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Sample records for belgian reactor 3

  1. Legal claims against Belgian reactors?

    International Nuclear Information System (INIS)

    Raetzke, Christian

    2016-01-01

    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. 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.)

  7. 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)

  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. 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...

  10. 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...

  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. 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)

  13. 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

  14. 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.

  15. 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)

  16. 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)

  17. 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.)

  18. 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)

  19. 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

  20. 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.

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

    Directory of Open Access Journals (Sweden)

    Lucas Dewalque

    2014-01-01

    Full Text Available 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.

  2. 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.

  3. 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

  4. 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

  5. 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)

  6. 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)

  7. 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)

  8. 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.

  9. 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

  10. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    1999-10-01

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

  12. 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)

  13. 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

  14. 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)

  15. 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)

  16. 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

  17. 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)

  18. 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

  19. Fracture toughness of reactor grade graphites, 3

    International Nuclear Information System (INIS)

    Sato, Sennosuke; Awaji, Hideo; Akuzawa, Hironobu; Kon, Junichi.

    1979-01-01

    In our recent papers, we presented a new technique for determining the thermal shock fracture toughness, using a disk specimen with an edge crack. The thermal shock fracture toughness is defined as K sub( ic)k/Eα(K sub( ic) standing for fracture toughness; k for thermal conductivity; E for Young's modulus; α for thermal expansion coefficient) and it can be determined en bloc by measuring the threshold electric power of the arc discharge heating produced when an edge crack propagates in the disk. The value obtained is the fracture toughness corresponding to the thermal shock resistance defined as σk/Eα (σ standing for tensile strength). The experimental data shown in the following discussion concern themselves with four kinds of reactor grade graphite and some varieties of electrode graphite. (author)

  20. 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.)

  1. 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

  2. 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

  3. 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

  4. The Belgian National Seismic Monitoring Network

    Science.gov (United States)

    van Camp, M.; Lecocq, T.; Vanneste, K.; Rapagnani, G.; Martin, H.; Devos, F.; Bukasa, B.; Hendrickx, M.; Collin, F.; Camelbeeck, T.

    2009-04-01

    The Royal Observatory of Belgium (ROB) is responsible for the seismic activity monitoring in Belgium. For this purpose the ROB operates a network of 24 seismic stations. In addition 18 accelerographs have been installed since 2001 in the most seismic active zones. Seismometers allow detecting and localizing any earthquake of magnitude larger than 1.0 in Belgium and surrounding regions. The location of the accelerometric stations is chosen in function of the type of sub-soil and in some places in function of the nearness of important infrastructures as well. Seven seismic stations are now sending their data in real time to the Observatory (in Uccle) using ADSL lines. This will be increased in a near future. Among them 3 broad-band stations are also sending data to the ORFEUS and IRIS data centres. IRIS also receives data from the Belgian superconducting gravimeter. In addition, in 2010, a broadband borehole seismometer is to be installed at the Princess Elizabeth Antarctic station (71°57' S - 23°20' E), on the bedrock, 180 km away from the coastline. Recently a low-cost seismic alert system was developed for the Belgian territory, based on the connection flow on the ROB website (http://www.seismology.be), in parallel to an automatic control of the "Did you feel it ?" macroseismic inquiries, implemented in 2002. The alert is then confirmed at the latest by the seismic signals from five seismic stations that appear on the website with a delay of more or less ten minutes. It was successfully tested during the earthquake sequence that has been observed in the region at the southwest of Brussels since July 2008.

  5. 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)

  6. Belgian Federalism after the Sixth State Reform

    Directory of Open Access Journals (Sweden)

    Goossens Jurgen

    2015-11-01

    Full Text Available This paper highlights the most important institutional evolutions of Belgian federalism stemming from the implementation of the sixth state reform (2012-2014. This reform inter alia included a transfer of powers worth 20 billion euros from the federal level to the level of the federated states, a profound reform of the Senate, and a substantial increase in fiscal autonomy for the regions. This contribution critically analyses the current state of Belgian federalism. Although the sixth state reform realized important and long-awaited changes, further evolutions are to be expected. Since the Belgian state model has reached its limits with regard to complexity and creativity, politicians and academics should begin to reflect on the seventh state reform with the aim of increasing the transparency of the current Belgian institutional labyrinth.

  7. 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.)

  8. 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.)

  9. 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)

  10. 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)

  11. 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

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

    International Nuclear Information System (INIS)

    Thomas, G.F.

    1983-11-01

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

  13. Pulsed neutron experiments on the graphite reactor EDF 3

    International Nuclear Information System (INIS)

    Fuster, S.; Tarabella, A.; Tellier, H.

    1967-04-01

    The use of pulsed neutron technique on the EDF 3 reactor has made possible to calibrate a rather great number of control elements. This technique which was tested on EDF 2 has given good results of which the interpretation is difficult due to the reactor size. As some processes ( geometrical location of sources and counters) were currently used to avoid this kind of difficulties another method for interpreting the results is proposed. The measurements are in good agreement with other experimental results obtained with the air-poisoning techniques and generally confirm the data obtained by preliminary computations. The experimental material was satisfying and seems particularly adapted to the problem to be solved. (author) [fr

  14. 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

  15. 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

  16. 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)

  17. 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

  18. 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)

  19. 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

  20. 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

  1. 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

  2. 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.)

  3. Reactor

    International Nuclear Information System (INIS)

    Evans, R.M.

    1976-01-01

    Disclosed is a neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch. 1 claim, 16 figures

  4. 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.

  5. 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)

  6. 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

  7. 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)

  8. Cultural Differences in Complex Addition: Efficient Chinese versus Adaptive Belgians and Canadians

    Science.gov (United States)

    Imbo, Ineke; LeFevre, Jo-Anne

    2009-01-01

    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…

  9. 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)

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

    International Nuclear Information System (INIS)

    Boulinier, C.; Faurot, P.; Sagot, M.; Teste du Bailler, A.

    1961-01-01

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

  11. 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.

  12. 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

  13. 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.)

  14. 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

  15. 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

  16. New Belgian position on containment leakage testing

    International Nuclear Information System (INIS)

    De Boeck, B.

    1986-01-01

    The containment leakage testing requirements (up to now 10 CFR 50 App. J) have recently been reevaluated in Belgium. The criterion for type A tests at half the accident pressure has been strengthened, but the periodicity has been relaxed. New overall leakage test at very low overpressure have been required after each extended cold shutdown period. A few items of the procedure for type A tests have been modified. It is felt that the new requirements improve the safety but also lower the burden of the containment leakage tests. Experimental results from tests performed in 1985 at Tihange reactor and Doel unit 3 are presented

  17. 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.

  18. 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.

  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. 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)

  2. Development of a model of the Danish research reactor DR3

    International Nuclear Information System (INIS)

    Nonboel, E.

    1985-12-01

    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 with heavy water to increase the thermal flux in the center of the reactor. (author)

  3. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. 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

  6. 3D alpha modes of a nuclear power reactor

    International Nuclear Information System (INIS)

    Verdu, Gumersindo; Ginestar, Damian; Roman, Jose; Vidal, Vicente

    2010-01-01

    This work is focused on the determination of the dominant α-modes corresponding to three time-eigenvalue problems. A nodal collocation method is used for the discretization of the differential eigenvalue problem and a restarted Krylov method has been used to solve the algebraic eigenvalue problem. Also, the relation between the λ-modes and α-modes is studied through the concept of prompt neutron generation time, which has several interpretations. To present numerical results of the different α-modes, we have studied a benchmark reactor and a configuration associated with Ringhals NPP. (author)

  7. 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)

  8. Enhancing earwig populations in Belgian orchards.

    Science.gov (United States)

    Gobin, B; Marien, A; Davis, S; Leirs, H

    2006-01-01

    Earwigs are key generalist predators to a variety of orchard pests. However, the once held believe that earwigs damage and spoil fruits led to control strategies and eventually the loss of large earwig populations in Belgian orchards. In recent years, Integrated and Organic fruit growers have tried to re-establish earwig populations, thus far with little success. We started a study linking various components of orchard management and the earwig life history to identify potential periods in which earwigs are vulnerable and management factors hazardous to earwigs. As a first step, detailed knowledge of earwig phenology in orchards is necessary to identify vulnerable stages in the life cycle. Here we describe the first results from organic apple orchards.

  9. Focal epilepsy in the Belgian shepherd

    DEFF Research Database (Denmark)

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

    2009-01-01

    OBJECTIVES: To establish the mode of inheritance and describe the clinical features of epilepsy in the Belgian shepherd, taking the outset in an extended Danish dog family (199 individuals) of Groenendael and Tervueren with accumulated epilepsy. METHODS: Epilepsy positive individuals (living...... and deceased) were ascertained through a telephone interview using a standardised questionnaire regarding seizure history and phenomenology. Living dogs were invited to a detailed clinical evaluation. Litters more than five years of age, or where epilepsy was present in all offspring before the age of five......, 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...

  10. Treatment of strong domestic sewage in a 96 m3 UASB reactor operated at ambient temperatures: two-stage versus single-stage reactor

    NARCIS (Netherlands)

    Halalsheh, M.M.I.; Sawajneh, Z.; Zu'bi, M.; Zeeman, G.; Lier, van J.B.; Fayyad, M.; Lettinga, G.

    2005-01-01

    A 96 m(3) UASB reactor was operated for 2.5 years under different conditions to assess the feasibility of treating strong sewage (CODtot = 1531 mg/l) at ambient temperatures, with averages of 18 and 25 degreesC for winter and summer respectively. During the first year, the reactor was operated as a

  11. 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

  12. 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)

  13. Transient analysis of a typical MTR fuel type research reactor by RELAP5/MOD 3.3

    International Nuclear Information System (INIS)

    Montazeri, A.; Bousbia Salah, A.; D'Auria, F.

    2004-01-01

    In this work the Transient Analysis for two different scenarios in a typical research reactor are considered. The first case is T he Spurious opening of the safety flapper valve, following reactor trip . The second case is the same as the first one, but the corresponding trip signal is not initiated, i.e. an anticipated transient without scram. The safety flapper valve is employed for natural convection in a typical swimming pool type research reactor. In both cases it is considered that the core is in steady state operation at full power. The reactor under consideration is a pool-type MTR core with maximum thermal capacity of 1 MW. Light water moderates the neutrons and a downward forced circulation of light water cools off the core as well. The analysis is performed using advanced computational tools, i.e. RELAP5/MOD 3.3

  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. 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.)

  16. 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

  17. Isolator arrangement design and 3D solid modeling for KALIMER reactor building

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Yoo, Bong; Koo, G. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The axial loads on the isolators under upper basemat of KALIMER reactor building vary according to the weight distribution of reactor building, so the unbalanced deflection and the stresses can be occurred on upper basemat supported by isolators. For evaluating the phenomena, the axial forces on the isolators are calculated using the finite element model of reactor building and isolators. Optimal design study of isolator arrangement is performed, and the structural integrity of isolators and upper basemat is evaluated. For the concept design of KALIMER building, 3-dimensional solid models using IDEAS program are built and some design modifications for reactor building are implemented for the latest design. 11 refs., 19 figs., 7 tabs. (Author)

  18. 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

  19. 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.

  20. On alteration of reactor installation (alteration of JRR-3 reactor facilities) in Tokai Research Establishment, Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    1985-01-01

    The Nuclear Safety Commission sent the report on this case to the Prime Minister on November 29, 1984, upon having received the report from the Committee on Examination of Nuclear Reactor Safety and after the deliberation. It was judged that the applicant has the technical capability for altering the reactor installation and properly operating the reactor. Also it was judged that the safety after the alteration of reactor installation can be ensured, and there is not any obstacle in the prevention of disasters caused by contaminated substances and the reactor. The main items of investigation and deliberation were the whole reactor facilities, aseismatic design, reactor and measurement/control system, reactor shutdown system, reactivity control system and safety protection system, reactor cooling system, reactor containment facilities, fuel handling and waste disposal system, radiation protection and control facilities, experimental utilization facilities, the safety of removal works and the soundness of existing facilities, the radiation exposure dose of public around the reactor facilities, the abnormal transient change in operation, accident and so on. (Kako, I.)

  1. 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

  2. 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

  3. Reactor vessel

    NARCIS (Netherlands)

    Makkee, M.; Kapteijn, F.; Moulijn, J.A.

    1999-01-01

    A reactor vessel (1) comprises a reactor body (2) through which channels (3) are provided whose surface comprises longitudinal inwardly directed parts (4) and is provided with a catalyst (6), as well as buffer bodies (8, 12) connected to the channels (3) on both sides of the reactor body (2) and

  4. 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

  5. Detection of noroviruses in shellfish and semiprocessed fishery products from a Belgian seafood company.

    Science.gov (United States)

    Li, Dan; Stals, Ambroos; Tang, Qing-Juan; Uyttendaele, Mieke

    2014-08-01

    Shellfish have been implicated in norovirus (NoV) infection outbreaks worldwide. This study presents data obtained from various batches of shellfish and fishery products from a Belgian seafood company over a 6-month period. For the intact shellfish (oysters, mussels, and clams), 21 of 65 samples from 12 of 34 batches were positive for NoVs; 9 samples contained quantitative NoV levels at 3,300 to 14,300 genomic copies per g. For the semiprocessed fishery products (scallops and common sole rolls with scallop fragments), 29 of 36 samples from all eight batches were positive for NoVs; 17 samples contained quantitative NoV levels at 200 to 1,800 copies per g. This convenience study demonstrated the performance and robustness of the reverse transcription quantitative PCR detection and interpretation method and the added value of NoV testing in the framework of periodic control of seafood products bought internationally and distributed by a Belgian seafood processing company to Belgian food markets.

  6. 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

  7. The CO2 cooling gas for the reactors G2/G3 (leaking, analysis, activity)

    International Nuclear Information System (INIS)

    Meiffren, J.; Dupay, F.

    1965-01-01

    The main objective of this study is to publicise the data obtained during five years operation of the reactor G2 and G3 at Marcoule as far as the cooling gas is concerned, from storage of reserves up to its slow escape into the atmosphere, and including all the stages of its practical use, its chemical examination, its nuclear behaviour and its possible physicochemical transformation. This work can not only yield information about the operations carried out at Marcoule but can also provide useful suggestions for improving the sealing and for decreasing the activity of the pressurized gas circuits in reactors similar to G2/G3. (authors) [fr

  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. Dietary intake of artificial sweeteners by the Belgian population.

    Science.gov (United States)

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

    2012-01-01

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

  10. 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

  11. 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)

  12. 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.

  13. Simulation of startup transients of the SLOWPOKE-3 nuclear heating reactor

    International Nuclear Information System (INIS)

    Tseng, C.M.; Frketich, G.

    1984-07-01

    A natural circulation, pool-type nuclear reactor, called SLOWPOKE-3, is being developed at the Chalk River Nuclear Laboratories. It is intended for producing 2 MW of hot water at atmospheric pressure for space heating and perhaps for generating some electricity. A simulation has been developed to investigate the rate-of-rise of reactor power and consequences of loss-of-regulation accidents during startup. The effects of fuel temperature and of coolant density on the rate-of-rise of reactor power were examined individually and in combination. The influence of different initial pool temperatures and input reactivity rates on startup transients was also investigated. The simulation is based on the point kinetic equations for neutrons and the conservation of mass, energy and momentum for the thermal-hydraulic circuit. The equations were solved digitally on CDC CYBER 175/6600 computers via the FORSIM computer program, developed at Chalk River Nuclear Laboratories

  14. Belgian Photography: Towards a Minor Photography.

    Directory of Open Access Journals (Sweden)

    Jan Baetens

    2011-08-01

    Full Text Available

    Abstract: This article investigates Belgian photography from within the national framework. Using the notion of "banal nationalism" (Billig, it explores how even in the case of a nation widely perceived as non-existing , a nation in which the “official” national culture has completely eroded, the relationship between  photography and national culture are worth wile addressing. Furthermore, this text shows that Belgian photography, repeatedly described as having no schools, no centre, no towering figures, nor strong institutions, in short as lacking any positive identity, can not be approached in any essentializing way. In order to study the photographic production made in Belgium, new questions are needed, questions reaching beyond the traditional aesthetic questions of styles, masters, evolutions; questions that address the medium in its multiplicity and its context sensitiveness. This article tries to develop such a new set of questions by proposing a transfer of the notion of “minor literature” as it was theorized by Gilles Deleuze and Felix Guattari to the study of the photographic medium.

     

    Résumé: Le présent article examine la photographie belge du point de vue des liens entre culture et nation. En partant du concept de

  15. 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

  16. Numerical evaluation of shielding analysing procedures for the upgraded JRR-3 research reactors

    International Nuclear Information System (INIS)

    Ise, Takeharu; Maruo, Takeshi; Umeda, Kentaro.

    1986-11-01

    Description is about the numerical evaluation of the analytical parameters in the shielding analysing procedures for the upgraded JRR-3 research reactor: number of energy groups, number of spatial meshes, number of angular quadratures and the bootstraps' width used in S N calculations; dose rate attenuation curves; and radiation source. (author)

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

    DEFF Research Database (Denmark)

    2001-01-01

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

  18. 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

  19. T-3, a packaging and transportation system for irradiated US breeder reactor experiments

    International Nuclear Information System (INIS)

    Hansen, L.J.; Berg, J.D.; Berger, J.D.; Weber, E.T.; Krupar, J.J.

    1980-01-01

    Irradiation testing of fast reactor fuels and materials in EBR-II, TREAT, and other US facilities required a packaging and transportation system that would ensure that changes in the specimens were irradiation induced and not the result of transport. The T-2 cask and the development of the T-3 cask are discussed

  20. 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

  1. 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)

  2. 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)

  3. 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.

  4. 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.

  5. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jasiulevicius, A. [Kaunas University of Technology, Dept. of Thermal and Nuclear Energy, Kaunas (Lithuania)

    2001-07-01

    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.

  7. 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)

  8. 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

  9. 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

  10. 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

  11. 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)

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

    International Nuclear Information System (INIS)

    Maki, Koichi; Seki, Yasushi; Kosako, Kazuaki; Kawasaki, Hiromitsu.

    1991-03-01

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

  13. 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.

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

    International Nuclear Information System (INIS)

    1979-10-01

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

  15. 3D Neutronic Analysis in MHD Calculations at ARIES-ST Fusion Reactors Systems

    Science.gov (United States)

    Hançerliogulları, Aybaba; Cini, Mesut

    2013-10-01

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

  16. 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

  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. 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)

  19. 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.

  20. Modernity and the Belgian Congo | Fraiture | Tydskrif vir letterkunde

    African Journals Online (AJOL)

    This article will explore the intellectual context in which French-Belgian colonial writing developed from the turn of the twentieth century to the late 1930s. This period is marked by a gradual shift from evolutionism to cultural relativism. The analysis will first focus on the Tervuren colonial exhibition of 1897 and the progressive ...

  1. 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)

  2. Wabbes and the Milan Triennales: Representing the Belgian Nation

    NARCIS (Netherlands)

    Floré, F.M.W.; Ferran-Wabbes, Marie; Strauven, Iwan

    2012-01-01

    The work of the Brussels based furniture designer and interior architect Jules Wabbes played a significant role in the Belgian section of two successive Milan Triennials, that of 1957 and 1960. His sophisticated modern furniture and lightning designs were well qualified to represent the nation.

  3. 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

  4. 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

  5. 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.

  6. 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)

  7. 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.

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

    Directory of Open Access Journals (Sweden)

    Mario Carelli

    2009-01-01

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

  9. 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.

  10. 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.

  11. 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.

  12. The mesozooplankton community of the Belgian shelf (North Sea)

    Science.gov (United States)

    Van Ginderdeuren, Karl; Van Hoey, Gert; Vincx, Magda; Hostens, Kris

    2014-01-01

    This manuscript presents the mesozooplankton community structure and its spatial and temporal variabilities in the Belgian part of the North Sea (BPNS), a first thorough study on this topic in nearly 40 years. Monthly sampling campaigns at ten stations in the BPNS in 2009 and 2010 yielded a total of 137 mesozooplankton taxa (46 holoplanktonic, 50 meroplanktonic and 41 tychoplanktonic), of which nine species had never been reported in the area. Smaller neritic copepods, especially Temora longicornis and Acartia clausi, were present in all samples and dominated zooplankton densities (66%), together with the appendicularian Oikopleura dioica (10%). They were joined by high numbers of meroplanktonic echinoderm larvae (9%) in spring and summer. Based on diversity alone, the mesozooplankton could be typified as one neritic zooplankton community, due to the ubiquitous presence in time and space of the dominant copepods. Yet, these neritic species were often joined by low numbers of oceanic species that are occasionally imported with the inflow of Atlantic oceanic water in the BPNS. Based on a combination of abundance and diversity, our results indicate distinct seasonal and spatial distribution patterns in the mesozooplankton. Months with highest average densities were May, June and July, lowest densities were noted in December and January. Only limited long-term zooplankton data are available for the BPNS from the Continuous Plankton Recorder surveys or the long-term monitoring stations in the vicinity of our research area. However, our data suggest that nowadays zooplankton species appear earlier in the BPNS, comparable with other areas in the North Sea. Densities varied between 150 and 15,000 ind.m- 3, and averaged highest at midshore stations, then nearshore and offshore. This is partially comparable with the spatial patterns recorded for other ecosystem components, such as demersal fish, epibenthos and macrobenthos, of which densities peak in a stretch almost

  13. 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

  14. 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

  15. Coupled neutronics and thermal hydraulics modelling in reactor dynamics codes TRAB-3D and HEXTRAN

    International Nuclear Information System (INIS)

    Kyrki-Rajamaeki, R.; Raety, H.

    1999-01-01

    The reactor dynamics codes for transient and accident analyses inherently include the coupling of neutronics and thermal hydraulics modelling. In Finland a number of codes with 1D and 3D neutronic models have been developed, which include models also for the cooling circuits. They have been used mainly for the needs of Finnish power plants, but some of the codes have also been utilized elsewhere. The continuous validation, simultaneous development, and experiences obtained in commercial applications have considerably improved the performance and range of application of the codes. The fast operation of the codes has enabled realistic analysis of 3D core combined to a full model of the cooling circuit even in such long reactivity scenarios as ATWS. The reactor dynamics methods are developed further and new more detailed models are created for tasks related to increased safety requirements. For thermal hydraulics calculations, an accurate general flow model based on a new solution method has been developed. Although mainly intended for analysis purposes, the reactor dynamics codes also provide reference solutions for simulator applications. As computer capability increases, these more sophisticated methods can be taken into use also in simulator environments. (author)

  16. Magnet system for catalyzed D-D and D-He3 reactors

    International Nuclear Information System (INIS)

    Usher, J.L.; Powell, J.R.; Hsieh, S.Y.

    1977-01-01

    A new concept for toroidal field magnets, called the DEALS (Demountable Externally Anchored Low Stress) magnet systems, has been conceived and developed for advanced fuel reactors. This concept should result in substantial cost reductions for toroidal magnets. In addition, the conductor stress is relatively low and compressive in nature, which is very important if brittle Nb 3 Sn conductors are to be successfully used, since advanced fuel Tokamak reactors require Nb 3 Sn conductor if reasonable power densities are to be achieved. A final very important benefit of the DEALS magnet system is demountability. Failed magnet segments can probably be replaced fairly rapidly (i.e., in approximately a month) without requiring movement and/or dissassembly of the blanket/shield regions and poloidal coils. Due to the high degree of non-circularity present in the plasmas of the low-β Tokamak reactor designs which are detailed in this report, a system of poloidal field coils is necessitated to provide both shape and positional stability. In addition, the uncertainty regarding the accessibility of the high-β regimes discussed may also necessitate extensive poloidal field systems. Volt second requirements are such that they must be alleviated also by this poloidal field system

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  18. 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.)

  19. 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.)

  20. 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.

  1. 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

  2. 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.

