melting point, U 3 O 8 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 % U 3 O 8 in aluminum. This content corresponds to a theoretical uranium density of 1.9 g/cc. Results indicate that the thermite reaction occurs at about 900 0 C in air without a violent effect. The heat of reaction was approximately 123 cal/g of U 3 O 8 -aluminum fuel. Tensile and compressive strength of the fuel tube section is low above 660 0 C. In tension, sections failed at about the aluminum melting point. In compression with 2-psi average axial stress, failure occurred at 917 0 C, while 7 psi average axial stress produced failure at 669 0 C

  6. 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.

  7. Knowledge and attitudes of nurses on pressure ulcer prevention: a cross-sectional multicenter study in Belgian hospitals.

    Science.gov (United States)

    Beeckman, Dimitri; Defloor, Tom; Schoonhoven, Lisette; Vanderwee, Katrien

    2011-09-01

    Evidence-based guidelines for pressure ulcer prevention have been developed and promoted by authoritative organizations. However, nonadherence to these guidelines is frequently reported. Negative attitudes and lack of knowledge may act as barriers to using guidelines in clinical practice. To study the knowledge and attitudes of nurses about pressure ulcer prevention in Belgian hospitals and to explore the correlation between knowledge, attitudes, and the application of adequate prevention. A cross-sectional multicenter study was performed in a random sample of 14 Belgian hospitals, representing 207 wards. Out of that group, 94 wards were randomly selected (2105 patients). Clinical observations were performed to assess the adequacy of pressure ulcer prevention and pressure ulcer prevalence. From each participating ward, a random selection of at least five nurses completed an extensively validated knowledge and attitude instrument. In total, 553 nurses participated. A logistic regression analysis was performed to evaluate the correlation between knowledge, attitudes, and the application of adequate prevention. Pressure ulcer prevalence (Category I-IV) was 13.5% (284/2105). Approximately 30% (625/2105) of the patients were at risk (Bradenscore pressure ulcer). Only 13.9% (87/625) of these patients received fully adequate prevention whilst in bed and when seated. The mean knowledge and attitude scores were 49.7% and 71.3%, respectively. The application of adequate prevention on a nursing ward was significantly correlated with the attitudes of the nurses (OR = 3.07, p = .05). No independent correlation was found between knowledge and the application of adequate prevention (OR = 0.75, p = .71). Knowledge of nurses in Belgian hospitals about the prevention of pressure ulcers is inadequate. The attitudes of nurses toward pressure ulcers are significantly correlated with the application of adequate prevention. No correlation was found between knowledge and the

  8. Improved SAFARI-1 research reactor irradiation position modeling in OSCAR-3 code system

    International Nuclear Information System (INIS)

    Moloko, L. E.; Belal, M. G. A. H.

    2009-01-01

    The demand on the availability of irradiation positions in the SAFARI-1 reactor is continuously increasing due to the commercial pressure to produce isotopes more efficiently. This calls for calculational techniques and modeling methods to be improved regularly to optimize irradiation services. The irradiation position models are improved using the OSCAR-3 code system, and results are compared to experimental measurements. It is concluded that the irradiation position models are essential if realistic core follow and reload studies are to be performed and most importantly, for the realization of improved agreement between experimental data and calculated results. (authors)

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

    International Nuclear Information System (INIS)

    Farias, Ruben; Gonzalez, S.J.; Bellino, A.; Sztenjberg, M.; Pinto, J.; Thorp, Silvia I.; Gadan, M.; Pozzi, Emiliano; Schwint, Amanda E.; Heber, Elisa M.; Trivillin, V.A.; Zarza, Leandro G.; Estryk, Guillermo; Miller, M.; Bortolussi, S.; Soto, M.S.; Nigg, D.W.

    2009-01-01

    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)

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

    International Nuclear Information System (INIS)

    Nicol, R.G.; Parrott, J.R.; Krichinsky, A.M.; Box, W.D.; Martin, C.W.; Whitson, W.R.

    1982-05-01

    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 233 U from rejected elements. This report describes the steps associated with conversion of purified uranyl nitrate (as solution) to U 3 O 8 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

  11. Numerical Simulation of Measurements during the Reactor Physical Startup at Unit 3 of Rostov NPP

    Science.gov (United States)

    Tereshonok, V. A.; Kryakvin, L. V.; Pitilimov, V. A.; Karpov, S. A.; Kulikov, V. I.; Zhylmaganbetov, N. M.; Kavun, O. Yu.; Popykin, A. I.; Shevchenko, R. A.; Shevchenko, S. A.; Semenova, T. V.

    2017-12-01

    The results of numerical calculations and measurements of some reactor parameters during the physical startup tests at unit 3 of Rostov NPP are presented. The following parameters are considered: the critical boron acid concentration and the currents from ionization chambers (IC) during the scram system efficiency evaluation. The scram system efficiency was determined using the inverse point kinetics equation with the measured and simulated IC currents. The results of steady-state calculations of relative power distribution and efficiency of the scram system and separate groups of control rods of the control and protection system are also presented. The calculations are performed using several codes, including precision ones.

  12. Electric failure on the reactor n.3 of the nuclear power plant of Dampierre

    International Nuclear Information System (INIS)

    2007-05-01

    This note of information resumes the progress of the electric failure on the reactor n.3 of the nuclear power plant of Dampierre, the organization during the incident, it establishes then a comparison with the incident arisen to Forsmark in 2006 and reminds that it lead in an inspection on behalf of the Asn which noticed that all the procedures had been respected by the operators and did not noticed any abnormality in the maintenance. This event was classified at the level 1 of the international nuclear event scale (INES). (N.C.)

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

    International Nuclear Information System (INIS)

    Baudron, A. M.; Doderlein, C.; Guerin, P.; Lautard, J. J.; Moreau, F.

    2007-01-01

    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)

  14. Data acquisition system for the 3 MW TRIGA reactor at AERE Savar

    International Nuclear Information System (INIS)

    Abudl Ahad, A.O.M.

    1998-01-01

    A 3 MW TRIGA Mark II research reactor control console has been studied in detail and the channels have been selected for monitoring, display and record using the microcomputer. Information from these channels are fed to the computer through hardware like buffer, AD converter, multiplexer, etc. for continues display and permanent records using video monitor, printer and diskettes. Besides, the information from the console, other information like operating time, power, total burnup of fuel, operating persons, etc. are also available, with very little modifications in both hardware and software, the data logging system is now running successfully. (author)

  15. The Quality of Work in the Belgian Service Voucher System.

    Science.gov (United States)

    Mousaid, Sarah; Huegaerts, Kelly; Bosmans, Kim; Julià, Mireia; Benach, Joan; Vanroelen, Christophe

    2017-01-01

    Several European countries implemented initiatives to boost the growth of the domestic cleaning sector. Few studies investigated the quality of work in these initiatives, although effects on workers' health and on social health inequalities can be expected. This study contributes to the scant research on this subject, by investigating the quality of work in the Belgian service voucher system - a subsidized system for domestic work. The applied research methodology includes a qualitative content analysis of parliamentary debates, legislation and previous research about the service voucher system and of 40 in-depth interviews with service voucher workers. The study shows that the legal framework that regulates the system must be further enhanced in order to improve the quality of work in the service voucher system. In addition, the actors involved must be better controlled, and sanctioned in case of non-compliance with legislation. © The Author(s) 2016.

  16. RECRUITING OLDER VOLUNTEERS: FINDINGS FROM THE BELGIAN AGEING STUDIES

    Directory of Open Access Journals (Sweden)

    Sarah DURY

    2010-01-01

    Full Text Available Although there is a significant body of work concerning voluntary work, hardly any attention is given to volunteering of older individuals. Moreover, the potential volunteers among older adults is even less examined. Next to volunteering among olde r adults, the neighbou rhood becomes more salient when people age and this due to their more intense use and time spent in the neighbourhood. In response to these lacunae, the main purpose of this contribution is to examine the impact of subjective neighbourhood features on the recruitment potential for volunteering among older people. This study uses data collected from the Belgian Ageing Studies. 59.977 adults aged sixty and over living self-reliantly in 127 Flemish municipalities in Belgium participated in this study. A binary logistic regression is ap plied to analyse the key va riables characterizing potential volunteers. Our findings stress the need for recognizing the crucial importance of the locality when recruiting older adults for volunteer activities.

  17. 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.

  18. Research on 3D power distribution of PWR reactor core based on RBF neural network

    International Nuclear Information System (INIS)

    Xia Hong; Li Bin; Liu Jianxin

    2014-01-01

    Real-time monitor for 3D power distribution is critical to nuclear safety and high efficiency of NPP's operation as well as the control system optimization. A method was proposed to set up a real-time monitor system for 3D power distribution by using of ex-core neutron detecting system and RBF neural network for improving the instantaneity of the monitoring results and reducing the fitting error of the 3D power distribution. A series of experiments were operated on a 300 MW PWR simulation system. The results demonstrate that the new monitor system works very well under condition of certain burnup range during the fuel cycle and reconstructs the real-time 3D distribution of reactor core power. The accuracy of the model is improved effectively with the help of several methods. (authors)

  19. CFX-10 and RELAP5-3D simulations of coolant mixing phenomena in RPV of VVER-1000 reactors

    International Nuclear Information System (INIS)

    Terzuoli, F.; Moretti, F.; Melideo, D.; D'Auria, F.; Shkarupa, O.

    2006-01-01

    The present paper deals with numerical analyses of coolant mixing in the reactor pressure vessel of a VVER-1000 reactor, performed with the ANSYS CFX-10 CFD code and with the RELAP5-3D system code. In particular, the attention focused on the 'swirl' effect that has been observed to take place in the downcomer of such kind of reactor, with the aim of assessing the capability of the codes to predict that effect, and to understand the reasons for its occurrence. The results have been compared against experimental data from V1000CT-2 Benchmark. (author)

  20. Survey of group data libraries for use of the DYN3D program for WWER type reactors

    International Nuclear Information System (INIS)

    Mittag, S.

    1994-06-01

    So-called few-group neutron data have to be used as input data in core models (such as DYN3D) calculating the reactor behaviour. A survey is given of qualified data libraries for the reactor cores of Russian VVER. The information about primary data used in group data generation and the accuracy reached by the cell codes is compiled in tables. To assess the quality of the data, comparisons have been made between measured and calculated reactor parameters. The information available does not show significant differences concerning the quality of the data libraries. (orig.) [de

  1. 3 MW TRIGA Research Reactor facility of BAEC and its Utilization

    International Nuclear Information System (INIS)

    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.

    1999-01-01

    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)

  2. 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)

  3. 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

    Energy Technology Data Exchange (ETDEWEB)

    Nonboel, E

    1985-07-01

    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)

  4. Treatment adherence in multiple sclerosis: a survey of Belgian neurologists

    Directory of Open Access Journals (Sweden)

    Decoo D

    2015-11-01

    Full Text Available Danny Decoo,1 Mathieu Vokaer2 1Department of Neurology and Neurorehab, AZ Alma, Sijsele, Belgium; 2Multiple Sclerosis Clinic, Edith Cavell Hospital, CHIREC group, Brussels, Belgium Background: Poor treatment adherence is common among patients with multiple sclerosis (MS. This survey evaluated neurologists’ perception of treatment adherence among MS patients.Materials and methods: This questionnaire-based survey of Belgian neurologists treating MS patients was conducted between June and July 2014. Face-to-face interviews with the neurologists were based on a semistructured questionnaire containing questions regarding the perception of the treatment-adherence level.Results: A total of 41 neurologists participated in the survey. Of these, 88% indicated frequent discussions about treatment adherence as beneficial for treatment efficacy. The mean time spent on the treatment-adherence discussion during the initial consultation was 11 minutes, with 24% of doctors spending 5 minutes and 24% of doctors spending 10 minutes discussing this issue. The majority of neurologists (56% perceived the adherence level in MS as good, and 12% perceived it as excellent. The majority of neurologists (64% indicated intolerance as a main cause of poor adherence, and all neurologists reported insufficient efficacy as a consequence of nonadherence. The importance of adherence in the neurologists’ practice was evaluated on a scale of 1–10, with 1= “not very important” and 10= “very important”: 44% of doctors indicated a score of 10, and the mean score was 9.0.Conclusion: Belgian neurologists consider treatment adherence in MS as essential for the benefits of therapies. However, although neurologists are aware of the consequences of nonadherence, they generally spend limited time discussing the importance of treatment adherence with their patients. Keywords: multiple sclerosis, treatment adherence, physician survey

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

    International Nuclear Information System (INIS)

    2014-02-01

    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)

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

    International Nuclear Information System (INIS)

    Nikonov, S.P.; Velkov, K.; Pautz, A.

    2011-01-01

    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)

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

    International Nuclear Information System (INIS)

    2007-03-01

    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)

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

    International Nuclear Information System (INIS)

    2009-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-02-15

    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)

  10. 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)

    Miyazaki, Osamu; Awa, Yasuaki; Isaka, Koji; Kutsukake, Kenichi; Komeda, Masao; Shibata, Ko; Hiyama, Kazuhisa; Suzuki, Mayu; Sone, Takuya; Ohuchi, Tomoaki; Terakado, Yuichi; Sataka, Masao

    2009-06-01

    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)

  11. Annual report of Department of Research Reactor and Tandem Accelerator, JFY2010. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and Tandem Accelerator

    International Nuclear Information System (INIS)

    Ishii, Tetsuro; Nakamura, Kiyoshi; Kawamata, Satoshi; Yamada, Yusuke; Kawashima, Kazuhiro; Asozu, Takuhiro; Nakamura, Takemi; Arai, Masaji; Yoshinari, Shuji; Sataka, Masao

    2012-03-01

    The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.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, 2010 and March 31, 2011. 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)

  12. Principle tests of reactor pumped laser of 3He-Ar-Xe system

    International Nuclear Information System (INIS)

    Chen Hande; Jin Xingxing; Yang Chengde

    1994-01-01

    A nuclear reactor-pumped laser was demonstrated firstly by using the 3 He(P,n) 3 H reaction to excite a 3 He-Ar-Xe laser. Lasing was observed on the 1.73 μm (5d (3/2) 1 -6p(5/2) 1 transition) in Xe I. The CAEP Pulsed Reactor (CFBR-II) was used as a source of fast neutrons moderated by a 50 mm thick polyethylene sleeve around the laser cell. A laser cell was constructed from K 9 glass tubing 600 mm long and Φ34 mm with each end cut at Brewster's angle (quartz). The laser cell was baked out to 10 -3 Pa prior to filling with research-grade mixture ( 3 He/Ar/Xe = 34.7: 34.7: 0.267 kPa). A dielectric-coated 2 m radius-curvature back mirror (99.7% reflectivity at 1.73 μm) and a flat output coupler (Φ30 mm) were used to form the optical cavity. The cavity optics were placed outside the cell as shown in Fig.3. A schematic of the experimental configuration is shown. Typical laser signals obtained with the Ge detector (-80 degree C) are also shown. Laser output lags the fast neutron pulse by the neutron thermalization time (∼278 μs). The observed light output was shown to be caused by stimulated emission, because the signal disappeared when the reflecting (flat) mirror was covered. This precludes the possibility that the phenomenon was the result of γ-radiation interaction with the detector or the air in the light path. The measured laser energy was 80 times the maximum possible fluorescence energy. The observed wavelength was measured to be 1.73 μm by using glass filters. Laser output duration was approximately 735 μs

  13. In-place thermal annealing of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Server, W.L.

    1985-04-01

    Radiation embrittlement of ferritic pressure vessel steels increases the ductile-brittle transition temperature and decreases the upper shelf level of toughness as measured by Charpy impact tests. A thermal anneal cycle well above the normal operating temperature of the vessel can restore most of the original Charpy V-notch energy properties. The Amry SM-1A test reactor vessel was wet annealed in 1967 at less than 343 0 C (650 0 F), and wet annealing of the Belgian BR-3 reactor vessel at 343 0 C (650 0 F) has recently taken place. An industry survey indicates that dry annealing a reactor vessel in-place at temperatures as high as 454 0 C (850 0 F) is feasible, but solvable engineering problems do exist. Economic considerations have not been totally evaluated in assessing the cost-effectiveness of in-place annealing of commercial nuclear vessels. An American Society for Testing and Materials (ASTM) task group is upgrading and revising guide ASTM E 509-74 with emphasis on the materials and surveillance aspects of annealing rather than system engineering problems. System safety issues are the province of organizations other than ASTM (e.g., the American Society of Mechanical Engineers Boiler and Pressure Vessel Code body)

  14. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 3. Programmer's manual. Final report

    International Nuclear Information System (INIS)

    Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

    1983-05-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear-reactor-core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This is Volume 3, the Programmer's Manual. It explains the codes' structures and the computer interfaces

  15. Simulation software of 3-D two-neutron energy groups for ship reactor with hexagonal fuel subassembly

    International Nuclear Information System (INIS)

    Zhang Fan; Cai Zhangsheng; Yu Lei; Gui Xuewen

    2005-01-01

    Core simulation software for 3-D two-neutron energy groups is developed. This software is used to simulate the ship reactor with hexagonal fuel subassembly after 10, 150 and 200 burnup days, considering the hydraulic and thermal feedback. It accurately simulates the characteristics of the fast and thermal neutrons and the detailed power distribution in a reactor under normal and abnormal operation condition. (authors)

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

    International Nuclear Information System (INIS)

    Govaerts, P.; Fieuw, G.; Deworm, J.P.; Zeevaert, Th.

    1986-01-01

    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.)

  17. Benchmark of the CASMO-3G/MICROBURN-B codes for Commonwealth Edison boiling water reactors

    International Nuclear Information System (INIS)

    Wheeler, J.K.; Pallotta, A.S.

    1992-01-01

    The Commonwealth Edison Company has performed an extensive benchmark against measured data from three boiling water reactors using the Studsvik lattice physics code CASMO-3G and the Siemens Nuclear Power three-dimensional simulator code MICROBURN-B. The measured data of interest for this benchmark are the hot and cold reactivity, and the core power distributions as measured by the traversing incore probe system and gamma scan data for fuel pins and assemblies. A total of nineteen unit-cycles were evaluated. The database included fuel product lines manufactured by General Electric and Siemens Nuclear Power, wit assemblies containing 7 x 7 to 9 x 9 pin configurations, several water rod designs, various enrichments and gadolina loadings, and axially varying lattice designs throughout the enriched portion of the bundle. The results of the benchmark present evidence that the CASMO-3G/MICROBURN-B code package can adequately model the range of fuel and core types in the benchmark, and the codes are acceptable for performing neutronic analyses of Commonwealth Edison's boiling water reactors

  18. Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3

    Energy Technology Data Exchange (ETDEWEB)

    Sudo, Y; Ando, H; Ikawa, H; Ohnishi, N [Department of Research Reactor Operation, Japan Atomic Energy Research Institute (JAERI), 319-11 Tokai-Mura, Ibaraki-Ken (Japan)

    1985-07-01

    This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)

  19. 3D printing in chemical engineering and catalytic technology: structured catalysts, mixers and reactors.