  3. 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)

  4. 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)

  5. The GAy MEn Sex StudieS erectile dysfunction among Belgian gay men

    Directory of Open Access Journals (Sweden)

    Vansintejan J

    2013-07-01

    Full Text Available Johan Vansintejan, Jan Vandevoorde, Dirk Devroey Department of Family Medicine, Vrije Universiteit Brussel (VUB, Brussels, Belgium Aim: To determine the prevalence of erectile dysfunction (ED in a sample of the Belgian men who have sex with men (MSM population, and to assess the relevance of major predictors such as age, relationship, and education. We investigated the use of phosphodiesterase type 5 (PDE5 inhibitors among Belgian MSM. Methods: An internet-based survey on sexual behavior and sexual dysfunctions, called GAy MEn Sex StudieS (GAMESSS, was administered to MSM, aged 18 years or older, between the months of April and December 2008. The questionnaire used was a compilation of the Kinsey's Heterosexual–Homosexual Rating Scale, Erection Quality Scale (EQS, and the shortened version of the International Index of Erectile Function (IIEF-5. Results: Of the 1752 participants, 45% indicated having some problems getting an erection. In this group of MSM, 71% reported mild ED; 22% mild to moderate ED; 6% moderate ED; and 2% severe ED. Independent predictors for the presence of ED were: age (odds ratio [OR] = 1.04, P < 0.0001, having a steady relationship (OR = 0.59, P < 0.0001, frequency of sex with their partner (OR = 1.22, P < 0.0001, versatile sex role (OR = 1.58, P = 0.016, passive sex role (OR = 3.12, P < 0.0001, problems with libido (OR = 1.15, P = 0.011, ejaculation problems (OR = 1.33, P < 0.0001, and anodyspareunia (OR = 0.87, P < 0.0001. Ten percent of the Belgian MSM used a PDE5 inhibitor (age 43 ± 11 years; mean ± standard deviation and 83% of them were satisfied with the effects. "Street drugs" were used by 43% of MSM to improve ED. Conclusion: Forty-five percent of participating Belgian MSM reported some degree of ED and 10% used a PDE5 inhibitor to improve erections. Older MSM reported more ED. MSM, who were in a steady relationship or frequently had sex with a partner, reported less ED. MSM with ejaculation problems

  6. Protocol of the Belgian food consumption survey 2014: objectives, design and methods.

    Science.gov (United States)

    Bel, Sarah; Van den Abeele, Sofie; Lebacq, Thérésa; Ost, Cloë; Brocatus, Loes; Stiévenart, Charlotte; Teppers, Eveline; Tafforeau, Jean; Cuypers, Koenraad

    2016-01-01

    Dietary patterns are one of the major determinants as far as health and burden of disease is concerned. Food consumption data are essential to evaluate and develop nutrition and food safety policies. The last national food consumption survey in Belgium took place in 2004 among the Belgian population aged 15 years and older. Since dietary habits are prone to change over time a new Belgian National Food Consumption Survey (BNFCS2014) was conducted in 2014-2015. The BNFCS2014 is a cross-sectional study. A representative sample (n = 3200) of the Belgian population aged 3 to 64 years old was randomly selected from the National Population Register following a multistage stratified sampling procedure. Data collection was divided equally over the four seasons and days of the week in order to incorporate seasonal effects and day-to-day variation in food intake. Information on food intake was collected in adults with two non-consecutive 24-h dietary recalls (using the GloboDiet® software). In children food intake was collected with two non-consecutive one-day food diaries followed by a completion interview with GloboDiet. Additional data on socio-demographic characteristics, eating habits, lifestyle, food safety (at household level), physical activity and sedentary behaviour were collected with a face-to-face questionnaire using a computer-assisted personal interviewing technique. In the time between the two visits, participants were asked to complete a self-administered food frequency questionnaire and health questionnaire. Height, weight and waist circumference were measured. In addition, children and adolescents were asked to wear an accelerometer and keep a logbook for seven consecutive days to objectively measure physical activity and sedentary behaviour. The main objective of the BNFCS2014 is to evaluate the habitual food, energy and nutrient intake in the Belgian population and to compare these with recommendations from the national dietary guidelines. A second

  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. A D-He3 fusion reactor based on a dipole magnetic field

    International Nuclear Information System (INIS)

    Hasegawa, Akira; Chen, Liu; Princeton Univ., NJ

    1989-07-01

    An innovative fusion reactor suitable for D-He 3 fuel is proposed, based on a dipole magnetic field produced by a simple one-turn coil with approx lt 16 T near-field intensity. The equilibrium plasma, phase-space density satisfies ∂ cflx f 0 (μ, J, ψ)/∂ψ = O, where ψ is the flux function, has a steep enough pressure profile for an efficient fusion reaction yet is stable for low frequency instabilities to local beta exceeding unity. The semi-open field configuration is particularly suitable for D-He 3 reactions. 8 refs., 1 fig

  9. 3-D design method for welding groove and seal weld of reactor CRDM adapter

    International Nuclear Information System (INIS)

    Ma Baiyong; Wang Xiaobin; Zhu Xiaoyong

    2008-01-01

    Based on the analysis of 2 D and 3 D shapes of welding groove and seal weld of reactor CRDM adapter, four intersecting curves are defined, and a method and gist to 3 D design of adapter welding groove and seal weld is proposed. Parameterized design of adapter welding groove and seal weld has been realized using UG software, and the main factors which affect the welding section areas have been analyzed. Compared with the measurement, the error of weld section area of each adapter created by spline fitting method is less than 0.8%. (authors)

  10. 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

  11. 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

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

    International Nuclear Information System (INIS)

    Oscar, D.S.; Attaway, S.W.; Miller, J.D.

    1991-08-01

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

  13. 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.

  14. Assessment of marine debris on the Belgian Continental Shelf.

    Science.gov (United States)

    Van Cauwenberghe, Lisbeth; Claessens, Michiel; Vandegehuchte, Michiel B; Mees, Jan; Janssen, Colin R

    2013-08-15

    A comprehensive assessment of marine litter in three environmental compartments of Belgian coastal waters was performed. Abundance, weight and composition of marine debris, including microplastics, was assessed by performing beach, sea surface and seafloor monitoring campaigns during two consecutive years. Plastic items were the dominant type of macrodebris recorded: over 95% of debris present in the three sampled marine compartments were plastic. In general, concentrations of macrodebris were quite high. Especially the number of beached debris reached very high levels: on average 6429±6767 items per 100 m were recorded. Microplastic concentrations were determined to assess overall abundance in the different marine compartments of the Belgian Continental Shelf. In terms of weight, macrodebris still dominates the pollution of beaches, but in the water column and in the seafloor microplastics appear to be of higher importance: here, microplastic weight is approximately 100 times and 400 times higher, respectively, than macrodebris weight. Copyright © 2013 Elsevier Ltd. All rights reserved.

  15. 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

  16. 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

  17. 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

  18. First steps towards a Zn/Co(III)sep-driven P450 BM3 reactor.

    Science.gov (United States)

    Zhao, Liqing; Güven, Güray; Li, Yin; Schwaneberg, Ulrich

    2011-08-01

    Cytochrome P450s are synthetically attractive hydroxylation catalysts. For cell-free applications, a constant supply of NAD(P)H can be very costly. Mediators such as Zn/Co(III)sep can be an alternative cofactor system to NAD(P)H. Several mutants of cytochrome P450 BM3 with improved electron transfer rate to Zn/Co(III)sep have been obtained in our group. P450 BM3 M7 (F87A V281G M354S R471C A1011T S1016G Q1022R) was immobilized on DEAE-650S, further entrapped with k-carrageenan together with zinc dust which function as electron source and catalase which removes produced hydrogen peroxide instantly. Immobilized P450 BM3 M7 were treated with 0.05% (v/v) glutaraldehyde to enhance operational stability. P450 BM3 M7 retained around 76% of its activity and conversions stayed above 80% in 10 batch cycles, indicating a high stability of immobilized P450 BM3 M7. To explore the synthetic potential, a small-scale bioreactor was developed to investigate the stability and efficiencies of P450 BM3 M9 (R47F F87A M238K V281G M354S D363H W575C A595T). P450 BM3 M9 was used for the continuous conversion of 3-phenoxytoluene in a plug flow reactor (PFR) since P450 BM3 M9 has a 3-fold higher activity for 3-phenoxytoluene compared to P450 BM3 M7 which was used for optimizing immobilization conditions with the highest activity for 12-pNCA assay. The reactor could be operated for 5 days with total turnover numbers (TTNs) over 2,000.

  19. 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...

  20. Bronchiolitis management by the Belgian paediatrician: discrepancies between evidence-based medicine and practice.

    Science.gov (United States)

    de Bilderling, G; Bodart, E

    2003-01-01

    There is no therapy with proven effect on bronchiolitis outcome. This leads to large variations in its management between different countries. In order to evaluate how this disease was managed in our country, a questionnaire was sent to all Belgian paediatricians. With a response rate above 40% of active paediatricians, we found that bronchodilators (74.7% vs. 77.2%), physiotherapy (76.2% vs. 85.6%) and antibiotics (63.8% vs. 74.4%) were still largely prescribed in in- and outpatient settings respectively, corticosteroids (orally or intravenously) being prescribed more often in hospitals (54.3% vs. 17.0%). There were also some variations in admission criteria (minimal age 2 months (75%) to 6 months (8.2%), lower limit for oxygen saturation: 90% (21.5%) to 95% (26.5%)) and 1/3 of the respondents did not use pulse oxymetry to evaluate hypoxaemia in infants with bronchiolitis. Logistic regression analyses allowed us to identify patterns of prescription based on age, type and level of activity and language. Many therapies with no proven effect are still used by Belgian paediatricians to treat children with bronchiolitis. Based on these results, we believe that publishing national guidelines will allow a reduction in the cost associated with this disease.

  1. 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.

  2. Design verification and 3D visualization of a nuclear transmutation reactor - PEACER

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C. H.; Hwang, I. S.; Suh, K. Y.; Kim, M. H.; Lee, K. J.; Shin, J. G.; Yi, K. W.; Jeong, S. H.; Lee, H. W. [Seoul National Univ., Seoul (Korea, Republic of)

    2004-07-01

    The conceptual design of a nuclear transmutation energy system named as the PEACER (Proliferation-resistant, Environment-friendly, Accident-tolerant, Continuable-energy, and Economical Reactor) has been verified through independent assessment. The lead-bismuth (Pb-Bi) cooled fast reactor core with a thermal trap region can burn transuranic (TRU) wastes and tabilize Tc-99 and I-129 produced from the water reactors. Pyrochemical partitioning processes are designed to achieve an overall decontamination factor of 10{sup 5} on the TRU elements by utilizing electrorefining and reductive salt purification. Proliferation-resistance is embedded using the technical barrier of the pyrochemical processing and the institutional barrier through multi-national operation. In summary, the original design is verified to have sound basis. Several issues, however, are identified that requires design improvements. A design shakedown effort has been initiated to address the needs. With the design improvement, all the PEACER goals can be met with no fundamental difficulties. An innovative 3D design tool has been developed by interfacing commercial 3D tools including CATIA and, VRML with nuclear solver codes. 3D visualizations have been customized for PEACER transparency on both geometry and functional performances. The new tool, designated as PEACER-VIEWER, can be used to enhance the proliferation resistance, accident-tolerance and economy of the PEACER design. In order to verify the expected benefit of PEACER design, the corrosion resistance and natural circulation capability are going to be demonstrated at a new large scale loop designated as HELIOS (Heavy Eutectic Liquid metal Loop for Investigation of Operability and Safety) that is being constructed at NUTRECK(Nuclear TRansmutation Energy research Center of Korea)

  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. 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

  5. U3O8 from resin for powder-metallurgy fabrication of reactor fuel

    International Nuclear Information System (INIS)

    Mosley, W.C. Jr.

    1982-01-01

    A study was conducted to determine if cation exchange resin could be used to produce U 3 O 8 powder suitable for use in powder metallurgy fabrication of fuel tubes for production reactors at the Savannah River Plant (SRP). U 3 O 8 powders have been produced from three cation exchange resins: DOWEX (Dow Chemical Co.) 50W, AG (Bio-Rad Laboratories, Richmond, California) MP-50, and Bio-REX (Bio-Rad Laboratories, Richmond, California). This study included characterization of the thermal decomposition of uranium-loaded resins, measurement of properties of resin-based U 3 O 8 powders, and metallographic examination of U 3 O 8 -Al cores in extruded fuel tubes. Results to date show that AG MP-50 appears to be the best resin for producing U 3 O 8 powder suitable for the PM process. 20 figures

  6. Belgium’s expansionist history between 1870 and 1930: imperialism and the globalisation of Belgian business

    OpenAIRE

    Abbeloos, Jan-Frederik

    2008-01-01

    This chapter considers if and how the political action of imperialism and the globalisation of business influenced each other in Belgium between 1870 and 1930. In addition to the role that Belgian King Leopold II played in the territorial partition of Africa and the opening up of China, the period sees a growing amount of capital and industrial know-how from Belgium being invested in markets outside Europe. Before World War I, the globalisation of Belgian business and Belgian imperialism oper...

  7. 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.)

  8. 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.

  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. 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)

  11. 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)

  12. 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)

  13. 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)

  14. Belgian surveillance plans to assess changes in Salmonella prevalence in meat at different production stages.

    Science.gov (United States)

    Ghafir, Yasmine; China, Bernard; Korsak, Nicolas; Dierick, Katelijne; Collard, Jean-Marc; Godard, Claudine; De Zutter, Lieven; Daube, Georges

    2005-11-01

    From 1997 to 1999, the prevalence of Salmonella was assessed at different stages through the pork, poultry, and beef meat production chains. Different dilutions of the initial sample suspension were analyzed to provide a semiquantitative evaluation of Salmonella contamination and to determine the most representative dilution necessary to detect a reduction in prevalence. An average of 300 samples for each type of meat were analyzed. According to Fisher's exact test, the dilution to be used to detect a reduction in prevalence was chosen based on an initial prevalence of 20 to 26%. Based on this introductory study, a new sampling plan representative of the nationwide Belgian meat production process was used from 2000 through to 2003. This study confirmed the consistently high rate and level of contamination of poultry meat: broiler and layer carcasses were the most contaminated samples followed by broiler fillets and poultry meat preparations. A constant and significant decrease in Salmonella prevalence was observed for pork carcasses, trimmings, and minced meat and for beef minced meat. Less than 3% of beef carcasses and trimming samples were positive for Salmonella. The Belgian plan, as utilized from 2000 to 2003, was suitable for monitoring of zoonoses because the sampling plan was representative of nationwide production processes, covered all periods of the year, and was executed by trained samplers and the analyses were carried out by recognized laboratories using an identical analytical method.

  15. 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)

  16. D-3He tokamak reactor with inductive a.c. and unidirectional current operation

    International Nuclear Information System (INIS)

    Mitarai, O.

    1995-01-01

    We propose inductively operated D- 3 He tokamak reactors (R=9.5m, a=2.8 and 3.2m, B t =10 and 7.5T, and I p =53.6 and 52.5MA respectively) with an a.c. or unidirectional current (UDC) mode, operated in the first stability regime, with a small current drive power of about 500kW in the plasma side, and about 2MW in the transformer side. The high electron temperature T e (0)∼62.5keV together with a high bootstrap current fraction greater than 70% decreases the plasma resistance (to less than 1nΩ) and the loop voltage (less than 10mV range) and then prolongs the pulse length to more than 15h. The ignition characteristics of a D- 3 He tokamak reactor are analyzed by the operation path method on the P ht τ E 2 -T plane and POPCON for different confinement scalings. ((orig.))

  17. 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)

  18. 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.

  19. Prevalence and molecular typing of Coxiella burnetii in bulk tank milk in Belgian dairy goats, 2009-2013.

    Science.gov (United States)

    Boarbi, Samira; Mori, Marcella; Rousset, Elodie; Sidi-Boumedine, Karim; Van Esbroeck, Marjan; Fretin, David

    2014-05-14

    Q fever, a worldwide zoonosis, is an arousing public health concern in many countries since the recent Dutch outbreak. An emerging C. burnetii clone, genotype CbNL01, was identified as responsible for the Dutch human Q fever cluster cases. Since 2009, Q fever surveillance in the goat industry was implemented by the Belgian authorities. The herd prevalence (December 2009-March 2013) ranged between 6.3 and 12.1%. Genotypic analysis highlighted the molecular diversity of the Belgian strains from goats and identified an emerging CbNL01-like genotype. This follow-up allowed the description of shedding profiles in positive farms which was either continuous (type I) and associated to the CbNL01-like genotype; or intermittent (type II) and linked to other genotypes. Despite the circulation of a CbNL01-like strain, the number of notified Belgian human cases was very low. The mandatory vaccination (in June 2011) on positive dairy goat farms in Belgium, contributed to a decrease in shedding. Copyright © 2014 Elsevier B.V. All rights reserved.

  20. 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.)

  1. 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

  2. 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.

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

    Science.gov (United States)

    Damian, F.; Brun, E.

    2014-06-01

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

  4. 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

  5. Loss estimation of TWDEC for a D-{sup 3}He FRC fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, M. [Institute of Engineering Mechanics, Univ. of Tsukuba, Tsukuba, Ibaraki (Japan); Kudo, T.; Yamane, T.; Shimizu, Y.; Tomita, Y.; Momota, H.

    1999-04-01

    Loss mechanisms of TWDEC (Traveling Wave Direct Energy Converter) for a D-{sup 3}He fueled FRC fusion reactor are studied with a numerical analysis. (1) The self-excitation of traveling wave has been attained by the one-dimensional analysis, which shows the conversion efficiency is about 70 to 73%. (2) Axisymmetric two-dimensional analyses have shown that the loss related to the two-dimensional effect without collision with grids is about 5%. (3) The loss related to the collision with grids is the most important, being about 11%. (4) Effects of secondary electrons produced by the collision with grids are rather small and sub-rings can suppress effects of secondary electrons. (5) The loss related to the leaked 3.6 MeV {sup 4}He is about 45 to 60% of the total energy of leaked {sup 4}He. (6) The gross efficiency of TWDEC is estimated over 60%. (author)

  6. 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

  7. 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)

  8. 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

  9. 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

  10. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

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

  11. Evidence for association between the HLA-DQA locus and abdominal aortic aneurysms in the Belgian population: a case control study

    Directory of Open Access Journals (Sweden)

    Sakalihasan Natzi

    2006-07-01

    Full Text Available Abstract Background Chronic inflammation and autoimmunity likely contribute to the pathogenesis of abdominal aortic aneurysms (AAAs. The aim of this study was to investigate the role of autoimmunity in the etiology of AAAs using a genetic association study approach with HLA polymorphisms. Methods HLA-DQA1, -DQB1, -DRB1 and -DRB3-5 alleles were determined in 387 AAA cases (180 Belgian and 207 Canadian and 426 controls (269 Belgian and 157 Canadian by a PCR and single-strand oligonucleotide probe hybridization assay. Results We observed a potential association with the HLA-DQA1 locus among Belgian males (empirical p = 0.027, asymptotic p = 0.071. Specifically, there was a significant difference in the HLA-DQA1*0102 allele frequencies between AAA cases (67/322 alleles, 20.8% and controls (44/356 alleles, 12.4% in Belgian males (empirical p = 0.019, asymptotic p = 0.003. In haplotype analyses, marginally significant association was found between AAA and haplotype HLA-DQA1-DRB1 (p = 0.049 with global score statistics and p = 0.002 with haplotype-specific score statistics. Conclusion This study showed potential evidence that the HLA-DQA1 locus harbors a genetic risk factor for AAAs suggesting that autoimmunity plays a role in the pathogenesis of AAAs.

  12. 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.

  13. 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

  14. 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)

  15. 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

  16. 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

  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. 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.

  19. 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)

  20. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Gurban, I.; Laaksoharju, M. [INTERA KB, Sollentuna (Sweden); Made, B.; Ledoux, E. [Ecole des Mines, Paris (France)

    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

  2. Cloning of a vacuolar invertase from Belgian endive leaves (Cichorium intybus).

    Science.gov (United States)

    Van Den Ende, Wim; Michiels, An; Le Roy, Katrien; Van Laere, André

    2002-08-01

    Although a lot of vacuolar invertase (EC 3.2.1.26) cDNAs are available from a diversity of plant species, up to now no sequence information is available on invertases from any dicot fructan-containing species. Therefore, we describe the cloning of vacuolar acid invertase cDNA from etiolated Belgian endive leaves (Cichorium intybus L. var. foliosum cv. Flash), formed throughout the forcing process of the witloof chicory roots. Full-length cDNA was obtained by a combination of RT-PCR, PCR and 5'- and 3' RACE RT-PCR, starting with primers based on conserved amino acid sequences. The cloned chicory acid invertase groups together with vacuolar type invertases and fructan biosynthetic enzymes. A putative role for vacuolar type invertases in fructan synthesizing plants is discussed.

  3. 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.

  4. 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

  5. Modeling and Analysis of a Lunar Space Reactor with the Computer Code Relap5-3D/Athena

    International Nuclear Information System (INIS)

    Carbajo, Juan J.; Qualls, A.L.

    2008-01-01

    The transient analysis 3-dimensional (3-D) computer code RELAP5-3D/ATHENA has been employed to model and analyze a space reactor of 180 kW(thermal), 40 kW (net, electrical) with eight Stirling engines (SEs). Each SE will generate over 6 kWe; the excess power will be needed for the pumps and other power management devices. The reactor will be cooled by NaK (a eutectic mixture of sodium and potassium which is liquid at ambient temperature). This space reactor is intended to be deployed over the surface of the Moon or Mars. The reactor operating life will be 8 to 10 years. The RELAP5-3D/ATHENA code is being developed and maintained by Idaho National Laboratory. The code can employ a variety of coolants in addition to water, the original coolant employed with early versions of the code. The code can also use 3-D volumes and 3-D junctions, thus allowing for more realistic representation of complex geometries. A combination of 3-D and 1-D volumes is employed in this study. The space reactor model consists of a primary loop and two secondary loops connected by two heat exchangers (HXs). Each secondary loop provides heat to four SEs. The primary loop includes the nuclear reactor with the lower and upper plena, the core with 85 fuel pins, and two vertical heat exchangers (HX). The maximum coolant temperature of the primary loop is 900 K. The secondary loops also employ NaK as a coolant at a maximum temperature of 877 K. The SEs heads are at a temperature of 800 K and the cold sinks are at a temperature of ∼400 K. Two radiators will be employed to remove heat from the SEs. The SE HXs surrounding the SE heads are of annular design and have been modeled using 3-D volumes. These 3-D models have been used to improve the HX design by optimizing the flows of coolant and maximizing the heat transferred to the SE heads. The transients analyzed include failure of one or more Stirling engines, trip of the reactor pump, and trips of the secondary loop pumps feeding the HXs of the

  6. 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)

  7. 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

  8. 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)

  9. Effect of HCO3- concentration on anammox nitrogen removal rate in a moving bed biofilm reactor.

    Science.gov (United States)

    Zekker, Ivar; Rikmann, Ergo; Tenno, Toomas; Vabamäe, Priit; Kroon, Kristel; Loorits, Liis; Saluste, Alar; Tenno, Taavo

    2012-01-01

    Anammox biomass enriched in a moving bed biofilm reactor (MBBR) fed by actual sewage sludge reject water and synthetically added NO2- was used to study the total nitrogen (TN) removal rate of the anammox process depending on bicarbonate (HCO3-) concentration. MBBR performance resulted in the maximum TN removal rate of 1100 g N m(-3) d(-1) when the optimum HCO3- concentration (910 mg L(-1)) was used. The average reaction ratio of NO2- removal, NO3- production and NH4+ removal were 1.18/0.20/1. When the HCO3- concentration was increased to 1760mg L(-1) the TN removal rate diminished to 270 g N m(-3) d(-1). The process recovered from bicarbonate inhibition within 1 week. The batch tests performed with biomass taken from the MBBR showed that for the HCO3- concentration of 615 mg L(-1) the TN removal rate was 3.3 mg N L(-1) h(-1), whereas for both lower (120 mg L(-1)) and higher (5750 mg L(-1)) HCO3- concentrations the TN removal rates were 2.3 (+/- 0.15) and 1.6 (+/- 0.12) mg N L(-1) d(-1), respectively. PCR and DGGE analyses resulted in the detection of uncultured Planctomycetales bacterium clone P4 and, surprisingly, low-oxygen-tolerant aerobic ammonia oxidizers. The ability of anammox bacteria for mixotrophy was established by diminished amounts of nitrate produced when comparing the experiments with an organic carbon source and an inorganic carbon source.

  10. 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

  11. 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.

  12. 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)

  13. Technical improvements in 19th century Belgian window glass production

    Science.gov (United States)

    Lauriks, Leen; Collette, Quentin; Wouters, Ine; Belis, Jan

    Glass was used since the Roman age in the building envelope, but it became widely applied together with iron since the 19th century. Belgium was a major producer of window glass during the nineteenth century and the majority of the produced window glass was exported all over the world. Investigating the literature on the development of 19th century Belgian window glass production is therefore internationally relevant. In the 17th century, wood was replaced as a fuel by coal. In the 19th century, the regenerative tank furnace applied gas as a fuel in a continuous glass production process. The advantages were a clean production, a more constant and higher temperature in the furnace and a fuel saving. The French chemist Nicolas Leblanc (1787-1793) and later the Belgian chemist Ernest Solvay (1863) invented processes to produce alkali out of common salt. The artificial soda ash improved the quality and aesthetics of the glass plates. During the 19th century, the glass production was industrialized, influencing the operation of furnaces, the improvement of raw materials as well as the applied energy sources. Although the production process was industrialized, glassblowing was still the work of an individual. By improving his work tools, he was able to create larger glass plates. The developments in the annealing process followed this evolution. The industry had to wait until the invention of the drawn glass in the beginning of the 20th century to fully industrialise the window glass manufacture process.

  14. Multiple sclerosis in Belgian children: A multicentre retrospective study.

    Science.gov (United States)

    Verhelst, Helene; De Waele, Liesbeth; Deconinck, Nicolas; Ceulemans, Berten; Willekens, Barbara; Van Coster, Rudy

    2017-03-01

    Although the diagnosis of multiple sclerosis (MS) in the paediatric population remains challenging, paediatric-onset MS is increasingly recognized worldwide. We report on the clinical and biochemical features of a Belgian multicentre cohort of paediatric MS patients in a national retrospective descriptive study. Twenty one paediatric MS patients from four Belgian University Hospitals were included. In nine patients, onset of MS was before the age of ten years which makes the study cohort of special interest. We report a higher incidence of acute disseminated encephalomyelitis (ADEM)-like first MS attacks and an overall higher proportion of polysymptomatic episodes than in adult and most paediatric cohorts reported in the literature. The clinical presentation in our cohort was rather severe with high median EDSS-score during the first clinical manifestation and barely more than half of our study patients showing full recovery after their first clinical manifestation. Also, a significant proportion of children in our cohort has severe disease progression despite disease modifying therapy and 9.5% of patients showed transition to secondary progressive multiple sclerosis during adolescence. An early and correct diagnosis of paediatric MS is essential to start early adequate treatment. As illustrated by our study cohort, current treatment options in childhood are unsatisfactory. Copyright © 2016 European Paediatric Neurology Society. Published by Elsevier Ltd. All rights reserved.

  15. 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.

  16. 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

  17. 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

  18. Tritium production in He-3 gas cells immersed in the Tokamak Fusion Test Reactor neutron field

    Energy Technology Data Exchange (ETDEWEB)

    Jassby, D.L.; Gentile, C.A.; Ascione, G.; Kugel, H.W.; Roquemore, A.L. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Kumar, A. [University of California, Los Angeles, California 90024 (United States)

    1999-01-01

    Tritium generated in an external cell by the reaction {sup 3}He(n,p)T can be used as a gauge of long-term fusion neutron production, because of the 12-year half-life of T and the relative ease of measuring the T content either by sampling or from the saturation current of the cell when operated as an ionization chamber. Two high-pressure {sup 3}He gas cells enclosed in polyethylene neutron moderators were exposed to Tokamak Fusion Test Reactor (TFTR) neutrons during high-power D{endash}T operation. The tritium produced in the cells was assayed by the Princeton Differential Atmospheric Tritium Sampler. The measured tritium generated per 10{sup 19} fusion neutrons was 510 pCi/cc at 2.3 m from the TFTR vessel and 1.3 m below the midplane, and 2020 pCi/cc at 1.0 m from the TFTR vessel in the midplane. Combining these results with previous measurements at a third location, we found 0.11 to 0.23 triton produced per neutron incident on the projected cell cross section, with an asymptotic local tritium breeding ratio of 0.32. {copyright} {ital 1999 American Institute of Physics.}

  19. 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.

  20. 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...

  1. 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...

  2. 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)

  3. Determining Reactor Neutrino Flux

    OpenAIRE

    Cao, Jun

    2011-01-01

    Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understa...

  4. 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.

  5. 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

  6. Modeling of a Pilot-Scale Fixed-Bed Reactor for Dehydration of 2,3-Butanediol to 1,3-Butadiene and Methyl Ethyl Ketone

    Directory of Open Access Journals (Sweden)

    Daesung Song

    2018-02-01

    Full Text Available A 1D heterogeneous reactor model accounting for interfacial and intra-particle gradients was developed to simulate the dehydration of 2,3-Butanediol (2,3-BDO to 1,3-Butadiene (1,3-BD and Methyl Ethyl Ketone (MEK over an amorphous calcium phosphate (a-CP catalyst in a pilot-scale fixed-bed reactor. The developed model was validated with experimental data in terms of a fluid temperature profile along with the length of the catalyst bed, 2,3-BDO conversion, and selectivity for the major products, 1,3-BD and MEK, at the outlet of the reactor. The fluid temperature profile obtained from the model along the length of the catalyst bed coincides satisfactorily with the experimental observations. The difference between the experimental data and the 1D heterogeneous reactor model prediction for 2,3-BDO conversion and selectivity of 1,3-BD and MEK were 0.1%, 9 wt %, and 2 wt %, respectively. In addition, valuable insights related to the feeding system of a commercial-scale plant were made through troubleshooting of the pilot tests. Notably, if the feed including only 2,3-BDO and furnaces that increase the temperature of the feed to the reaction temperature were used in a commercial plant, the feeding system could not be operational because of the presence of heavy chemicals considered oligomers of 2,3-BDO.

  7. 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)

  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. Modeling of neutron diffractometry facility of Tehran Research Reactor using Vitess 3.3a and MCNPX codes

    Directory of Open Access Journals (Sweden)

    Z. Gholamzadeh

    2018-02-01

    Full Text Available The neutron powder diffractometer (NPD is used to study a variety of technologically important and scientifically driven materials such as superconductors, multiferroics, catalysts, alloys, ceramics, cements, colossal magnetoresistance perovskites, magnets, thermoelectrics, zeolites, pharmaceuticals, etc. Monte Carlo–based codes are powerful tools to evaluate the neutronic behavior of the NPD. In the present study, MCNPX 2.6.0 and Vitess 3.3a codes were applied to simulate NPD facilities, which could be equipped with different optic devices such as pyrolytic graphite or neutron chopper. So, the Monte Carlo–based codes were used to simulate the NPD facility of the 5 MW Tehran Research Reactor. The simulation results were compared to the experimental data. The theoretical results showed good conformity to experimental data, which indicates acceptable performance of the Vitess 3.3a code in the neutron optic section of calculations. Another extracted result of this work shows that application of neutron chopper instead of monochromator could be efficient to keep neutron flux intensity higher than 106 n/s/cm2 at sample position.

  10. The functioning of the reactors G2-G3 at Marcoule and E.D.F. 1

    International Nuclear Information System (INIS)

    Boussard, R.; Conte, F.

    1964-01-01

    After resuming briefly the characteristics of the installations G2-G3 at Marcoule and EDF 1 at Chinon, the authors review the main aspects of the tests, the starting and the exploitation of these reactors. Among the various points examined, particular emphasis is given to the devices of original nature such as tubular fuel elements, flattening of the neutron flux by stuffing, behaviour of the reactor tanks and the cooling circuits, the blowers, unloading devices, regulation and functioning of the informations. This analysis deals equally with the performances obtained and the difficulties and the various incidents experienced during the initial starting period. Among the more interesting results, the progressive increase in the power of the Marcoule reactors is mentioned, obtained through a better knowledge of the parameters covering the functioning of the reactors such as the distribution of the flux and the temperatures etc... acquired during the course of the exploitation of the reactor. The conclusion reached by the authors is that the experience gained on these installations has shown: - that during an initial period, adjustments became necessary, all of which turned out to be possible, - that an analysis of their functioning has permitted the progressive movement towards a truly industrial exploitation. (authors) [fr

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

    Energy Technology Data Exchange (ETDEWEB)

    Noel Duckwitz

    2011-06-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  12. 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.

  13. 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.

  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. 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.

  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. In-Core-Instrumentation Methods for 3-Dimensional Distribution Information of Reactor Core Temperatures and Melt-down

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Yeong Cheol [KHNP, Daejeon (Korea, Republic of); Eun, Myoung; Kim, Sung Jun [Woojin Inc., Hwaseong (Korea, Republic of)

    2014-08-15

    The tsunami-induced nuclear accident at Japanese Fukushima power plants in March 2011 has revealed some weaknesses in the severe accident monitoring system. The plant instrumentation did not provide utility, safety experts, and government officials with adequate and reliable information. The information on the reactor core damage and coolability is critical for making decisions correctly as well as in a timely manner during the course of the mitigation of severe accidents. Current Pressurized Water Reactor (PWR)s have an In-Core-Instrumentation (ICI) system that measures the temperature distribution of the top surface (i.e. Core Exit Temperatures) of the reactor core mainly to indicate when to begin Severe Accident Mitigation Guidelines (SAMG). This design concept giving only the core exit temperature has limitations in terms of sufficiency as well as availability of the information necessary for diagnosis on the status of the degraded core and the effectiveness of the measures taken as mitigation strategies. The reactor core exit temperatures are not sufficient to support the assessment of the degree of the core damage and the location of the molten core debris and recognition whether the core damage progresses on or it is mitigated. The ICI location being at the top of the reactor core also makes the ICI thermocouples vulnerable to melt-down because the upper part of the reactor core uncovers first, thereby melt down at the early stage of the accident. This means that direct indication of reactor core temperature will be lost and unavailable during the later stages of severe accident. To address the aforementioned weaknesses of the current ICIs, it is necessary to develop a new ICI system that provides information that is more expanded and more reliable for accident mitigation. With the enhanced information available, the SAMG can be prepared in more refined and effective way based on the direct and suitable indication of status of damages and the 3-dimensional

  19. Incidence and risk factors of lower leg fractures in Belgian soccer players.

    Science.gov (United States)

    Vanlommel, Luc; Vanlommel, Jan; Bollars, Peter; Quisquater, Laurent; Van Crombrugge, Kris; Corten, Kristoff; Bellemans, Johan

    2013-12-01

    Soccer is the world's most popular sport and one that is physically demanding and highly competitive. Consequently, the rate of injuries resulting from this sport is only increasing. It is estimated that 2-20% of all such injuries are fractures, one-third of which are located in the lower extremities. The aim of this epidemiological study was to investigate the incidence of lower-leg fractures (LLFs) in Belgian soccer players and determine the possible risk factors that lead to them. All injuries of players associated with the Royal Belgium Football Association (RBFA) were reported and collected in a nationwide registry. We retrospectively compared the incidence rate of and risk factors for LLFs in Belgian soccer players during two seasons, 1999-2000 and 2009-2010. In total, 1600 fractures (3%) were located in the lower leg. After a decade, the number of LLFs remained unchanged. Ankle fractures were the most common (37%), followed by foot and tibia fractures (33% and 22%, respectively). The least common were fibula fractures, which accounted for just 9%. A higher incidence of every type of LLF was observed in older and amateur-level soccer players, when compared with their younger and professional counterparts. Male players experienced more tibia and foot fractures, whereas the incidences of ankle and fibula fractures were comparable with those in female soccer players. The vast majority of fractures occurred during soccer games. Ankle fractures and foot fractures represented two-thirds of all fractures noted in this analysis. Male gender, recreational level and adult age were important risk factors for LLFs. After 10 years, the incidence of LLFs did not decrease. Given the socioeconomic impact of these injuries, improved prevention techniques are required to reduce their incidence, particularly with regard to the frequently occurring ankle and foot fractures in this population. Copyright © 2013 Elsevier Ltd. All rights reserved.

  20. 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

  1. 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)

  2. 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

  3. Reactor safety issues resolved by the 2D/3D Program. International Agreement Report

    Energy Technology Data Exchange (ETDEWEB)

    Damerell, P.S.; Simons, J.W. [eds.] [MPR Associates, Inc., Washington, DC (United States)

    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.

  4. 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.

  5. 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.)

  6. 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

  7. 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

  8. Propulsion reactors

    International Nuclear Information System (INIS)

    Anon.

    1999-01-01

    A nuclear reactor equips the recently constructed French aircraft- carrier Charles-De-Gaulle, in a few months the second nuclear submarine (SNLE) of new generation will be operational. In last october the government launched the program Barracuda which consists of 6 submarines (SNA) whose series head will be operational in 2010. The main asset of nuclear propulsion is to allow an almost unlimited autonomy: soft water, air are produced inside the submarine and the maximum time spent underwater is only limited by human capacity to cope with confinement. CEA has 3 missions concerning country defence. First the designing, the fabrication and the maintenance of weapons, secondly the supplying of fissile materials and thirdly the nuclear propulsion. A new generation of propulsion reactors is being studied and a ground installation involving a test reactor equivalent to that on board is being built. This test reactor (RES) will simulate any type of on-board reactors by adjusting temperature, pressure, flowrate and even equipment such as steam generator. This reactor will validate the technological choices for the Barracuda program. (A.C.)

  9. The German and Belgian accreditation models for diabetic foot services.

    Science.gov (United States)

    Morbach, Stephan; Kersken, Joachim; Lobmann, Ralf; Nobels, Frank; Doggen, Kris; Van Acker, Kristien

    2016-01-01

    The International Working Group on the Diabetic Foot recommends that auditing should be part of the organization of diabetic foot care, the efforts required for data collection and analysis being balanced by the expected benefits. In Germany legislature demands measures of quality management for in- and out-patient facilities, and, in 2003, the Germany Working Group on the Diabetic Foot defined and developed a certification procedure for diabetic foot centres to be recognized as 'specialized'. This includes a description of management facilities, treatment procedures and outcomes, as well as the organization of mutual auditing visits between the centres. Outcome data is collected at baseline and 6 months on 30 consecutive patients. By 2014 almost 24,000 cases had been collected and analysed. Since 2005 Belgian multidisciplinary diabetic foot clinics could apply for recognition by health authorities. For continued recognition diabetic foot clinics need to treat at least 52 patients with a new foot problem (Wagner 2 or more or active Charcot foot) per annum. Baseline and 6-month outcome data of these patients are included in an audit-feedback initiative. Although originally fully independent of each other, the common goal of these two initiatives is quality improvement of national diabetic foot care, and hence exchanges between systems has commenced. In future, the German and Belgian accreditation models might serve as templates for comparable initiatives in other countries. Just recently the International Working Group on the Diabetic Foot initiated a working group for further discussion of accreditation and auditing models (International Working Group on the Diabetic Foot AB(B)A Working Group). Copyright © 2016 John Wiley & Sons, Ltd.

  10. 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)

  11. 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

  12. 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.

  13. 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.)

  14. 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)

  15. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-09-01

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

  17. 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

  18. A longitudinal study of survival in belgian shepherds with genetic epilepsy

    DEFF Research Database (Denmark)

    Gulløv, Christina Hedal; Toft, Nils; Berendt, Mette

    2012-01-01

    Belgian Shepherds have focal genetic epilepsy. The prevalence of epilepsy has been estimated as 9.5% in the breed and as 33% in the family investigated. Dogs with epilepsy might have an increased risk of premature death.......Belgian Shepherds have focal genetic epilepsy. The prevalence of epilepsy has been estimated as 9.5% in the breed and as 33% in the family investigated. Dogs with epilepsy might have an increased risk of premature death....

  19. Extension of RAPTOR-M3G to r-θ-z Geometry for Use in Reactor Dosimetry Applications

    Science.gov (United States)

    Hunter, Melissa A.; Longoni, Gianluca; Anderson, Stanwood L.

    2009-08-01

    The RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3-D Geometries) is a new deterministic radiation transport code that was originally developed for x-y-z geometry. The development of the r-θ-z version of RAPTOR-M3G and its application to determine ex-vessel neutron dosimetry responses in the cavity of a typical 2-loop pressurized water reactor is presented. The neutron dosimetry responses determined from RAPTOR-M3G and TORT 3-D r-θ-z calculations are compared to actual measured responses.

  20. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

    Energy Technology Data Exchange (ETDEWEB)

    Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

  1. Enhancement of REBUS-3/DIF3D for whole-core neutronic analysis of prismatic very high temperature reactor (VHTR).

    Energy Technology Data Exchange (ETDEWEB)

    Lee, C. H.; Zhong, Z.; Taiwo, T.A.; Yang, W.S.; Khalil, H.S.; Smith, M.A.; Nuclear Engineering Division

    2006-10-13

    Enhancements have been made to the REBUS-3/DIF3D code suite to facilitate its use for the design and analysis of prismatic Very High Temperature Reactors (VHTRs). A new cross section structure, using table-lookup, has been incorporated to account for cross section changes with burnup and fuel and moderator temperatures. For representing these cross section dependencies, three new modules have been developed using FORTRAN 90/95 object-oriented data structures and implemented within the REBUS-3 code system. These modules provide a cross section storage procedure, construct microscopic cross section data for all isotopes, and contain a single block of banded scattering data for efficient data management. Fission products other than I, Xe, Pm, and Sm, can be merged into a single lumped fission product to save storage space, memory, and computing time without sacrificing the REBUS-3 solution accuracy. A simple thermal-hydraulic (thermal-fluid) feedback model has been developed for prismatic VHTR cores and implemented in REBUS-3 for temperature feedback calculations. Axial conduction was neglected in the formulation because of its small magnitude compared to radial (planar) conduction. With the simple model, the average fuel and graphite temperatures are accurately estimated compared to reference STAR-CD results. The feedback module is currently operational for the non-equilibrium fuel cycle analysis option of REBUS-3. Future work should include the extension of this capability to the equilibrium cycle option of the code and additional verification of the feedback module. For the simulation of control rods in VHTR cores, macroscopic cross section deviations (deltas) have been defined to account for the effect of control rod insertion. The REBUS-3 code has been modified to use the appropriately revised cross sections when control rods are inserted in a calculation node. In order to represent asymmetric core blocks (e.g., fuel blocks or reflector blocks containing

  2. 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

  3. 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.

  4. 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.

  5. 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

  6. 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.

  7. 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.

  8. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

    The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

  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. 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.

  11. 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.

  12. 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

  13. 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

  14. 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.

  15. RB Research nuclear reactor RB reactor, Annual report for 2000

    International Nuclear Information System (INIS)

    Milosevic, M.

    2000-12-01

    Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

  16. Modeling and analysis of a lunar space reactor with the computer code RELAP5-3D/ATHENA

    International Nuclear Information System (INIS)

    Carbajo, J. J.; Quails, L.

    2008-01-01

    The transient analysis three-dimensional (3-D) computer code RELAP5-3D/ATHENA has been employed to model and analyze a fission-based power system for application on the surface of the Moon or Mars. The reactor power is 180 kW (thermal) and 40 kW (net, electrical) generated by eight Stirling converters. The reactor is cooled by a eutectic mixture of sodium and potassium (NaK) that is liquid at ambient temperature. The model consists of a primary loop and two independent secondary loops connected by two heat exchangers. Each secondary loop provides power to four Stirling converters. The maximum and minimum coolant temperatures of the primary loop are set at 902 K and 851 K. The secondary loops also employ NaK as the coolant at a maximum temperature of 877 K and a minimum temperature of 842 K. The Stirling converter hot ends are at an assumed temperature of 800 K. The heat exchangers surrounding the Stirling converter hot ends are of annular design and have been modeled using 3-D volumes. The transients analyzed include failure of one of the Stirling converters, trip of the reactor pumps, and trips of the secondary loop pumps. The transients studied have demonstrated the inherent safety and the operability of the system; however, the results are preliminary. (authors)

  17. 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

  18. Neutronic reactor

    International Nuclear Information System (INIS)

    Carleton, J.T.

    1977-01-01

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment. 3 claims, 6 figures

  19. 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

  20. Neutronic and thermal-hydraulic coupling for 3D reactor core modeling combining MCB and fluent

    Directory of Open Access Journals (Sweden)

    Królikowski Igor P.

    2015-09-01

    Full Text Available Three-dimensional simulations of neutronics and thermal hydraulics of nuclear reactors are a tool used to design nuclear reactors. The coupling of MCB and FLUENT is presented, MCB allows to simulate neutronics, whereas FLUENT is computational fluid dynamics (CFD code. The main purpose of the coupling is to exchange data such as temperature and power profile between both codes. Temperature required as an input parameter for neutronics is significant since cross sections of nuclear reactions depend on temperature. Temperature may be calculated in thermal hydraulics, but this analysis needs as an input the power profile, which is a result from neutronic simulations. Exchange of data between both analyses is required to solve this problem. The coupling is a better solution compared to the assumption of estimated values of the temperatures or the power profiles; therefore the coupled analysis was created. This analysis includes single transient neutronic simulation and several steady-state thermal simulations. The power profile is generated in defined points in time during the neutronic simulation for the thermal analysis to calculate temperature. The coupled simulation gives information about thermal behavior of the reactor, nuclear reactions in the core, and the fuel evolution in time. Results show that there is strong influence of neutronics on thermal hydraulics. This impact is stronger than the impact of thermal hydraulics on neutronics. Influence of the coupling on temperature and neutron multiplication factor is presented. The analysis has been performed for the ELECTRA reactor, which is lead-cooled fast reactor concept, where the coolant fl ow is generated only by natural convection

  1. Influence of apixaban on commonly used coagulation assays: results from the Belgian national External Quality Assessment Scheme.

    Science.gov (United States)

    Van Blerk, M; Bailleul, E; Chatelain, B; Demulder, A; Devreese, K; Douxfils, J; Jacquemin, M; Jochmans, K; Mullier, F; Wijns, W; China, B; Vernelen, K; Soumali, M R

    2017-08-01

    The Belgian national External Quality Assessment Scheme performed a survey to assess the effect of the direct oral anticoagulant apixaban on the coagulation assays prothrombin time (PT), activated partial thromboplastin time (aPTT), fibrinogen and antithrombin as performed with a large number of reagent/instrument combinations. Four lyophilized plasma samples spiked with apixaban (0, 41, 94 and 225 ng/mL) were sent to the 195 Belgian and Luxembourg clinical laboratories performing coagulation testing. PT and aPTT were barely influenced at the concentrations tested. At 225 ng/mL apixaban, PT and aPTT clotting times were only 1.15 times longer than at 0 ng/mL. Among PT reagents, RecombiPlasTin 2G ® showed a slightly higher sensitivity with 225 ng/mL apixaban prolonging the PT clotting time 1.3-fold. Among aPTT reagents, there was no appreciable difference in sensitivity. Fibrinogen results were unaffected by the presence of apixaban, but antithrombin activity was considerably overestimated when measured with a FXa-based assay. At 225 ng/mL apixaban, the median percentage increase in antithrombin level was 31% when measured with the Liquid Antithrombin ® reagent and 44% with the Innovance Antithrombin ® reagent. Our data provide clinical laboratories with useful information on the impact of apixaban on their routine coagulation assays. © 2017 John Wiley & Sons Ltd.

  2. 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 power 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.)

  3. 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.)

  4. Reactor Physics Programme

    International Nuclear Information System (INIS)

    De Raedt, C.

    2000-01-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

  5. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  6. Evaluation of the Use of Existing RELAP5-3D Models to Represent the Actinide Burner Test Reactor

    International Nuclear Information System (INIS)

    C. B. Davis

    2007-01-01

    The RELAP5-3D code is being considered as a thermal-hydraulic system code to support the development of the sodium-cooled Actinide Burner Test Reactor as part of Global Nuclear Energy Partnership. An evaluation was performed to determine whether the control system could be used to simulate the effects of non-convective mechanisms of heat transport in the fluid that are not currently represented with internal code models, including axial and radial heat conduction in the fluid and subchannel mixing. The evaluation also determined the relative importance of axial and radial heat conduction and fluid mixing on peak cladding temperature for a wide range of steady conditions and during a representative loss-of-flow transient. The evaluation was performed using a RELAP5-3D model of a subassembly in the Experimental Breeder Reactor-II, which was used as a surrogate for the Actinide Burner Test Reactor. An evaluation was also performed to determine if the existing centrifugal pump model could be used to simulate the performance of electromagnetic pumps

  7. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  8. 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.

  9. Proba "spacecraft family" small autonomous satellites - a Belgian innovative exportproduct

    Science.gov (United States)

    Bermyn, J.; Du Pre, T.; Bernaerts, Dirk; Baudoux, Dominique

    2004-11-01

    After the successful realisation of Proba 1, a 100 kg small autonomous satellite for ESA technology demonstration purposes now generating for more then 2,5 years splendid earth observation images, and the first steps set in the development of its successor PROBA 2, Verhaert and Spacebel teamed for the further worldwide commercialisation of this innovative Belgian product. In PROBA 1, Verhaert was the prime contractor and acted as the small systems integrator, where Spacebel was responsible for the on-board software development. Verhaert and Spacebel offer now a complete small satellite system solution to the user community. The combined company team deals with the system aspects, while Verhaert is responsible for the satellite platform realisation and the launch procurement. Spacebel realizes the control and exploitation ground segment and is responsible for all data management aspects (on board and ground software). Other elements, such as payload development and operations, are covered by other companies on a case-by-case basis in function of the client's wishes.

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

    Science.gov (United States)

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

    2015-01-01

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

  11. RA Research nuclear reactor, Part 1, RA reactor operation and maintenance in 1993, with comparative review for the period 1991 - 1993, Annex 3

    International Nuclear Information System (INIS)

    Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Mikic, N.; Tanaskovic, M.

    1993-01-01

    RA reactor was not operated during 1993 because of the complete instrumentation exchange. Although it has been planned to exchange the complete instrumentation until the end of 1993, and to start reactor operation in the first half of 1993 this was not fulfilled because the instrumentation was not delivered until the end of 1993. Main activities during past seven years were related to construction of the emergency cooling system; repair and reconstruction of the system for handling the spent fuel and improvement of spent fuel storage conditions; exchange of the aged instrumentation. Other reactor components and systems, reactor core, primary coolant loop and gas circulation system are in good condition concerning future start-up [sr

  12. Conceptual analysis of a preliminary model for instability study in normal operation of a natural circulation reactor type EBWR, using Relap5/Mod 3.2

    International Nuclear Information System (INIS)

    Ojeda S, J.; Morales S, J.; Chavez M, C.

    2009-10-01

    This work intends a model using the code Relap5/Mod 3.2, for the instability study in normal operation of a natural circulation reactor type ESBWR. A conceptual analysis is considered because all the information was obtained of the open literature, and some of reactor operation or dimension (not available) parameters were approached. As starting point was took the pattern developed for reactor type BWR, denominated Browns Ferry and changes were focused in elimination of bonds of forced recirculation, in modification of operation parameters, dimensions and own control parameters, according to internal code structure. Additionally the nodalization outline is described analyzing for separate the four fundamental areas employees in peculiar geometry of natural circulation reactor. Comparative analysis of results of stability behavior obtained with those reported in the open literature were made, by part of commercial reactor designer ESBWR. (Author)

  13. 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

  14. SIMMER-3 and SIMMER-4 safety code development for reactors with transmutation capability

    International Nuclear Information System (INIS)

    Maschek, W.; Rineiski, A.; Suzuki, T.; Wang, S.; Mori, M.; Wiegner, E.; Wilhelm, D.; Kretzschmar, F.; Tobita, Y.; Yamano, H.; Fujita, S.; Coste, P.; Pigny, S.; Henriques, A.; Cadiou, T.; Morita, K.; Bandini, G.

    2005-01-01

    The SIMMER-code family has played an outstanding role in the framework and history of mechanistic safety code development for liquid metal cooled reactors with fast neutron spectrum. Its unmatched position is related to the extended simulation range from normal operation conditions, via transients and accidents, core melting and core destruction to in-vessel relocation phenomena and post accident heat removal conditions. SIMMER-III is a two-dimensional, SIMMER-IV a three-dimensional, multi-velocity-field, multi-phase, multi-component, Eulerian, fluid-dynamics code system coupled with a structure model for fuel-pins, hex-cans and general structures, and a space-, time- and energy-dependent transport theory neutron dynamics model. An elaborate analytical equation-of-state closes the fluid-dynamics conservation equations. The code has originally been developed for the severe accident simulation of sodium cooled fast reactors. However, the general philosophy behind the SIMMER development was to generate a very versatile and flexible tool, applicable for the safety analysis of various reactor types with different neutron spectra and coolants. The multi-physics and multi-scale code family SIMMER is developed in a joint effort of different laboratories, where currently the main application area in Europe is in the framework of accelerator-driven systems. The most recent code developments and SIMMER application areas are discussed in this paper. (authors)

  15. PERIGEE computer codes for reactor simulation in 3 dimensions, using 1 or 2 neutron velocity groups

    International Nuclear Information System (INIS)

    Olson, A.P.

    1964-02-01

    PERIGEE is a code written in SNAP for the G-20 computer. It solves the one- or two-group neutron diffusion equations by finite-difference methods on a three-dimensional, uniform mesh having a common spacing in the two directions normal to the fuel channels. The positions of mesh points along a fuel channel, relative to points in adjacent channels, may correspond to either NPD or CANDU fuel bundle positions. The extrapolated flux boundary may be specified in sufficient detail to represent a tapered or stepped circumferential reflector, a variable axial length and, for a reactor with axis horizontal, a variable moderator level and a variable plane bottom surface equivalent to the CANDU dump structure. The neutron flux may be normalized to give a specified power output from the hottest fuel bundle or hottest channel, or to give a total thermal power limited by the turbine and generator. Reactor operation may be simulated in finite time steps, taking into account any fuel shifts, any changes in moderator level and the change in nuclear properties of the fuel with increasing irradiation. The appropriate properties are obtained by interpolation from tables supplied for as many as 8 types of fuel bundle. The mean fuel exit burnup can be calculated at equilibrium for a reactor in which the exit burnups for two zones may be adjusted to give radial power flattening and the fuelling schedules may be designed to give axial power flattening in one or both zones. (author)

  16. Licensed operating reactors. Status summary report, data as of 3-31-82

    International Nuclear Information System (INIS)

    1982-04-01

    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 Management and Program Analysis, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the Errata page. 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

  17. Democracy and expertise: The waste problem of nuclear power before the loading of the Ringhals-3 reactor. Demokrati och expertis: Kaenkraftens avfallsproblem infoer laddningen a Ringhals 3

    Energy Technology Data Exchange (ETDEWEB)

    Sundqvist, G.