    Science.gov (United States)

    Parra-Cabrera, Cesar; Achille, Clement; Kuhn, Simon; Ameloot, Rob

    2018-01-02

    Computer-aided fabrication technologies combined with simulation and data processing approaches are changing our way of manufacturing and designing functional objects. Also in the field of catalytic technology and chemical engineering the impact of additive manufacturing, also referred to as 3D printing, is steadily increasing thanks to a rapidly decreasing equipment threshold. Although still in an early stage, the rapid and seamless transition between digital data and physical objects enabled by these fabrication tools will benefit both research and manufacture of reactors and structured catalysts. Additive manufacturing closes the gap between theory and experiment, by enabling accurate fabrication of geometries optimized through computational fluid dynamics and the experimental evaluation of their properties. This review highlights the research using 3D printing and computational modeling as digital tools for the design and fabrication of reactors and structured catalysts. The goal of this contribution is to stimulate interactions at the crossroads of chemistry and materials science on the one hand and digital fabrication and computational modeling on the other.

  20. Research program in reactor core diagnostics with neutron noise methods: Stage 3. Final report

    International Nuclear Information System (INIS)

    Pazsit, I.; Garis, N.S.; Karlsson, J.; Racz, A.

    1997-09-01

    Stage 3 of the program has been executed 96-04-12. The long term goal is to develop noise methods for identification and localization of perturbations in reactor cores. The main parts of the program consist of modelling the noise source, calculation of the space- and frequency dependent transfer function, calculation of the neutron noise via a convolution of the transfer function of the system and the noise source, i.e. the perturbation, and finally finding an inversion or unfolding procedure to determine noise source parameters from the neutron noise. Most previous work is based on very simple (analytical) reactor models for the calculation of the transfer function as well as analytical unfolding methods. The purpose of this project is to calculate the transfer function in a more realistic model as well as elaborating powerful inversion methods that do not require analytical transfer functions. The work in stage 3 is described under the following headlines: Further investigation of simplified models for the calculation of the neutron noise; Further investigation of methods based on neural networks; Further investigation of methods for detecting the vibrations and impacting of detectors; Application of static codes for determination of the neutron noise using the adiabatic approximation

  1. Collective occupational dose for nuclear reactors of the 2., 3. and 4. generation

    International Nuclear Information System (INIS)

    Guidez, J.; Saturnin, A.

    2016-01-01

    In France during reactor operation the individual occupational doses are collected and recorded according to the law. When you sum up all the individual doses you get the yearly collective dose expressed in Man.Sv/year. This piece of information can be used to make comparisons between various types of reactors and between reactors of the same type. The results show a steady decrease of the collective dose for all types of reactors over the time except for CANDU reactors for which a slight increase of the dose has appeared since the years 1996-1998. The decrease is due to the continuous improvement of reactor operating and to changes in the reactor design. There is also a constant gap over time between the collective dose for a BWR reactor (1.12 Man.Sv/y) and a PWR reactor 0.60 Man.Sv/y), this gap is certainly due to N 16 nuclide that is created in the primary circuit and transported to turbines in the case of a BWR reactor. For sodium-cooled fast reactors (RNR-Na) the collective dose is below 0.40 Man.Sv/y except for the BN-600 reactor. (A.C.)

  2. Examination of a failed reactor coolant pump rotating assembly from Crystal River Unit 3

    International Nuclear Information System (INIS)

    Hayner, G.O.; Lubnow, T.; Clary, M.

    1990-01-01

    On January 18, 1989, the A reactor coolant pump rotating assembly at the Crystal River Unit 3 Nuclear Power Plant failed during operation. A rotating assembly from this pump had previously failed in 1986. The reactor coolant pump was fabricated by Byron Jackson Pump Division of Borg-Warner Ind. Products, Inc. from UNS S66286 superalloy (Alloy A286). A root cause failure analysis examination was performed on the pump shaft and other components. The failure analysis included shaft vibrational mode and stress analyses, pump clearance and alignment analyses, and detailed destructive examination of the shaft and hydrostatic bearing assemblies. Based on the detailed physical examination of the shaft it was concluded that cracks initiated in the pump shaft at two sites approximately 180 0 apart in a band of shallow, thermally induced fatigue cracks. The cracks initiated at the bottom edge of the motor end shrink fit pad under the shrink fit sleeve supporting the hydrostatic bearing journal. The band of thermally induced fatigue cracks was apparently caused by mixing of cold seal injection water and hot reactor coolant in gaps between the pump shaft and sleeve. The motor end shrink fit was apparently not effective in preventing introduction of the seal injection water to this area. Initial crack propagation occurred by fatigue due to lateral vibration; however, the majority of crack propagation occurred by abnormal torsional fatigue loading induced by contact and sticking between the rotating and stationary portions of the hydrostatic bearing. Final fracture of the shaft occurred by torsional overload. Metallurgical characteristics and mechanical properties of the shaft were within design specification and probably did not significantly influence the cracking process

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

    International Nuclear Information System (INIS)

    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

  4. Empty pledges: A content analysis comparing Belgian and Dutch child-targeting food websites

    OpenAIRE

    Neyens, Evy; Smits, Tim

    2015-01-01

    Background: In the EU Pledge, the food industry has vowed to retain from unhealthy food promotion to children under the age of twelve. Nonetheless, food brands increasingly lure children to branded websites packed with unhealthy food and beverage advertising. This study first explores the prevalence of online marketing strategies on 49 Belgian and Dutch child-targeting food websites. Second, it examines the nutrient content of the advertised foods. Third, it scrutinizes whether Belgian and Du...

  5. 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.

  6. Maintenance Energy Requirements of Double-Muscled Belgian Blue Beef Cows

    OpenAIRE

    Fiems, Leo O.; De Boever, Johan L.; Vanacker, José M.; De Campeneere, Sam

    2015-01-01

    Simple Summary Double-muscled Belgian Blue animals are extremely lean, characterized by a deviant muscle fiber type with more fast-glycolytic fibers, compared to non-double-muscled animals. This fiber type may result in lower maintenance energy requirements. On the other hand, lean meat animals mostly have a higher rate of protein turnover, which requires more energy for maintenance. Therefore, maintenance requirements of Belgian Blue cows were investigated based on a zero body weight gain. T...

  7. Operating experiences and utilization programmes of the BAEC 3 MW TRIGA Mark-II research reactor of Bangladesh

    International Nuclear Information System (INIS)

    Haque, M.M.; Soner, M.A.M.; Saha, P.K.; Salam, M.A.; Zulquarnain, M.A.

    2008-01-01

    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 ( 131 I, 99m Tc, 46 Sc), various R and D activities, manpower training and education. The reactor has been operated successfully since its 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 that time, several modifications of the reactor cooling system along with its associated structures were also implemented and then necessary testing and commissioning of the newly installed component/equipment were carried out. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50g of TeO 2 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 ( 131 I). 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. The facility has so far been used to train up a total of 27 personnel including several foreign nationals to the level of Senior Reactor Operator (SRO) and Reactor Operator (RO). The

  8. Cooking up a culinary identity for Belgium. Gastrolinguistics in two Belgian cookbooks (19th century).

    Science.gov (United States)

    Parys, Nathalie

    2013-12-01

    The notion of cookbooks as socio-historic markers in a society is generally accepted within food studies. As both representations and prescriptions of food practices, perceived habits and attitudes towards food, they represent a certain identity for their readers. This paper investigates the nature of the identity that Belgian cookbooks constructed through their rhetoric. An important part of this study is to explore how and to what extent explicit reference to Belgium was made. To this end recipe titles/labels and recipe comments used in two leading bourgeois cookbooks from nineteenth-century Belgium were subjected to a quantitative and qualitative content analysis. The analysis showed that clear attention was paid to national culinary preferences. In terms of a domestic culinary corpus, it became apparent that both the Dutch and French editions of these cookbooks promoted dishes that were ascribed a Belgian origin. Internationality, however, was also an important building block of Belgian culinary identity. It was part of the desire of Belgian bourgeoisie to connect with an international elite. It fit into the 'search for sophistication', which was also expressed through the high representation of the more costly meats and sweet dishes. In addition, other references associated with bourgeois norms and values, such as family, convenience and frugality, were additional building blocks of Belgian culinary identity. Other issues such as tradition, innovation and health, were also matters of concerns to these Belgian cookbooks. Copyright © 2013 Elsevier Ltd. All rights reserved.

  9. Conclusions of the experts group of the RA reactor at the meeting held on November 2 and 3 1964 - Annex 12a

    International Nuclear Information System (INIS)

    Pavicevic, M.

    1964-01-01

    Conclusions of the experts group of the RA reactor are related to: analyses of reactor operation at 6.5 MW power with heavy water coolant flow of 250 m 3 /h (2 pumps rotation speed 1500 rotations/min); decisions of future operation; further preparation activities related to reactor operation in forced regime and reduced cooling conditions

  10. MCNP calculation of the critical H_3BO_3 concentrations for the first fuel loading into the reactor core of NPP MO-3-4 units

    International Nuclear Information System (INIS)

    Vrban, B.; Lueley, J.; Farkas, G.; Hascik, J.; Hinca, R.; Petriska, M.; Slugen, V.

    2012-01-01

    The purpose of the analysis was the determination of critical H_3BO_3 concentrations for the first fuel loading into the reactor core of MO34 units using 2"n"d generation fuel during the first start-up of new unit using calculation code MCNP 1.60. H_3BO_3 concentrations were computed for the given temperature of the primary circuit and position of the 6"t"h safety control rod group. Because of the very first start-up of these units, detailed analyses of active-core parameters are required by National Regulatory Authority and needed for safe operation of nuclear facility. (authors)

  11. Nuclear computerized library for assessing reactor reliability (NUCLARR): User's guide: Part 3, NUCLARR system description

    International Nuclear Information System (INIS)

    Gilmore, W.E.; Gentillon, C.D.; Gertman, D.I.; Beers, G.H.; Galyean, W.J.; Gilbert, B.G.

    1988-06-01

    The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is an automated data base management system for processing and storing human error probability and hardware component failure data. The NUCLARR system software resides on an IBM (or compatible) personal micro-computer. NUCLARR can be used by the end user to furnish data inputs for both human and hardware reliability analysis in support of a variety of risk assessment activities. The NUCLARR system is documented in a five-volume series of reports. Volume IV of this series is the User's Guide for operating the NUCLARR software and is presented in three parts. This document, Part 3: NUCLARR System Description, provides an in-depth discussion of the design characteristics and special features of the NUCLARR software. Part 3 also presents the organization of the data base structures and techniques used to manipulate the data

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

    International Nuclear Information System (INIS)

    Londen, S.O.

    1966-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Londen, S O

    1966-01-15

    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.

  14. 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)

    Ferri, M. G.; D'Auria, F.; Forasassi, G.; Giot, M.

    2000-01-01

    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)

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

    International Nuclear Information System (INIS)

    Huda, M.Q.; Chakrobortty, T.K.; Rahman, M.; Sarker, M.M.; Mahmood, M.S.

    2003-05-01

    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)

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

    International Nuclear Information System (INIS)

    Klein, M.; Demeulemeester, Y.; Ponnet, M.; Emond, M.; Emond, O.; Dadoumont, J.; Massaut, V.

    2000-01-01

    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)

  17. Spectral history modeling in the reactor dynamics code DYN3D

    International Nuclear Information System (INIS)

    Bilodid, Yurii

    2014-01-01

    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.

  18. Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

    International Nuclear Information System (INIS)

    Sekhri, Abdelkrim; Graham, Andy; D'Arcy, Alan; Oliver, Melissa

    2008-01-01

    The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

  19. Atomization of U3Si2/U3Si for research reactor fuel

    International Nuclear Information System (INIS)

    Kuk, Il Hiun

    2004-01-01

    Instead of comminuting, U 3 Si 2 /U 3 Si powders are produced by atomizating directly from the molten alloys. Many benefits are introduced by applying the atomization technique: reduction of the process, homogeneous alloy composition within a particle and between particles, increase of the thermal conductivity and decrease of the chemical reactivity with aluminium due to particle's spherical shape. (author)

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

    International Nuclear Information System (INIS)

    Price, M.

    2004-01-01

    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

  1. Contamination levels observed on the Belgian territory subsequent to the Chernobyl accident

    Energy Technology Data Exchange (ETDEWEB)

    Hoof, J van [State University of Ghent, Faculty of Veterinary Medicine, Gent (Belgium); Maghuin-Rogister, G [Universite de Liege, Brussels (Belgium)

    1986-07-01

    Contaminated air masses reached the Belgian territory from the South during the night of the first to the second of May. At this stage however the origin of this contamination was already identified through earlier observations over the Scandinavian area and the subsequent message about the reactor accident at the Chernobyl site. Later on radioactive clouds were also detected over the central part of Europe, demonstrating the persistent nature of the emissions from the damaged reactor. Consequently the influence on the Belgian territory was not unexpected. The authorities called on the SCK/CEN at Mol, and the IRE at Fleurus to assist the IHE at Brussels in collecting the necessary data for judging the radiological situation in our country. The KMI/IRM at Brussels was involved for the follow-up of meteorological conditions and analysis of the trajectories of contaminated air masses. Early detection possibilities for the arrival of contaminated air were provided by the continuous environmental monitoring apparatus for ambient {gamma}-dose rate or for {beta} activity of airborne dust, available at nuclear institutions and nuclear power plants. On detection of enhanced air radioactivity, the sampling period of routine air dust samplers was significantly shortened to allow for the hour to hour renewal of data for gross {beta} activity as a general indication of the evolution of the air contamination. {gamma}-spectrometric analysis of those filters provided the necessary data for the estimation of the dose equivalent due to inhalation. Ground deposition data at the location of the participating institutions were obtained by daily analysis of the radioactivity contents of a water container collecting both dust and rainwater. Field gamma spectrometry was used later on at a number of other locations, to estimate the integrated ground deposition of radioactivity and its distribution over the country. As the grazing season was just started or was about to be started in the

  2. Contamination levels observed on the Belgian territory subsequent to the Chernobyl accident

    International Nuclear Information System (INIS)

    Hoof, J. van; Maghuin-Rogister, G.

    1986-01-01

    Contaminated air masses reached the Belgian territory from the South during the night of the first to the second of May. At this stage however the origin of this contamination was already identified through earlier observations over the Scandinavian area and the subsequent message about the reactor accident at the Chernobyl site. Later on radioactive clouds were also detected over the central part of Europe, demonstrating the persistent nature of the emissions from the damaged reactor. Consequently the influence on the Belgian territory was not unexpected. The authorities called on the SCK/CEN at Mol, and the IRE at Fleurus to assist the IHE at Brussels in collecting the necessary data for judging the radiological situation in our country. The KMI/IRM at Brussels was involved for the follow-up of meteorological conditions and analysis of the trajectories of contaminated air masses. Early detection possibilities for the arrival of contaminated air were provided by the continuous environmental monitoring apparatus for ambient γ-dose rate or for Β activity of airborne dust, available at nuclear institutions and nuclear power plants. On detection of enhanced air radioactivity, the sampling period of routine air dust samplers was significantly shortened to allow for the hour to hour renewal of data for gross Β activity as a general indication of the evolution of the air contamination. γ-spectrometric analysis of those filters provided the necessary data for the estimation of the dose equivalent due to inhalation. Ground deposition data at the location of the participating institutions were obtained by daily analysis of the radioactivity contents of a water container collecting both dust and rainwater. Field gamma spectrometry was used later on at a number of other locations, to estimate the integrated ground deposition of radioactivity and its distribution over the country. As the grazing season was just started or was about to be started in the following days for

  3. Belgian coca-cola-related outbreak: intoxication, mass sociogenic illness, or both?

    Science.gov (United States)

    Gallay, A; Van Loock, F; Demarest, S; Van der Heyden, J; Jans, B; Van Oyen, H

    2002-01-15

    An epidemic of health complaints occurred in five Belgian schools in June 1999. A qualitative investigation described the scenario. The role of soft drinks was assessed by using a case-control study. Cases were students complaining of headache, dizziness, nausea, vomiting, abdominal pain, diarrhea, or trembling. Controls were students present at school on the day of the outbreak but not taken ill. An analysis was performed separately for school A, where the outbreak started, and was pooled for schools B-E. In school A, the attack rate (13.2%) was higher than in schools B-E (3.6%, relative risk = 3.6, 95% confidence interval (CI): 2.5, 5.3). Exclusive consumption of regular Coca-Cola (school A: odds ratio (OR) = 29.7, 95% CI: 1.32, 663.6; schools B-E: OR = 7.3, 95% CI: 2.9, 18.0) and low mental health score (school A: OR = 16.1, 95% CI: 1.3, 201.9; schools B-E: OR = 3.1, 95% CI: 1.5, 6.6) were independently associated with the illness. In schools B-E, consumption of Fanta, consumption of Coca-Cola light, and female gender were also associated with the illness. It seems reasonable to attribute the first cases of illness in school A to regular Coca-Cola consumption. However, mass sociogenic illness could explain the majority of the other cases.

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

    International Nuclear Information System (INIS)

    2007-01-01

    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

  5. New Statute of the Nuclear Energy Research Centre (CEN), copy of Act No. 1429, amended by Act No. 1145 and Act No. 5652, respectively published in the annexes of the Belgian Official Gazette of 28 March 1963, 3 March 1966 and 17 September 1970

    International Nuclear Information System (INIS)

    1963-01-01

    The Statute of the Belgian Nuclear Research Centre (CEN) lays down the objects of the Centre which consist of undertaking research on the applications of nuclear energy, promoting scientific and technical studies of such applications, carrying out advanced technological research on other topics in the country's scientific and technological programmes; and at the public authorities' request co-ordinating such research by all means available to it. The CEN is run by a Board of Directors selected by representatives of the nuclear industry, scientific and university circles as well as by the Government. The Statute also provides for the rules of procedure of the Board as well as for the Centre's budget and management under the responsibility of a Director General appointed by the Board. Finally, the CEN is declared as being an establishment created in the public interest, and any amendment to its Statute must obtain the prior approval of the King. (NEA) [fr

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

    International Nuclear Information System (INIS)

    Gurban, I.; Laaksoharju, M.; Made, B.; Ledoux, E.

    1999-12-01

    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

  7. The radiological impact of the Belgian phosphate industry

    Energy Technology Data Exchange (ETDEWEB)

    Vanmarcke, H.; Paridaens, J. [Belgian Nuclear Research Centre, SCK.CEN, Boeretang 200, 2400 Mol (Belgium)

    2006-07-01

    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

  8. Predicted irradiation behavior of U3O8-Al dispersion fuels for production reactor applications

    International Nuclear Information System (INIS)

    Cronenberg, A.W.; Rest, J.

    1990-01-01

    Candidate fuels for the new heavy-water production reactor include uranium/aluminum alloy and U 3 O 8 -Al dispersion fuels. The U 3 O 8 -Al dispersion fuel would make possible higher uranium loadings and would facilitate uranium recycle. Research efforts on U 3 O 8 -Al fuel include in-pile irradiation studies and development of analytical tools to characterize the behavior of dispersion fuels at high-burnup. In this paper the irradiation performance of U 3 O 8 -Al is assessed using the mechanistic Dispersion Analysis Research Tool (DART) code. Predictions of fuel swelling and alteration of thermal conductivity are presented and compared with experimental data. Calculational results indicate good agreement with available data where the effects of as-fabricated porosity and U 3 O 8 -Al oxygen exchange reactions are shown to exert a controlling influence on irradiation behavior. The DART code is judged to be a useful tool for assessing U 3 O 8 -Al performance over a wide range of irradiation conditions

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

    International Nuclear Information System (INIS)

    Ermutlu, H.E.

    1993-01-01

    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

  10. Study of the effect of slight variants to a 3-loop pressurized water nuclear reactor design in order to improve the reactor safety

    International Nuclear Information System (INIS)

    Castiglia, F.; Oliveri, E.; Taibi, S.; Vella, G.

    1992-01-01

    In order to improve the safety features of a 3-loop pressurized water nuclear reactor we propose a slight design variant consisting in the introduction of a bypass hole in the divider plate of the coolant chambers of the steam generators. The aim is to reduce both the extent and the duration of the core exposure and thus the maximum value of the peak cladding temperature, in case of a hypothetical cold leg small break loss of coolant accident. The proposal, as attested by a preliminary RELAP5/MOD3 analysis, seems to deserve some attention. (6 figures) (Author)

  11. Control of xenon oscillations in large PWRs result of control in Mihama Unit No.3 Reactor

    International Nuclear Information System (INIS)

    Shimazu, Yoichiro; Itahara, Kuniyuki; Nishimura, Takeshi.

    1977-01-01

    Recently, constant axial offset control method for power distribution control is applied to PWRs (of Westinghouse type). In this mode of operation, Xe induced spatial oscillations which could cause the problem from the power distribution point of view hardly break out. Under special conditions such as tests, however, Xe oscillations can break out. It is required that power distribution can be controlled without any special operational actions. Several control methods for Xe oscillations have been reported through many theoretical studies and operational experiences. Among them is First Overtone Control Method. This requires quite a simple control action but can effectively control Xe oscillations. The authors have shown the effectiveness of First Overtone Control Method in Mihama Unit No. 3 Reactor. In this report the detailed procedures are described. The outline of Constant Axial Offset Control Method is described in Appendix. (auth.)

  12. Once-through thorium fuel cycle evaluation for TVA's Browns Ferry-3 Boiling Water Reactor

    International Nuclear Information System (INIS)

    Hopkins, G.C.

    1982-05-01

    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 ThO 2 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/ThO 2 fuel reactivity, and, therefore, the results of the preliminary evaluations completed under the program should be conservative

  13. Alize 3 - first critical experiment for the franco-german high flux reactor - calculations

    International Nuclear Information System (INIS)

    Scharmer, K.

    1969-01-01

    The results of experiments in the light water cooled D 2 O 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 k eff 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 D 2 O 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) [fr

  14. 3-D seismic response of a base-isolated fast reactor

    International Nuclear Information System (INIS)

    Kitamura, S.; Morishita, M.; Iwata, K.

    1992-01-01

    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)

  15. 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.

  16. Finnish EPR Olkiluoto 3. The world's first third-generation reactor now under construction

    International Nuclear Information System (INIS)

    2007-01-01

    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 CO 2 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)

  17. Refurbishment, Modernization and Ageing Management Program of The 3MW TRIGA Mark-II Research Reactor of Bangladesh

    International Nuclear Information System (INIS)

    Salam, M. A.

    2013-01-01

    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

  18. 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

  19. New sparse matrix solver in the KIKO3D 3-dimensional reactor dynamics code

    International Nuclear Information System (INIS)

    Panka, I.; Kereszturi, A.; Hegedus, C.

    2005-01-01

    The goal of this paper is to present a more effective method Bi-CGSTAB for accelerating the large sparse matrix equation solution in the KIKO3D code. This equation system is obtained by using the factorization of the improved quasi static (IQS) method for the time dependent nodal kinetic equations. In the old methodology standard large sparse matrix techniques were considered, where Gauss-Seidel preconditioning and a GMRES-type solver were applied. The validation of KIKO3D using Bi-CGSTAB has been performed by solving of a VVER-1000 kinetic benchmark problem. Additionally, the convergence characteristics were investigated in given macro time steps of Control Rod Ejection transients. The results have been obtained by the old GMRES and new Bi-CGSTAB methods are compared. (author)

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

    International Nuclear Information System (INIS)

    Peacock, H.B.

    1983-08-01

    Heating tests using 53 wt % U 3 O 8 -Al pellets show that an exothermic reaction occurs between 875 and 1000 0 C and takes 10 to 20 seconds to reach maximum temperature. The maximum temperature is a function of particle size of the U 3 O 8 with large particles exhibiting lower peak temperatures. The calculated energy release was 123 cal/g of U 3 O 8 -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 U 3 O 8 -Al reaction is not an important energy source. The compressive and tensile strengths of U 3 O 8 tubes above 660 0 C are low. In compression, sections with 2 psi average axial stress failed at 917 0 C, while sections with 7 psi failed at 669 0 C. Tubes with U-Al alloy cores failed at about 670 0 C 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

  1. Development of a 3D-Multigroup program to simulate anomalous diffusion phenomena in the nuclear reactors

    International Nuclear Information System (INIS)

    Maleki Moghaddam, Nader; Afarideh, Hossein; Espinosa-Paredes, Gilberto

    2015-01-01

    Highlights: • The new version of neutron diffusion equation for simulating anomalous diffusion is presented. • Application of fractional calculus in the nuclear reactor is revealed. • A 3D-Multigroup program is developed based on the fractional operators. • The super-diffusion and sub-diffusion phenomena are modeled in the nuclear reactors core. - Abstract: The diffusion process is categorized in three parts, normal diffusion, super-diffusion and sub-diffusion. The classical neutron diffusion equation is used to model normal diffusion. A new scheme of derivatives is required to model anomalous diffusion phenomena. The fractional space derivatives are employed to model anomalous diffusion processes where a plume of particles spreads at an inconsistent rate with the classical Brownian motion model. In the fractional diffusion equation, the fractional Laplacians are used; therefore the statistical jump length of neutrons is unrestricted. It is clear that the fractional Laplacians are capable to model the anomalous phenomena in nuclear reactors. We have developed a NFDE-3D (neutron fractional diffusion equation) as a core calculation code to model normal and anomalous diffusion phenomena. The NFDE-3D is validated against the LMW-LWR reactor. The results demonstrate that reactors exhibit complex behavior versus order of the fractional derivatives which depends on the competition between neutron absorption and super-diffusion phenomenon

  2. Modelling and thermal hydraulic analysis of the Angra-2 nuclear reactor using RELAP5-3D code

    International Nuclear Information System (INIS)

    González Mantecón, Javier

    2015-01-01

    The evaluation of Nuclear Power Plants (NPPs) performance during steady-state and accident conditions has been one of the main research subjects in the nuclear field. In order to simulate the behavior of water-cooled reactors, several complex thermal-hydraulic codes systems have been developed. Particularly, the RELAP5 code, developed by the Idaho National Laboratory, is a best-estimate thermal-hydraulic analysis tool and one of the most used in nuclear industry. The RELAP5-3D 3.0.0 code was used to develop a detailed model of Angra 2 nuclear reactor using reference data from the Final Safety Analysis Report. Angra 2 is the second Brazilian NPP, which began commercial operation in 2001. The plant is equipped with a Pressurized Water Reactor (PWR) type with 3771.0 MWt. Simulations of the reactor behavior during normal operation conditions and postulated accident conditions were performed. Results achieved in the reactor steady-state simulation were compared with nominal parameters of the NPP. These results proved to be in good agreement, with relative errors less than 1%. In the transient simulation, the obtained results were coherent and satisfactory. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal-hydraulic behavior of the Angra-2 PWR during diverse operation conditions and it can contribute for the process of the plant safety analysis. (author)

  3. Variability of patient safety culture in Belgian acute hospitals.

    Science.gov (United States)

    Vlayen, Annemie; Schrooten, Ward; Wami, Welcome; Aerts, Marc; Barrado, Leandro Garcia; Claes, Neree; Hellings, Johan

    2015-06-01

    The aim of this study was to measure differences in safety culture perceptions within Belgian acute hospitals and to examine variability based on language, work area, staff position, and work experience. The Hospital Survey on Patient Safety Culture was distributed to hospitals participating in the national quality and safety program (2007-2009). Hospitals were invited to participate in a comparative study. Data of 47,136 respondents from 89 acute hospitals were used for quantitative analysis. Percentages of positive response were calculated on 12 dimensions. Generalized estimating equations models were fitted to explore differences in safety culture. Handoffs and transitions, staffing, and management support for patient safety were considered as major problem areas. Dutch-speaking hospitals had higher odds of positive perceptions for most dimensions in comparison with French-speaking hospitals. Safety culture scores were more positive for respondents working in pediatrics, psychiatry, and rehabilitation compared with the emergency department, operating theater, and multiple hospital units. We found an important gap in safety culture perceptions between leaders and assistants within disciplines. Administration and middle management had lower perceptions toward patient safety. Respondents working less than 1 year in the current hospital had more positive safety culture perceptions in comparison with all other respondents. Large comparative databases provide the opportunity to identify distinct high and low scoring groups. In our study, language, work area, and profession were identified as important safety culture predictors. Years of experience in the hospital had only a small effect on safety culture perceptions.

  4. Estimation of furan contamination across the Belgian food chain.

    Science.gov (United States)

    Scholl, G; Scippo, M-L; De Pauw, E; Eppe, G; Saegerman, C

    2012-01-01

    This paper provides an estimate of the furan content of Belgian foods. The objective of the study was to achieve the best food chain coverage with a restricted number of samples (n = 496). The geographic distribution, different market chains and labels, and consumption frequencies were taken into account in the construction of the sampling plan. Weighting factors such as contamination levels, consumption frequency and the diversity of food items were applied to set up the model. The very low detection capabilities (CC(β)) of the analytical methods used (sub-ppb) allowed reporting of 78.2% of the overall dataset above CC(β) and, in particular, 96.7% for the baby food category. The highest furan levels were found in powdered roasted bean coffee (1912 µg kg(-1)) with a mean of 756 µg kg(-1) for this category. Prepared meat, pasta and rice, breakfast cereals, soups, and baby food also showed high mean furan contents ranging from 16 to 43 µg kg(-1). Comparisons with contamination surveys carried out in other countries pointed out differences for the same food group and therefore contamination levels are related to the geographical origin of food items.

  5. 3D modeling of the primary circuit in the reactor pressure vessel of a PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramajo, Damian, E-mail: dramajo@santafe-conicet.gov.ar; Corzo, Santiago; Schiliuk, Nicolas; Nigro, Norberto

    2013-12-15

    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

  6. Development of a 3D consistent 1D neutronics model for reactor core simulation

    International Nuclear Information System (INIS)

    Lee, Ki Bog; Joo, Han Gyu; Cho, Byung Oh; Zee, Sung Quun

    2001-02-01

    In this report a 3D consistent 1D model based on nonlinear analytic nodal method is developed to reproduce the 3D results. During the derivation, the current conservation factor (CCF) is introduced which guarantees the same axial neutron currents obtained from the 1D equation as the 3D reference values. Furthermore in order to properly use 1D group constants, a new 1D group constants representation scheme employing tables for the fuel temperature, moderator density and boron concentration is developed and functionalized for the control rod tip position. To test the 1D kinetics model with CCF, several steady state and transient calculations were performed and compared with 3D reference values. The errors of K-eff values were reduced about one tenth when using CCF without significant computational overhead. And the errors of power distribution were decreased to the range of one fifth or tenth at steady state calculation. The 1D kinetics model with CCF and the 1D group constant functionalization employing tables as a function of control rod tip position can provide preciser results at the steady state and transient calculation. Thus it is expected that the 1D kinetics model derived in this report can be used in the safety analysis, reactor real time simulation coupled with system analysis code, operator support system etc.

  7. RELAP5-3D version 4.0.3: installation and tests for applications to space reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lobo, Paulo D.C.; Braz Filho, Francisco A.; Borges, Eduardo M.; Guimaraes, Lamartine N.F., E-mail: plobo.a@uol.com.br, E-mail: fbraz@ieav.cta.br, E-mail: eduardo@ieav.cta.br, E-mail: guimarae@ieav.cta.br [Instituto de Estudos Avancados (IEAv), Sao Jose dos Campos, SP (Brazil); Sabundjian, Gaiane, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    To attend the TERRA project (Tecnologia de Reatores Rapidos Avancados), currently conducted by the Nuclear Energy Division (ENU) of the IEAv, this work presents the RELAP5-3D, Version 4.0.3, prepared in July 12, 2012, also known as r3d403is, received recently by the IEAv from the Idaho National Laboratory (INL). This version of RELAP5-3D is configured for the International User Group source Code Group and is developed and maintained at the INL for the US Department of Energy. RELAP5-3D, the latest in the series of RELAP5 codes, is a highly generic code that, in addition to calculating the behavior of a reactor coolant system during a transient, can be used for simulation of a wide variety of hydraulic and thermal transients in both nuclear and nonnuclear systems involving mixtures of vapor, liquid, noncondensable gases, and nonvolatile solute. Enhancements include all features and models previously available in the ATHENA configuration version of the code which are as follows: addition of new work fluids and a magneto-hydrodynamic mode. Following the instructions from the README file, the RELAP5-3D, version 4.0.3 was installed creating the necessaries subdirectories, by using the LINUX platform and applying both Intel Fortran 95 and C-language compilers. Many input examples were executed and the same results were observed as compared to the received documentation. A sample of the Edwards-O'Brien test was evaluated to verify if the code could simulate a LOCA type accident properly. The test executed by the RELAP5-3D demonstrated good agreement with test data including a new output involving the mass flow during the test. (author)

  8. RELAP5-3D version 4.0.3: installation and tests for applications to space reactors

    International Nuclear Information System (INIS)

    Lobo, Paulo D.C.; Braz Filho, Francisco A.; Borges, Eduardo M.; Guimaraes, Lamartine N.F.; Sabundjian, Gaiane

    2013-01-01

    To attend the TERRA project (Tecnologia de Reatores Rapidos Avancados), currently conducted by the Nuclear Energy Division (ENU) of the IEAv, this work presents the RELAP5-3D, Version 4.0.3, prepared in July 12, 2012, also known as r3d403is, received recently by the IEAv from the Idaho National Laboratory (INL). This version of RELAP5-3D is configured for the International User Group source Code Group and is developed and maintained at the INL for the US Department of Energy. RELAP5-3D, the latest in the series of RELAP5 codes, is a highly generic code that, in addition to calculating the behavior of a reactor coolant system during a transient, can be used for simulation of a wide variety of hydraulic and thermal transients in both nuclear and nonnuclear systems involving mixtures of vapor, liquid, noncondensable gases, and nonvolatile solute. Enhancements include all features and models previously available in the ATHENA configuration version of the code which are as follows: addition of new work fluids and a magneto-hydrodynamic mode. Following the instructions from the README file, the RELAP5-3D, version 4.0.3 was installed creating the necessaries subdirectories, by using the LINUX platform and applying both Intel Fortran 95 and C-language compilers. Many input examples were executed and the same results were observed as compared to the received documentation. A sample of the Edwards-O'Brien test was evaluated to verify if the code could simulate a LOCA type accident properly. The test executed by the RELAP5-3D demonstrated good agreement with test data including a new output involving the mass flow during the test. (author)

  9. Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2007

    International Nuclear Information System (INIS)

    2012-03-01

    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)

  10. Deployment of Smart 3D Subsurface Contaminant Characterization at the Brookhaven Graphite Research Reactor

    International Nuclear Information System (INIS)

    Sullivan, T.; Heiser, J.; Kalb, P.; Milian, L.; Newson, C.; Lilimpakas, M.; Daniels, T.

    2002-01-01

    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

  11. Transportable nuclear power plant T3C-M with two reactor plants of improved safety

    International Nuclear Information System (INIS)

    Ogloblin, B.G.; Gromov, B.F.