    1991-01-01

    The article describes the political processes that followed upon the application to start operating the Swedish Ringhals 3 reactor. According to the low, the applicant had to shown how and where a totally safe final deposition of the radioactive waste generated could be performed. The political and scientific discussions and buck-passing during the time form the application (Dec 1977) to the authorization (Mar 1980) are analysed and discussed by the author. (44 refs.).

  18. 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

  19. Characterization of a dopamine transporter polymorphism and behavior in Belgian Malinois.

    Science.gov (United States)

    Lit, Lisa; Belanger, Janelle M; Boehm, Debby; Lybarger, Nathan; Haverbeke, Anouck; Diederich, Claire; Oberbauer, Anita M

    2013-05-30

    The Belgian Malinois dog breed (MAL) is frequently used in law enforcement and military environments. Owners have reported seizures and unpredictable behavioral changes including dogs' eyes "glazing over," dogs' lack of response to environmental stimuli, and loss of behavioral inhibition including owner-directed biting behavior. Dogs with severe behavioral changes may be euthanized as they can represent a danger to humans and other dogs. In the dog, the dopamine transporter gene (DAT) contains a 38-base pair variable number tandem repeat (DAT-VNTR); alleles have either one or two copies of the 38-base pair sequence. The objective of this study was to assess frequency of DAT-VNTR alleles, and characterize the association between DAT-VNTR alleles and behavior in MAL and other breeds. In an American sample of 280 dogs comprising 26 breeds, most breeds are predominantly homozygous for the DAT-VNTR two-tandem-repeat allele (2/2). The one-tandem-repeat allele is over-represented in American MAL (AM-MAL) (n = 144), both as heterozygotes (1/2) and homozygotes (1/1). All AM-MAL with reported seizures (n = 5) were 1/1 genotype. For AM-MAL with at least one "1" allele (1/1 or 1/2 genotype, n = 121), owners reported higher levels of attention, increased frequency of episodic aggression, and increased frequency of loss of responsiveness to environmental stimuli. In behavior observations, Belgian Military Working Dogs (MWD) with 1/1 or 1/2 genotypes displayed fewer distracted behaviors and more stress-related behaviors such as lower posture and increased yawning. Handlers' treatment of MWD varied with DAT-VNTR genotype as did dogs' responses to handlers' behavior. For 1/1 or 1/2 genotype MWD, 1) lower posture after the first aversive stimulus given by handlers was associated with poorer obedience performance; 2) increased aversive stimuli during protection exercises were associated with decreased performance; 3) more aversive stimuli during obedience were associated

  20. 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.

  1. Synthesis of the IRSN report related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor. Referral of the Permanent Group of Experts for nuclear reactors (GPR), examination of probabilistic level-2 safety studies (EPS 2) and severe accidents (AG) of the Flamanville reactor nr 3. Opinion related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor (FA3). Electronuclear reactors - EDF - Flamanville 3 EPR reactor. Severe accidents and probabilistic level 2 studies

    International Nuclear Information System (INIS)

    2015-01-01

    This document gathers several documents. The first one recalls the main arrangements implemented on the FA3 EPR reactor regarding accidents with core fusion, reports the analysis made by the IRSN about the sizing of these arrangements to reach a controlled status of the installation after a severe accident, regarding the probabilistic level-2 safety assessment, regarding the radiological impact of a severe accident on the population and on the environment, regarding those aimed at facing a total and long duration loss of electric power sources and cold sources, and about the situation of the reactor with respect to WENRA positions on severe accidents for new reactors. The second document is a letter sent by the ASN to the Permanent Group of Experts for nuclear reactors (GPR) to address probabilistic level-2 safety studies (EPS2) and severe accidents for the Flamanville 3 reactor. The third one reports the opinion of the GPR on these both issues and proposes a set of recommendations. The next document is a letter sent by the ASN to the Flamanville 3 project manager at EDF which recalls the related objectives, the ASN opinion on the implemented arrangements for severe accidents (de-pressurization of the primary circuit, management of hydrogen-related risks, corium recovery and cooling outside the vessel, limitation of vapour explosion risks outside the vessel, heat evacuation system, containment enclosure, management of the risk of a return to criticality), to face a total and long duration loss of electricity sources and cold sources, and other aspects addressed in the IRSN analysis. Requests and remarks formulated by the ASN are provided in an appendix to this last document

  2. Impact of grazing on carbon balance of a Belgian grassland

    Science.gov (United States)

    Jérôme, Elisabeth; Beckers, Yves; Bodson, Bernard; Moureaux, Christine; Dumortier, Pierre; Beekkerk van Ruth, Joran; Aubinet, Marc

    2013-04-01

    This work analyzes the impact of grazing on the carbon balance of a grassland grazed by the Belgian Blue breed of cattle. The research was run at the Dorinne terrestrial observatory (DTO). The experimental site is a permanent grassland of ca. 4.2 ha located in the Belgian Condroz (50° 18' 44" N; 4° 58' 07" E; 248 m asl.). Other studies are conducted at the DTO including measurements of methane (CH4) and nitrous oxide fluxes (Dumortier et al., Geophysical Research Abstracts, Vol. 15, EGU2013-2083-1, 2013; Beekkerk van Ruth et al., Geophysical Research Abstracts, Vol. 15, EGU2013-3211, 2013, respectively). Grassland carbon budget (Net Biome Productivity, NBP) was calculated from Net Ecosystem Exchange (NEE) measured by eddy covariance by taking imports and exports of organic C and losses of carbon as CH4 into account. After 2 years of measurements (May 2010 - May 2012), the grassland behaved on average as a CO2 source (NEE = 73 ±31 g C m-2 y-1). After inclusion of all the C inputs and outputs the site was closed to equilibrium (NBP = 23 ±34 g C m-2 y-1). To analyze the impact of grazing on CO2 fluxes, we studied the temporal evolution of gross maximal photosynthetic capacity GPPmax and dark respiration Rd (deduced from the response of daytime fluxes to radiation over 5-day windows). We calculated GPPmax and Rd variation between the end and the beginning of grazing or non-grazing periods (ΔGPPmax and ΔRd, respectively). We observed a significant decrease of GPPmax during grazing periods and measured a ΔGPPmax dependence on the average stocking rate. This allows us to quantify the assimilation reduction due to grass consumption by cattle. On the contrary, no Rd decrease was observed during grazing periods. Moreover, we found that cumulated monthly NEE increased significantly with the average stocking rate. In addition, a confinement experiment was carried out in order to analyze livestock contribution to Total Ecosystem Respiration. Each experiment extended over

  3. Analysis of pellet cladding interaction and creep of U 3SIi2 fuel for use in light water reactors

    Science.gov (United States)

    Metzger, Kathryn E.

    Following the accident at the Fukushima plant, enhancing the accident tolerance of the light water reactor (LWR) fleet became a topic of serious discussion. Under the direction of congress, the DOE office of Nuclear Energy added accident tolerant fuel development as a primary component to the existing Advanced Fuels Program. The DOE defines accident tolerant fuels as fuels that "in comparison with the standard UO2- Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, as well as design-basis and beyond design-basis events." To be economically viable, proposed accident tolerant fuels and claddings should be backward compatible with LWR designs, provide significant operating cost improvements such as power uprates, increased fuel burnup, or increased cycle length. In terms of safety, an alternative fuel pellet must have resistance to water corrosion comparable to UO2, thermal conductivity equal to or larger than that of UO2, and a melting temperature that allows the material to remain solid under power reactor conditions. Among the candidates, U3Si2 has a number of advantageous thermophysical properties, including; high density, high thermal conductivity at room temperature, and a high melting temperature. These properties support its use as an accident tolerant fuel while its high uranium density is capable of supporting uprates to the LWR fleet. This research characterizes U3Si2 pellets and analyzes U3Si2 under light water reactor conditions using the fuel performance code BISON. While some thermophysical properties for U3Si2 have been found in the literature, the irradiation behavior is sparse and limited to experience with dispersion fuels. Accordingly, the creep behavior for U3Si2 has been unknown, making it

  4. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  5. Focusing and polarized neutron ultra-small-angle scattering spectrometer (SANS-J-II) at Research Reactor JRR3, Japan

    International Nuclear Information System (INIS)

    Koizumi, Satoshi; Iwase, Hiroki; Suzuki, Jun-ichi; Oku, Takayuki; Motokawa, Ryuhei; Sasao, Hajime; Tanaka, Hirokazu; Yamaguchi, Daisuke; Shimizu, Hirohiko M.; Hashimoto, Takeji

    2006-01-01

    By employing focusing lenses of sextupole permanent magnet and biconcave MgF 2 crystal, and high-resolution photomultiplier, SANS-J (pinhole small-angle neutron scattering spectrometer at research reactor JRR3, Tokai) was reconstructed to focusing and polarized neutron ultra-small-angle scattering spectrometer (SANS-J-II). Consequently, an accessible minimum wave number q min was improved from 3x10 -3 A -1 to 3x10 -4 A -1 . Furthermore, we added 3 He sub-detectors with an analyzer super mirror at a sample position. With this setup, we perform polarization analysis at high q (>0.2 A -1 ) in order to quantitatively discriminate spin incoherent scattering from hydrogen or to perform spin contrast variation, by a dynamic nuclear polarization method (>0.2 A -1 )

  6. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 3, Sessions 12-16

    Energy Technology Data Exchange (ETDEWEB)

    Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

    1995-09-01

    This document, Volume 3, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, ad the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected abstracts have been indexed separately for inclusion in the Energy Science and Technology Database.

  7. 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.

  8. Basic Experimental Simulation of a Traveling Wave Direct Energy Converter for a D-3He Fusion Reactor

    Science.gov (United States)

    Takeno, Hiromasa; Ikeda, Youichi; Yamada, Takashi; Noda, Kazuyuki; Yasaka, Yasuyoshi

    2000-09-01

    We simulated a traveling wave direct energy converter for application in a D-3He fusion reactor. The deceleration of a simulated ion beam was studied, and the dependence of the efficiency of the decelerator on its length was examined. The effectiveness of spatial variation in the phase velocity of the traveling wave was demonstrated experimentally. Optimization of the structure of the decelerator is discussed with the aid of numerical simulations. By employing the proper decelerator structure, the efficiency per unit length of about 0.2 with the normalized decelerator length of about 2.5 is expected.

  9. SIEX3: A correlated computer code for prediction of fast reactor mixed oxide fuel and blanket pin performance

    International Nuclear Information System (INIS)

    Baker, R.B.; Wilson, D.R.

    1986-04-01

    The SIEX3 computer program was developed to calculate the fuel and cladding performance of oxide fuel and oxide blanket pins irradiated in the fast neutron environment of a liquid metal cooled reactor. The code is uniquely designed to be accurate yet quick running and use a minimum of computer core storage. This was accomplished through the correlation of physically based models to very large data bases of irradiation test results. Data from over 200 fuel pins and over 800 transverse fuel microscopy samples were used in the calibrations

  10. Analysis of the metallic containment integrity of Angra 2/3 reactor under the effects of the design basis accident

    International Nuclear Information System (INIS)

    Costa, J.R.

    1981-06-01

    The application of Condru 4 computer code, developed to determine the maximum values of pressure and temperature that occur inside the metallic containment building of PWR nuclear power plants, in case of a hypothetic accident - LOCA - considered as a Design Basic Accident - DBA. The hypothesis, input and results for the simulation of a loss of coolant in the hot leg of the Angra-2/3 reactors, considered as the most critical case for that Kind of project, are presented. The analysis was made with input provided by the manufacturer. (Author) [pt

  11. Basic experimental stimulation of a traveling wave direct energy converter for a D-3He fusion reactor

    International Nuclear Information System (INIS)

    Takeno, Hiromasa; Ikeda, Youichi; Yamada, Takashi; Noda, Kazuyuki; Yasaka, Yasuyoshi

    2000-01-01

    We simulated a traveling wave direct energy converter for application in a D- 3 He fusion reactor. The deceleration of a simulated ion beam was studied, and the dependence of the efficiency of the decelerator on its length was examined. The effectiveness of spatial variation in the phase velocity of the traveling wave was demonstrated experimentally. Optimization of the structure of the decelerator is discussed with the aid of numerical simulations. By employing the proper decelerator structure, the efficiency per unit length of about 0.2 with the normalized decelerator length of about 2.5 is expected. (author)

  12. 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.

  13. Neutron transport in hexagonal reactor cores modeled by trigonal-geometry diffusion and simplified P{sub 3} nodal methods

    Energy Technology Data Exchange (ETDEWEB)

    Duerigen, Susan

    2013-05-15

    The superior advantage of a nodal method for reactor cores with hexagonal fuel assemblies discretized as cells consisting of equilateral triangles is its mesh refinement capability. In this thesis, a diffusion and a simplified P{sub 3} (or SP{sub 3}) neutron transport nodal method are developed based on trigonal geometry. Both models are implemented in the reactor dynamics code DYN3D. As yet, no other well-established nodal core analysis code comprises an SP{sub 3} transport theory model based on trigonal meshes. The development of two methods based on different neutron transport approximations but using identical underlying spatial trigonal discretization allows a profound comparative analysis of both methods with regard to their mathematical derivations, nodal expansion approaches, solution procedures, and their physical performance. The developed nodal approaches can be regarded as a hybrid NEM/AFEN form. They are based on the transverse-integration procedure, which renders them computationally efficient, and they use a combination of polynomial and exponential functions to represent the neutron flux moments of the SP{sub 3} and diffusion equations, which guarantees high accuracy. The SP{sub 3} equations are derived in within-group form thus being of diffusion type. On this basis, the conventional diffusion solver structure can be retained also for the solution of the SP{sub 3} transport problem. The verification analysis provides proof of the methodological reliability of both trigonal DYN3D models. By means of diverse hexagonal academic benchmark and realistic detailed-geometry full-transport-theory problems, the superiority of the SP{sub 3} transport over the diffusion model is demonstrated in cases with pronounced anisotropy effects, which is, e.g., highly relevant to the modeling of fuel assemblies comprising absorber material.

  14. Radial distribution of UO2 and Gd2O3 in fuel cells of a BWR Reactor

    International Nuclear Information System (INIS)

    Montes, J.L.; Ortiz, J.J.; Perusquia del C, R.; Francois, J.L.; Martin del Campo M, C.

    2008-01-01

    The fuel system that is used at the moment in a power plant based on power reactors BWR, includes as much like the one of its substantial parts to the distribution of the fissile materials like a distribution of burnt poisons within each one of the cells which they constitute the fuel assemblies, used for the energy generation. Reason why at the beginning of a new operation cycle in a reactor of this type, the reactivity of the nucleus should be compensated by the exhaustion of the assemblies that it moves away of the nucleus for their final disposition. This compensation is given by means of the introduction of the recharge fuel, starting from the UO 2 enriched in U 2 35, and of the Gadolinium (Gd 2 O 3 ). The distribution of these materials not only defines the requirements of energy generation, but in certain measures also the form in that the margins will behave to the limit them thermal during the operation of the reactor. These margins must be taken into account for the safe and efficient extraction of the energy of the fuel. In this work typical fuel cells appear that are obtained by means of the use of a emulation model of an ants colony. This model allows generating from a possible inventory of values of enrichment of U 2 35, as well as of concentration of Gadolinium a typical fuel cell, which consists of an arrangement of lOxlO rods, of which 92 contain U 2 35, some of these rods contain a concentration of Gd 2 O 3 and 8 of the total contain only water. The search of each cell finishes when the value of the Local Peak Power Factor (LPPF) in the cell reaches a minimal value, or when a pre established value of iterations is reached. The cell parameters are obtained from the results of the execution of the code HELIOS, which incorporates like a part integral of the search algorithm. (Author)

  15. High-resolution mapping of the NO2 spatial distribution over Belgian urban areas based on airborne APEX remote sensing

    Science.gov (United States)

    Tack, Frederik; Merlaud, Alexis; Iordache, Marian-Daniel; Danckaert, Thomas; Yu, Huan; Fayt, Caroline; Meuleman, Koen; Deutsch, Felix; Fierens, Frans; Van Roozendael, Michel

    2017-05-01

    We present retrieval results of tropospheric nitrogen dioxide (NO2) vertical column densities (VCDs), mapped at high spatial resolution over three Belgian cities, based on the DOAS analysis of Airborne Prism EXperiment (APEX) observations. APEX, developed by a Swiss-Belgian consortium on behalf of ESA (European Space Agency), is a pushbroom hyperspectral imager characterised by a high spatial resolution and high spectral performance. APEX data have been acquired under clear-sky conditions over the two largest and most heavily polluted Belgian cities, i.e. Antwerp and Brussels on 15 April and 30 June 2015. Additionally, a number of background sites have been covered for the reference spectra. The APEX instrument was mounted in a Dornier DO-228 aeroplane, operated by Deutsches Zentrum für Luft- und Raumfahrt (DLR). NO2 VCDs were retrieved from spatially aggregated radiance spectra allowing urban plumes to be resolved at the resolution of 60 × 80 m2. The main sources in the Antwerp area appear to be related to the (petro)chemical industry while traffic-related emissions dominate in Brussels. The NO2 levels observed in Antwerp range between 3 and 35 × 1015 molec cm-2, with a mean VCD of 17.4 ± 3.7 × 1015 molec cm-2. In the Brussels area, smaller levels are found, ranging between 1 and 20 × 1015 molec cm-2 and a mean VCD of 7.7 ± 2.1 × 1015 molec cm-2. The overall errors on the retrieved NO2 VCDs are on average 21 and 28 % for the Antwerp and Brussels data sets. Low VCD retrievals are mainly limited by noise (1σ slant error), while high retrievals are mainly limited by systematic errors. Compared to coincident car mobile-DOAS measurements taken in Antwerp and Brussels, both data sets are in good agreement with correlation coefficients around 0.85 and slopes close to unity. APEX retrievals tend to be, on average, 12 and 6 % higher for Antwerp and Brussels, respectively. Results demonstrate that the NO2 distribution in an urban environment, and its fine

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

    Science.gov (United States)

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

    2015-04-01

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

  17. Concerning modification of plan for installation of reactors (addition of No.4 reactor and modification of No.1, No.2 and No.3 reactor facilities) at Hamaoka nuclear power plant of Chubu Electric Power Co., Ltd. (reply to inquiry)

    International Nuclear Information System (INIS)

    1988-01-01

    The Nuclear Safety Commission on Oct. 22, 1987 directed the Nuclear Reactor Safety Expert Group to carry out a study on it, made deliberations after receiving a report from the Group on Jun. 28, and submitted the findings to the Minister of International Trade and Industry on Jul. 14. The study and deliberations were intended to determine the conformity of the modifications to the applicable laws relating to control of nuclear material, nuclear fuel and nuclear reactor. The investigation on the location covered the site conditions, geology, seismic environment, meteorology, hydrology, and social environment. The investigations on the safety design of the reactor facilities addressed the design of the overall reactor facilities, anti-earthquake design, and design of other equipment including reactor core, fuel equipment, monitors, reactivity controllers, safety protection equipment, and emergency core cooler. Other investigations included exposure evaluation and accident analysis. It was concluded that the modifications would not have adverse effects on the safety of the reactor facilities. (Nogami, K.)

  18. Integrity evaluation of the pressure vessels of Angra-2 and Angra-3 reactors by stress analysis

    International Nuclear Information System (INIS)

    Gomes, E.

    1978-01-01

    The integrity of the reactor pressure vessel of the unit II/III of the Nuclear Power Station at 'Angras do Reis' is evaluated by stress analysis, through the dynamics relaxation method. For the solution of the problem an axisymmetric model is fixed. Initially, the data of the Oak Ridge Vessel V-7 is compared with those obtained by two computer programs used in this study. The methods used in the computer programs are FEM and DEM. A11 the results are compared with the ASME Code Section III 1974 edition. The range deviation is determined to 99% confidence limit, in order to minimize the error probabilities. Finally, the equivalent intensity stress obtained is calculated and compared with the acceptable values of the ASME Code Section III, 1974 edition [pt

  19. Replacement Nuclear Research Reactor. Supplement to Draft Environmental Impact Statement. Volume 3

    International Nuclear Information System (INIS)

    1999-01-01

    The Draft Environmental Impact Statement for a replacement research reactor at Lucas Heights, was available for public examination and comment for some three months during 1998. A Supplement to the Draft Environmental Impact Statement (Draft EIS) has been completed and was lodged with Environment Australia on 18 January 1999. The Supplement is an important step in the overall environmental assessment process. It reviews submissions received and provides the proponent's response to issues raised in the public review period. General issues extracted from submissions and addressed in the Supplement include concern over liability issues, Chernobyl type accidents, the ozone layer and health issues. Further studies, relating to issues raised in the public submission process, were undertaken for the Supplementary EIS. These studies confirm, in ANSTO's view, the findings of the Draft EIS and hence the findings of the Final EIS are unchanged from the Draft EIS

  20. Replacement Nuclear Research Reactor. Supplement to Draft Environmental Impact Statement. Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-01-01

    The Draft Environmental Impact Statement for a replacement research reactor at Lucas Heights, was available for public examination and comment for some three months during 1998. A Supplement to the Draft Environmental Impact Statement (Draft EIS) has been completed and was lodged with Environment Australia on 18 January 1999. The Supplement is an important step in the overall environmental assessment process. It reviews submissions received and provides the proponent`s response to issues raised in the public review period. General issues extracted from submissions and addressed in the Supplement include concern over liability issues, Chernobyl type accidents, the ozone layer and health issues. Further studies, relating to issues raised in the public submission process, were undertaken for the Supplementary EIS. These studies confirm, in ANSTO`s view, the findings of the Draft EIS and hence the findings of the Final EIS are unchanged from the Draft EIS

  1. 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

    Energy Technology Data Exchange (ETDEWEB)

    Vandecasteele, C.M. [FANC, Ravenstein Street 36, B-1000 Brussels (Belgium)]. E-mail: christian.vandecasteele@fanc.fgov.be; Hardeman, F. [SCK.CEN, Boeretang 200, B-2400 Mol (Belgium); Pauwels, O. [CGCCR, rue Ducale, B-1000 Brussels (Belgium); Bernaerts, M. [CGCCR, rue Ducale, B-1000 Brussels (Belgium); Carle, B. [SCK.CEN, Boeretang 200, B-2400 Mol (Belgium); Sombre, L. [FANC, Ravenstein Street 36, B-1000 Brussels (Belgium)

    2005-07-01

    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.

  2. 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

  3. Mechanical behavior of a fast reactor core: Application of the 3D codes to SUPER PHENIX 1

    International Nuclear Information System (INIS)

    Bernard, A.; Masoni, P.; Dorsselaere, J.P. van

    1983-01-01

    The series of the 3-dimensional mechanical codes of a fast reactor core was used for the first time within the framework of a design study of an industrial reactor: SUPER-PHENIX 1. These codes are the following ones: - ARGOH which calculates the behavior of an isolated subassembly. - HARMONIE which calculates the core mechanical equilibrium - TRACAR which yields a graphic visualization of HARMONIE results, and calculates the handling forces and support reactions - HARMOREA which calculates the reactivity variations between given equilibrium states (for instance: pads effect and diagrid effect); now at the end of its development. The calculations were performed on 1/3 of the SPX1 core. Their purpose is double: - on the one hand, to check design criteria, and provide the loadings for the subassembly mechanical design studies; on the other hand, to evaluate the reactivity effects, related to the horizontal core deformations, and useful for operation and safety studies. The results of these calculations showed that the design criteria were verified for the contractual lifetime of the subassemblies. (orig.)

  4. arXiv Performance of a full scale prototype detector at the BR2 reactor for the SoLid experiment

    CERN Document Server

    Abreu, Y.; Arnold, L.; Ban, G.; Beaumont, W.; Bongrand, M.; Boursette, D.; Castle, B.C.; Clark, K.; Coupé, B.; Cussans, D.; De Roeck, A.; D'Hondt, J.; Durand, D.; Fallot, M.; Ghys, L.; Giot, L.; Guillon, B.; Ihantola, S.; Janssen, X.; Kalcheva, S.; Kalousis, L.N.; Koonen, E.; Labare, M.; Lehaut, G.; Manzanillas, L.; Mermans, J.; Michiels, I.; Moortgat, C.; Newbold, D.; Park, J.; Pestel, V.; Petridis, K.; Piñera, I.; Pommery, G.; Popescu, L.; Pronost, G.; Rademacker, J.; Ryckbosch, D.; Ryder, N.; Saunders, D.; Schune, M.-H.; Simard, L.; Vacheret, A.; Van Dyck, S.; Van Mulders, P.; van Remortel, N.; Vercaemer, S.; Verstraeten, M.; Weber, A.; Yermia, F.

    The SoLid collaboration has developed a new detector technology to detect electron anti-neutrinos at close proximity to the Belgian BR2 reactor at surface level. A 288$\\,$kg prototype detector was deployed in 2015 and collected data during the operational period of the reactor and during reactor shut-down. Dedicated calibration campaigns were also performed with gamma and neutron sources. This paper describes the construction of the prototype detector with a high control on its proton content and the stability of its operation over a period of several months after deployment at the BR2 reactor site. All detector cells provide sufficient light yields to achieve a target energy resolution of better than 20%/$\\sqrt{E(MeV)}$. The capability of the detector to track muons is exploited to equalize the light response of a large number of channels to a precision of 3% and to demonstrate the stability of the energy scale over time. Particle identification based on pulse-shape discrimination is demonstrated with calibr...