    1993-01-01

    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

  12. Evaluation on activation activity of reactor in JRR-2 applied 3 dimensional model to neutron flux calculation

    International Nuclear Information System (INIS)

    Kishimoto, Katsumi; Arigane, Kenji

    2005-03-01

    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 1.9x10 14 Bq. (author)

  13. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR)

    Science.gov (United States)

    Brovchenko, Mariya; Dechenaux, Benjamin; Burn, Kenneth W.; Console Camprini, Patrizio; Duhamel, Isabelle; Peron, Arthur

    2017-09-01

    The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR). The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  14. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR

    Directory of Open Access Journals (Sweden)

    Brovchenko Mariya

    2017-01-01

    Full Text Available The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR. The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  15. Characteristics of Flameless Combustion in 3D Highly Porous Reactors under Diesel Injection Conditions

    Directory of Open Access Journals (Sweden)

    M. Weclas

    2013-01-01

    Full Text Available The heat release process in a free volume combustion chamber and in porous reactors has been analyzed under Diesel engine-like conditions. The process has been investigated in a wide range of initial pressures and temperatures simulating engine conditions at the moment when fuel injection starts. The resulting pressure history in both porous reactors and in free volumes significantly depends on the initial pressure and temperature. At lower initial temperatures, the process in porous reactors is accelerated. Combustion in a porous reactor is characterized by heat accumulation in the solid phase of the porous structure and results in reduced pressure peaks and lowered combustion temperature. This depends on reactor heat capacity, pore density, specific surface area, pore structure, and heat transport properties. Characteristic modes of a heat release process in a two-dimensional field of initial pressure and temperature have been selected. There are three characteristic regions represented by a single- and multistep oxidation process (with two or three slopes in the reaction curve and characteristic delay time distribution has been selected in five characteristic ranges. There is a clear qualitative similarity of characteristic modes of the heat release process in a free volume and in porous reactors. A quantitative influence of porous reactor features (heat capacity, pore density, pore structure, specific surface area, and fuel distribution in the reactor volume has been clearly indicated.

  16. Power measurement of the RA-3 reactor using the neutron noise technique and {sup 16}N; Medicion de la potencia del reactor RA-3, mediante la tecnica de ruido neutronico y nitrogeno 16

    Energy Technology Data Exchange (ETDEWEB)

    Gomez, Angel [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. Reactores y Centrales Nucleares

    2003-07-01

    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 {sup 16}N 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 {sup 16}N measuring system, an estimator, that is a function of the ratio of this currents, is proposed in order to detect flow changes. (author)

  17. Analysis of a possible experimental assessment of a prototype fuel element containing burnable poison in the RA-3 reactor

    International Nuclear Information System (INIS)

    Lerner, Ana Maria; Madariaga, Marcelo

    2002-01-01

    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)

  18. CESAR5.3: Isotopic depletion for Research and Testing Reactor decommissioning

    Science.gov (United States)

    Ritter, Guillaume; Eschbach, Romain; Girieud, Richard; Soulard, Maxime

    2018-05-01

    CESAR stands in French for "simplified depletion applied to reprocessing". The current version is now number 5.3 as it started 30 years ago from a long lasting cooperation with ORANO, co-owner of the code with CEA. This computer code can characterize several types of nuclear fuel assemblies, from the most regular PWR power plants to the most unexpected gas cooled and graphite moderated old timer research facility. Each type of fuel can also include numerous ranges of compositions like UOX, MOX, LEU or HEU. Such versatility comes from a broad catalog of cross section libraries, each corresponding to a specific reactor and fuel matrix design. CESAR goes beyond fuel characterization and can also provide an evaluation of structural materials activation. The cross-sections libraries are generated using the most refined assembly or core level transport code calculation schemes (CEA APOLLO2 or ERANOS), based on the European JEFF3.1.1 nuclear data base. Each new CESAR self shielded cross section library benefits all most recent CEA recommendations as for deterministic physics options. Resulting cross sections are organized as a function of burn up and initial fuel enrichment which allows to condensate this costly process into a series of Legendre polynomials. The final outcome is a fast, accurate and compact CESAR cross section library. Each library is fully validated, against a stochastic transport code (CEA TRIPOLI 4) if needed and against a reference depletion code (CEA DARWIN). Using CESAR does not require any of the neutron physics expertise implemented into cross section libraries generation. It is based on top quality nuclear data (JEFF3.1.1 for ˜400 isotopes) and includes up to date Bateman equation solving algorithms. However, defining a CESAR computation case can be very straightforward. Most results are only 3 steps away from any beginner's ambition: Initial composition, in core depletion and pool decay scenario. On top of a simple utilization architecture

  19. 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)

    Adorni, M.; Salah, A.B.; Di Maro, B.; Pierro, F.; D'Auria, F.; Hamidouche, T.

    2004-01-01

    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

  20. Evaluation of the trial design studies for an advanced marine reactor, (3)

    International Nuclear Information System (INIS)

    Ambo, Noriaki; Yokomura, Takeyoshi.

    1988-03-01

    JAERI have carried out four core designs for three different type reactors (Semi-Integrated, Integrated and Integrated (self-pressured) type reactors), as the trial designs of an Advanced Marine Reactor for three years (1983 ∼ 1985). This report describes the result of comparison and studies of the core specific characteristics of these four cores, which include core concept, specifications, core life, specific power density, burn-up, reactivity control and etc. In conclusion, it was found that the Integrated type reactor core and the Semi-Integrated type reactor core designs satisfy the conditions of long core life (four years), high specific power density (50 ∼ 61 kw/l) and high burn-up (30,000 ∼ 32,000 MWD/t), so these two cores will be optimum designs based on the present technologies. (author)

  1. Fully 3D printed integrated reactor array for point-of-care molecular diagnostics.

    Science.gov (United States)

    Kadimisetty, Karteek; Song, Jinzhao; Doto, Aoife M; Hwang, Young; Peng, Jing; Mauk, Michael G; Bushman, Frederic D; Gross, Robert; Jarvis, Joseph N; Liu, Changchun

    2018-06-30

    Molecular diagnostics that involve nucleic acid amplification tests (NAATs) are crucial for prevention and treatment of infectious diseases. In this study, we developed a simple, inexpensive, disposable, fully 3D printed microfluidic reactor array that is capable of carrying out extraction, concentration and isothermal amplification of nucleic acids in variety of body fluids. The method allows rapid molecular diagnostic tests for infectious diseases at point of care. A simple leak-proof polymerization strategy was developed to integrate flow-through nucleic acid isolation membranes into microfluidic devices, yielding a multifunctional diagnostic platform. Static coating technology was adopted to improve the biocompatibility of our 3D printed device. We demonstrated the suitability of our device for both end-point colorimetric qualitative detection and real-time fluorescence quantitative detection. We applied our diagnostic device to detection of Plasmodium falciparum in plasma samples and Neisseria meningitides in cerebrospinal fluid (CSF) samples by loop-mediated, isothermal amplification (LAMP) within 50 min. The detection limits were 100 fg for P. falciparum and 50 colony-forming unit (CFU) for N. meningitidis per reaction, which are comparable to that of benchtop instruments. This rapid and inexpensive 3D printed device has great potential for point-of-care molecular diagnosis of infectious disease in resource-limited settings. Copyright © 2018 Elsevier B.V. All rights reserved.

  2. Safety Design Strategy for the Advanced Test Reactor Diesel Bus (E-3) and Switchgear Replacement Project

    International Nuclear Information System (INIS)

    Duckwitz, Noel

    2011-01-01

    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.

  3. Characteristics of biohydrogen production by ethanoligenens R{sub 3} isolated from continuous stirred tank reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jiao, A.Y.; Liu, K. [Northeast Forestry Univ., Harbin (China). School of Forestry; Li, Y.F. [Northeast Forestry Univ., Harbin (China). School of Forestry; Shanghai Univ. of Engineering Science (China). College of Chemistry and Chemical Engineering; Liu, B. [Northeast Forestry Univ., Harbin (China). School of Material Science and Engineering; Xu, J.L. [Shanghai Univ. of Engineering Science (China). College of Chemistry and Chemical Engineering

    2010-07-01

    This study investigated the fermentative hydrogen production characteristics of ethanoligenens R{sub 3} isolated from anaerobic sludge in a continuous stirred tank reactor. The effects of the initial pH value, the proportion of carbon and nitrogen sources, and the effects of fermentation temperature were investigated in a series of batch experiments. Substrates for the hydrogen production of glucose and peptone were used as carbon and nitrogen sources. Results of the experiments showed that a maximum hydrogen production yield of 834 mlH{sub 2}/L culture was obtained with a fermentation temperature of 35 degrees C and an initial pH value of 5.5. The maximum average hydrogen production rate of 10.87 mmolH{sub 2}/g cell dry weight per hour was obtained at a carbon-nitrogen source ratio of 3.3. The degradation efficiency of the glucose used as a carbon source ranged from 91.5 to 95.43 per cent during the conversion of glucose to hydrogen by the bacteria.

  4. The usefulness of Belgian formulae in third molar-based age assessment of Indians.

    Science.gov (United States)

    Bhowmik, Biyas; Acharya, Ashith B; Naikmasur, Venkatesh G

    2013-03-10

    The third molars are one of few useful predictors for assessing the degree of maturity in adolescence and young adulthood. It has application in age estimation in the age group of 14-23 years, in general, and in juvenile/adult status prediction, in particular. Using a 10-stage grading of third molars, Gunst et al. developed regression formulae on a large sample of Belgians (n=2513) for estimating age. Their research has been recommended as a 'reference study' in age estimation guidelines. The present study has ventured to determine if estimating age in Indians using the Belgian formulae produced results comparable to those reported in the Belgian study; in addition, this study attempts to determine if the same formulae predicted juvenile/adult status (age aged between 14 and 23 years. The OPGs included a mix of one, two, three and four third molars. In total, 916 teeth were assessed using the same 10-stage grading. Age in each OPG was estimated by applying the relevant Belgian regression formulae (regression formulae are available for one, two, three and four third molars). To determine if the formulae produced age estimates comparable to those in the Belgian study, the percentage of Indian subjects whose actual age fell within the 68% confidence interval (CI) (calculated from the ± 1 S.D. value available for each Belgian formula) was ascertained. If ≥ 68% of Indian subjects' age fell inside this interval, it indicates that the Belgian formulae are applicable in Indians. To assess the suitability of the Belgian formulae in predicting juvenile/adult status in Indians, the accuracy of the age estimation per se was not considered, rather, the number of correct age predictions only was noted. Overall, ≈ 74% of Indian subjects' actual age fell within the 68% CI; with regards to the Belgian formulae being able to correctly predict juvenile/adult status, 78% of all subjects were categorized to the correct age group (age estimation per se of Indians; however, the

  5. Prevalence and correlates of poor sleep quality and daytime sleepiness in Belgian truck drivers.

    Science.gov (United States)

    Braeckman, Lutgart; Verpraet, Rini; Van Risseghem, Marleen; Pevernagie, Dirk; De Bacquer, Dirk

    2011-03-01

    Sleepiness and sleep complaints are common among professional drivers. Sleepiness is a considerable problem not only because it affects the drivers' well-being, but also because of the consequences for performance and safety. Assessment of the (self-reported) prevalence and research into the risk factors are thus an important health issue and are also indispensable to prevent productivity loss and work-related accidents and injuries. Therefore, the aim of this study was to describe sleeping, driving, and health characteristics of Belgian truck drivers and to determine occupational and individual factors associated with poor sleep quality and daytime sleepiness. Cross-sectional data were collected using a self-administered questionnaire that included the Pittsburgh Sleep Quality Index (PSQI), Epworth Sleepiness Scale (ESS), and Berlin Questionnaire (BQ). The mean (SD) age of the 476 studied truck drivers was 42.7 (10.2) yrs and the mean (SD) body mass index was 27.3 (5.1) kg/m(2). Approximately 47% declared that they drove >50 h/wk and found their work schedule unrealistic. The mean (SD) PSQI score was 4.45 (2.7); poor quality of sleep (PSQI >5) was found in 27.2%. The mean (SD) ESS score was 6.79 (4.17); 18% had a score >10. The BQ indicated that 21.5% had a higher risk on obstructive sleep apnea. In multiple logistic regression analysis, low educational level (odds ratio [OR] 1.86), current smoking (OR 1.75), unrealistic work schedule (OR 1.75), and risk for obstructive sleep apnea (OR 2.97) were found to be independent correlates of daytime sleepiness. Poor sleep quality was significantly associated with poor self-perceived health (OR 1.95), unrealistic work schedule (OR 2.85), low job satisfaction (OR 1.91), and less driving experience (OR 1.73). These results show that poor sleep quality and daytime sleepiness were prevalent in Belgian truck drivers. Taking into account that several significant correlates with respect to these sleep problems were identified

  6. A packed bed membrane reactor for the oxidative dehydrogenation of propane on a Ga2O3 / MoO3 based catalyst

    NARCIS (Netherlands)

    Kotanjac, Ž.S.; Sint Annaland, van M.; Kuipers, J.A.M.

    2010-01-01

    Oxidative dehydrogenation of propane has been studied over a Ga2O3/MoO3 based catalyst. Using a differentially operated packed bed reactor with premixed oxygen and propane feed, the kinetic parameters for the main reaction and the consecutive and parallel reactions were experimentally determined. It

  7. NEK1 genetic variability in a Belgian cohort of ALS and ALS-FTD patients.

    Science.gov (United States)

    Nguyen, Hung Phuoc; Van Mossevelde, Sara; Dillen, Lubina; De Bleecker, Jan L; Moisse, Matthieu; Van Damme, Philip; Van Broeckhoven, Christine; van der Zee, Julie

    2018-01-01

    We evaluated the genetic impact of the amyotrophic lateral sclerosis (ALS) risk gene never in mitosis gene a-related kinase 1 (NEK1) in a Belgian cohort of 278 patients with ALS (n = 245) or ALS with frontotemporal dementia (ALS-FTD, n = 33) and 609 control individuals. We identified 2 ALS patients carrying a loss-of-function (LOF) mutation, p.Leu854Tyrfs*2 and p.Tyr871Valfs*17, that was absent in the control group. A third LOF variant p.Ser1036* was present in 2 sibs with familial ALS but also in an unrelated control person. Missense variants were common in both patients (3.6%) and controls (3.0%). The missense variant, p.Arg261His, which was previously associated with ALS risk, was detected with a minor allele frequency of 0.90% in patients compared to 0.33% in controls. Taken together, NEK1 LOF variants accounted for 1.1% of patients, although interpretation of pathogenicity and penetrance is complicated by the observation of occasional LOF variants in unaffected individuals (0.16%). Furthermore, enrichment of additional ALS gene mutations was observed in NEK1 carriers, suggestive of a "second hit" model were NEK1 variants may modify disease presentation of driving mutations. Copyright © 2017 The Author(s). Published by Elsevier Inc. All rights reserved.

  8. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1

    International Nuclear Information System (INIS)

    Reyes F, M. C.; Del Valle G, E.; Filio L, C.

    2013-10-01

    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 2 and P 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)

  9. Effectiveness and tolerability of second-line therapy with vildagliptin versus other oral agents in type 2 diabetes (EDGE): post-hoc subanalysis of the Belgian data.

    Science.gov (United States)

    Hoste, J; Daci, E; Mathieu, C

    2014-06-01

    To assess the efficacy and safety of vildagliptin versus other oral glucose-lowering drugs added to antidiabetic monotherapy in Belgian patients with type 2 diabetes mellitus, in comparison to the global EDGE study results. This is a pre-specified post-hoc subanalysis of the Belgian patient cohort from a worldwide 1-year observational study that compared the effectiveness and tolerability of vildagliptin to other oral antidiabetic agents in type 2 diabetes patients failing monotherapy with oral glucose-lowering agents (EDGE). A total of 1793 Belgian patients were enrolled. Physicians could add any oral antidiabetic drug and patients entered either into the vildagliptin or the comparator cohort. The primary effectiveness and tolerability endpoint was defined as the proportion of patients having a treatment response (HbA1c reduction from baseline to month 12 endpoint >0·3%) without hypoglycemia, weight gain, peripheral oedema, or gastrointestinal side-effects. In the Belgian population, 37·8% of patients in the vildagliptin group and 32·8% in the comparator group had a decrease in HbA1c of >0·3% without the predefined tolerability issues of hypoglycemia, weight gain, oedema or, gastrointestinal complaints (primary endpoint), resulting in an unadjusted odds ratio of 1·24 (95% CI: 0·96-1·61). Mean HbA1c change from baseline was -0·81% in the vildagliptin cohort and -0·75% in the comparator cohort. Overall, vildagliptin was well tolerated with similarly low incidences of total adverse events (14·9% versus 14·5% in the compactor group) and serious adverse events (2·7% versus 2·5% in the comparator group). In this EDGE subgroup of Belgian patients with type 2 diabetes who do not achieve the glycemic targets with monotherapy, a similar trend as in the global EDGE study was observed. Adding vildagliptin as a second oral glucose-lowering agent resulted in lowering HbA1c to <7% without weight gain, hypoglycemia or peripheral oedema in a higher proportion of

  10. Mutations in Splicing Factor Genes Are a Major Cause of Autosomal Dominant Retinitis Pigmentosa in Belgian Families

    Science.gov (United States)

    Coppieters, Frauke; Roels, Dimitri; De Jaegere, Sarah; Flipts, Helena; De Zaeytijd, Julie; Walraedt, Sophie; Claes, Charlotte; Fransen, Erik; Van Camp, Guy; Depasse, Fanny; Casteels, Ingele; de Ravel, Thomy

    2017-01-01

    Purpose Autosomal dominant retinitis pigmentosa (adRP) is characterized by an extensive genetic heterogeneity, implicating 27 genes, which account for 50 to 70% of cases. Here 86 Belgian probands with possible adRP underwent genetic testing to unravel the molecular basis and to assess the contribution of the genes underlying their condition. Methods Mutation detection methods evolved over the past ten years, including mutation specific methods (APEX chip analysis), linkage analysis, gene panel analysis (Sanger sequencing, targeted next-generation sequencing or whole exome sequencing), high-resolution copy number screening (customized microarray-based comparative genomic hybridization). Identified variants were classified following American College of Medical Genetics and Genomics (ACMG) recommendations. Results Molecular genetic screening revealed mutations in 48/86 cases (56%). In total, 17 novel pathogenic mutations were identified: four missense mutations in RHO, five frameshift mutations in RP1, six mutations in genes encoding spliceosome components (SNRNP200, PRPF8, and PRPF31), one frameshift mutation in PRPH2, and one frameshift mutation in TOPORS. The proportion of RHO mutations in our cohort (14%) is higher than reported in a French adRP population (10.3%), but lower than reported elsewhere (16.5–30%). The prevalence of RP1 mutations (10.5%) is comparable to other populations (3.5%-10%). The mutation frequency in genes encoding splicing factors is unexpectedly high (altogether 19.8%), with PRPF31 the second most prevalent mutated gene (10.5%). PRPH2 mutations were found in 4.7% of the Belgian cohort. Two families (2.3%) have the recurrent NR2E3 mutation p.(Gly56Arg). The prevalence of the recurrent PROM1 mutation p.(Arg373Cys) was higher than anticipated (3.5%). Conclusions Overall, we identified mutations in 48 of 86 Belgian adRP cases (56%), with the highest prevalence in RHO (14%), RP1 (10.5%) and PRPF31 (10.5%). Finally, we expanded the molecular

  11. Efficient H2O2/CH3COOH oxidative desulfurization/denitrification of liquid fuels in sonochemical flow-reactors.

    Science.gov (United States)

    Calcio Gaudino, Emanuela; Carnaroglio, Diego; Boffa, Luisa; Cravotto, Giancarlo; Moreira, Elizabeth M; Nunes, Matheus A G; Dressler, Valderi L; Flores, Erico M M

    2014-01-01

    The oxidative desulfurization/denitrification of liquid fuels has been widely investigated as an alternative or complement to common catalytic hydrorefining. In this process, all oxidation reactions occur in the heterogeneous phase (the oil and the polar phase containing the oxidant) and therefore the optimization of mass and heat transfer is of crucial importance to enhancing the oxidation rate. This goal can be achieved by performing the reaction in suitable ultrasound (US) reactors. In fact, flow and loop US reactors stand out above classic batch US reactors thanks to their greater efficiency and flexibility as well as lower energy consumption. This paper describes an efficient sonochemical oxidation with H2O2/CH3COOH at flow rates ranging from 60 to 800 ml/min of both a model compound, dibenzotiophene (DBT), and of a mild hydro-treated diesel feedstock. Four different commercially available US loop reactors (single and multi-probe) were tested, two of which were developed in the authors' laboratory. Full DBT oxidation and efficient diesel feedstock desulfurization/denitrification were observed after the separation of the polar oxidized S/N-containing compounds (S≤5 ppmw, N≤1 ppmw). Our studies confirm that high-throughput US applications benefit greatly from flow-reactors. Copyright © 2013 Elsevier B.V. All rights reserved.

  12. Review of in-service thermal annealing of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Server, W.L.

    1984-01-01

    Radiation embrittlement of ferritic pressure vessel steels increases the ductile-brittle transition temperature and decreases the upper-shelf level of toughness as measured by Charpy impact tests. A thermal anneal cycle well above the normal operating temperature of the vessel can restore most of the original Charpy V-notch energy properties. A test reactor pressure vessel has been wet annealed at less than 343 0 C (650 0 F), and annealing of the Belgian BR-3 reactor vessel has recently taken place. An industry survey indicates that dry annealing a reactor vessel in-place is feasible, but solvable engineering problems do exist. The materials with highest radiation sensitivity in the older reactor vessels are submerged-arc weld metals with high copper and nickel concentrations. The limited Charpy V-notch and fracture toughness data available for five such welds were reviewed. The review suggested that significant recovery results from annealing at 454 0 C (850 0 F) for one week. Two of the main concerns with a localized heat treatment at 454 0 C (850 0 F) are the degree of distortion that may occur after the annealing cycle and the extent of residual stresses. A thermal and structural analysis of a reactor vessel for distortions and residual stresses found no problems with the reactor vessel itself but did indicate a rotation at the nozzle region of the vessel that would plastically deform the attached primary piping. Further analytical studies are needed. An American Society for Testing and Materials (ASTM) task group is upgrading and revising the ASTM Recommended Guide for In-Service Annealing of WaterCooled Nuclear Reactor Vessels (E 509-74) with emphasis on the materials and surveillance aspects of annealing rather than system engineering problems. System safety issues are the province of organizations other than ASTM (for example, the American Society of Mechanical Engineers Boiler and Pressure Vessel Code body)

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

    Energy Technology Data Exchange (ETDEWEB)

    Baugnet, J M; Roedt, Ch de; Gubel, P; Koonen, E [Centre d' Etude de I' Energie Nucleaire, Studiecentrum voor Kernenergie, C.E.N./S.C.K., Mol (Belgium)

    1990-05-01

    The BR2 reactor (Mol, Belgium) is a high-flux materials testing reactor. The fuel is 93% {sup 235}U 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 10{sup 15} n/cm{sup 2} s; fast (E > 0.1 MeV) : 8.4 x 10{sup 14} n /cm{sup 2} 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)

  14. MULTI - A multigroup or multipoint P{sub 3} programme for calculating thermal neutron spectra in a reactor cell

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M V [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1968-06-15

    Programme MULTI calculates the space energy distribution of thermal neutrons in a multizone, cylindrical, infinitely long reactor lattice by using the multigroup or multipoint P{sub 3} approximation. This report presents a short description of the algorithm and the programme and gives the instructions for its exploitation. (author)

  15. TRIO a general computer code for reactor 3-D flows analysis. Application to a LMFBR hot plenum

    International Nuclear Information System (INIS)

    Magnaud, J.P.; Rouzaud, P.

    1985-09-01

    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.

  16. Modular 3D printed lab-on-a-chip bio-reactor for the biochemical energy cascade of microorganisms

    International Nuclear Information System (INIS)

    Podwin, Agnieszka; Dziuban, Jan A

    2017-01-01

    The paper presents the sandwiched polymer 3D printed lab-on-a-chip bio-reactor for the biochemical energy cascade of microorganisms. Euglenas and yeast were separately and simultaneously cultured for 10 d in the chip. As a result of the experiments, euglenas, light-initialized and nourished by CO 2 —a product of ethanol fermentation handled by yeast—generated oxygen, based on the photosynthesis process. The presence of oxygen in the bio-reactor was confirmed by the colorimetric method—a bicarbonate (pH) indicator. Preliminary studies towards the obtainment of an effective source of oxygen are promising and further research should be done to enable the utility of the bio-reactor in, for instance, microbial fuel cells. (paper)

  17. Modular 3D printed lab-on-a-chip bio-reactor for the biochemical energy cascade of microorganisms

    Science.gov (United States)

    Podwin, Agnieszka; Dziuban, Jan A.

    2017-10-01

    The paper presents the sandwiched polymer 3D printed lab-on-a-chip bio-reactor for the biochemical energy cascade of microorganisms. Euglenas and yeast were separately and simultaneously cultured for 10 d in the chip. As a result of the experiments, euglenas, light-initialized and nourished by CO2—a product of ethanol fermentation handled by yeast—generated oxygen, based on the photosynthesis process. The presence of oxygen in the bio-reactor was confirmed by the colorimetric method—a bicarbonate (pH) indicator. Preliminary studies towards the obtainment of an effective source of oxygen are promising and further research should be done to enable the utility of the bio-reactor in, for instance, microbial fuel cells.

  18. Recommendations of the Reactor Safety Commission (RSK) 1975-1977. Vol. 3

    International Nuclear Information System (INIS)

    Schneider, M.

    1978-08-01

    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 [de

  19. Modeling flow stress constitutive behavior of SA508-3 steel for nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Sun Mingyue, E-mail: mysun@imr.ac.cn [Shenyang National Laboratory for Materials Science, Institute of Metal Research, Chinese Academy of Sciences, 72 Wenhua Road, Shenyang 110016 (China); Luhan, Hao; Shijian, Li; Dianzhong, Li; Yiyi, Li [Shenyang National Laboratory for Materials Science, Institute of Metal Research, Chinese Academy of Sciences, 72 Wenhua Road, Shenyang 110016 (China)

    2011-11-15

    Highlights: > A series of flow stress constitutive equations for SA508-3 steel were successfully established. > The experimental results under different conditions have validated the constitutive equations. > An industrial application of the model was present to simulate a large conical shell forging process. - Abstract: Based on the measured stress-strain curves under different temperatures and strain rates, a series of flow stress constitutive equations for SA508-3 steel were firstly established through the classical theories on work hardening and softening. The comparison between the experimental and modeling results has confirmed that the established constitutive equations can correctly describe the mechanical responses and microstructural evolutions of the steel under various hot deformation conditions. We further represented a successful industrial application of this model to simulate a forging process for a large conical shell used in a nuclear steam generator, which evidences its practical and promising perspective of our model with an aim of widely promoting the hot plasticity processing for heavy nuclear components of fission reactors.

  20. Safety analysis of an irradiation device for 99Mo production in RA-3 reactor

    International Nuclear Information System (INIS)

    Lerner, Ana Maria; Madariaga, Marcelo; Waldman, Ricardo

    2000-01-01

    The Argentine RA-3 research reactor (5 MW) has been converted to LEU fuel more than nine years ago. Since then, it has been operating with LEU fuel, which has been designed and fabricated at the National Atomic Energy Commission (CNEA). The Nuclear Regulatory Authority (ARN) is the institution in charge of the installation safety control. It is under this framework that the ARN has elaborated a neutronic calculation model for the RA-3 core, paying special attention to the device presently used for the irradiation of (HEU) 235 U targets required to obtain 9 '9Mo as a fission product. A regulatory analysis of results is carried out in the framework of ARN standards for fixed experiments. For such purpose, calculated reactivity values associated with such device are compared with recently measured values at the installation. Finally, and according to guidelines established in the first part of this work, a calculation model for a new device proposed by CNEA for the irradiation of metallic (LEU) uranium targets and still at its design stage, is here analysed. (author)

  1. Development of Nb3Sn based multi-filamentary superconductor wires for fusion reactor magnets

    International Nuclear Information System (INIS)

    Kundu, Sayandeep; Singh, A.K.; Hussain, M.M.

    2016-01-01

    Nb 3 Sn 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 Nb 3 Sn 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)

  2. Modeling flow stress constitutive behavior of SA508-3 steel for nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Sun Mingyue; Hao Luhan; Li Shijian; Li Dianzhong; Li Yiyi

    2011-01-01

    Highlights: → A series of flow stress constitutive equations for SA508-3 steel were successfully established. → The experimental results under different conditions have validated the constitutive equations. → An industrial application of the model was present to simulate a large conical shell forging process. - Abstract: Based on the measured stress-strain curves under different temperatures and strain rates, a series of flow stress constitutive equations for SA508-3 steel were firstly established through the classical theories on work hardening and softening. The comparison between the experimental and modeling results has confirmed that the established constitutive equations can correctly describe the mechanical responses and microstructural evolutions of the steel under various hot deformation conditions. We further represented a successful industrial application of this model to simulate a forging process for a large conical shell used in a nuclear steam generator, which evidences its practical and promising perspective of our model with an aim of widely promoting the hot plasticity processing for heavy nuclear components of fission reactors.

  3. Reactor safety issues resolved by the 2D/3D program

    International Nuclear Information System (INIS)

    1995-09-01

    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)

  4. Reactor safety issues resolved by the 2D/3D Program

    International Nuclear Information System (INIS)

    Damerell, P.S.; Simons, J.W.

    1993-07-01

    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

  5. Reactor safety issues resolved by the 2D/3D program

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    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).

  6. Detection of the contamination of air by tritiated water vapour around the reactor EL3

    International Nuclear Information System (INIS)

    Lebouleux, P.; Ardellier, A.; Valero, S.

    1968-01-01

    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) [fr

  7. Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

    Directory of Open Access Journals (Sweden)

    Guidez Joel

    2017-01-01

    In the case of sodium-cooled fast reactors (SFRs, the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction. From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

  8. Computational analysis of Bangladesh 3 MW TRIGA research reactor using MCNP4C, JENDL-3.3 and ENDF/B-Vl data libraries

    International Nuclear Information System (INIS)

    Huda, M.Q.

    2006-01-01

    The three-dimensional continuous energy Monte Carlo code MCNP4C 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 JENDL-3.3 and ENDF/BVI continuous energy cross-section data for MCNP4C was performed against some well-known benchmark lattices. For TRIGA analysis, data from JENDL-3.3 and ENDF/B-VI in combination with the JENDL-3.2 and ENDF/B-V data files (for nat Zr, nat Mo, nat Cr, nat Fe, nat Ni, nat Si, and nat Mg) at 300 K 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 TRIGA reactor. There is ∼20.0% decrease of thermal neutron flux occurs when the thermal library is removed during the calculation. Effect of erbium isotope that is present in the TRIGA fuel was also studied. In addition to the effective multiplication values, the well-known integral parameters: δ 28 , δ 25 , ρ 25 , and C * were calculated and compared for both JENDL3.3 and ENDF/B-VI libraries and were found to be in very good agreement. Results are also reported for most of the analyses performed by JENDL-3.2 and ENDF/B-V data libraries

  9. Water holding capacity of meat from rabbits (Belgian Giant breed

    Directory of Open Access Journals (Sweden)

    Gabriela Tărnăuceanu Frunză

    2016-03-01

    Full Text Available Abstract. Meat science has always been most interested in practical applications and macroscopic effects of internal/external factors in relation to water holding capacity (WHC. Research has been motivated by technological and sensory aspects, both finally linked to economic benefits. The objective of this study was to assess the effects of gender and different muscle groups from rabbits on WHC of their meat. Determining WHC was carried out by compression of the meat over filter paper between two plates. The biological material was collected from 56 Belgian Giant breed rabbits (25 males and 31 females. Aged 11-12 months, the rabbits had an average body weight of 11.5 kg. Measurements were performed on the muscles Longissimus dorsi (LD, Psoas major (PM and Semimembranosus (SM 24 hours after slaughter. The percentage of WHC was calculated as ratio (per cent of weight of released water to intact meat. WHC for females, in LD, had an average value of 8%, in PM it had an average value of 9.31%, and in SM it had an average value of 12.91%. For males, WHC in LD was 7.6%, in PM 8.23% and in SM 11.43%. The average value for WHC was higher for females than for males. Regarding the statistical significance of differences by gender, distinct significant differences for SM and very significant differences for PM were recorded. For PM, a higher average value of WHC is probably due to the smaller diameter of muscle fibres and also to water higher percentage of them.

  10. The DR 3 reactor at Risoe, Denmark and its associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the DR 2 reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of seven information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  11. User's guide of DETRAS system-3. Description of the simulated reactor plant

    International Nuclear Information System (INIS)

    Yamaguchi, Yukichi

    2006-12-01

    DETRAS system is a PWR reactor simulator system for operation trainings whose distinguished feature is that it can be operated from the remote place of the simulator site. The document which is the third one of a series of three volumes of the user's guide of DETRAS, describes firstly an outline of the simulated reactor system then a user's interface needed for operation of the simulator of interest and finally a series of procedure for startup of the simulated reactor and shutdown of it from its rated operation state. (author)

  12. Characteristics of UV-MicroO3 Reactor and Its Application to Microcystins Degradation during Surface Water Treatment

    Directory of Open Access Journals (Sweden)

    Guangcan Zhu

    2015-01-01

    Full Text Available The UV-ozone (UV-O3 process is not widely applied in wastewater and potable water treatment partly for the relatively high cost since complicated UV radiation and ozone generating systems are utilized. The UV-microozone (UV-microO3, a new advanced process that can solve the abovementioned problems, was introduced in this study. The effects of air flux, air pressure, and air humidity on generation and concentration of O3 in UV-microO3 reactor were investigated. The utilization of this UV-microO3 reactor in microcystins (MCs degradation was also carried out. Experimental results indicated that the optimum air flux in the reactor equipped with 37 mm diameter quartz tube was determined to be 18∼25 L/h for efficient O3 generation. The air pressure and humidity in UV-microO3 reactor should be low enough in order to get optimum O3 output. Moreover, microcystin-RR, YR, and LR (MC-RR, MC-YR, and MC-LR could be degraded effectively by UV-microO3 process. The degradation of different MCs was characterized by first-order reaction kinetics. The pseudofirst-order kinetic constants for MC-RR, MC-YR, and MC-LR degradation were 0.0093, 0.0215, and 0.0286 min−1, respectively. Glucose had no influence on MC degradation through UV-microO3. The UV-microO3 process is hence recommended as a suitable advanced treatment method for dissolved MCs degradation.

  13. The impact of gendered friendship patterns on the prevalence of homophobia among belgian late adolescents.

    Science.gov (United States)

    Hooghe, Marc

    2011-06-01

    In order to assess the determinants of homophobia among Belgian adolescents, a shortened version of the Homophobia scale (Wright et al., 1999) was included in a representative survey among Belgian adolescents (n = 4,870). Principal component analysis demonstrated that the scale was one-dimensional and internally coherent. The results showed that homophobia is still widespread among Belgian adolescents, despite various legal reforms in the country aiming to combat discrimination of gay women and men. A multivariate regression analysis demonstrated that boys, ethnic minorities, individuals with high levels of ethnocentrism and an instrumental worldview, Muslim minorities, and those with low levels of associational involvement scored significantly higher on the scale. While among boys an extensive friendship network was associated with higher levels of homophobia, the opposite phenomenon was found among girls. We discuss the possible relation between notions of masculinity within predominantly male adolescent friendship networks and social support for homophobia.

  14. Emerging contaminants in Belgian marine waters: single toxicant and mixture risks of pharmaceuticals.

    Science.gov (United States)

    Claessens, Michiel; Vanhaecke, Lynn; Wille, Klaas; Janssen, Colin R

    2013-06-15

    Knowledge on the effects of pharmaceuticals on aquatic marine ecosystems is limited. The aim of this study was therefore to establish the effect thresholds of pharmaceutical compounds occurring in the Belgian marine environment for the marine diatom Phaeodactylum tricornutum, and subsequently perform an environmental risk assessment for these substances. Additionally, a screening-level risk assessment was performed for the pharmaceutical mixtures. No immediate risk for acute toxic effects of these compounds on P. tricornutum were apparent at the concentrations observed in the Belgian marine environment. In two Belgian coastal harbours however, a potential chronic risk was observed for the β-blocker propranolol. No additional risks arising from the exposure to mixtures of pharmaceuticals present in the sampling area could be detected. However, as risk characterization ratios for mixtures of up to 0.5 were observed, mixture effects could emerge should more compounds be taken into account. Copyright © 2013 Elsevier Ltd. All rights reserved.

  15. Summary of failed reactor coolant pump rotating assembly experience at Crystal River Unit 3

    International Nuclear Information System (INIS)

    Hayner, G.O.; Clary, M.D.