  5. 3-dimensional earthquake response analysis of embedded reactor building using hybrid model of boundary elements and finite elements

    International Nuclear Information System (INIS)

    Muto, K.; Motosaka, M.; Kamata, M.; Masuda, K.; Urao, K.; Mameda, T.

    1985-01-01

    In order to investigate the 3-dimensional earthquake response characteristics of an embedded structure with consideration for soil-structure interaction, the authors have developed an analytical method using 3-dimensional hybrid model of boundary elements (BEM) and finite elements (FEM) and have conducted a dynamic analysis of an actual nuclear reactor building. This paper describes a comparative study between two different embedment depths in soil as elastic half-space. As the results, it was found that the earthquake response intensity decreases with the increase of the embedment depth and that this method was confirmed to be effective for investigating the 3-D response characteristics of embedded structures such as deflection pattern of each floor level, floor response spectra in high frequency range. (orig.)

  6. Loss mechanisms of travelling wave direct energy converters for D-{sup 3}He FRC fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, M. E-mail: misikawa@kz.tsukuba.ac.jp; Yamane, T.; Shimizu, Y.; Momota, H.; Tomita, Y.; Miley, G.H

    2000-11-01

    Loss mechanisms of TWDEC for a D-{sup 3}He fuelled FRC fusion reactor are studied with numerical analysis. (1) The self-excitation of a travelling wave has been attained with external electric circuits, which shows the conversion efficiency is about 70-73%. (2) Three-dimensional behavior has been studied by axisymmetric approximation, showing that the loss related to the radial non-uniformity effect without collision with grids is about 5%. (3) The loss related to the collision with grids is the most important, being about 11%. (4) Effects of secondary electrons produced by the collision with the grids are rather small and subrings can suppress the effects of secondary electrons. (5) The loss related to leaked 3.6 MeV {alpha} particles is about 45-60% of the total energy of leaked {alpha} particles. (6) The gross efficiency of TWDEC is estimated to be over 60%.

  7. Nuclear reactor

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.; Struensee, S.

    1976-01-01

    The invention concerns the use of burnable poisons in a nuclear reactor, especially in PWRs, in order to improve the controllability of the reactor. An unsymmetrical arrangement in the lattice is provided, if necessary also by insertion of special rods for these additions. It is proposed to arrange the burnable poisons in fuel elements taken over from a previous burn-up cycle and to distribute them, going out from the side facing the control rods, over not more than 20% of the lenth of the fuel elements. It seems sufficient, for the burnable poisons to bind an initial reactivity of only 0.1% and to become ineffective after normal operation of 3 to 4 months. (ORU) [de

  8. 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

  9. Mathematical Modelling of Catalytic Fixed-Bed Reactor for Carbon Dioxide Reforming of Methane over Rh/Al2O3 Catalyst

    Directory of Open Access Journals (Sweden)

    New Pei Yee

    2008-04-01

    Full Text Available A one-dimensional mathematical model was developed to simulate the performance of catalytic fixed bedreactor for carbon dioxide reforming of methane over Rh/Al2O3 catalyst at atmospheric pressure. The reactionsinvolved in the system are carbon dioxide reforming of methane (CORM and reverse water gas shiftreaction (RWGS. The profiles of CH4 and CO2 conversions, CO and H2 yields, molar flow rate and molefraction of all species as well as reactor temperature along the axial bed of catalyst were simulated. In addition,the effects of different reactor temperature on the reactor performance were also studied. The modelscan also be applied to analyze the performances of lab-scale micro reactor as well as pilot-plant scale reactorwith certain modifications and model verification with experimental data. © 2008 BCREC UNDIP. All rights reserved.[Received: 20 August 2008; Accepted: 25 September 2008][How to Cite: N.A.S. Amin, I. Istadi, N.P. Yee. (2008. Mathematical Modelling of Catalytic Fixed-Bed Reactor for Carbon Dioxide Reforming of Methane over Rh/Al2O3 Catalyst. Bulletin of Chemical Reaction Engineering and Catalysis, 3 (1-3: 21-29. doi:10.9767/bcrec.3.1-3.19.21-29

  10. YBa2Cu3O7-δ thin films deposited by MOCVD vertical reactor with a flow guide

    International Nuclear Information System (INIS)

    Sujiono, E.H.; Negeri Makassar; Sani, R.A.; Saragi, T.; Arifin, P.; Barmawi, M.

    2001-01-01

    The effect of a flow guide in a vertical MOCVD reactor on the deposition uniformity and growth rate of thin YBCO films has been studied. Without the flow guide the growth rates are low, have a poor uniformity and the film composition is not stoichiometric. The growth rate of the films grown using a reactor with the flow guide was approximately twice that without the flow guide. Using this flow guide the growth rates were 0.4-0.7 μm/h for growth temperatures varying between 600 and 750 C, and the crystalline quality as well as the surface morphology of YBCO films on MgO substrates is improved. For films grown at temperatures above 650 C the composition of Y:Ba:Cu is 1:2:3, as confirmed by EDAX spectra. Films deposited without and with the flow guide at 700 C have critical temperatures around 85 and 88 K, respectively. The reduction in ΔT c (T c,zero -T c,onset ) also shows an improvement of the superconducting properties of YBCO thin films deposited with a flow guide. (orig.)

  11. Fabrication and nuclear reactor tests of ultra-thin 3D silicon neutron detectors with a boron carbide converter

    Science.gov (United States)

    Fleta, C.; Guardiola, C.; Esteban, S.; Jumilla, C.; Pellegrini, G.; Quirion, D.; Rodríguez, J.; Lousa, A.; Martínez-de-Olcoz, L.; Lozano, M.

    2014-04-01

    We describe the design, fabrication process and characterization of a thermal neutron detector based on ultra-thin silicon PIN diodes with 3D electrodes and a 10B4C neutron converter layer. The sensors were fabricated on SOI silicon with an active thickness of 20 μm which allows for a low gamma sensitivity, while the 3D structure of the electrodes results in a lower capacitance that in the equivalent planar sensor. The 2.7 μm 10B4C converter layer was deposited through RF magnetron sputtering on a whole silicon wafer, opening the path for mass-production. The detectors were tested in a thermal neutron beam at the nuclear reactor at the Instituto Superior Técnico in Lisbon and their intrinsic detection efficiency for themal neutrons and the gamma sensitivity as a function of the energy threshold were obtained.

  12. Study on Utilization of Super Grade Plutonium in Molten Salt Reactor FUJI-U3 using CITATION Code

    Science.gov (United States)

    Wulandari, Cici; Waris, Abdul; Pramuditya, Syeilendra; Asril, Pramutadi AM; Novitrian

    2017-07-01

    FUJI-U3 type of Molten Salt Reactor (MSR) has a unique design since it consists of three core regions in order to avoid the replacement of graphite as moderator. MSR uses floride as a nuclear fuel salt with the most popular chemical composition is LiF-BeF2-ThF4-233UF4. ThF4 and 233UF4 are the fertile and fissile materials, respectively. On the other hand, LiF and BeF2 working as both fuel and heat transfer medium. In this study, the super grade plutonium will be utilized as substitution of 233U since plutonium is easier to be obtained compared to 233U as main fuel. Neutronics calculation was performed by using PIJ and CITATION modules of SRAC 2002 code with JENDL 3.2 as nuclear data library.

  13. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1988-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab

  14. 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

  15. Draft Genome Sequence of Taylorella equigenitalis Strain MCE529, Isolated from a Belgian Warmblood Horse

    OpenAIRE

    Hébert, Laurent; Touzain, Fabrice; de Boisséson, Claire; Breuil, Marie-France; Duquesne, Fabien; Laugier, Claire; Blanchard, Yannick; Petry, Sandrine

    2014-01-01

    Taylorella equigenitalis is the causative agent of contagious equine metritis (CEM), a sexually transmitted infection of horses. We herein report the genome sequence of T. equigenitalis strain MCE529, isolated in 2009 from the urethral fossa of a 15-year-old Belgian Warmblood horse in France.

  16. Draft Genome Sequence of Taylorella equigenitalis Strain MCE529, Isolated from a Belgian Warmblood Horse.

    Science.gov (United States)

    Hébert, Laurent; Touzain, Fabrice; de Boisséson, Claire; Breuil, Marie-France; Duquesne, Fabien; Laugier, Claire; Blanchard, Yannick; Petry, Sandrine

    2014-11-26

    Taylorella equigenitalis is the causative agent of contagious equine metritis (CEM), a sexually transmitted infection of horses. We herein report the genome sequence of T. equigenitalis strain MCE529, isolated in 2009 from the urethral fossa of a 15-year-old Belgian Warmblood horse in France. Copyright © 2014 Hébert et al.

  17. Irradiation-assisted stress corrosion cracking[1997 Scientific Report of the Belgian Nuclear Research Centre

    Energy Technology Data Exchange (ETDEWEB)

    Moons, F.

    1998-07-01

    The programme on corrosion at the Belgian Nuclear Research Centre SCK-CEN started in 1996 and focusses on modelling irradiation-assisted stress corrosion cracking and on developing in-pile electrochemical sensors and diagnostic equipment. The objective of this programme is to predict the behaviour of LWR core internals with respect to IASCC. Progress for 1997 is summarised.

  18. Late onset of Strongyloides stercoralis meningitis in a retired Belgian miner.

    Science.gov (United States)

    Pypen, Y; Oris, E; Meeuwissen, J; Vander Laenen, M; Van Gompel, F; Coppens, G

    2015-12-01

    We report a rare case of Strongyloides stercoralis meningitis in an immunocompromised patient treated for a lung carcinoma. Despite his Belgian origin, he was infected with S. stercoralis due to his former work as a miner. Although mostly prevalent in (sub)tropical areas, there are temperate regions where this nematode can occur.

  19. A Belgian margarine manufacturer is testing the green cogeneration; Un margarinier belge experimente la cogeneration verte

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2006-01-15

    Aigremont, an independent Belgian margarine manufacturer, is exploiting a 'green cogeneration' unit fueled with vegetal and animal fats. This unit generates 770 kW electrical power which is injected into the power grid and the same quantity of thermal power which is consumed by the fabrication process. Short paper. (J.S.)

  20. The Use of English as Ad Hoc Institutional Standard in the Belgian Asylum Interview

    Science.gov (United States)

    Maryns, Katrijn

    2017-01-01

    In institutional settings of globalization, labelled languages are generally preferred over multilingual repertoires and mobile language resources. Drawing on linguistic-ethnographic analysis of the way English is treated as an invariable "ad hoc" idiom in the Belgian asylum interview, this article demonstrates how institutional measures…

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

    NARCIS (Netherlands)

    Minnaert, A

    1996-01-01

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

  2. 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.…

  3. 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

  4. Estimation of price-cost margins and union bargaining power for Belgian manufacturing

    NARCIS (Netherlands)

    Dobbelaere, S.

    2004-01-01

    This paper extends Hall's (1988) [Hall, R.E., 1988. The relationship between price and marginal cost in US industry, Journal of Political Economy, 96, 921-947] methodology to analyze imperfections in both the product and the labor market for firms in the Belgian manufacturing industry over the

  5. 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...

  6. Avoiding Compliance and Resistance through Collaboration? A Belgian Teaching Portfolio Case

    Science.gov (United States)

    Wouters, Pascale; Clement, Mieke; Frenay, Mariane; Buelens, Herman; Gilis, Annelies

    2014-01-01

    In this paper, the authors describe the implementation process of a teaching portfolio at a Belgian university. The case is intriguing because it departs substantially from what others have described as the typical antagonistic way in which academic developers interact with formal leaders. Rather than being caught in an edgy game of compliance and…

  7. 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

  8. Local multigrid mesh refinement in view of nuclear fuel 3D modelling in pressurised water reactors

    International Nuclear Information System (INIS)

    Barbie, L.

    2013-01-01

    The aim of this study is to improve the performances, in terms of memory space and computational time, of the current modelling of the Pellet-Cladding mechanical Interaction (PCI), complex phenomenon which may occurs during high power rises in pressurised water reactors. Among the mesh refinement methods - methods dedicated to efficiently treat local singularities - a local multi-grid approach was selected because it enables the use of a black-box solver while dealing few degrees of freedom at each level. The Local Defect Correction (LDC) method, well suited to a finite element discretization, was first analysed and checked in linear elasticity, on configurations resulting from the PCI, since its use in solid mechanics is little widespread. Various strategies concerning the implementation of the multilevel algorithm were also compared. Coupling the LDC method with the Zienkiewicz-Zhu a posteriori error estimator in order to automatically detect the zones to be refined, was then tested. Performances obtained on two-dimensional and three-dimensional cases are very satisfactory, since the algorithm proposed is more efficient than h-adaptive refinement methods. Lastly, the LDC algorithm was extended to nonlinear mechanics. Space/time refinement as well as transmission of the initial conditions during the re-meshing step were looked at. The first results obtained are encouraging and show the interest of using the LDC method for PCI modelling. (author) [fr

  9. Conceptual design study for the demonstration reactor of JSFR. (3) Safety design and evaluation

    International Nuclear Information System (INIS)

    Tani, Akihiro; Shimakawa, Yoshio; Kubo, Shigenobu; Fujimura, Ken; Yamano, Hidemasa

    2011-01-01

    This paper describes the result of conceptual safety design and evaluation for the demonstration plant of Japan sodium-cooled fast reactor (JSFR), which was preliminarily conducted for providing information necessary to decide the plant specification for further design study. The plant major specifications except for output power and safety design concept are almost the same as those of the commercial JSFR. A set of safety evaluation for typical design basis events (DBEs) is mainly focused here, which was conducted for the 750 MWe design. Safety analyses for DBEs evaluation were performed on the basis of conservative assumptions using a one-dimensional flow network code with point kinetics. For representative DBEs, transient over power type events and loss of flow type events were analyzed. The long-term loss-of-offsite power event was also calculated to evaluate the natural circulation decay heat removal system. All analytical results showed to meet tentative safety criteria, thus it was confirmed that the safety design concept of JSFR is feasible against DBEs. (author)

  10. Draft environmental impact statement for the siting, construction, and operation of New Production Reactor capacity. Volume 3, Sections 7-12, Appendices A-C

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

    This Environmental Impact Statement (EIS) assesses the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation`s nuclear defense requirements well into the 21st century. A capacity equivalent to that of about a 3,000-megawatt (thermal) heavy-water reactor was assumed as a reference basis for analysis in this EIS; this is the approximate capacity of the existing production reactors at DOE`s Savannah River Site near Aiken, South Carolina. The EIS programmatic alternatives address Departmental decisions to be made on whether to build new production facilities, whether to build one or more complexes, what size production capacity to provide, and when to provide this capacity. Project-specific impacts for siting, constructing, and operating new production reactor capacity are assessed for three alternative sites: the Hanford Site near Richland, Washington; the Idaho National Engineering Laboratory near Idaho Falls, Idaho; and the Savannah River Site. For each site, the impacts of three reactor technologies (and supporting facilities) are assessed: a heavy-water reactor, a light-water reactor, and a modular high-temperature gas-cooled reactor. Impacts of the no-action alternative also are assessed. The EIS evaluates impacts related to air quality; noise levels; surface water, groundwater, and wetlands; land use; recreation; visual environment; biotic resources; historical, archaeological, and cultural resources; socioeconomics; transportation; waste management; and human health and safety. The EIS describes in detail the potential radioactive releases from new production reactors and support facilities and assesses the potential doses to workers and the general public. This volume contains references; a list of preparers and recipients; acronyms, abbreviations, and units of measure; a glossary; an index and three appendices.

  11. RB research nuclear reactor - Annual report for 1985, I - III

    International Nuclear Information System (INIS)

    Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Jevremovic, M.; Ilic, I.

    1985-01-01

    This paper describes the operation of the RB reactor during 1985. The report includes 3 parts: Engineering description and operation of the RB reactor, reactor components and maintenance, and RB reactor utilization in 1985 [sr

  12. 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

  13. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

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

  14. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

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

  15. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

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

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

    Science.gov (United States)

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

    2015-04-01

    Chocolate confectionery fillings are generally regarded as microbiologically stable. The stability of these fillings is largely due to the general practice of adding either alcohol or preservatives. Consumer demands are now stimulating producers to move away from adding alcohol or other preservatives to their confectionery fillings and instead to search for innovative formulations. Such changes in composition can influence the shelf life of the product and may lead to spoilage by xerophilic fungi. The aim of this study was to test whether the production environment of Belgian chocolate confectionery factories and common ingredients of chocolate confectioneries could be potential sources of contamination with xerophilic fungal species. In the factory environment, the general and strictly xerophilic fungal spore load was determined using an RCS Air Sampler device in combination with DG18 and MY50G medium, respectively. Four basic ingredients of chocolate confectionery fillings were also examined for fungal spore levels using a direct plating technique. Detected fungi were identified to species level by a combination of morphological characterization and sequence analysis. Results indicated a general fungal spore load in the range of 50-250 colony forming units per cubic meter of air (CFU/m(3) air) and a more strict xerophilic spore load below 50 CFU/m(3) air. These results indicate rather low levels of fungal spores present in the factory environment. The most prevalent fungi in the factory environment were identified as Penicillium spp., particularly Penicillium brevicompactum. Examination of the basic ingredients of confectionery fillings revealed nuts to be the most likely potential source of direct contamination. In nuts, the most prevalent fungal species identified were Eurotium, particularly Eurotium repens. Copyright © 2014 Elsevier Ltd. All rights reserved.

  17. Multipurpose research reactors

    International Nuclear Information System (INIS)

    1988-01-01

    The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  18. Irradiation induced tensile property change of SA 508 Cl.3 reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Chi, Se-Hwan; Hong, Jun-Hwa; Kuk, Il-Hiun

    1998-01-01

    Irradiation induced tensile property change of four kinds of reactor pressure vessel steels manufactured by different steel refining process was compared based on the differences in the unirradiated and irradiated microstructure. Microvickers hardness, indentation, and miniature tensile specimen tests were conducted for mechanical property measurement and optical microscope (OM) and transmission electron microscope (TEM) were used for microstructural characterization. Specimens were 2 irradiated to a neutron fluence of 2.7x10 19 n/cm 2 (E ≥ 1 MeV) at 288 deg. C. Investigation on the unirradiated microstructures showed largely a same microstructure in that tempered acicular bainite and ferrite with bainitic phase prevailing in the unirradiated condition. Band-shaped segregations were also clearly observed except a kind of materials. A large difference in the unirradiated microstructure appeared in the grain size and carbide microstructure. Of carbide microstructures, noticeable differences were observed in the size and distribution of cementite, and bainitic lath microstructures. No noticeable changes were observed in the optical and thin film TEM microstructures after irradiation. Complicated microstructural. state of heat treated bainitic low alloy microstructure prevents easy quantification of microstructural changes due to irradiation. Apparent differences, however, were observed in the results of mechanical testing. Results of tensile testing and hardness measurement show that a steel refined by vacuum carbon deoxidation(VCD) method exhibits the highest radiation hardening behavior. Some of mechanical testing results on irradiated materials were possible to understand based on the initial microstructure, but further investigations using a wide array of sophisticated tools (for example, SANS, APFIM) are required to understand and characterize irradiation induced defects that are responsible for irradiation hardening behavior but are not revealed by

  19. Experimental fast reactor Joyo emergency operation act in Northeastern Japan Earthquake 3.11

    International Nuclear Information System (INIS)

    Karube, Koji; Aita, Tsuyoshi; Okawa, Toshikatsu

    2012-01-01

    Experimental fast reactor 'Joyo' under facility's periodic inspection received damage at its power incoming unit due to the 2011 off the Pacific Coast of Tohoku Earthquake, which incurred loss-of-offsite-power. Immediately, two emergency diesel generators (D/G) automatically started, which supplied emergency system power. During eight days before the power incoming unit got provisional restoration, power supply through D/G continued. During this period, emergency measures for fuel and cooling water securement for D/G were taken, which forced unexperienced long-term load operation for D/G. This paper reports the change of plant conditions of Joyo, as well as the measures taken for keeping fuel and cooling water for continuing the operation of D/G. Since the main machines of the primary system and secondary system were functioning normally after the earthquake, the plant could be maintained in a stable state. As for the fuel for D/G, reduction in fuel consumption based on load suppression was implemented, and the cooperation of trading firms, related facilities, and local suppliers secured the fuel that supported the long-term operation of D/G. As for the cooling water, the cooperation of a self-defense fire brigade allowed to secure the required amount by utilizing the reserved water in the water tank for fire prevention. From these series of experiences of handling the plant, it was possible to extract the future challenges against the time when a massive earthquake or prolonged power failure has occurred. (A.O.)

  20. 3-D thermal stress analysis of hot spots in reactor piping using BEM

    International Nuclear Information System (INIS)

    Bains, R.S.; Sugimoto, Jun

    1994-08-01

    A three-dimensional steady state thermoelastic analysis has been conducted on the hot leg of a pressurized water reactor(PWR) containing localized hot spots resulting from fission product aerosol deposition occurring during a hypothetical severe accident. The boundary element method (BEM) of numerical solution was successfully employed to investigate the structural response of the hot leg. Convergence of solution can be realized provided sufficiently large number of elements are employed and correct modelling of the temperature transition region (TTR) adjacent to the hot spot on the inner surface is conducted. The only correct temperature field across the TTR is that which can be represented by the interpolation functions employed in the BEM code. Further, incorrect solutions can also be generated if the TTR is too thin. The nature of the deformation at the hot spot location depends on whether the thermal boundary condition on the outer surface of the hot leg is one of constant temperature or adiabatic. The analysis shows that at the location of the hot spot on the inner surface large compressive stresses can be established. On the outer surface at the same location, large tensile stresses can be established. The presence of these large stress elevations in the vicinity of the hot spot could be detrimental to the integrity of the hot leg. The tensile stresses are extremely important since they can act as sites of crack initiation and subsequent propagation. Once a crack propagates through the thickness, leak worthiness of the hot leg comes into question. Consequently, additional analysis incorporating the effects of plasticity and temperature dependence of the material properties must be conducted to ascertain the integrity of the hot leg. (J.P.N.)

  1. A 3D heat conduction model for block-type high temperature reactors and its implementation into the code DYN3D

    International Nuclear Information System (INIS)

    Baier, Silvio; Kliem, Soeren; Rohde, Ulrich

    2011-01-01

    The gas-cooled high temperature reactor is a concept to produce energy at high temperatures with a high level of inherent safety. It gets special attraction due to e.g. high thermal efficiency and the possibility of hydrogen production. In addition to the PBMR (Pebble Bed Modular Reactor) the (V)HTR (Very high temperature reactor) concept has been established. The basic design of a prismatic HTR consists of the following elements. The fuel is coated with four layers of isotropic materials. These so-called TRISO particles are dispersed into compacts which are placed in a graphite block matrix. The graphite matrix additionally contains holes for the coolant gas. A one-dimensional model is sufficient to describe (the radial) heat transfer in LWRs. But temperature gradients in a prismatic HTR can occur in axial as well as in radial direction, since regions with different heat source release and with different coolant temperature heat up are coupled through the graphite matrix elements. Furthermore heat transfer into reflector elements is possible. DYN3D is a code system for coupled neutron and thermal hydraulics core calculations developed at the Helmholtzzentrum Dresden-Rossendorf. Concerning neutronics DYN3D consists of a two-group and multi-group diffusion approach based on nodal expansion methods. Furthermore a 1D thermal-hydraulics model for parallel coolant flow channels is included. The DYN3D code was extensively verified and validated via numerous numerical and experimental benchmark problems. That includes the NEA CRP benchmarks for PWR and BWR, the Three-Miles-Island-1 main steam line break and the Peach Bottom Turbine Trip benchmarks, as well as measurements carried out in an original-size VVER-1000 mock-up. An overview of the verification and validation activities can be found. Presently a DYN3D-HTR version is under development. It involves a 3D heat conduction model to deal with higher-(than one)-dimensional effects of heat transfer and heat conduction in

  2. Biomass granulation in an upflow anaerobic sludge blanket reactor treating 500 m3/day low-strength sewage and post treatment in high-rate algal pond.

    Science.gov (United States)

    Chatterjee, Pritha; Ghangrekar, M M

    2017-09-01

    A pilot-scale upflow anaerobic sludge blanket-moving bed biofilm (UASB-MBB) reactor followed by a high-rate algal pond (HRAP) was designed and operated to remove organic matter, nutrients and pathogens from sewage and to facilitate reuse. For an influent chemical oxygen demand (COD) concentration of 233 ± 20 mg/L, final effluent COD was 50 ± 6 mg/L. Successful biomass granulation was observed in the sludge bed of the upflow anaerobic sludge blanket (UASB) reactor after 5 months of operation. Ammonia removal in HRAP was 85.1 ± 2.4% with average influent and effluent ammonia nitrogen concentrations of 20 ± 3 mg/L and 3 ± 1 mg/L, respectively. Phosphate removal after treatment in the HRAP was 91 ± 1%. There was a 2-3 log scale pathogen removal after treatment in HRAP with most probable number (MPN) of the final effluent being 600-800 per 100 mL, which is within acceptable standards for surface irrigation. The blackwater after treatment in UASB-MBBR-HRAP is being reused for gardening and landscaping. This proper hydro-dynamically designed UASB reactor demonstrated successful granulation and moving bed media improved sludge retention in UASB reactor. This combination of UASB-MBB reactor followed by HRAP demonstrated successful sewage treatment for a year covering all seasons.

  3. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (CAPSULE 2) of Ulchin Nuclear Power Plant Unit 3

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2006-12-15

    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 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejon after the capsule was transported from Ulchin 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 Ulchin unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.674E 18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.920 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.913E 18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 9.249E 18 and 1.834E 19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  4. Radio-activity measurements inside the pressure-vessel of the reactor G 3 after 4 years operation

    International Nuclear Information System (INIS)

    Chassany, J.Ph.; Guillermin, P.; Delmar, J.