    1992-01-01

    Four reactor coolant pump (RCP) rotating assemblies (shafts) have failed or have severely cracked during operation at the Crystal River Unit 3 (CR-3) Nuclear Power Plant. The two failed shafts removed from RCP-1A have been extensively examined. All of the RCP shafts (except the D shaft) were fabricated from UNS S66286 superalloy (Alloy A-286). The D shaft was fabricated from UNS S20910 (Alloy XM-19/Nitronic 50). Torsional strain gauge analysis was performed on the RCP-1A shaft during the 1990 refueling outage. This type of analysis has not been performed previously on an operating RCP. Several results were found including: (1) the primary components of alternating torsional stress during normal RCP operation are impeller vane pass and a sub-2X torsional resonance with maximum components of ∼±0.8 ksi; (2) a typical vane pass cycle is initiated by an abrupt unloading of the shaft followed by a reload past equilibrium and a damped return to equilibrium; (3) a higher (compared to normal four pump operation) alternating torsional stress range resulted from solo operation of RCP-1A at low temperature and pressure (normal startup conditions); (4) the 2/0 combination produced the highest mean torsional stresses and the lowest alternating stresses and (5) a startup of a secured RCP with three operating pumps results in significantly higher alternating stress than a cold startup. The root cause RCP failure mechanism appears to involve RCP startup sequence at CR-3, peculiarities that necessitate this sequence and complex shaft stresses just above or under the journal bearing. The 1986 impeller bolt failure is not considered to be a root cause effect. It was also determined that fatigue cracking has always been responsible for both shaft initiation and propagation mechanisms and cracking can occur independent of shaft material

  16. Carcass and meat quality in double-muscled Belgian Blue bulls and cows.

    Science.gov (United States)

    Fiems, L O; De Campeneere, S; Van Caelenbergh, W; De Boever, J L; Vanacker, J M

    2003-03-01

    Carcass and meat quality of 37 bulls and 91 cows of the Belgian Blue breed (double-muscled type) were compared. Age at slaughter averaged 648±73 and 1820±689 days, respectively. Both groups of cattle were finished on maize silage supplemented with concentrate, and were slaughtered at about 750 kg live weight. Females had a lower (P=0.004) cold carcass weight (469.7 kg) in comparison with bulls (500.8 kg), due to a reduced dressing percentage (63.8 vs. 66.6; P M. longissimus thoracis (LT) muscle (2.3 vs. 1.1%; P <0.001) were higher for females than for males. The LT of cows was darker (lower L* and higher a*-value; P <0.001), had a better waterholding capacity (P⩽0.063) and was slightly more tender (P=0.120) than the LT of bulls. Increasing parity reduced dressing percentage and increased LT lightness (L*-value) in cows. Several carcass (SEUROP-grading, composition, LT-area) and meat quality traits (protein and fat contents, drip and cooking losses, a*-value) were better correlated with carcass weight than parity. It is concluded that meat quality of the aged LT of cows is not negatively affected by age, while some carcass quality traits decreased with advancing age. Carcass quality traits adjusted for age at slaughter were better for bulls, but LT meat quality characteristics were at least as good for females as for males.

  17. 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.

  18. Management of radioactive liquid and solid wastes at the Research Reactor Institute, Kyoto University, (3)

    International Nuclear Information System (INIS)

    Tsutsui, Tenson; Shimoura, K.; Koyama, A.

    1977-11-01

    In this report, the management of radioactive liquid and solid wastes at the Research Reactor Institute, Kyoto University during past 6 years, from April in 1971 to March in 1977 are reviewed. (auth.)

  19. Control rod interaction models for use in 2D and 3D reactor geometries

    International Nuclear Information System (INIS)

    Bannerman, R.C.

    1985-11-01

    Control rod interaction models are developed for use in two-dimensional and three-dimensional reactor geometries. These models allow the total worth of any combination of control rods inserted to be determined from the individual worths in conjunction with an algorithm representing interaction effects between them. The validity of the assumptions is demonstrated for fast and thermal systems showing modelling errors of 1#percent# or less in inserted control rod worths. The models are ideally suited for most reactor systems. (UK)

  20. Overview of European Community (Activity 3) work on materials properties of fast reactor structural materials

    International Nuclear Information System (INIS)

    Wood, D.S.

    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

  1. Licensed operating reactors. Status summary report, data as of February 28, 1986. Volume 10, No. 3

    International Nuclear Information System (INIS)

    1986-04-01

    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

  2. Technology assessment HTR. Part 3. Economics of new concept of the modular High Temperature Reactor

    International Nuclear Information System (INIS)

    Lako, P.

    1996-06-01

    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

  3. Calculation-measurement comparison for control rods reactivity in RA-3 nuclear reactor

    International Nuclear Information System (INIS)

    Estryk, Guillermo; Gomez, Angel

    2002-01-01

    The RA-3 Nuclear Reactor of the Atomic Energy National Commission from Argentina, begun working with high enrichment fuel elements in 1967, and turned to low enrichment by 1990. During 1999 it was found out that several fuel elements had problems, so more than 50 % of them had to be removed from the core. Because of this, it was planned to go from core 93 to core 94 with special care from nuclear safety point of view. Core 94 was preceded by other five, T-1 to T-5, only as transitory ones. The care implied several nuclear parameters measurements: core reactivity excess, calibration of control rods, etc. Calculations were performed afterwards to simulate those measurements using the neutron diffusion code PUMA. The comparison shows a good agreement for more than 80% of the cases with differences lower than 10% in reactivity. The greatest differences were found in the last part of the control rods calibration and a better calculation of cell constants is planned to be done in order to improve the adjustment. (author)

  4. Irradiation tests of THTR fuel elements in the DRAGON reactor (irradiation experiment DR-K3)

    International Nuclear Information System (INIS)

    Burck, W.; Duwe, R.; Groos, E.; Mueller, H.

    1977-03-01

    Within the scope of the program 'Development of Spherical Fuel Elements for HTR', similar fuel elements (f.e.) have been irradiated in the DRAGON reactor. The f.e. were fabricated by NUKEM and were to be tested under HTR conditions to scrutinize their employability in the THTR. The fuel was in the form of coated particles moulded into A3 matrix. The kernels of the particles were made of mixed oxide of uranium and thorium with an U 235 enrichment of 90%. One aim of the post irradiation examination was the investigation of irradiation induced changes of mechanical properties (dimensional stability and elastic behaviour) and of the corrosion behaviour which were compared with the properties determined with unirradiated f.e. The measurement of the fission gas release in annealing tests and ceramografic examinations exhibited no damage of the coated particles. The measured concentration distribution of fission metals led to conclusions about their release. All results showed, that neither the coated particles nor the integral fuel spheres experienced any significant changes that could impair their utilization in the THTR. (orig./UA) [de

  5. MC2-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Yang, W. S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2013-11-08

    The MC2-3 code is a Multigroup Cross section generation Code for fast reactor analysis, developed by improving the resonance self-shielding and spectrum calculation methods of MC2-2 and integrating the one-dimensional cell calculation capabilities of SDX. The code solves the consistent P1 multigroup transport equation using basic neutron data from ENDF/B data files to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (~2000) or hyperfine (~400,000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified isotopic temperatures. The pointwise cross sections are directly used in the hyperfine group calculation whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for two-dimensional whole-core problems to generate region-dependent broad-group cross sections. Multigroup cross sections are written in the ISOTXS format for a user-specified group structure. The code is executable on UNIX, Linux, and PC Windows systems, and its library includes all isotopes of the ENDF/BVII. 0 data.

  6. Design of a New Research Reactor: Preliminary Conceptual Design (3rd Year)

    International Nuclear Information System (INIS)

    Park, Cheol; Lee, B. C.; Chae, H. T. and others

    2006-01-01

    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

  7. OECD MCCI project enhancing instrumentation for reactor materials experiments, Rev. 0 September 3, 2002

    International Nuclear Information System (INIS)

    Lomperski, S.; Basu, S.

    2011-01-01

    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 ∼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 delay would

  8. LEU fuel cycle analyses for the Belgian BR2 Research Reactor

    International Nuclear Information System (INIS)

    Deen, J.R.; Snelgrove, J.L.

    1988-01-01

    Equilibrium fuel cycle characteristics were calculated for reference HEU and two proposed LEU fuel cycles using an 11-group diffusion-theory neutron flux solution in hexagonal-Z geometry. The diffusion theory model was benchmarked with a detailed Monte Carlo core model. The two proposed LEU fuel designs increased the 235 U loading 20% and the fuel meat volume 51%. The first LEU design used 10 B as a burnable absorber. Either proposed LEU fuel element would provide equilibrium fuel cycle characteristics similar to those of the HEU fuel cycle. Irradiation rates of Co control followers and Ir disks in the center of the core were reduced 6 ± 1% in the LEU equilibrium core compared to reference HEU core. 11 refs., 4 figs., 5 tabs

  9. Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2008

    International Nuclear Information System (INIS)

    2014-02-01

    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)

  10. Nuclear reactor

    International Nuclear Information System (INIS)

    Schulze, I.; Gutscher, E.

    1980-01-01

    The core contains a critical mass of UN or U 2 N 3 in the form of a noncritical solution with melted Sn being kept below a N atmosphere. The lining of the reactor core consists of graphite. If fission progresses part of the melted metal solution is removed and cleaned from fission products. The reactor temperatures lie in the range of 300 to 2000 0 C. (Examples and tables). (RW) [de

  11. Examination policy concerning the additional installation of No. 3 and No. 4 reactors in Takahama Nuclear Power Station and No. 3 and No. 4 reactors in Fukushima No. 2 Nuclear Power Station

    International Nuclear Information System (INIS)

    1980-01-01

    The Nuclear Safety Commission decided the annual examination policy on the modification of reactor installation in Takahama Nuclear Power Station to construct No. 3 and No. 4 reactors inquired under date of November 26, 1979, by the Minister of International Trade and Industry, so that the examination results of the accident in Three Mile Island nuclear power station are reflected to the examination for the purpose of improving reactor safety. The examination results of the accident in Three Mile Island power station are being investigated by the Committee on Examination of Reactor Safety, based on the policy shown in ''On the second report of the special committee examining the accident in a nuclear power station in the U.S.'' determined by the Nuclear Safety Commission under date of September 13, 1979. Though the Committee will further clarify the past guideline about the items concerning the criteria, design and operation management, the Committee decided the tentative policy to reflect it to safety examination. Further, a table is attached, in which 52 items to be reflected to the security measures are classified from the viewpoint of necessity to reflect them to the final examination. This table includes 13 items of criteria and examination, 7 items related to design, 10 items related to operation management, 10 antidisaster items, and 12 items related to safety research. (Wakatsuki, Y.)

  12. 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)

  13. PASC-1, Petten AMPX-II/SCALE-3 Code System for Reactor Neutronics Calculation

    International Nuclear Information System (INIS)

    Yaoqing, W.; Oppe, J.; Haas, J.B.M. de; Gruppelaar, H.; Slobben, J.

    1995-01-01

    1 - Description of program or function: The Petten AMPX-II/SCALE-3 Code System PASC-1 is a reactor neutronics calculation programme system consisting of well known IBM-oriented codes, that have been translated into FORTRAN-77, for calculations on a CDC-CYBER computer. Thus, the portability of these codes has been increased. In this system, some AMPX-II and SCALE-3 modules, the one-dimensional transport code ANISN and the 1 to 3-dimensional diffusion code CITATION are linked together on the CDC-CYBER/855 computer. The new cell code XSDRNPM-S and the old XSDRN code are included in the system. Starting from an AMPX fine group library up to CITATION, calculations can be performed for each individual module. Existing AMPX master interface format libraries, such as CSRL-IV, JEF-1, IRI and SCALE-45, and the old XSDRN-formatted libraries such as the COBB library can be used for the calculations. The code system contains the following modules and codes at present: AIM, AJAX, MALOCS, NITAWL-S, REVERT-I, ICE-2, CONVERT, JUAN, OCTAGN, XSDRNPM-S, XSDRN, ANISN and CITATION. The system will be extended with other SCALE modules and transport codes. 2 - Method of solution: The PASC-1 system is based on AMPX-II/SCALE-3 modules. Except for some SCALE-3 modules taken from the SCALIAS package, the original AMPX-II modules were IBM versions written in FORTRAN IV. These modules have been translated into CDC FORTRAN V. In order to test these modules and link them with some codes, some of the sample problem calculations have been performed for the whole PASC-1 system. During these calculations, some FORTRAN-77 errors were found in MALOCS, REVERT, CONVERT and some subroutines of SUBLIB (FORTRAN-77 subroutine library). These errors have been corrected. Because many corrections were made for the REVERT module, it is renamed as REVERT-I (improved version of REVERT). After these corrections, the whole system is running on a CDC-CYBER Computer (NOS-BE operating system). 3 - Restrictions on the

  14. 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)

    Vandecasteele, C.M.; Hardeman, F.; Pauwels, O.; Bernaerts, M.; Carle, B.; Sombre, L.

    2005-01-01

    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

  15. 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)

    Khan, Jahirul Haque

    2013-01-01

    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

  16. Investigations related to a one-piece removal of the reactor block in the frame of the JRR-3 reconstruction program

    International Nuclear Information System (INIS)

    Onishi, N.; Kanenari, A.; Futamura, Y.; Sakurai, H.; Suzuki, S.; Nagase, T.; Iwatani, A.; Otsubo, F.

    1987-01-01

    In the Japan Atomic Energy Research Institute (JAERI), an outdated research reactor (Japan Research Reactor No.3; JRR-3) was removed to a storage facility between October 14th and November 7th, 1986. The removal of the 2250-ton reactor block (10 x 10 x 10 m) was performed as a part of a program to replace the JRR-3's core (10-MW thermal) with an upgraded research reactor core. The heavy water and fuel elements were taken out from the JRR-3 before removal work began. The reactor block was raised about 3.7 meters, using a 12-cubic meter steel frame and a center-hole jack system. The reactor block was then transported horizontally about 34 meters on steel rails, using four 100-ton jacks, to a storage facility. Finally, the reactor block was lowered 14 meters into the storage facility. After the reactor block was stored, a new 20-MW thermal, light-water moderated and cooled JRR-3 core will be built, with criticality targeted for 1989

  17. Antibiotic use and resistance in animals: Belgian initiatives.

    Science.gov (United States)

    Daeseleire, Els; De Graef, Evelyne; Rasschaert, Geertrui; De Mulder, Thijs; Van den Meersche, Tina; Van Coillie, Els; Dewulf, Jeroen; Heyndrickx, Marc

    2016-05-01

    The widespread use of antibiotics in animals is causing concerns about the growing risk for development and the spread of antibiotic-resistant bacteria. Antibiotic consumption is higher in animals than in humans as reported in a joint publication of EFSA (European Food Safety Agency), ECDC (European Centre for Disease Prevention and Control), and EMA (European Medicines Agency) using data from 2011 and 2012. Both in humans and animals, positive associations between the consumption of antibiotics and resistant bacteria are observed. Responsible use of antibiotics in humans and animals should therefore be promoted. In this paper some general aspects of antibiotic resistance such as microbiological versus clinical resistance, intrinsic versus acquired resistance, resistance mechanisms, and transfer of resistance are briefly introduced. In 2012, the Belgian Center of Expertise on Antimicrobial Consumption and Resistance in Animals (AMCRA) was founded. Its mission is to collect and analyze all data related to antibiotic use and resistance in animals in Belgium and to communicate these findings in a neutral and objective manner. One of AMCRA's 10 objectives is a 50% reduction in antibiotic consumption in veterinary medicine in Belgium by 2020. The aim of this paper is to report on the achievements of this national project. The Institute for Agricultural and Fisheries Research (ILVO, Merelbeke-Melle), in collaboration with Ghent University, is currently working on three nationally funded projects on antibiotic resistance in animal husbandry. In the first project, an in vitro model is used to study the influence of low antibiotic concentrations due to carry-over after production and usage of medicated feed on the development of resistance in the pig gut. Part of that project is to develop a quantitative risk assessment model. A second project focuses on tracking excreted antibiotics used in pig rearing and their influence on the development of antibiotic resistance in pig

  18. Future fuel cycle and reactor strategies. Key issue paper no. 3

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-06-01

    The scope of this paper includes those issues that are expected to remain or become important in the time period from 2015 to 2050. Events in this time frame are difficult to predict with any certainty, so the framework of this paper is necessarily somewhat speculative. The paper includes consideration of all facets of nuclear energy utilization, from ore extraction to final disposal of waste products. The paper first addresses the factors influencing the choice of reactor and fuel cycle. It then goes on to address the quantitatively largest category of reactor types expected to be important during the period; that is, thermal reactors burning uranium and plutonium fuel in various forms. The fast reactor type is then discussed both as stand-alone technology and as technology used in combination with thermal reactors. Thorium fuel use is discussed briefly. This paper is concentrated on the ``medium variant`` energy growth scenario identified in Key Issue Paper No. 1. The effects of either higher or lower growth could, of course, profoundly change the future development of the nuclear power industry. 31 refs, 5 figs, 4 tabs.

  19. Cross-cultural differences in lay attitudes and utilisation of antibiotics in a Belgian and a Dutch city.

    Science.gov (United States)

    Deschepper, Reginald; Vander Stichele, Robert H; Haaijer-Ruskamp, Flora M

    Cultural differences are probably an important factor in the considerable variation in antibiotic use between countries. The objective of this study was to explore local cultural differences in the lay perspective on coping with URTD and using antibiotics. We interviewed 30 persons in a Dutch and a Belgian city. Twenty-one were interviewed a second time after 3 months. Between the first and second interview, they noted in a diary all URTD episodes experienced by themselves and their family members (N=69) and how they coped with them. The Dutch participants labelled most URTD episodes as "common cold" or "flu". The Flemish participants labelled most of their URTD episodes as "bronchitis" and used more antibiotics. Four categories of antibiotic users could be distinguished. Participants with a Protestant background were more sceptical about medicines than those with a Catholic background. A thorough understanding of the cultural context is necessary to design effective campaigns to promote rational antibiotic use.

  20. Validation and application of 3D-methods for the design and safety analysis of high temperature reactors

    International Nuclear Information System (INIS)

    Bader, J.; Lapins, J.; Buck, M; Bernnat, W.; Laurien, E.