    1965-01-01

    At the end of the piping coming into the vessel, the dose rate reached 75 mR/hr and 100 mR/hr near the deflector. On the other side of this deflector it was still 100 mR/hr and then increased rapidly to over 1 R/hr at 1 metre distance from the starting-up chambers. On the sides, the flux tended to decrease (80 mR/hr) and was 2 R/hr at a height of 3 metres. This dose rate could certainly have been decreased by discharging the peripheral zone of the reactor. Consequently it should be possible to intervene if necessary, on condition that great care is taken to avoid contamination and that the total dose is followed as precisely as possible during the operations. (authors) [fr

  5. Simulation of the fault transitory of the feedwater controller in a Boiling water reactor with the Ramona-3B code

    International Nuclear Information System (INIS)

    Hernandez M, J.L.; Ortiz V, J.

    2005-01-01

    The obtained results when carrying out the simulation of the fault transitory of the feedwater controller (FCAA) with the Ramona-3B code, happened in the Unit 2 of the Laguna Verde power plant (CNLV), in September of the year 2000 are presented. The transitory originates as consequence of the controller's fault of speed of a turbo pump of feedwater. The work includes a short description of the event, the suppositions considered for the simulation and the obtained results. Also, a discussion of the impact of the transitory event is presented on aspects of reactor safety. Although the carried out simulation is limited by the capacities of the code and for the lack of available information, it was found that even in a conservative situation, the power was incremented only in 12% above the nominal value, while that the thermal limit determined by the minimum reason of the critical power, MCPR, always stayed above the limit values of operation and safety. (Author)

  6. Heat transfer tests conducted on full-scale model, to investigate cooling conditions of EL.3 experimental reactor

    International Nuclear Information System (INIS)

    Raievski, R.; Bousquet, M.; Braudeau, M.; Milliat, M.

    1958-01-01

    For such high heat flux density as is released in the channels of EL3 reactor (2.10 6 kcal/m 2 h on the hottest point) cooling conditions have proved to be satisfactory, that is free from nucleate boiling. The arrangements provided for these tests and the technique used for measurements (of temperature particularly) are specified. Two fields have been investigated: in the former (forced convection without nucleate boiling) a good agreement is found with Colburn's formula. The influence of the ratio L/D is pointed out. The latter field is of forced convection with beginning of nucleate boiling; there the observed raise of the transfer coefficient has been shown occurring with some delay. (author) [fr

  7. Propagation of epithermal neutrons in the empty mesh of the EDF3 reactor. Application of Zeus IV

    International Nuclear Information System (INIS)

    Brisbois, Jacques; Sonnet, Alain; Rafalovitch, Armand

    1969-10-01

    The authors report the verification of the Zeus 4 Monte Carlo code by applying it to an actual example of neutron propagation in a large enclosure, the empty mesh of the EDF3 reactor, supplied through the box holes (for the passage of heat transfer fluid ducts). This enclosure is represented by 105 parallelepipedic meshes in which are placed 4 plane and square sources of same cross section as the holes, and emitting inwards in cos θ. Source intensity is determined by flow rate measurements during tests in the air, interpreted by a Zeus-based calculation and channel power measurements during a 387 MW level operation. The presence of ducts is taken into account by applying a global attenuation coefficient to the sources. Calculation and experiment results agree very well in view of geometry complexity [fr

  8. Modelling 3D crack propagation in ageing graphite bricks of Advanced Gas-cooled Reactor power plant

    Directory of Open Access Journals (Sweden)

    Thi-Tuyet-Giang Vo

    2015-10-01

    Full Text Available In this paper, crack propagation in Advanced Gas-cooled Reactor (AGR graphite bricks with ageing properties is studied using the eXtended Finite Element Method (X-FEM. A parametric study for crack propagation, including the influence of different initial crack shapes and propagation criteria, is conducted. The results obtained in the benchmark study show that the crack paths from X-FEM are similar to the experimental ones. The accuracy of the strain energy release rate computation in a heterogeneous material is also evaluated using a finite difference approach. Planar and non-planar 3D crack growth simulations are presented to demonstrate the robustness and the versatility of the method utilized. Finally, this work contributes to the better understanding of crack propagation behaviour in AGR graphite bricks and so contributes to the extension of the AGR plants’ lifetimes in the UK by reducing uncertainties.

  9. Measuring Neutron Spectrum at MIT Research Reactor Utilizing He-3 Bonner Cylinder Approach with an Unfolding Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Leder, A. [MIT; Anderson, A. J. [Chicago U., KICP; Billard, J. [Lyon, IPN; Figueroa-Feliciano, E. [Northwestern U.; Formaggio, J. A. [MIT; Hasselkus, C. [Wisconsin U., Madison; Newman, E. [MIT; Palladino, K. [Wisconsin U., Madison; Phuthi, M. [MIT; Winslow, L. [MIT; Zhang, L. [MIT

    2017-10-02

    The Ricochet experiment seeks to measure Coherent (neutral-current) Elastic Neutrino-Nucleus Scattering using dark-matter-style detectors with sub-keV thresholds placed near a neutrino source, such as the MIT (research) Reactor (MITR), which operates at 5.5 MW generating approximately 2.2e18 neutrinos/second at the core. Currently, Ricochet is characterizing the backgrounds at MITR, the main component of which comes in the form of neutrons emitted from the core simultaneous with the neutrino signal. To characterize this background, we wrapped a Bonner cylinder around a He-3 thermal neutron detector, whose data was then unfolded to produce a neutron energy spectrum across several orders of magnitude. We discuss the resulting spectrum and its implications for deploying Ricochet in the future at the MITR site as well as the feasibility of reducing this background level via the addition of polyethylene shielding around the detector setup.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hopkins, G.C. (comp.)

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

  11. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  12. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

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

  13. Development of the body size in stallions of selected Bohemian-Moravian Belgian horse, Silesian Noriker and Noriker breeds in the Czech Republic

    Directory of Open Access Journals (Sweden)

    Jan NAVRÁTIL

    2016-06-01

    Full Text Available The aim of this work was to evaluate the basic body measurements (stick height at withers - KVH, tape height at withers - KVP, chest perimeter - OHR, shin perimeter - OHOL from data available from the 40th years of the 20th century in the Bohemian-Moravian Belgian horse, Silesian Noriker and Noriker breeds. The evaluation included a total of 1,080 stallions aged 2-3 years, used in mating and breeding. Processing and evaluation of a data set was done using Microsoft Office Excel and the statistical program SAS 9.3. Numerous statistically significant differences were found among the evaluated breeds (P < 0.05 to 0.01. The highest values of KVH, KVP and OHR were achieved by stallions of Bohemian-Moravian Belgian horse. The increase of the basic dimensions from 40´s years to 70´s – 90´s years was found during a detailed evaluation focusing on decade of stallions’ birth year. After year 2000, there was a dramatic +drop in the values of the fundamental physical dimensions. Statistically significant differences (P < 0.05 to 0.01 were evaluated primarily between stallions born in the 70´s years, and most other decades. A declining tendency in all assessed dimensions was observed during evaluation of the race and decade interaction effect mainly for the Czech-Moravian Belgian horse stallions born from 70´s, respectively 80´s. Apparent downward trend between the 90´s and stallions born after year 2000 was detected especially for stallions of Silesian Noriker breed. Decline of body size parameters was not observed for stallions of Noriker breed. The more likely slightly increased tendencies for KVH, KVP, OHR and OHOL were determined in Noriker stallions.

  14. The Shepherds' Tale: A Genome-Wide Study across 9 Dog Breeds Implicates Two Loci in the Regulation of Fructosamine Serum Concentration in Belgian Shepherds.

    Directory of Open Access Journals (Sweden)

    Simon K G Forsberg

    Full Text Available Diabetes mellitus is a serious health problem in both dogs and humans. Certain dog breeds show high prevalence of the disease, whereas other breeds are at low risk. Fructosamine and glycated haemoglobin (HbA1c are two major biomarkers of glycaemia, where serum concentrations reflect glucose turnover over the past few weeks to months. In this study, we searched for genetic factors influencing variation in serum fructosamine concentration in healthy dogs using data from nine dog breeds. Considering all breeds together, we did not find any genome-wide significant associations to fructosamine serum concentration. However, by performing breed-specific analyses we revealed an association on chromosome 3 (pcorrected ≈ 1:68 × 10-6 in Belgian shepherd dogs of the Malinois subtype. The associated region and its close neighbourhood harbours interesting candidate genes such as LETM1 and GAPDH that are important in glucose metabolism and have previously been implicated in the aetiology of diabetes mellitus. To further explore the genetics of this breed specificity, we screened the genome for reduced heterozygosity stretches private to the Belgian shepherd breed. This revealed a region with reduced heterozygosity that shows a statistically significant interaction (p = 0.025 with the association region on chromosome 3. This region also harbours some interesting candidate genes and regulatory regions but the exact mechanisms underlying the interaction are still unknown. Nevertheless, this finding provides a plausible explanation for breed-specific genetic effects for complex traits in dogs. Shepherd breeds are at low risk of developing diabetes mellitus. The findings in Belgian shepherds could be connected to a protective mechanism against the disease. Further insight into the regulation of glucose metabolism could improve diagnostic and therapeutic methods for diabetes mellitus.

  15. Weld residual stress predictions in reactor vessel head penetrations

    Energy Technology Data Exchange (ETDEWEB)

    Suanno, Rodolfo L.M.; Ferrari, Lucio D.B. [ELETROBRAS Termonuclear S.A. - ELETRONUCLEAR, Rio de Janeiro, RJ (Brazil). Stress Analysis Department], e-mail: rsuanno@eletronuclear.gov.br; Teughels, Anne; Malekian, Christian [Tractebel Engineering, Brussels (Belgium). Reliability Nuclear Department], e-mail: anne.teughels@tractebel.com

    2009-07-01

    The penetrations in the early Pressurized Water Reactors Vessels are characterized by Alloy 600 tubes, welded by Alloy 182/82. The Alloy 600 tubes have been shown to be susceptible to PWSCC (Primary Water Stress Corrosion Cracking) which may lead to crack forming. The cracking mechanism is driven mainly by the welding residual stress and, in a second place, by the operational stress in the weld region. It is therefore of big interest to quantify the weld residual stress field correctly. In this paper the weld residual stress field is calculated by finite elements using a sequentially coupled approach, which is well known in literature. It includes a transient thermal analysis simulating the heating during the multipass welding, followed by a transient thermo-mechanical analysis for the determination of the stresses involved with it. The welding consists of a sequence of weld beads, each of which is deposited in its entirety, at once, instead of gradually. Central as well as eccentric sidehill nozzles on the vessel head are analyzed in the paper. For the former a 2- dimensional axisymmetrical finite element model is used, whereas for the latter a 3-dimensional model is set up. Different positions on the vessel head are compared and the influence of the sidehill effect is illustrated. In the framework of a common project for Angra 1, Tractebel Engineering (Belgium) and ELETRONUCLEAR (Nuclear Utility, Brazil) had the opportunity to compare their analysis method, which they applied to the Belgian and the Brazilian nuclear reactors, respectively. The global approach in both cases is very similar but is applied to different configurations, specific for each plant. In the article the results of both cases are compared. (author)

  16. Insertion of control systems models in the Almod 3 computer code for the simulation of Angra I reactor start-up tests

    International Nuclear Information System (INIS)

    Camargo, C.T.M.

    1981-09-01

    The Almod 3 computer code was modified, aiming at the simulation of Angra I nuclear power plant behavior during some reactor start-up tests. The results obtained with the modified computer code (Almod 3W) are compared with those obtained with the Retran computer code. (E.G.) [pt

  17. Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 17, 2011, 3:00 PM status

    International Nuclear Information System (INIS)

    2011-01-01

    This situation note is established according to the information gained on March 17, 2011, at 3:00 PM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of all 6 reactors of the Fukushima I site (Dai-ichi) and of their spent fuel pools, as well as the situation of the reactors No. 1, 2, 3 and 4 of the Fukushima II site (Daini), and of the Onagawa and Tokai power plants is briefly presented with the progress of the accident management actions. (J.S.)

  18. Regulating the Emission Spectrum of CsPbBr3 from Green to Blue via Controlling the Temperature and Velocity of Microchannel Reactor

    OpenAIRE

    Yong Tang; Hanguang Lu; Longshi Rao; Zongtao Li; Xinrui Ding; Caiman Yan; Binhai Yu

    2018-01-01

    The ability to precisely obtain tunable spectrum of lead halide perovskite quantum dots (QDs) is very important for applications, such as in lighting and display. Herein, we report a microchannel reactor method for synthesis of CsPbBr3 QDs with tunable spectrum. By adjusting the temperature and velocity of the microchannel reactor, the emission peaks of CsPbBr3 QDs ranging from 520 nm to 430 nm were obtained, which is wider than that of QDs obtained in a traditional flask without changing hal...

  19. Integral data testing of JENDL-3.2 for fusion reactor and shielding applications

    International Nuclear Information System (INIS)

    Oyama, Yukio

    1995-01-01

    Integral data testing of JENDL-3.2 is being performed in the activities of two working groups of the Japanese Nuclear Data Committee. The continuous and group-wise libraries prepared from JENDL-3.2 are planned to be tested by the working groups. In this paper, the continuous library FSXLIB-J3R2 processed from JENDL-3.2 for MCNP was tested for fission and fusion neutrons using data of integral experiments and compared to the results of JENDL-3.1. The results of integral data testing of JENDL-3.2 for fusion and shielding application are reviewed. (author)

  20. Increasing incidence of carbapenemase-producing Escherichia coli and Klebsiella pneumoniae in Belgian hospitals.

    Science.gov (United States)

    De Laveleye, M; Huang, T D; Bogaerts, P; Berhin, C; Bauraing, C; Sacré, P; Noel, A; Glupczynski, Y

    2017-01-01

    Carbapenemase-producing Enterobacteriaceae are increasingly reported worldwide. The aim of the study was to determine the incidence and molecular epidemiology of carbapenemase-producing (CP) Escherichia coli and Klebsiella pneumoniae (CP-E/K) in Belgium. Eleven hospital-based laboratories collected carbapenem non-susceptible (CNS) isolates of E. coli and K. pneumoniae detected in clinical specimens from January 2013 to December 2014. All CNS strains were tested for carbapenemase production and typed by multilocus sequence typing (MLST) for a 6-month period as part of the European Survey on Carbapenemase-Producing Enterobacteriaceae in Europe (EuSCAPE) structured survey. In addition, an equal number of carbapenem-susceptible isolates collected were preserved as a control group for risk factor analysis. The overall incidence rate of CP-E/K isolates in hospitals increased from 0.124 in 2013 to 0.223 per 1000 admissions in 2014. From November 2013 to April 2014, 30 CP K. pneumoniae [OXA-48 (n = 16), KPC (n = 13), OXA-427 (n = 1)] and five CP E. coli [OXA-48 (n = 3), NDM (n = 1), OXA-427 (n = 1)] isolates were detected in ten hospitals. The 16 OXA-48-producing K. pneumoniae strains were distributed into eight sequence types (STs), while the 13 KPC-producing K. pneumoniae clustered into three STs dominated by ST512 (n = 7) and ST101 (n = 5). Compared to controls, we observed among CP-E/K carriers significantly higher proportion of males, respiratory origins, previous hospitalization, nosocomial setting, and a significantly lower proportion of bloodstream infections. Our study confirms the rapid spread of CP-E/K in Belgian hospitals and the urgent need for a well-structured and coordinated national surveillance plan in order to limit their dissemination.

  1. Molecular source tracking of predominant lactic acid bacteria in traditional Belgian sourdoughs and their production environments.

    Science.gov (United States)

    Scheirlinck, I; Van der Meulen, R; De Vuyst, L; Vandamme, P; Huys, G

    2009-04-01

    To investigate the circulation of predominant sourdough lactic acid bacteria (LAB) species in the production environment of two Belgian artisan sourdough bakeries. Isolates were collected from sourdoughs, flour, hands of the baker and air in the bakery setting and taxonomically characterized using repetitive element sequence-based PCR fingerprinting, pheS and/or 16S rRNA gene sequencing and amplified fragment length polymorphism (AFLP) analysis. In parallel, PCR-DGGE (denaturing gradient gel electrophoresis) analysis of V3-16S rDNA amplicons was applied to visualize the predominant bacterial population in the sourdoughs and the corresponding bakery environment (flour, hands of the baker, air and bakery equipment). Both approaches revealed that sourdoughs produced at D01 and D10 were mainly dominated by Lactobacillus spicheri and L. plantarum and by L. sanfranciscensis, respectively, and that these LAB species also circulated in the corresponding bakery environment. Furthermore, AFLP fingerprinting demonstrated that sourdough and bakery environment isolates of these species were genetically indistinguishable. For more sensitive source-tracking, SYBR Green-based real-time PCR assays were developed using species-specific primers targeting the pheS gene of L. plantarum and L. sanfranciscensis, detected in air samples from D01 and D10, respectively. The results obtained in this study indicate that specific strains of LAB persist in artisan doughs over years and circulate in the bakery environment. Furthermore, the importance of air as a potential carrier of LAB in artisan bakery environments was demonstrated. PheS-based real-time PCR can be used to detect, quantify and/or monitor specific LAB species (e.g. starter cultures) in sourdough and bakery environment samples.

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

    Science.gov (United States)

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

    2015-01-01

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

  3. Synergism of the method of characteristic, R-functions and diffusion solution for accurate representation of 3D neutron interactions in research reactors using the AGENT code system

    International Nuclear Information System (INIS)

    Hursin, Mathieu; Xiao Shanjie; Jevremovic, Tatjana

    2006-01-01

    This paper summarizes the theoretical and numerical aspects of the AGENT code methodology accurately applied for detailed three-dimensional (3D) multigroup steady-state modeling of neutron interactions in complex heterogeneous reactor domains. For the first time we show the fine-mesh neutron scalar flux distribution in Purdue research reactor (that was built over forty years ago). The AGENT methodology is based on the unique combination of the three theories: the method of characteristics (MOC) used to simulate the neutron transport in two-dimensional (2D) whole core heterogeneous calculation, the theory of R-functions used as a mathematical tool to describe the true geometry and fuse with the MOC equations, and one-dimensional (1D) higher-order diffusion correction of 2D transport model to account for full 3D heterogeneous whole core representation. The synergism between the radial 2D transport and the 1D axial transport (to take into account the axial neutron interactions and leakage), called the 2D/1D method (used in DeCART and CHAPLET codes), provides a 3D computational solution. The unique synergism between the AGENT geometrical algorithm capable of modeling any current or future reactor core geometry and 3D neutron transport methodology is described in details. The 3D AGENT accuracy and its efficiency are demonstrated showing the eigenvalues, point-wise flux and reaction rate distributions in representative reactor geometries. The AGENT code, comprising this synergism, represents a building block of the computational system, called the virtual reactor. Its main purpose is to perform 'virtual' experiments and demonstrations of various mainly university research reactor experiments

  4. The ISIS operation: Robotics repair work on the CHINON A3 natural uranium, carbon dioxide cooled, graphite moderated reactor

    International Nuclear Information System (INIS)

    Hilmoine, R.M.E.

    1989-01-01

    After describing the upper internal support structures of the CHINON A3 reactor, the problems resulting from their degradation due to corrosion and to the difficulties of the ISIS operation are presented here. The repair method is as follows: all tools and repair parts reach the working area by the feeding-pipes drilled through the 7 m thick concrete vessel surrounding the reactor core; the robots handle into the reactor, the tool heads and the repair parts which are automatically positioned and welded around the corroded structure, thus restoring the support of measurement devices. The parts are either linked together or to the existing structure by means of 2 studs of 12 mm in diameter. The different phases to sort out a problem are: in-core topography, reconforming of the full-scale mock-up with the repair area, learning on this mock-up and in-core repair. The technical specificities of the robots used are the following: they have an 11 meter long, 0.22 meter across telescopic mast with jointed arms reaching a radius of 2.7 m. Then the useful load is 70 daN and the repeatability 0.1 mm. Different tool heads can be handled by the robot: telemeter and laser reconstruction: it allows to locate the in core points and to materialize them on the mock-up by a laser crossed-beams locating technique; scouring: it cleans the corroded parts of the structures before welding; welding: it allows the parts handling and the carried studs welding; screwing; tensile test: carried out when the stud welds are defective. A high level computerized control system is organized around a central unit which calculates the displacements of robots and synchronises the actions of different tools by communicating with several local units. A 100,000 hour designing, a 200,000 hour building and assembling and a 450,000 hour operating on working area were necessary to repair 15 out of the 102 corroded structures by fitting and welding 205 repair parts. 10 figs

  5. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.; Lineberry, M.J.

    1990-01-01

    Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs

  6. Analyzing the loss of coolant accident in PWR nuclear reactors with elevation change in cold leg by RELAP5/MOD3.2 system code

    International Nuclear Information System (INIS)

    Kheshtpaz, H.; Alison, C.

    2006-01-01

    As, the Russian designed VVER-1000 reactor of the Bushehr Nuclear Power Plant by taking into account the change from German technology to that of Russian technology, and with the design of elevation change in the cold legs has been developed; therefore safety assessment of these systems for loss of coolant accident in elevation change in the cold legs and comparison results for non change elevation in the cold legs for a typical reactor (normal design of nuclear reactors) is the main important factor to be considered for the safe operation. In this article, the main objective is the simulation of the loss of coolant accident scenario by the RELAP5/MOD3.2 code in two different cases; first, the elevation change in the cold legs, and the second, non change in it. After comparing and analyzing these two code calculations the results have been generalized for a new design feature of Bushehr reactor. The design and simulation of the elevation change in the cold legs process with RELAP5/MOD3.2 code for PWR reactor is performed for the first time in the country, where it is introducing several important results in this respect

  7. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Martens, Frederick H. [Argonne National Laboratory; Jacobson, Norman H.

    1968-09-01

    This booklet discusses research reactors - reactors designed to provide a source of neutrons and/or gamma radiation for research, or to aid in the investigation of the effects of radiation on any type of material.

  8. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

    Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

  9. Development of a general coupling interface for the fuel performance code TRANSURANUS – Tested with the reactor dynamics code DYN3D

    International Nuclear Information System (INIS)

    Holt, L.; Rohde, U.; Seidl, M.; Schubert, A.; Van Uffelen, P.; Macián-Juan, R.

    2015-01-01

    Highlights: • A general coupling interface was developed for couplings of the TRANSURANUS code. • With this new tool simplified fuel behavior models in codes can be replaced. • Applicable e.g. for several reactor types and from normal operation up to DBA. • The general coupling interface was applied to the reactor dynamics code DYN3D. • The new coupled code system DYN3D–TRANSURANUS was successfully tested for RIA. - Abstract: A general interface is presented for coupling the TRANSURANUS fuel performance code with thermal hydraulics system, sub-channel thermal hydraulics, computational fluid dynamics (CFD) or reactor dynamics codes. As first application the reactor dynamics code DYN3D was coupled at assembly level in order to describe the fuel behavior in more detail. In the coupling, DYN3D provides process time, time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in case of the two-way coupling approach transfers parameters like fuel temperature and cladding temperature back to DYN3D. Results of the coupled code system are presented for the reactivity transient scenario, initiated by control rod ejection. More precisely, the two-way coupling approach systematically calculates higher maximum values for the node fuel enthalpy. These differences can be explained thanks to the greater detail in fuel behavior modeling. The numerical performance for DYN3D–TRANSURANUS was proved to be fast and stable. The coupled code system can therefore improve the assessment of safety criteria, at a reasonable computational cost

  10. Reactor Neutrinos

    OpenAIRE

    Kim, Soo-Bong; Lasserre, Thierry; Wang, Yifang

    2013-01-01

    We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very ...

  11. BOILING REACTORS

    Science.gov (United States)

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  12. 3. Research Coordination Meeting (RCM) of the Coordinated Research Project (CRP) on 'Studies of advanced reactor technology options for effective incineration of radioactive waste'. Working material

    International Nuclear Information System (INIS)

    2007-01-01

    To meet expressed Member States' needs, the IAEA has initiated a Coordinated Research Project (CRP) on 'Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste'. The final goal of the CRP is to deepen the understanding of the dynamics of transmutation systems, e.g. the accelerator driven system, especially systems with deteriorated safety parameters, qualify the available methods, specify the range of validity of methods, and formulate requirements for future theoretical developments. Should transient experiments be available, the CRP will pursue experimental benchmarking work. In any case, based on the results, the CRP will conclude on the potential need of transient experiments and make appropriate proposals for experimental programs. The Technical Meeting in Chennai was the 3rd Research Coordination Meeting (RCM) of the CRP The man objectives of the RCM were to: - Discuss and perform inter-comparisons of the various benchmark results; - Prepare the first draft of the final CRP Report Status of the analyses and inter-comparisons of the results. The main objective of the CRP was to study innovative technology options for incinerating/utilizing radioactive wastes. The CRP's benchmarking exercises focused on eight innovative transmutation 'Domains', which correspond to different critical and sub-critical concepts or groups of concepts: I. Critical fast reactor, solid fuel, with fertile; II. Critical fast reactor, solid fuel, fertile-free; III. ADS, solid fuel, with fertile; IV. ADS, solid fuel, fertile-free; V. Critical reactor and ADS, molten salt fuel, with fertile; VI. Critical reactor and ADS, molten salt fuel, fertile-free; VII. Critical fast reactor and ADS, gas cooled; VIII. Fusion/fission hybrid system. For each of these Domains, the discussions and inter-comparisons considered the following issues: - Reactor-models; - Scenarios/phenomena; - Static analyses; - Dynamic analyses; - Methods; - Codes; - Neutronic data base

  13. Integrating 3D CAD data for manufacturing and fabrication the core model of reactor TRIGA PUSPATI

    International Nuclear Information System (INIS)

    Abu Bakar Harun

    2005-01-01

    This paper describe the intrigue integration of digital 3 Dimensional Computer Aided Design (3D CAD) data manipulation for the Core Model fabrication of REAKTOR TRIGA PUSPATI and ready for mass manufacturing. 3 Dimensional CAD data from Computer Aided Design program will be used as an interpreter in the fabrication of this project. The Core Model of REAKTOR TRIGA PUSPATI will be fabricated with the aid of 3D CAD drawings and digital files. The components will be segregated and divided into 2 categories namely Conventional d Rapid Fabrication. (Author)

  14. Translation and adaption of the interRAI suite to local requirements in Belgian hospitals

    Directory of Open Access Journals (Sweden)

    Wellens Nathalie IH

    2012-09-01

    Full Text Available Abstract Background The interRAI Suite contains comprehensive geriatric assessment tools designed for various healthcare settings. Although each instrument is developed for a particular population, together they form an integrated health evaluation system. The interRAI Acute Care Minimum Data Set (interRAI AC is tailored for hospitalized older persons. Our aim in this study was to translate and adapt the interRAI AC to the Belgian hospital context, where it can be used together with the interRAI Home Care (HC and the interRAI Long Term Care Facility (LTCF. Methods A systematic, comprehensive, and rigorous 10-step approach was used to adapt the interRAI AC to local requirements. After linguistic translation by an official translator, five researchers assessed the translation for appropriate hospital jargon. Three researchers double-checked for translation accuracy and proposed additional items. A provisional version was converted into the three official languages of Belgium—Flemish, French, and German. Next, a multidisciplinary panel of nine experts judged item relevance to the Belgian care context and advised which country-specific items should be added. After these suggestions were incorporated into the interRAI AC, hospital staff from nine Flemish hospitals field-tested the tool in their practice. After evaluating field-test results, we compared the interRAI AC with Belgian versions of the interRAI HC and interRAI LTCF. Next, the Flemish, French, and German versions of the Belgian interRAI portfolio were harmonized. Finally, we submitted the Belgian interRAI AC to the interRAI organization for ratification. Results Eighteen administrative items of the interRAI AC were adapted to the Belgian healthcare context (e.g., usual residence, formal community services prior to admission. Fourteen items assessing the ‘informal caregiver’, and 17 items, including country-specific items, were added (e.g., advanced directive for euthanasia

  15. Translation and adaption of the interRAI Suite to local requirements in Belgian hospitals.

    Science.gov (United States)

    Wellens, Nathalie I H; Flamaing, Johan; Moons, Philip; Deschodt, Mieke; Boonen, Steven; Milisen, Koen

    2012-09-07

    The interRAI Suite contains comprehensive geriatric assessment tools designed for various healthcare settings. Although each instrument is developed for a particular population, together they form an integrated health evaluation system. The interRAI Acute Care Minimum Data Set (interRAI AC) is tailored for hospitalized older persons. Our aim in this study was to translate and adapt the interRAI AC to the Belgian hospital context, where it can be used together with the interRAI Home Care (HC) and the interRAI Long Term Care Facility (LTCF). A systematic, comprehensive, and rigorous 10-step approach was used to adapt the interRAI AC to local requirements. After linguistic translation by an official translator, five researchers assessed the translation for appropriate hospital jargon. Three researchers double-checked for translation accuracy and proposed additional items. A provisional version was converted into the three official languages of Belgium-Flemish, French, and German. Next, a multidisciplinary panel of nine experts judged item relevance to the Belgian care context and advised which country-specific items should be added. After these suggestions were incorporated into the interRAI AC, hospital staff from nine Flemish hospitals field-tested the tool in their practice. After evaluating field-test results, we compared the interRAI AC with Belgian versions of the interRAI HC and interRAI LTCF. Next, the Flemish, French, and German versions of the Belgian interRAI portfolio were harmonized. Finally, we submitted the Belgian interRAI AC to the interRAI organization for ratification. Eighteen administrative items of the interRAI AC were adapted to the Belgian healthcare context (e.g., usual residence, formal community services prior to admission). Fourteen items assessing the 'informal caregiver', and 17 items, including country-specific items, were added (e.g., advanced directive for euthanasia). The interRAI AC was adapted to local requirements using a

  16. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  17. Status of French reactors

    International Nuclear Information System (INIS)

    Ballagny, A.

    1997-01-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm 3 . The OSIRIS reactor has already been converted to LEU. It will use U 3 Si 2 as soon as its present stock of UO 2 fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU

  18. Effect of supplementation on the performance of grazing Belgian Blue double-muscled heifers.

    Science.gov (United States)

    Fiems, L O; De Boever, J L; Vanacker, J M

    2013-11-01

    Six experiments were conducted to investigate the effect of a feed supplement on the performance of grazing Belgian Blue double-muscled (BBDM) heifers with an initial weight and age of 195 ± 43 kg and 190 ± 52 days. Treatments included were: Exp. 1: supplementation with beet pulp (BP): 2 kg/day per head v. ad libitum intake; Exp. 2: supplementation ad libitum with BP v. a mixture of BP and soybean meal (SBM; BP/SBM ratio of 80/20; FW (fresh weight) basis); Exp. 3: supplementation with 4 kg/day per head of a mixture of BP/SBM (80/20; FW basis) v. BP/formaldehyde-treated SBM (BP/FSBM); Exp. 4: supplementation with 4 kg/day per head of a mixture with a similar protein content (125 g DVE per kg dry matter (DM)), consisting of 80/20 BP/SBM v. 92/8 BP/FSBM; Exp. 5: supplementation with 3 kg/day per head of a mixture of BP/SBM (80/20; FW basis) v. BP/DDGS (dried distillers grains and solubles; 70/30, FW basis); and Exp. 6: supplementation with 3 kg/day per head of 80/20 BP/SBM v. maize silage (MS) and SBM, on the basis of a similar protein concentration in the DM as the 80/20 BP/SBM supplement, and fed at a similar amount of DM as in the BP/SBM group. Supplementing BP ad libitum did not affect daily gain (0.54 v. 0.48 kg) and partial feed conversion (3.62 kg on average) compared with 2 kg/day. Supplying SBM besides BP increased growth rate compared with BP (0.87 v. 0.62 kg/day; P 0.10), but blood urea levels were reduced by FSBM (P daily gain, but partial feed conversion tended to be higher (3.21 v. 3.60 kg/kg body weight (BW) gain; P = 0.062). Increasing the supplement (80/20 BP/SBM) level from 3 to 4 kg daily, corresponding to 1.02% and 1.18% of the mean BW, respectively, resulted in a tendency (P = 0.121) for an increased growth rate. Grazing BBDM heifers of <1 year of age necessitate extra protein besides an energy supplement to improve their performance. DDGS can replace SBM and BP can be replaced by MS.

  19. Estimation of inbreeding rates and extinction risk of forty one Belgian chicken breeds in 2005 and 2010

    OpenAIRE

    Moula, Nassim; Philippe, François-Xavier; Antoine-Moussiaux, Nicolas; Leroy, Pascal; Michaux, Charles

    2014-01-01

    In Belgium, as generally in Europe, the dominant position of the high producing commercial strains specialized in meat or eggs production threats of extinction the local traditional breeds. In this work, a follow up of the changes in populations size, and the rates of inbreeding of the Belgian poultry breeds, has been carried out in 2005 and 2010. About forty breeds were concerned. The Belgian hen breeds being overwhelmingly under threat of extinction, because of the low number of individu...

  20. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  1. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

    This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

  2. Reactor building 3D-model for evaluating the pressures on concrete regularization and foundation waterproofing membrane

    International Nuclear Information System (INIS)

    Mello Junior, Glauco J.T.; Cardoso, Tarcisio de F.; Prates, Carlos L.M.

    2009-01-01

    Angra dos Reis site in Brazil has already 2 operating PWR NPPs. Unit 3, with identical design to Unit 2, also a 1350 MW PWR, is expected to have its construction started in 2009. This new plant shall be founded directly on sound rock. The first step is to prepare this rock surface with a concrete regularization and a foundation waterproofing membrane. This study presents a 3D model approach of the corresponding reactor building to verify the maximum pressure acting on this surface. The 3D model permits to show a more realistic pressure distribution at every foundation specific detail. A static analysis is performed using ANSYS Mechanical Release 11.1. Dead weight, permanent and live loads, Safe Shutdown Earthquake (SSE) combined with Burst Pressure Wave (BPW) from the Feedwater Tank (SSB=SSE+BPW) and differences of temperature are taken into account. Considering all foundation nodes , the pressure distribution on the waterproofing membrane for each load case is obtained for vertical and horizontal directions, which corresponds to compression and tangential reaction loads. The maximum values occur in distinct positions for each load case. The maximum results are obtained according to DIN 25449 (2008) load combination criteria. The results are compared to a simplified analysis performed before, showing a good agreement in global values. (author)

  3. Heat Transfer Reactor Experiment (HTRE)-3 Container Storage Unit Resource Conservation Recovery Act closure plan

    International Nuclear Information System (INIS)

    Spry, M.J.

    1992-11-01

    This document describes the closure of the HTRE-3 Container Storage Unit under the requirements of the Resource Conservation and Recovery Act. The unit's location, size, history, and current status are described. The document also summarizes the decontamination and decommissioning efforts performed in 1983 and provides an estimate of,waste residues remaining in the HTRE-3 assembly. A risk evaluation was performed that demonstrates that the residue does not pose a hazard to public health or the environment. Based on the risk evaluation, it is proposed that the HTRE-3 Container Storage Unit be closed in its present condition, without further decontamination or removal activities

  4. Inertial confinement fusion reactor systems

    International Nuclear Information System (INIS)

    Frank, T.G.; Bohachevsky, I.O.; Pendergrass, J.H.

    1980-01-01

    A variety of reactor cavity concepts, drivers, and energy conversion mechanisms are being considered to realize commercial applications of ICF. Presented in this paper are: (1) a review of reactor concepts with estimates of practically achievable pulse repetition rates; (2) a survey of drivers with estimates of the requirements on reactor conditions imposed by beam propagation characteristics; and (3) an assessment of compatible driver-reactor combinations

  5. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Nagatomi, Shozo.

    1976-01-01

    Object: To provide a jet and missile protective wall of a configuration being inflated toward the center of a reactor container on the inside of a body of the reactor container disposed within a biological shield wall to thereby increase safety of the reactor container. Structure: A jet and missile protective wall comprised of curved surfaces internally formed with a plurality of arch inflations filled with concrete between inner and outer iron plates and shape steel beam is provided between a reactor container surrounded by a biological shield wall and a thermal shield wall surrounding the reactor pressure vessel, and an adiabatic heat insulating material is filled in space therebetween. (Yoshino, Y.)

  6. RA-3 reactor core with uranium silicide fuel elements P-07 type

    International Nuclear Information System (INIS)

    Abbate, Maximo J.; Sbaffoni, Maria M.

    2003-01-01

    Following the studies on the utilization of fuel elements (FE) containing uranium silicide, core of the RA-3 was analyzed with several calculation models. At first, the present situation, i.e. the core charged with normal FE (U 3 O 8 ), has been analyzed to validate the simulation methodology comparing with experimental results and to establish reference data to 5 and 10 MW able to be compared with future new situations. Also, CITVAP's nuclear data libraries to be used in irradiation experiment planning were completed. The results were satisfactory and were applied to the study of the core containing P-07 FE [U 3 Si 2 ], in face of a future core change. Comparing with the performance of the U 3 O 8 FE, the silicides ones show the following advantages: - average burnup: 45 % greater; -extraction burnup increase 12 %; and, -the residence time [in full power days] could be a 117 % greater. (author)

  7. Advanced converter reactors

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1979-01-01

    Advanced converter reactors (ACRs) of primary US interest are those which can be commercialized within about 20 years, and are: Advanced Light-Water Reactors, Spectral-Shift-Control Reactors, Heavy-Water Reactors (CANDU type), and High-Temperature Gas-Cooled Reactors. These reactors can operate on uranium, thorium, or uranium-thorium fuel cycles, but have the greatest fuel utilization on thorium type cycles. The water reactors tend to operate more economically on uranium cycles, while the HTGR is more economical on thorium cycles. Thus, the HTGR had the greatest practical potential for improving fuel utilization. If the US has 3.4 to 4 million tons U 3 O 8 at reasonable costs, ACRs can make important contributions to maintaining a high nuclear power level for many decades; further, they work well with fast breeder reactors in the long term under symbiotic fueling conditions. Primary nuclear data needs of ACRs are integral measurements of reactivity coefficients and resonance absorption integrals

  8. Development and Applications of a General Coupled Thermal-hydraulic/Neutronic Model for the Ringhals-3 Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Staalek, Mathias

    2008-03-01

    Coupled calculations are important for the simulation of nuclear power plants when there is a strong feedback between the neutron kinetics and the thermal-hydraulics. A general coupled model of the Ringhals-3 Pressurized Water Reactor has been developed for this purpose. The development is outlined in the thesis with details given in the appended papers. A PARCS model was developed for the core calculations and a RELAP5 model for the thermal-hydraulic calculations. The RELAP5 model has 157 channels for modelling the flow in the fuel assemblies. This means that there is a one-one correspondence radially between the neutronic and thermal-hydraulic nodalization. This detailed mapping between the neutron kinetics and the thermal-hydraulics makes it possible to use the model for all kinds of transient. To provide realistic material data to the PARCS model, a cross-section interface was developed. With this interface one can import material data from a binary CASMO-4 library file into PARCS. Due to the one-to-one mapping, any any core loading can easily be considered. The PARCS model was benchmarked against measurements of the steady-state power distribution of Ringhals-3. The power shape was well reproduced by the model. Validational work for steady-state conditions of the thermal-hydraulic was also successfully performed. The most challenging part of the validation of a coupled model is for transients. This is much more difficult since the dynamics of the system becomes very important. Two transients that occurred at Ringhals-3 were chosen for the validational work. The first transient was a Load Rejection Transient. In general the model gave good results but some problems were experienced, e.g. the pressurizer pressure turned out to be more difficult to be correctly simulated. The second transient was a Loss of Feed Water transient. A malfunctioning feed water control valve closed, and therefore shut down the feed water supply to the steam generator in one of the

  9. Operando UV-Vis spectroscopy of a catalytic solid in a pilos-scale reactor: deactivation of a CrOx/Al2O3 propane dehydrogenation catalyst

    NARCIS (Netherlands)

    Sattler, J.J.H.B.; Gonzalez-Jimenez, I.D.; Mens, A.J.M.; Arias, M.J.; Visser, T.; Weckhuysen, B.M.

    2013-01-01

    A novel operando UV-Vis spectroscopic set-up has been constructed and tested for the investigation of catalyst bodies loaded in a pilot-scale reactor under relevant reaction conditions. Spatiotemporal insight into the formation and burning of coke deposits on an industrial CrOx/Al2O3 catalyst during

  10. Real-Time Quantitative Operando Raman Spectroscopy of a CrOx/Al2O3 Propane Dehydrogenation Catalyst in a Pilot-Scale Reactor

    NARCIS (Netherlands)

    Sattler, Jesper J. H. B.|info:eu-repo/dai/nl/328235601; Mens, Ad M.; Weckhuysen, Bert M.|info:eu-repo/dai/nl/285484397

    2014-01-01

    Combined operando UV/vis-Raman spectroscopy has been used to study the deactivation of CrOx/Al2O3 catalyst extrudates in a pilot scale propane dehydrogenation reactor. For this purpose, UV/vis and Raman optical fiber probes have been designed, constructed and tested. The light absorption measured by

  11. Design of the fuel element 'snow-flake' in uranium oxide, canned with aluminium, for the experimental reactor EL 3 (1960)

    International Nuclear Information System (INIS)

    Gauthron, M.; Guibert, B.

    1960-01-01

    This report sums up the main studies have been carried out on the fuel element 'Snowflake' (uranium oxide, canned with aluminium), designed to replace the present element of the experimental reactor EL3 in order to increase the reactivity without modifying the neutron flux/thermal power ratio. (author) [fr

  12. Thermohydraulic Analysis of Loss of Flow Accident of Multipurpose ReactorG.A. Siwabessy with U3Si2-Al Fuel

    International Nuclear Information System (INIS)

    Azizul-Khakim

    2000-01-01

    The use of U 3 Si 2 -Al fuel to replace U 3 O 8 -Al fuel in MultipurposeReactor G.A Siwabessy will give better performance and safety margin. Thecomparative strong points of U 3 Si 2 -Al to U 3 O 8 -Al such as, its higherdensity and better thermal conductivity underlie the consideration forreplacing U 3 O 8 -Al with U 3 Si 2 -Al. Two cooling modes used in MPR G.ASiwabessy are forced convection cooling mode during high power operation andnatural circulation during low power and shut down condition. CONVEC codecalculated natural circulation with the safety margin of Onset of NucleateBoiling (ONB) giving result of 1.104 MW (t), namely the maximum power thatshould not be exceeded during natural circulation cooling mode. Let thesafety factor be 20 %, consequently the reactor can operate safely at 0.883MW (t) or less. The analyses of reactor transient due to primary pumpsfailure was carried out by PARET/ANL code, by assuming that the pumps losttheir electric power leading to the decrease of the coolant flow andsubsequently stopped and reverse flow occurred. The Reactor Protection System(RPS) responded to scram when the coolant flow reached 85 % of the nominalflow. The pumps moment inertia made the decrease of coolant flow rate runslowly and gave time interval of 76.78 s for the reactor core to lower itspower into the allowable level to undergo natural circulation cooling mode.After the coolant flow was stagnant, the coolant, clad and fuel temperaturewent up abruptly to 102.25 o C, 110.16 o C and 110.26 o C respectively. Theassociated pressure was 199.72 Kpa with the saturated temperature of 121 o C,meaning that boiling did not occur and the core was still safe. (author)

  13. Seismic qualification of the primary loop of the non-seismically designed Belgian NPP

    International Nuclear Information System (INIS)

    Detroux, P.; Van Vyve, J.

    1989-01-01

    This paper describes the non linear time history analysis performed to seismically qualify the primary loop of the three first Belgian PWR plants. These plants (Chooz A 300 MWe - commissioned in 1966, and Doel 1/2 2 x 390 MWe - commissioned in 1974 and 1975) were not seismically designed, no seismic event being specified at that time. But, after respectively 20 years and 10 years of operation, the safety authorities required to seismically qualify the primary loop of these plants, for a SSE event. Both primary loops and, particularly, the Chooz A primary loop, have a supporting system that differs from the other seismically qualified Belgian NPP. A seismic qualification by analogy was thus not possible, but they are equipped with restraints designed for LOCA/SLB that will certainly be active in case of an SSE

  14. The influence of a smoking ban on the profitability of Belgian restaurants.

    Science.gov (United States)

    De Schoenmaker, Sofie; Van Cauwenberge, Philippe; Vander Bauwhede, Heidi

    2013-05-01

    To examine whether the nationwide smoking ban, imposed in 2007, had an impact on the profitability of Belgian restaurants. Objective financial reporting data on 1613 restaurants were analysed with return on assets as the outcome measure. The data were collected from the Belfirst database and cover the period 2004-2009. To assess the impact of the smoking ban, a differences-in-differences estimation method was used, with bars serving as the control group. The regression model was estimated, while controlling for firm-specific characteristics and unobserved firm-level heterogeneity. The variable of interest is the interaction between the smoking ban dummy and the dummy for the treatment group. The coefficient of this variable is insignificant. The adoption of the nationwide smoking ban did not affect the profitability of Belgian restaurants.

  15. Application of American and French rules for the next belgian PWR

    International Nuclear Information System (INIS)

    Roch, M.; Cavaco, A.

    1987-01-01

    The licensing practice in Belgium is evolving from the precedent compliance with the USNRC rules (as applied to the 4 last Belgian PWRs) to a more sophisticated approach applied to the next Belgian PWR (N8), which incorporates a mixed compliance with the USNRC or with French rules, depending on the equipment, the structure or the system considered. In this paper, we present the approach concerning the licensing rules applicable to N8. The following aspects are covered: rules applicable to the NSSS; rules applicable to the BOP (codes of design for systems and structures); rules applicable to the equipment (code of construction for mechanical and electrical components); impact on the lay-out of the plant. Some examples of application of this methodology are given. (author)

  16. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  17. Wigner effect in graphite stack: G2 and G3 reactors

    International Nuclear Information System (INIS)

    Artozoul, M.; L'Homme, M.

    1982-11-01

    This text describes work carried out between 1978 and 1980 by a COGEMA/CEA team responsible for a report on the feasibility, effectiveness and possible hazards likely to be encountered in the nuclear annealing of G2 and in changing the operating conditions of G3 [fr

  18. The Electrodeless Discharge Lamp: A Prospective Tool for Photochemistry Part 3. The Microwave Photochemistry Reactor

    Czech Academy of Sciences Publication Activity Database

    Klán, P.; Hájek, Milan; Církva, Vladimír

    2001-01-01

    Roč. 140, č. 3 (2001), s. 185-189 ISSN 1010-6030 Institutional research plan: CEZ:AV0Z4072921 Keywords : photochemistry * electrodeless discharge lamp * microwave Subject RIV: CH - Nuclear ; Quantum Chemistry Impact factor: 1.038, year: 2001

  19. Hydrogenation of furfural at the dynamic Cu surface of CuOCeO2/Al2O3 in vapor phase packed bed reactor

    Science.gov (United States)

    The hydrogenation of furfural to furfuryl alcohol over a CuOCeO2/'-Al2O3 catalyst in a flow reactor is reported. The catalyst was prepared by the wet impregnation of Cu onto a CeO2/'-Al2O3 precursor. The calcined catalyst was then treated with HNO3 to remove surface CuO resulting in a mixed CuCe oxi...

  20. Characterization of a dopamine transporter polymorphism and behavior in Belgian Malinois

    OpenAIRE

    Lit, Lisa; Belanger, Janelle M; Boehm, Debby; Lybarger, Nathan; Haverbeke, Anouck; Diederich, Claire; Oberbauer, Anita M

    2013-01-01

    Abstract Background The Belgian Malinois dog breed (MAL) is frequently used in law enforcement and military environments. Owners have reported seizures and unpredictable behavioral changes including dogs’ eyes “glazing over,” dogs’ lack of response to environmental stimuli, and loss of behavioral inhibition including owner-directed biting behavior. Dogs with severe behavioral changes may be euthanized as they can represent a da...

  1. How does the average Belgian view nuclear technology? Barometer assesses knowledge and risk perception

    International Nuclear Information System (INIS)

    Turcanu, C.; Perko, T.

    2011-01-01

    Irrespective of the choices made in the future, questions about nuclear safety, radiation protection and waste management will always be of topical interest. SCK-CEN has consciously opted for social research on risk perception, sustainable development, and communication. SCK-CEN founded the Barometer in 2002 in order to keep both feet firmly planted in society: a regular, large-scale survey amongst the Belgian population on radiation and nuclear technology. The findings of the third edition were published in 2010.

  2. Human African trypanosomiasis in a Belgian traveller returning from the Masai Mara area, Kenya, February 2012.

    Science.gov (United States)

    Clerinx, J; Vlieghe, E; Asselman, V; Van de Casteele, S; Maes, M B; Lejon, V

    2012-03-08

    A Belgian traveller was diagnosed with human African trypanosomiasis (HAT) due to Trypanosoma brucei rhodesiense nine days after visiting the Masai Mara area in Kenya. He presented with an inoculation chancre and was treated with suramin within four days of fever onset. Two weeks earlier, HAT was also reported in a German traveller who had visited the Masai Mara area. Because no cases have occurred in the area for over 12 years, this may indicate a focal cluster of HAT.

  3. Study of heat transfer in 3D fuel rods of the EPRI-9R reactor modified; Estudo da transferencia de calor em varetas combustiveis 3D do reator EPRI-9R 3D modificado

    Energy Technology Data Exchange (ETDEWEB)

    Affonso, Renato Raoni Werneck; Lava, Deise Diana; Borges, Diogo da Silva; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes, E-mail: raoniwa@yahoo.com.br, E-mail: deisedy@gmail.com, E-mail: diogosb@outlook.com, E-mail: sampaio@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    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.

  4. Identification of a novel idiopathic epilepsy locus in Belgian Shepherd dogs.

    Directory of Open Access Journals (Sweden)

    Eija H Seppälä

    Full Text Available Epilepsy is the most common neurological disorder in dogs, with an incidence ranging from 0.5% to up to 20% in particular breeds. Canine epilepsy can be etiologically defined as idiopathic or symptomatic. Epileptic seizures may be classified as focal with or without secondary generalization, or as primary generalized. Nine genes have been identified for symptomatic (storage diseases and one for idiopathic epilepsy in different breeds. However, the genetic background of common canine epilepsies remains unknown. We have studied the clinical and genetic background of epilepsy in Belgian Shepherds. We collected 159 cases and 148 controls and confirmed the presence of epilepsy through epilepsy questionnaires and clinical examinations. The MRI was normal while interictal EEG revealed abnormalities and variable foci in the clinically examined affected dogs. A genome-wide association study using Affymetrix 50K SNP arrays in 40 cases and 44 controls mapped the epilepsy locus on CFA37, which was replicated in an independent cohort (81 cases and 88 controls; combined p = 9.70×10⁻¹⁰, OR = 3.3. Fine mapping study defined a ∼1 Mb region including 12 genes of which none are known epilepsy genes or encode ion channels. Exonic sequencing was performed for two candidate genes, KLF7 and ADAM23. No variation was found in KLF7 but a highly-associated non-synonymous variant, G1203A (R387H was present in the ADAM23 gene (p = 3.7×10⁻⁸, OR = 3.9 for homozygosity. Homozygosity for a two-SNP haplotype within the ADAM23 gene conferred the highest risk for epilepsy (p = 6.28×10⁻¹¹, OR = 7.4. ADAM23 interacts with known epilepsy proteins LGI1 and LGI2. However, our data suggests that the ADAM23 variant is a polymorphism and we have initiated a targeted re-sequencing study across the locus to identify the causative mutation. It would establish the affected breed as a novel therapeutic model, help to develop a DNA test for breeding

  5. GAPCON-THERMAL-3 verification and comparison to in-reactor data.

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Panisko, F.E.; Mohr, C.L.

    1978-01-01

    The GAPCON-THERMAL-3 computer code is an outgrowth of the GAPCON series, written in response to a need to predict the interacting path-dependent thermal and mechanical behavior of oxide fuel rods. Previous GAPCON versions did not include a comprehensive path-dependent prediction of cladding stress and strain that was connected to the thermal history. GAPCON-THERMAL-3 includes an incremental, finite element solution for the cladding that includes elastic, plastic, and creep contributions. The stress-strain increments for each time step reflect the thermal-mechanical state of the fuel at that time, so that the evolution of fuel and cladding displacements and temperatures through time is truly interconnected and path dependent.