    2011-01-01

    Some of the concepts for future nuclear reactors are high-temperature gas-cooled reactors. Previous simulation codes for their cores were often based on one- or two-dimensional models, but today's increasing computer capabilities make an advance to 3D-codes possible now. Our thermal-hydraulic code ATTICA3D (Advanced Thermal-hydraulic Tool for In-vessel and Core Analysis in 3 Dimensions) is based on the porous media approach, including 3-D models of heat conduction and gas flow, using a coarse-grid integration method for the time-dependent conservation equations of mass, momentum and energy. Results of numerical calculations for various validation cases are presented: First, the test facility SANA is chosen, which has been used to study heat transfer phenomena inside a coolant-gas filled pebble-bed core, which was heated by embedded electrical heating elements. Calculations were carried out for different tests taken from the experimental database. Measured and calculated temperatures at different positions are compared and found in good agreement. Second, our code was used to simulate a depressurized loss of forced cooling experiment with simulated decay heat in the AVR Experimental Reactor. Due to its design with the shut-down rods located inside columnar noses, which extend into the pebble bed of the core, geometry and power distribution are genuinely three-dimensional. The power distribution was calculated by the 3D-Neutronic Diffusion Code CITATION in conjunction with the spectral code MICROX-2. The neutronics and thermal-hydraulics calculations were carried out for a 3D, 45°-degree section of the reactor. It is demonstrated, that the experimental results could be qualitatively reproduced. (author)

  1. Effect of cull potatoes in the diet for finishing Belgian Blue double-muscled cows.

    Science.gov (United States)

    Fiems, L O; De Boever, J L; Vanacker, J M; De Brabander, D L

    2013-01-01

    The use of culled potatoes was investigated in Belgian Blue double-muscled finishing cows, confined in tie stalls. The control diet (Treatment 1) consisted of concentrate and maize silage (50/50 on a dry matter (DM) basis). Potatoes either replaced 60% maize silage (Treatment 2) or 60% concentrate (Treatment 3). Diets were formulated to be isocaloric and isonitrogenous. They were fed ad libitum. Approximately 18 kg potatoes were fed daily in Treatments 2 and 3. Daily gain was not significantly altered; it decreased from 1.09 kg (Treatment 1) to 1.04 kg (Treatment 2) or increased to 1.20 kg (Treatment 3), although potatoes stimulated DM intake by 5% to 8% (P < 0.05). Feed conversion was unaffected in comparison with the control diet, when expressed in terms of DM, but energy efficiency (MJ/kg live weight gain) was substantially lower for Treatment 2 compared with Treatment 1 (89.9 v. 79.0; P = 0.046). Carcass weight, grading and composition were not affected by treatments, but potatoes increased dressing percentage (P = 0.009). Treatment had no significant effect on meat quality parameters. However, potatoes (Treatments 2 and 3) tended to decrease moisture content (P = 0.090) and tended to increase drip loss (P = 0.059) compared with Treatment 1. Because of a better animal performance and a lower feed cost, it is most appropriate to use potatoes as a replacement for concentrate. Feeding large amounts of potatoes besides concentrate may have an adverse effect on the fibrousness of the diet, resulting in a tendency (-5%) for a reduced daily gain and a lower energy efficiency (P < 0.05).

  2. 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)

  3. Licensed operating reactors. Status summary report as of February 29, 1984. Volume 8, No. 3

    International Nuclear Information System (INIS)

    1984-04-01

    The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Resource Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement (IE), from NRC's Regional Offices, and from utilities. 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

  4. Advanced 3D Characterization and Reconstruction of Reactor Materials FY16 Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Fromm, Bradley [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hauch, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sridharan, Kumar [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-12-01

    A coordinated effort to link advanced materials characterization methods and computational modeling approaches is critical to future success for understanding and predicting the behavior of reactor materials that operate at extreme conditions. The difficulty and expense of working with nuclear materials have inhibited the use of modern characterization techniques on this class of materials. Likewise, mesoscale simulation efforts have been impeded due to insufficient experimental data necessary for initialization and validation of the computer models. The objective of this research is to develop methods to integrate advanced materials characterization techniques developed for reactor materials with state-of-the-art mesoscale modeling and simulation tools. Research to develop broad-ion beam sample preparation, high-resolution electron backscatter diffraction, and digital microstructure reconstruction techniques; and methods for integration of these techniques into mesoscale modeling tools are detailed. Results for both irradiated and un-irradiated reactor materials are presented for FY14 - FY16 and final remarks are provided.

  5. Advanced 3D Characterization and Reconstruction of Reactor Materials FY16 Final Report

    International Nuclear Information System (INIS)

    Fromm, Bradley; Hauch, Benjamin; Sridharan, Kumar

    2016-01-01

    A coordinated effort to link advanced materials characterization methods and computational modeling approaches is critical to future success for understanding and predicting the behavior of reactor materials that operate at extreme conditions. The difficulty and expense of working with nuclear materials have inhibited the use of modern characterization techniques on this class of materials. Likewise, mesoscale simulation efforts have been impeded due to insufficient experimental data necessary for initialization and validation of the computer models. The objective of this research is to develop methods to integrate advanced materials characterization techniques developed for reactor materials with state-of-the-art mesoscale modeling and simulation tools. Research to develop broad-ion beam sample preparation, high-resolution electron backscatter diffraction, and digital microstructure reconstruction techniques; and methods for integration of these techniques into mesoscale modeling tools are detailed. Results for both irradiated and un-irradiated reactor materials are presented for FY14 - FY16 and final remarks are provided.

  6. Decontamination and decommissioning project of the TRIGA Mark-2 and 3 research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jung, K J; Baik, S T; Chung, U S; Jung, K H; Park, S K; Lee, B J; Kim, J K; Yang, S H

    2000-01-01

    During the review on the decommissioning plan and environmental impact assessment report by the KINS, the number of the inquired items were two hundred and fifty one, and the answers were made and sent until September 10, 1999, as the screened review results were reported to Ministry of Science and Technology(MOST) in December 14, 1999, all the reviews on the licence were over. Radioactive liquid wastes of 400 tons generated during the operation of the research reactors including reactor vessels are stored in the facility of the research reactor 1 and 2. Those liquid wastes have the low-level-radioactivity which can be discharged to the surroundings, but was wholly treated to be vaporized naturally by means of the increased numbers of the natural vaporization disposal facilities with the annual capacity of 200 tons for the purpose of the minimized environmental contamination.

  7. RA Reactor operation and maintenance (I-IX), Part IV, Task 3.08/04, Refurbishment of the RA reactor; Pogon i odrzavanje reaktora RA (I-IX), IV Deo, Zadatak 3.08/04 Remont reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    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.

  8. Presence and mechanisms of acquired antimicrobial resistance in Belgian Brachyspira hyodysenteriae isolates belonging to different clonal complexes.

    Science.gov (United States)

    Mahu, M; Pasmans, F; Vranckx, K; De Pauw, N; Vande Maele, L; Vyt, Philip; Vandersmissen, Tamara; Martel, A; Haesebrouck, F; Boyen, F

    2017-08-01

    Swine dysentery (SD) is an economically important disease for which antimicrobial treatment still occupies an important place to control outbreaks. However, acquired antimicrobial resistance is increasingly observed in Brachyspira hyodysenteriae. In this study, the Minimal Inhibitory Concentrations (MIC) of six antimicrobial compounds for 30 recent Belgian B. hyodysenteriae isolates were determined using a broth microdilution method. In addition, relevant regions of the 16S rRNA, 23S rRNA and the L3 protein encoding genes were sequenced to reveal mutations associated with acquired resistance. Finally, a phylogeny was reconstructed using minimal spanning tree analysis of multi locus sequence typing of the isolates. For lincomycin, doxycycline, tylosin and tylvalosin, at least 70% of the isolates did not belong to the wild-type population and were considered to have acquired resistance. For valnemulin and tiamulin, this was over 50%. In all isolates with acquired resistance to doxycycline, the G1058C mutation was present in their 16S rRNA gene. All isolates showing acquired resistance to lincomycin and both macrolides displayed the A2058T mutation in their 23S rRNA gene. Other mutations in this gene and the N148S mutation in the L3 protein were present in both wild-type isolates and isolates considered to have acquired resistance. Multi locus sequence analysis revealed a previously undescribed clonal complex, with 4 novel sequence types in which the majority of isolates showed acquired resistance to all tested antimicrobial products. In conclusion, acquired antimicrobial resistance is widespread among Belgian B. hyodysenteriae isolates. The emergence of multi-resistant clonal complexes can pose a threat to swine industry. Copyright © 2017 Elsevier B.V. All rights reserved.

  9. Critical heat flux experiments for high conversion light water reactor, (3)

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Okubo, Tsutomu; Suemura, Takayuki; Hiraga, Fujio; Murao, Yoshio

    1990-03-01

    As a part of the thermal-hydraulic feasibility study of a high conversion light water reactor (HCLWR), critical heat flux (CHF) experiments were performed using triangular array rod bundles under steady-state and flow reduction transient conditions. The geometries of test sections were: rod outer diameter 9.5 mm, number of rods 4∼7, heated length 0.5∼1.0 m, and pitch to diameter ratio (P/D) 1.126∼1.2. The simulated fuel rod was a stainless steel tube and uniformly heated electrically with direct current. In the steady-state tests, pressures ranged: 1.0∼3.9 Mpa, mass velocities: 460∼4270 kg/s·m 2 , and exit qualities: 0.02∼0.35. In the transient tests, the times to CHF detection ranged from 0.5 to 25.4 s. The steady-state CHF's for the 4-rod test sections were higher than those for the 7-rod test sections with respect to the bundle averaged flow conditions. The measured CHF's increased with decreasing the heated length and decreased with decreasing the P/D. Based on the local flow conditions obtained with the subchannel analysis code COBRA-IV-I, KfK correlation agreed with the CHF data within 20 %, while WSC-2, EPRI-B and W, EPRI-Columbia and Kattor correlations failed to give satisfactory agreements. Under flow reduction rates less than 6 %/s, no significant difference in the onset conditions of DNB (departure from nucleate boiling) was recognized between the steady-state and transient conditions. At flow reduction rates higher than 6 %/s, on the other hand, the DNB occurred earlier than the DNB time predicted with the steady-state experiments. (author)

  10. Analysis of the VVER-440 reactor steam generator secondary side with the RELAP5/MOD3 code

    International Nuclear Information System (INIS)

    Tuunanen, J.

    1993-01-01

    Nuclear Engineering Laboratory of the Technical Research Centre of Finland has widely used RELAP5/MOD2 and -MOD3 codes to simulate horizontal steam generators. Several models have been developed and successfully used in the VVER-safety analysis. Nevertheless, the models developed have included only rather few nodes in the steam generator secondary side. The secondary side has normally been divided into about 10 to 15 nodes. Since the secondary side at the steam generators of VVER-440 type reactors consists of a rather large water pool, these models were only roughly capable to predict secondary side flows. The paper describes an attempt to use RELAP5/MOD3 code to predict secondary side flows in a steam generator of a VVER-440 reactor. A 2D/3D model has been developed using RELAP5/MOD3 codes cross-flow junctions. The model includes 90 volumes on the steam generator secondary side. The model has been used to calculate steady state flow conditions in the secondary side of a VVER-440 reactor steam generator. (orig.) (1 ref., 9 figs., 2 tabs.)

  11. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual. Part 3: Hardware component failure data; Volume 5, Revision 4

    International Nuclear Information System (INIS)

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

    1994-09-01

    This data manual contains a hard copy of the information in the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) Version 3.5 database, which is sponsored by the US Nuclear Regulatory Commission. NUCLARR was designed as a tool for risk analysis. Many of the nuclear reactors in the US and several outside the US are represented in the NUCLARR database. NUCLARR includes both human error probability estimates for workers at the plants and hardware failure data for nuclear reactor equipment. Aggregations of these data yield valuable reliability estimates for probabilistic risk assessments and human reliability analyses. The data manual is organized to permit manual searches of the information if the computerized version is not available. Originally, the manual was published in three parts. In this revision the introductory material located in the original Part 1 has been incorporated into the text of Parts 2 and 3. The user can now find introductory material either in the original Part 1, or in Parts 2 and 3 as revised. Part 2 contains the human error probability data, and Part 3, the hardware component reliability data

  12. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others

    1997-07-01

    This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

  13. 3-DB, 3-D Multigroup Diffusion, X-Y-Z, R-Theta-Z, Triangular-Z Geometry, Fast Reactor Burnup

    International Nuclear Information System (INIS)

    Hardie, R.W.; Little, W.W. Jr.; Mroz, W.

    1974-01-01

    1 - Description of problem or function: 3DB is a three-dimensional (x-y-z, r-theta-z, triangular-z) multigroup diffusion code for use in detailed fast-reactor criticality and burnup analysis. The code can be used to - (a) compute k eff and perform criticality searches on time absorption, reactor composition, and reactor dimensions by means of either a flux or an adjoint model, (b) compute material burnup using a flexible material shuffling scheme, and (c) compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Eigenvalues are computed by standard source- iteration techniques. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Adjoint solutions are obtained by inverting the input data and redefining the source terms. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy-averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes are formed by the user. The code does not contain built- in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated

  14. Contribution of VPS35 genetic variability to LBD in the Flanders-Belgian population

    NARCIS (Netherlands)

    Verstraeten, Aline; Wauters, Eline; Crosiers, David; Meeus, Bram; Corsmit, Ellen; Elinck, Ellen; Mattheijssens, Maria; Peeters, Karin; Cras, Patrick; Pickut, Barbara; Vandenberghe, Rik; Engelborghs, Sebastiaan; De Deyn, Peter Paul; Van Broeckhoven, Christine; Theuns, Jessie

    VPS35 was recently identified as a novel autosomal dominant gene for Parkinson disease. In this study, we aimed to determine the contribution of simple and complex VPS35 variations to the genetic etiology of the spectrum of Lewy body disorders (LBD) in a Flanders-Belgian patient cohort (n = 677). We

  15. 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.

  16. Proceedings of the 9th Dutch-Belgian Information Retrieval Workshop

    NARCIS (Netherlands)

    Aly, Robin; Hauff, C.; den Hamer, Ida; Hiemstra, Djoerd; Huibers, Theo W.C.; de Jong, Franciska M.G.

    Welcome to the 9th Dutch-Belgian Information Retrieval Workshop (DIR). I very well remember the DIR workshop in 2001 that was also organized in Twente. It took place exactly one day before my PhD defense, to give us the opportunity to have one of the PhD committee members, Stephen Robertson, as the

  17. Where is the Frame? : Victims and Intruders in the Belgian Press Coverage of the Asylum Issue

    NARCIS (Netherlands)

    Gorp, B. van

    2005-01-01

    In this article an empirically oriented conceptualization of frames is developed, using the issue of asylum and illegal immigration in the Belgian press as a test case. The methodological focus of this study is on the question of how these frames can be detected in the coverage. How can they be

  18. Clonal expansion of the Belgian Phytophthora ramorum populations based on new microsatellite markers

    Science.gov (United States)

    A. Vercauteren; I. De Dobbelaere; N. J. Grünwald; P. Bonants; E. Van Bockstaele; M. Maes; K. Heungens

    2010-01-01

    Co-existence of both mating types A1 and A2 within the EU1 lineage of Phytophthora ramorum has only been observed in Belgium, which begs the question whether sexual reproduction is occurring. A collection of 411 Belgian P. ramorum isolates was established during a 7-year survey. Our main objectives were genetic characterization of this population to test for sexual...

  19. The labelling and reporting of euthanasia by Belgian physicians: a study of hypothetical cases

    NARCIS (Netherlands)

    Smets, T.; Cohen, J.; Bilsen, J.; van Wesemael, Y.; Rurup, M.L.; Deliens, L.

    2012-01-01

    Background: Belgium legalized euthanasia in 2002. Physicians must report each euthanasia case to the Federal Control and Evaluation Committee. This study examines which end-of-life decisions (ELDs) Belgian physicians label 'euthanasia', which ELDs they think should be reported and the physician

  20. Survey of the natural radiation of Belgian territory as determined by different methods

    International Nuclear Information System (INIS)

    Deworm, J.P.; Slegers, W.; Gillard, J.; Flemal, J.M.; Culst, J.P.

    1988-01-01

    Measurement of the environmental exposure to natural radiation was performed by the Institute of Hygiene and Epidemiology and the Nuclear Research Centre in Mol. The aim of the study was the estimation of the external doses from natural radioactivity received by the Belgian population and the setting up on a map of the territory of natural exposure rates measured using different methods. (author)

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

    NARCIS (Netherlands)

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

    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

  2. Dust suppression in Belgian coal mines. Situation at the beginning of 1978

    Energy Technology Data Exchange (ETDEWEB)

    Preat, B; Vanstraelen, M

    1978-01-01

    Gives a general view of dust control in Belgian mines. Statistical data are presented in tables. Prior spraying, wet cutting, and water infusion and pneumatic drills with sprays are used; in some cases, two or more of these techniques are used together in the same face. (In French and In Dutch)

  3. 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' achievement.…

  4. 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

  5. High efficiency algorithm for 3D transient thermo-elasto-plastic contact problem in reactor pressure vessel sealing system

    International Nuclear Information System (INIS)

    Xu Mingyu; Lin Tengjiao; Li Runfang; Du Xuesong; Li Shuian; Yang Yu

    2005-01-01

    There are some complex operating cases such as high temperature and high pressure during the operating process of nuclear reactor pressure vessel. It is necessary to carry out mechanical analysis and experimental investigation for its sealing ability. On the basis of the self-developed program for 3-D transient sealing analysis for nuclear reactor pressure vessel, some specific measures are presented to enhance the calculation efficiency in several aspects such as the non-linear solution of elasto-plastic problem, the mixed solution algorithm for contact problem as well as contract heat transfer problem and linear equation set solver. The 3-D transient sealing analysis program is amended and complemented, with which the sealing analysis result of the pressure vessel model can be obtained. The calculation results have good regularity and the calculation efficiency is twice more than before. (authors)

  6. Seventh regular meeting of the International Working Group on Reliability of Reactor Pressure Components, Vienna, 3-5 September 1985

    International Nuclear Information System (INIS)

    1986-07-01

    The seventh regular meeting of the IAEA International Working Group on Reliability of Reactor Pressure Components was held at the Agency's Headquarters in Vienna from 3 to 5 September 1985. The representatives of Member States and of the Commission of the European Communities reported the status of the research programmes in this field (12 presentations). A separate abstract was prepared for each of the presentations

  7. Decontamination and decommissioning project status of the TRIGA mark-2±3 research reactors

    International Nuclear Information System (INIS)

    Jung, K. J.; Baek, S. T.; Jung, W. S.; Park, S. K.; Jung, K. H.

    1999-01-01

    TRIGA Mark-II, the first research reactor in Korea, has operated since 1962, and the second one, TRIGA Mark-III since 1972. Both of them had their operation phased out in 1995 due to their lives and operation of the new research reactor, HANARO at the Korea Atomic Energy Research Institute (KAERI) in Taejeon. Decontamination and decommissioning (D and D) project of the TRIGA Mark-II and Mark-III was started in January 1997 and will be completed in December 2002. In the first year of the project, work was performed in preparation of the decommissioning plan, start of the environmental impact assessment and setup licensing procedure and documentation for the project with cooperation of Korea Institute of Nuclear Safety (KINS). In 1998, Hyundai Engineering Company (HEC) is the main contractor to do design and licensing documentation for the D and D of both reactors. British Nuclear Fuels plc (BNFL) is technical assisting partner of HEC. The decommissioning plan document was submitted to the Ministry of Science and Technology (MOST) for the decommissioning license in December 1998, and it expecting to be issued a license at the end of September 1999. The goal of this project is to release the reactor site and buildings as an unrestricted area. This paper summarizes current status and future plan for the D and D project

  8. Gas-Cooled Thermal Reactor Program. Semiannual technical progress report, October 1, 1982-March 3, 1983

    International Nuclear Information System (INIS)

    1983-06-01

    This report provides descriptions and results of the technical effort during the first half of FY 83 on the Gas-Cooled Thermal Reactor Program. The work on Integration and Management (WBS 01) includes the preparation of the Advanced Systems Concept Evaluation Plan and the Advanced Systems Technology Development Plan in addition to the program management activities. The Market Definition (WBS 03) efforts considered the application of the Modular Reactor System with reforming (MRS-R) to the production of methanol and ammonia and the refining of petroleum. Within the Plant Technology (WBS 13) task there were activities to develop anlytical methods for investigation of Coolant Transport Behavior and to define methods and criteria for High Temperature Structural Engineering design. In addition to the work on the advanced HTGR for process heat users, new activities were initiated in support of the HTGR-SC/C Lead plant Protect (WBS 30 and 31). The Plant Simulation task (WBS 31) was initiated to develop a computer code for simulation of plant operation and for plant transient systems analysis. The efforts on the advanced HTGR systems was performed under the Modular Systems task (WBS 41) to study the potential for multiple small reactors to provide lower costs, improved safety, and higher availability than the large monolithic core reactors

  9. Taxonomic structure and stability of the bacterial community in belgian sourdough ecosystems as assessed by culture and population fingerprinting.

    Science.gov (United States)

    Scheirlinck, Ilse; Van der Meulen, Roel; Van Schoor, Ann; Vancanneyt, Marc; De Vuyst, Luc; Vandamme, Peter; Huys, Geert

    2008-04-01

    A total of 39 traditional sourdoughs were sampled at 11 bakeries located throughout Belgium which were visited twice with a 1-year interval. The taxonomic structure and stability of the bacterial communities occurring in these traditional sourdoughs were assessed using both culture-dependent and culture-independent methods. A total of 1,194 potential lactic acid bacterium (LAB) isolates were tentatively grouped and identified by repetitive element sequence-based PCR, followed by sequence-based identification using 16S rRNA and pheS genes from a selection of genotypically unique LAB isolates. In parallel, all samples were analyzed by denaturing gradient gel electrophoresis (DGGE) of V3-16S rRNA gene amplicons. In addition, extensive metabolite target analysis of more than 100 different compounds was performed. Both culturing and DGGE analysis showed that the species Lactobacillus sanfranciscensis, Lactobacillus paralimentarius, Lactobacillus plantarum, and Lactobacillus pontis dominated the LAB population of Belgian type I sourdoughs. In addition, DGGE band sequence analysis demonstrated the presence of Acetobacter sp. and a member of the Erwinia/Enterobacter/Pantoea group in some samples. Overall, the culture-dependent and culture-independent approaches each exhibited intrinsic limitations in assessing bacterial LAB diversity in Belgian sourdoughs. Irrespective of the LAB biodiversity, a large majority of the sugar and amino acid metabolites were detected in all sourdough samples. Principal component-based analysis of biodiversity and metabolic data revealed only little variation among the two samples of the sourdoughs produced at the same bakery. The rare cases of instability observed could generally be linked with variations in technological parameters or differences in detection capacity between culture-dependent and culture-independent approaches. Within a sampling interval of 1 year, this study reinforces previous observations that the bakery environment

  10. Taxonomic Structure and Stability of the Bacterial Community in Belgian Sourdough Ecosystems as Assessed by Culture and Population Fingerprinting▿ †

    Science.gov (United States)

    Scheirlinck, Ilse; Van der Meulen, Roel; Van Schoor, Ann; Vancanneyt, Marc; De Vuyst, Luc; Vandamme, Peter; Huys, Geert

    2008-01-01

    A total of 39 traditional sourdoughs were sampled at 11 bakeries located throughout Belgium which were visited twice with a 1-year interval. The taxonomic structure and stability of the bacterial communities occurring in these traditional sourdoughs were assessed using both culture-dependent and culture-independent methods. A total of 1,194 potential lactic acid bacterium (LAB) isolates were tentatively grouped and identified by repetitive element sequence-based PCR, followed by sequence-based identification using 16S rRNA and pheS genes from a selection of genotypically unique LAB isolates. In parallel, all samples were analyzed by denaturing gradient gel electrophoresis (DGGE) of V3-16S rRNA gene amplicons. In addition, extensive metabolite target analysis of more than 100 different compounds was performed. Both culturing and DGGE analysis showed that the species Lactobacillus sanfranciscensis, Lactobacillus paralimentarius, Lactobacillus plantarum, and Lactobacillus pontis dominated the LAB population of Belgian type I sourdoughs. In addition, DGGE band sequence analysis demonstrated the presence of Acetobacter sp. and a member of the Erwinia/Enterobacter/Pantoea group in some samples. Overall, the culture-dependent and culture-independent approaches each exhibited intrinsic limitations in assessing bacterial LAB diversity in Belgian sourdoughs. Irrespective of the LAB biodiversity, a large majority of the sugar and amino acid metabolites were detected in all sourdough samples. Principal component-based analysis of biodiversity and metabolic data revealed only little variation among the two samples of the sourdoughs produced at the same bakery. The rare cases of instability observed could generally be linked with variations in technological parameters or differences in detection capacity between culture-dependent and culture-independent approaches. Within a sampling interval of 1 year, this study reinforces previous observations that the bakery environment

  11. Development of clinical pharmacy in Belgian hospitals through pilot projects funded by the government.

    Science.gov (United States)

    Somers, A; Spinewine, A; Spriet, I; Steurbaut, S; Tulkens, P; Hecq, J D; Willems, L; Robays, H; Dhoore, M; Yaras, H; Vanden Bremt, I; Haelterman, M

    2018-04-30

    Objectives The goal is to develop clinical pharmacy in the Belgian hospitals to improve drug efficacy and to reduce drug-related problems. Methods From 2007 to 2014, financial support was provided by the Belgian federal government for the development of clinical pharmacy in Belgian hospitals. This project was guided by a national Advisory Working Group. Each funded hospital was obliged to describe yearly its clinical pharmacy activities. Results In 2007, 20 pharmacists were funded in 28 pilot hospitals; this number was doubled in 2009 to 40 pharmacists over 54 institutions, representing more than half of all acute Belgian hospitals. Most projects (72%) considered patient-related activities, whereas some projects (28%) had a hospital-wide approach. The projects targeted patients at admission (30%), during hospital stay (52%) or at discharge (18%). During hospital stay, actions were mainly focused on geriatric patients (20%), surgical patients (15%), and oncology patients (9%). Experiences, methods, and tools were shared during meetings and workshops. Structure, process, and outcome indicators were reported and strengths, weaknesses, opportunities, and threats were described. The yearly reports revealed that the hospital board was engaged in the project in 87% of the cases, and developed a vision on clinical pharmacy in 75% of the hospitals. In 2014, the pilot phase was replaced by structural financing for clinical pharmacy in all acute Belgian hospitals. Conclusion The pilot projects in clinical pharmacy funded by the federal government provided a unique opportunity to launch clinical pharmacy activities on a broad scale in Belgium. The results of the pilot projects showed clear implementation through case reports, time registrations, and indicators. Tools for clinical pharmacy activities were developed to overcome identified barriers. The engagement of hospital boards and the results of clinical pharmacy activities persuaded the government to start structural

  12. Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

    Science.gov (United States)

    Guidez, Joel; Saturnin, Anne

    2017-11-01

    During the operation of a nuclear reactor, the external individual doses received by the personnel are measured and recorded, in conformity with the regulations in force. The sum of these measurements enables an evaluation of the annual collective dose expressed in man·Sv/year. This information is a useful tool when comparing the different design types and reactors. This article discusses the evolution of the collective dose for several types of reactors, mainly based on publications from the NEA and the IAEA. The spread of good practices (optimization of working conditions and of the organization, sharing of lessons learned, etc.) and ongoing improvements in reactor design have meant that over time, the doses of various origins received by the personnel have decreased. In the case of sodium-cooled fast reactors (SFRs), the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction). From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

  13. Effects of thermal effluents from the Unterweser reactor (KKU) on biocenoses in the Unterweser. Pt. 3

    International Nuclear Information System (INIS)

    Vobach, M.; Feldt, W.

    1991-01-01

    Between August 1975 and November 1982, the influence of thermal pollution in the biocenosis of the Unterweser from cooling water dischanged by the Unterweser reactor was investigated. The state of the parts of the river not yet stressed by cooling water is compared with conditions after the start-up of the reactor (September 1978). The seasonal cycle of water temperature has not changed after the start-up as compared to the time before. A warming of the river water because of cooling water discharged from the reactor is recognizable in the area immediately surrounding the month of the discharge system. Benthal investigations show the composition of species and number of individuals to be unchanged after the start-up of the reactor. Phytoplankton, too, continues to have its population maximum in May and August. Zooplankton, being present abundantly and in clusters, has retained its original composition of species. Now as before the reactor's start-up, flounder, smelt, stickleback, sprat and gudgeon which between them account for 97 per cent of the total catch, continue to be the five major fish species. Variations in the composition of catch are not to be explained by changes of temperature. The slight temperature increase does not modify the spectrum of species; there is no temperature stimulus. The seasonal cycle of water temperatures, which are important for a diapause, i.e. a slowing in the developmental cycle with a reduced metabolism, thus safeguarding the survival of certain species, do continue to occur. Any observed changes are to be interpreted as expressions of longer-term biological cycles. (orig./BBR) [de

  14. Doses of radioiodine administered for hyperthyroidism: a sampling of Belgian nuclear medicine physician's attitudes

    International Nuclear Information System (INIS)

    Tondeur Dejonckheere, Marianne; Glinoer, Daniel; Verelst, Jean; Sand, Alain; Ham, Hamphrey

    2005-01-01

    Full text: While radioiodine (RI) is a well established treatment for hyperthyroidism, there is no consensus regarding the administration of fixed or calculated doses. Guidelines from scientific societies do not specify the preferable approach, nor the parameters to be used in order to calculate the latter. Therefore, the doses might, for the same patient, be different with regard to the chosen procedure. This study was undertaken to assess the variability of RI amounts administered in Belgium in various cases of hyperthyroidism. 21 Belgian nuclear medicine physicians issued from different departments and universities participated into the study. They received a file with clinical and biological data, iodine turnover rate, scintigraphic images and calculated thyroid surfaces from 10 patients (8 females, 2 males), 30-77 yrs suffering from hyperthyroidism of various etiologies: 7 patients had clinically overt hyperthyroidism and 3 subclinical hyperthyroidism; 7 patients had toxic goiters of various size (Graves' disease), 2 multi nodular goiter and 1 toxic nodule. None suffered from cardiac anomalies or ophthalmopathy. Participants were asked to define the amount of RI they would give in each case. Answers were received during a 8-week period. Analysing data from case 1 to case 10, the ranges of the proposed doses varied between 8 and 22 milli Curies (mCi) (sd : 2.4 - 6.07). Considering all the patients, the proposed doses varied between 2 mCi and 25 mCi. Analysing answers among the 21 participants, mean proposed doses varied between 4.5 and 17.3 mCi (sd: 0.69 - 7.99). Conclusion: These results demonstrate a wide variability among nuclear medicine physicians in the proposed RI doses and confirm that in Belgium there is no uniformity in the procedure used to determine the amount of RI to administer for various causes of hyperthyroidism. This emphasizes the notion that the determination of the amount of RI to be administered remains a matter of debate. (author)

  15. Annual report of Department of Research Reactor and Tandem Accelerator, JFY2011. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and tandem accelerator

    International Nuclear Information System (INIS)

    Ishii, Tetsuro; Nakamura, Kiyoshi; Kawamata, Satoshi; Ishikuro, Yasuhiro; Kawashima, Kazuhito; Kabumoto, Hiroshi; Nakamura, Takemi; Tamura, Itaru; Kawasaki, Sayuri; Sataka, Masao

    2013-03-01

    The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.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, 2011 and March 31, 2012. The activities were categorized into six service/development fields: (1) Recovery from the Great East Japan Earthquake, (2) Operation and maintenance of research reactors and tandem accelerator, (3) Utilization of research reactors and tandem accelerator, (4) Upgrading of utilization techniques of research reactors and tandem accelerator, (5) Safety administration for research reactors and tandem accelerator, (6) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, number of staff members dispatched to Fukushima for the technical assistance, commendation, outcomes in service and technical developments and so on. (author)

  16. The first critical experiment with a new type of fuel assemblies IRT-3M on the training reactor VR-I

    International Nuclear Information System (INIS)

    Matejka, Karel; Sklenka, Lubomir

    1997-01-01

    The paper 'The first critical experiment with a new type of fuel assemblies IRT-3M on training reactor VR-1 presents basic information about the replacement of fuel on the reactor VR-1 run on FJFI CVUT in Prague. In spring 1997 the IRT-2M fuel type used till then was replaced by the IRT-3M type. When the fuel was replaced, no change in its enrichment was made, i.e. its level remained as 36% 235 U. The replacement itself was carried out in tight co-operation with the Nuclear Research Institute Rez plc., as related to the operation of the research reactor LVR-15. The fuel replacement on the VR-I reactor is a part of the international program RERTR (Reduced Enrichment for Research and Test Reactors) in which the Czech Republic participates. (author)

  17. Development of Pneumatic Transfer Irradiation Facility (PTS no.3) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-04-15

    A pneumatic transfer system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide. The pneumatic transfer irradiation system (PTS no.3) involving a manual system and an semi-automatic system were reconstructed with new designs of a functional improvement at the HANARO research reactor and NAA laboratory of RI building in 2006. In this technical report, the design, operation and control of these system (PTS no.3) was described. Also the experimental results and the characteristic parameters measured from a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  18. Evaluation of LOCA in a swimming-pool type reactor using the 3D-AIRLOCA code

    International Nuclear Information System (INIS)

    Nagler, A.; Gilat, J.; Hirshfeld, H.

    1991-01-01

    The 3D-AIRLOCA code was used to calculate core temperature evolution curves in the wake of a full LOCA in a swimming pool type reactor, resulting in complete core exposure and dryout within about 1000 sec of the initiating event. The results show that fuel integrity loss thresholds (450 C for softening and 650 C for melting) are reached and exceeded over large fractions of the core at powr levels as low as 2 MW. At 4.5 MW, the softening threshold is reached even when the accident occurs up to 12 hours after reactor shutdown for continuous operation, and up to 2 hrs after shutdown for intermittent (6 hrs/day, 4 days a week) operation. The situation is even more severe in blockage cases, when the air flow through the core is blocked by residual water at the grid plate level. It is concluded that substantial fission product releases are quite likely in this class of accidents. (orig.)

  19. The importance of using the mixed neutron flux in activation analysis of D-3He fueled reactors

    International Nuclear Information System (INIS)

    Khater, H.Y.; Sawan, M.E.

    1992-01-01

    This paper reports on the D-D and D-T secondary reactions in D- 3 He reactors which provide the neutron source term for most of the radioactivity produced in the structure of the reactor. radionuclides are produced as a result of neutron interactions with their parent nuclides. The amount of activity produced by any radionuclide depends on the number of its parent atoms present at any given time. One approach to account for the activity induced by both neutron sources in any activation analysis is to add their individual contributions. Performing two separate calculations for the D-D and D-T neutron flux components and adding their contributions yields conservative results due to underestimating the destruction of the parent atoms. The overestimation is more pronounced for short and intermediate lived nuclides, long operation time, large neutron flux and large destruction cross section for the parent atoms. In the steel first wall of a typical d- 3 He reactor, adding the individual contributions of the tow neutron sources results in overestimating the activities produced by most of the radioactive isotopes of Ag, Lu, Ta, W and Re. After 30 years of reactor operation, the activity of 187 W, which is a major source of safety concern in case of an accident, is more than an order of magnitude higher than its value if the mixed neutron flux is used. The activity of 188 Re, which is an important source of offsite does in case of accidental release, is overestimated by more than a factor of two

  20. Earthquakes as collapse precursors at the Han-sur-Lesse Cave in the Belgian Ardennes

    Science.gov (United States)

    Camelbeeck, Thierry; Quinif, Yves; Verheyden, Sophie; Vanneste, Kris; Knuts, Elisabeth

    2018-05-01

    Collapse activation is an ongoing process in the evolution of karstic networks related to the weakening of cave vaults. Because collapses are infrequent, few have been directly observed, making it challenging to evaluate the role of external processes in their initiation and triggering. Here, we study the two most recent collapses in the Dôme chamber of the Han-sur-Lesse Cave (Belgian Ardenne) that occurred on or shortly after 3rd December 1828 and between the 13th and 14th of March 1984. Because of the low probability that the two earthquakes that generated the strongest ground motions in Han-sur-Lesse since 1800, on 23rd February 1828 (Mw = 5.1 in Central Belgium) and 8th November 1983 (Mw = 4.8 in Liège) occurred by coincidence less than one year before these collapses, we suggest that the collapses are related to these earthquakes. We argue that the earthquakes accelerated the cave vault instability, leading to the collapses by the action of other factors weakening the host rock. In particular, the 1828 collapse was likely triggered by a smaller Mw = 4.2 nearby earthquake. The 1984 collapse followed two months of heavy rainfall that would have increased water infiltration and pressure in the rock mass favoring destabilization of the cave ceiling. Lamina counting of a stalagmite growing on the 1828 debris dates the collapse at 1826 ± 9 CE, demonstrating the possibility of dating previous collapses with a few years of uncertainty. Furthermore, our study opens new perspectives for studying collapses and their chronology both in the Han-sur-Lesse Cave and in other karstic networks. We suggest that earthquake activity could play a stronger role than previously thought in initiating cave collapses.