  6. Health physics during work on the G. 2 and G. 3 reactor exchanges

    International Nuclear Information System (INIS)

    Rodier, J.; Chassany, J.; Guillermin, P.

    1965-01-01

    During this work and its preparation, which took place first at G. 2 and then at G. 3 over a period of 11 months, 15000 measurement results were obtained. Their analysis, together with a consideration of the organisation on the site and of the conclusions drawn from the experiment, shows the various factors which determine the importance of the radio-active dangers. (authors) [fr

  7. Quality control of the structural design of Angra 2 and 3 power reactor

    International Nuclear Information System (INIS)

    Merino, J.L.E.; Carvalho, S.R.B. de

    1985-01-01

    The quality control requirements for nuclear power plants are done by CNEN and might be followed by the organizations that are responsible for those activities, such as: design, suppliers, construction, fabrication, mounting, installation, commissioning, operation, maintenance, modification and decommissioning. The Promon's experience for using a quality control program in the structural design of Angra 2 and 3, and the verification system adopted, are presented. (E.G.) [pt

  8. Synthesis and characterization of Nd2O3 nanoparticles in a radiofrequency thermal plasma reactor

    Science.gov (United States)

    Dhamale, G. D.; Mathe, V. L.; Bhoraskar, S. V.; Sahasrabudhe, S. N.; Dhole, S. D.; Ghorui, S.

    2016-02-01

    The synthesis of nanocrystalline Nd2O3 through an inductively coupled radiofrequency thermal plasma route is reported. Unlike in conventional synthesis processes, plasma-synthesized nanoparticles are directly obtained in a stable hexagonal crystal structure with a faceted morphology. The synthesized nanoparticles are highly uniform with an average size around 20 nm. The nanoparticles are characterized in terms of phase formation, crystallinity, morphology, size distribution, nature of chemical bonds and post-synthesis environmental effects using standard characterization techniques. X-ray diffraction, transmission electron microscopy, and scanning electron microscopy are used for structural and morphological studies. The thermo-gravimetric technique, using a differential scanning calorimeter, is used to investigate the purity of phase. Fourier transform infrared spectroscopy is used to investigate the nature of existing bonds. The optical response of the nanoparticles is investigated through the electronic transition of Nd3+ ions in its crystalline structure via UV-visible spectroscopy. The presence of defect states and corresponding activation energies in the nanocrystalline Nd2O3 compared to those of the precursors are studied using thermoluminescence.

  9. Flaw evaluation methodology for class 2, 3 components in light water reactors

    International Nuclear Information System (INIS)

    Miura, Naoki; Kashima, Koichi; Miyazaki, Katsumasa; Hasegawa, Kunio; Oritani, Naohiko

    2006-01-01

    It is quite important to validate the structural integrity of operating plant components as aged LWR plants are gradually increasing in Japan. The rules on fitness-for-service for nuclear power plants constituted by the JSME provides flaw evaluation methodology. They are mainly focused on Class 1 components, while flaw evaluation criteria for Class 2, 3 components are not consolidated. As such, they also required from the viewpoints of in-service inspection request, reduction of operating cost and systematization of consistent code/standard. In this study, basic concept of flaw evaluation for Class 2, 3 piping was considered, and it is concluded that the same evaluation procedure as Class 1 piping in the current rules is applicable. Some technical issues on practical flaw evaluation for Class 2, 3 piping were listed up, and a countermeasure for each issue was devised. Especially, both allowable flaw sizes in acceptance standards and critical flaw sizes in acceptance criteria have to be determined in consideration of degraded fracture toughness. (author)

  10. Determination of contamination pathways of phthalates in food products sold on the Belgian market.

    Science.gov (United States)

    Van Holderbeke, Mirja; Geerts, Lieve; Vanermen, Guido; Servaes, Kelly; Sioen, Isabelle; De Henauw, Stefaan; Fierens, Tine

    2014-10-01

    As numerous studies have indicated that food ingestion is the most important exposure pathway to several phthalates, this study aimed to determine possible contamination pathways of phthalates in food products sold on the Belgian market. To do this, concentrations of eight phthalates (dimethyl phthalate (DMP), diethyl phthalate (DEP), diisobutyl phthalate (DiBP), di-n-butyl phthalate (DnBP), benzylbutyl phthalate (BBP), dicyclohexyl phthalate (DCHP), di(2-ethylhexyl) phthalate (DEHP) and di-n-octyl phthalate (DnOP)) were determined in 591 foods and 30 packaging materials. In general, the four most prominent phthalates in Belgian food products were DEHP, DiBP, DnBP and BBP. Special attention was given to the origin of these phthalates in bread, since high phthalate concentrations (especially DEHP) were determined in this frequently consumed food product. Phthalates seemed to occur in Belgian bread samples due to the use of contaminated ingredients (i.e. use of contaminated flour) as well as due to migration from phthalate containing contact materials used during production (e.g. coated baking trays). Also the results of the conducted concentration profiles of apple, bread, salami and two cheese types revealed the important role of processing - and not packaging - on phthalate contents in foods. Copyright © 2014 Elsevier Inc. All rights reserved.

  11. Last developments in the Belgian disposal programme for low and intermediate short-lived waste

    International Nuclear Information System (INIS)

    Boyazis, Jean-Paul

    2006-01-01

    After an historical reminder of the several phases of the Belgian program for the disposal of low and medium level short-lived waste since the creation of ONDRAF/NIRAS and the bad results obtained in the 90's by using a pure technical approach, the presentation will explain the main lines of the new methodology developed, as a consequence of the government decision of 16 January 1998 in ONDRAF/NIRAS to improve local acceptance for the disposal project. The way local partnerships were created with four nuclear municipalities under the form of a non-profit organization with a clear mission, the functioning, on a voluntary base, of the different partnerships during four to six years and the concrete results obtained until now using this very innovative method will be addressed. The last developments of the Belgian program for the disposal of low and medium level and short-lived waste will be presented, including the recent and very important decision of the Belgian government of 23 June 2006 to dispose of the low and medium active short-lived waste in a surface disposal installation on the territory of the municipality Dessel. (author)

  12. Microbial diversity and metabolite composition of Belgian red-brown acidic ales.

    Science.gov (United States)

    Snauwaert, Isabel; Roels, Sanne P; Van Nieuwerburg, Filip; Van Landschoot, Anita; De Vuyst, Luc; Vandamme, Peter

    2016-03-16

    Belgian red-brown acidic ales are sour and alcoholic fermented beers, which are produced by mixed-culture fermentation and blending. The brews are aged in oak barrels for about two years, after which mature beer is blended with young, non-aged beer to obtain the end-products. The present study evaluated the microbial community diversity of Belgian red-brown acidic ales at the end of the maturation phase of three subsequent brews of three different breweries. The microbial diversity was compared with the metabolite composition of the brews at the end of the maturation phase. Therefore, mature brew samples were subjected to 454 pyrosequencing of the 16S rRNA gene (bacteria) and the internal transcribed spacer region (yeasts) and a broad range of metabolites was quantified. The most important microbial species present in the Belgian red-brown acidic ales investigated were Pediococcus damnosus, Dekkera bruxellensis, and Acetobacter pasteurianus. In addition, this culture-independent analysis revealed operational taxonomic units that were assigned to an unclassified fungal community member, Candida, and Lactobacillus. The main metabolites present in the brew samples were L-lactic acid, D-lactic acid, and ethanol, whereas acetic acid was produced in lower quantities. The most prevailing aroma compounds were ethyl acetate, isoamyl acetate, ethyl hexanoate, and ethyl octanoate, which might be of impact on the aroma of the end-products. Copyright © 2016 Elsevier B.V. All rights reserved.

  13. Air crew exposure on board of long-haul flights of the Belgian airlines

    International Nuclear Information System (INIS)

    Verhaegen, F.; Poffijn, A.

    2000-01-01

    New European radiation protection recommendations state that measures need to be taken for flight crew members whose annual radiation exposure exceeds 1 mSv. This will be the case for flight crew members who accumulate most of their flying hours on long-haul flights. The Recommendations for the Implementation of the Basic Safety Standards Directive states that for annual exposure levels between 1 and 6 mSv individual dose estimates should be obtained, whereas for annual exposures exceeding 6 mSv, which might rarely occur, record keeping with appropriate medical surveillance is recommended. To establish the exposure level of Belgian air crews, radiation measurements were performed on board of a total of 44 long-haul flights of the Belgian airlines. The contribution of low linear energy transfer (LET) radiation (photons, electrons, protons) was assessed by using TLD-700H detectors. The exposure to high-LET radiation (mostly neutrons) was measured with bubble detectors. Results were compared to calculations with an adapted version of the computer code CARI. For the low-LET radiation the calculations were found to be in good agreement with the measurements. The measurements of the neutron dose were consistently lower than the calculations. With the current flight schedules used by the Belgian airlines, air crew members are unlikely to receive annual doses exceeding 4 mSv. (author)

  14. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, 3rd Quarterly Report

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth

    2002-06-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

  15. 3D Monte-Carlo transport calculations of whole slab reactor cores: validation of deterministic neutronic calculation routes

    International Nuclear Information System (INIS)

    Palau, J.M.

    2005-01-01

    This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U 235 , U 238 , Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)

  16. Basic numerical simulation of TWDEC for a D-{sup 3}He fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, M.; Kudo, T.; Hayashi, S.; Hayashi, S.; Yamane, T. [Kyoto Univ. (Japan). Dept. of Electrical Engineering; Tomita, Y.; Momota, H. [National Institute for Fusion Science, Furo-cho, Chikusa-ku, Nagoya 464-01 (Japan)

    1998-09-01

    The objective of the present paper is to study the behaviour of a travelling wave direct energy converter which is supposed to convert the energy of fusion proton ions produced in a FRC D-{sup 3}H{sub e} fusion plasma. Numerical experiments with a one and two-dimensional approximation give the following results: (1) an external circuit has been proposed for the travelling wave, and time-dependent, one-dimensional computations of the momentum conservation equation together with Poisson`s equation have been carried out, resulting in autonomous oscillation of the travelling wave. About 72.3% conversion efficiency has been obtained with a grid voltage of 1.0 MV. (2) With a two-dimensional approximation, the efficiency and grid voltage with newly proposed cylindrical grids result in 65.5% and 1.7 MV when the collision loss with the grids is neglected, while 54.7% and 1.4 MV are obtained when the collision loss is included. (orig.) 7 refs.

  17. A new Tone's method in APOLLO3® and its application to fast and thermal reactor calculations

    Directory of Open Access Journals (Sweden)

    Li Mao

    2017-09-01

    Full Text Available This paper presents a newly developed resonance self-shielding method based on Tone's method in APOLLO3® for fast and thermal reactor calculations. The new method is based on simplified models, the narrow resonance approximation for the slowing down source and Tone's approximation for group collision probability matrix. It utilizes mathematical probability tables as quadrature formulas in calculating effective cross-sections. Numerical results for the ZPPR drawer calculations in 1,968 groups show that, in the case of the double-column fuel drawer, Tone's method gives equivalent precision to the subgroup method while markedly reducing the total number of collision probability matrix calculations and hence the central processing unit time. In the case of a single-column fuel drawer with the presence of a uranium metal material, Tone's method obtains less precise results than those of the subgroup method due to less precise heterogeneous–homogeneous equivalence. The same options are also applied to PWR UOX, MOX, and Gd cells using the SHEM 361-group library, with the objective of analyzing whether this energy mesh might be suitable for the application of this methodology to thermal systems. The numerical results show that comparable precision is reached with both Tone's and the subgroup methods, with the satisfactory representation of intrapellet spatial effects.

  18. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  19. Characterization of mandibular molar root and canal morphology using cone beam computed tomography and its variability in Belgian and Chilean population samples

    Energy Technology Data Exchange (ETDEWEB)

    Torres, Andres; Jacobs, Reinhilde; Lambrechts, Paul [Katholieke Universiteit Leuven, Leuven (Belgium); Brizuela, Claudia; Cabrera, Carolina; Concha, Guillermo; Pedemonte, Maria Eugenia [Universidad de los Andes, Santiago (Chile)

    2015-06-15

    This study used cone-beam computed tomography (CBCT) to characterize mandibular molar root and canal morphology and its variability in Belgian and Chilean population samples. We analyzed the CBCT images of 515 mandibular molars (257 from Belgium and 258 from Chile). Molars meeting the inclusion criteria were analyzed to determine (1) the number of roots; (2) the root canal configuration; (3) the presence of a curved canal in the cross-sectional image of the distal root in the mandibular first molar and (4) the presence of a C-shaped canal in the second mandibular molar. A descriptive analysis was performed. The association between national origin and the presence of a curved or C-shaped canal was evaluated using the chi-squared test. The most common configurations in the mesial root of both molars were type V and type III. In the distal root, type I canal configuration was the most common. Curvature in the cross-sectional image was found in 25% of the distal canals of the mandibular first molars in the Belgian population, compared to 11% in the Chilean population. The prevalence of C-shaped canals was 10% or less in both populations. In cases of unclear or complex root and canal morphology in the mandibular molars, CBCT imaging might assist endodontic specialists in making an accurate diagnosis and in treatment planning.

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

    International Nuclear Information System (INIS)

    Kondo, Shunsuke

    2013-01-01

    • New nuclear energy strategy is still in the process of deliberation in Japan. Though many of idled plants will be restarted after renovation of their safety features in accordance with new safety rules set by the NRA, the contribution of nuclear power in Japan will probably not return to the level before 3.11. • Japan is in the process of reviewing its strategy for SFR R&D with a view to making it compatible with the new situation to be realized under new safety regulation as well as new energy strategy to be formulated within a year. • It is contemplated that major emphasis of the SFR R&D should be on a) the completion and use of MONJU, b) FR safety and c) waste minimization, not to mention the making effective use of our knowledge basis and research and engineering capabilities cultivated through the FaCT project in the development of sustainable nuclear energy systems worldwide for future generations of mankind. • Promotion of international cooperation should be an essential ingredient of the strategy

  1. Physical study of the 'snow flake' version of the reactor EL-3 (1963)

    International Nuclear Information System (INIS)

    Abillon, E.; Genthon, J.P.

    1963-01-01

    A systematic study has been made of the different thermodynamic and neutronic aspects of various versions of the 'snow flake' likely to fulfil the aims which were set out (improvement of the reactivity, of the experimental volume, improvement in the working conditions of EL-3). After this systematic study, a choice was made in principle of one version amongst the several available (solutions using 8 or 10 pencils-various methods of heat transfer: corrugation, cartridges which may or may not be filled with helium... etc., variable enrichments... ). The choice is that of a version which can already at the present be considered as feasible, which represents already a notable improvement of the pile and which does not imply the existence of a major problem as far as the possible change-over later to an improved version is concerned. This choice defines a fuel element and a configuration for the core. A detailed recapitulation has been made of all the research being carried out for the possible solutions in this particular case. (authors) [fr

  2. Utilization of research reactors - A global perspective

    International Nuclear Information System (INIS)

    Muranaka, R.G.

    1988-01-01

    This paper presents 1) a worldwide picture of research reactors, operable, shutdown, under construction and planned, 2) statistics on utilization of research reactors including TRIGA reactors, and 3) some results of a survey conducted during 1988 on the utilization of research reactors in developing Member States in the Asia-Pacific Region

  3. Heterogeneous reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author) [pt

  4. Research on 3-D base isolation system applied to new power reactor 3-D seismic isolation device with rolling seal type air spring: Pt.2

    International Nuclear Information System (INIS)

    Junji Suhara; Ryoichiro Matsumoto; Shinsuke Oguri; Yasuo Okada; Kazuhiko Inoue; Kenji Takahashi

    2005-01-01

    A three dimensional seismic base isolation device was developed for heavy structures and buildings such as nuclear power reactor buildings. The device realizes 3-D isolation by combining a LRB (laminated rubber bearing) for horizontal isolation with an air spring for vertical isolation in series. In this study, scale models of the 3-D base isolation device were made and were tested to examine the dynamic properties and ultimate strengths of the device. The performance of the device under earthquake excitation was examined through shaking table tests of 1/7 scale models. As the results, it was confirmed that the device worked smoothly under the horizontal and vertical excitations, and that the theoretical formulae of the orifice damping could explain the test results. The high-pressure air springs of trial production were forced to burst to find out which factor influenced ultimate strength. It was confirmed from results of the burst test that the strength of the air spring depended upon the diameter of rolling part of the bellows and the number of layers of the reinforcing fibers. Judging from the results of the shaking table test and the burst test, the developed 3-D base isolation device was applicable to a nuclear power plant building. (authors)

  5. Nuclear reactor (1960)

    International Nuclear Information System (INIS)

    Maillard, M.L.

    1960-01-01

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [fr

  6. The integral fast reactor concept

    International Nuclear Information System (INIS)

    Chang, Yoon I.; Marchaterre, J.F.

    1987-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) an integral fuel cycle, based on pyrometallurgical processing and injection-cast fuel fabrication, with the fuel cycle facility collocated with the reactor, if so desired. This paper gives a review of the IFR concept

  7. On the energy gain enhancement of DT+D3He fuel configuration in nuclear fusion reactor driven by heavy ion beams

    Directory of Open Access Journals (Sweden)

    S Khoshbinfar

    2016-09-01

    Full Text Available It is expected that advanced fuels be employed in the second generation of nuclear fusion reactors. Theoretical calculations show that in such a fuel, a high plasma temperature about 100 keV is a requisite for reaction rate improvement of nuclear fusion. However, creating such a temporal condition requires a more powerful driver than we have today. Here, introducing an optimal fuel configuration consisting of DT and D-3He layers, suitable for inertial fusion reactors and driven by heavy ion beams, the optimal energy gain conditions have been simulated and derived for 1.3 MJ system. It was found that, in this new fuel configuration, the ideal energy gain, is 22 percent more comparing with energy gain in corresponding single DT fuel layer. Moreover, the inner DT fuel layer contributed as an ignition trigger, while the outer D3He fuel acts as particle and radiation shielding as well as fuel layer.

  8. Investigation of tetrafluoromethane as a plasma gas in a very high pressure/low current dc batch reactor by means of 3D MHD modelling

    International Nuclear Information System (INIS)

    Lebouvier, A; Iwarere, S A; Ramjugernath, D; Fulcheri, L

    2013-01-01

    This paper deals with 3D MHD modelling of the behaviour of a tetrafluoromethane (CF 4 ) plasma arc in a batch reactor under peculiar conditions of low current (0.35 A) and very high pressure (50 atm). The first part of the manuscript presents results for a horizontal configuration of the reactor, as is undertaken experimentally. The model has led to the understanding of the instabilities observed experimentally for such unusual operating conditions. The curved shape of the arc and the sliding of the anodic arc root along the electrode have been revealed to be the source of the experimental instabilities. The latter part of the manuscript investigates the effect of two vertical configurations of the reactor; with a cathode at the top and cathode at the bottom to overcome the instabilities. In these reactor configurations, the arc is much more stable and stays centred in the middle of the electrodes. These configurations are more suitable for the stability of the arc discharge, but have to be verified experimentally. (paper)

  9. Investigation of tetrafluoromethane as a plasma gas in a very high pressure/low current dc batch reactor by means of 3D MHD modelling

    Science.gov (United States)

    Lebouvier, A.; Iwarere, S. A.; Ramjugernath, D.; Fulcheri, L.

    2013-09-01

    This paper deals with 3D MHD modelling of the behaviour of a tetrafluoromethane (CF4) plasma arc in a batch reactor under peculiar conditions of low current (0.35 A) and very high pressure (50 atm). The first part of the manuscript presents results for a horizontal configuration of the reactor, as is undertaken experimentally. The model has led to the understanding of the instabilities observed experimentally for such unusual operating conditions. The curved shape of the arc and the sliding of the anodic arc root along the electrode have been revealed to be the source of the experimental instabilities. The latter part of the manuscript investigates the effect of two vertical configurations of the reactor; with a cathode at the top and cathode at the bottom to overcome the instabilities. In these reactor configurations, the arc is much more stable and stays centred in the middle of the electrodes. These configurations are more suitable for the stability of the arc discharge, but have to be verified experimentally.

  10. Manned space flight nuclear system safety. Volume 3: Reactor system preliminary nuclear safety analysis. Part 1: Reference Design Document (RDD)

    Science.gov (United States)

    1972-01-01

    The Reference Design Document, of the Preliminary Safety Analysis Report (PSAR) - Reactor System provides the basic design and operations data used in the nuclear safety analysis of the Rector Power Module as applied to a Space Base program. A description of the power module systems, facilities, launch vehicle and mission operations, as defined in NASA Phase A Space Base studies is included. Each of two Zirconium Hydride Reactor Brayton power modules provides 50 kWe for the nominal 50 man Space Base. The INT-21 is the prime launch vehicle. Resupply to the 500 km orbit over the ten year mission is provided by the Space Shuttle. At the end of the power module lifetime (nominally five years), a reactor disposal system is deployed for boost into a 990 km high altitude (long decay time) earth orbit.

  11. RAZORBACK - A Research Reactor Transient Analysis Code Version 1.0 - Volume 3: Verification and Validation Report.

    Energy Technology Data Exchange (ETDEWEB)

    Talley, Darren G.

    2017-04-01

    This report describes the work and results of the verification and validation (V&V) of the version 1.0 release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This V&V effort was intended to confirm that the code shows good agreement between simulation and actual ACRR operations.

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    Mysels, K.J.; Shenoy, A.S.

    1976-01-01

    A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

  13. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

    On April 3rd, 1986, two demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the integral fast reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics also makes possible a simplified close fuel cycle and waste process improvements. The paper describes the IFR concept, the inherent safety, tests, and status of IFR development today

  14. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

    On April 3rd, 1986, two dramatic demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the Integral Fast Reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics and also makes possible a simplified closed fuel cycle and waste process improvements

  15. Reactor container cooling device

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1995-11-10

    The device of the present invention efficiently lowers pressure and temperature in a reactor container upon occurrence of a severe accident in a BWR-type reactor and can cool the inside of the container for a long period of time. That is, (1) pipelines on the side of an exhaustion tower of a filter portion in a filter bent device of the reactor container are in communication with pipelines on the side of a steam inlet of a static container cooling device by way of horizontal pipelines, (2) a back flow check valve is disposed to horizontal pipelines, (3) a steam discharge valve for a pressure vessel is disposed closer to the reactor container than the joint portion between the pipelines on the side of the steam inlet and the horizontal pipelines. Upon occurrence of a severe accident, when the pressure vessel should be ruptured and steams containing aerosol in the reactor core should be filled in the reactor container, the inlet valve of the static container cooling device is closed. Steams are flown into the filter bent device of the reactor container, where the aerosols can be removed. (I.S.).

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

    A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

  17. Evaluation of Selected Chemical Processes for Production of Low-cost Silicon, Phase 3. [using a fluidized bed reactor

    Science.gov (United States)

    Blocher, J. M., Jr.; Browning, M. F.

    1979-01-01

    The construction and operation of an experimental process system development unit (EPSDU) for the production of granular semiconductor grade silicon by the zinc vapor reduction of silicon tetrachloride in a fluidized bed of seed particles is presented. The construction of the process development unit (PDU) is reported. The PDU consists of four critical units of the EPSDU: the fluidized bed reactor, the reactor by product condenser, the zinc vaporizer, and the electrolytic cell. An experimental wetted wall condenser and its operation are described. Procedures are established for safe handling of SiCl4 leaks and spills from the EPSDU and PDU.

  18. Application of the Synthesis method to the calculations of neutron flow in 3D in the enveloping of a BWR reactor with the DORT code

    International Nuclear Information System (INIS)

    Xolocostli M, J.V.; Gomez T, A.M.; Palacios H, J.C.

    2006-01-01

    The surveillance program of the vessel materials of a BWR reactor requires the determination of the neutron flux in 3D in the core enveloping. To carry out these calculations of the neutron flux, the Regulatory Guide 1.190 of the NRC recommends the use of the following codes: MCNP, TORT and DORT. In the case of using the DORT code, the one which solves the transport equation in discreet coordinates and in two dimensions (xy, rθ, and rz), the regulatory guide in reference, requires to make an approach of the flow in three dimensions by means of the call Synthesis Method. It is denominated like this due to that a flow representation in 3D is achieved 'combining' or 'synthesizing' the calculated flows by DORT in rθ, rz and r. In this work the application of the Synthesis Method it is presented, according to the Regulatory Guide 1.190, to determine the 3D flows in a BWR reactor. To achieve the above mentioned it was implemented the Synthesis Method in a computer program developed in the ININ to which is denominated SYNTHESIS. This program applies the synthesis method, and is 'coupled' with the DORT code to determine by this way the neutronic fluxes in 3D on the enveloping of a BWR reactor. (Author)

  19. Effect of Co3O4 and CeO2 Infiltration on the Activity of a LSM15/GDC10 Highly Porous Electrochemical Reactor

    DEFF Research Database (Denmark)

    Ippolito, Davide; Kammer Hansen, Kent

    2014-01-01

    . The effect of the infiltration on the electrochemical properties and catalytic activity of the reactor was investigated by electrochemical impedance spectroscopy (EIS) and gas analysis. Figure 1 shows the SEM cross-section micrograph of the electrochemical reactor made of 11 alternating layers of electrode...... matter, which lead to the formation of ozone in urban and regional areas [1]. The electrocatalytic activity of a porous electrochemical reactor, made of La0.85Sr0.15MnO3±δ (LSM) as electrode and Ce0.9Gd0.1O1.95 (GDC) as electrolyte, was studied for the electrochemical oxidation of propene (C3H6), a major...... (LSM) and electrolyte (GDC). Figure 2 shows the Nyquist plot of the impedance spectra of Co3O4 infiltrated backbone recorded at OCP with 10% O2 and 10% O2 + 1000 ppm C3H6, 2 L/h, 400 °C. [1] R. Atkinson, Atmospheric Chemistry of VOCs and NOx, Atmos Environ. 34 (2000) 2063; [Formula]...

  20. Prevalence and typing of Listeria monocytogenes in ready-to-eat food products on the Belgian market.

    Science.gov (United States)

    Van Coillie, Els; Werbrouck, Hadewig; Heyndrickx, Marc; Herman, Lieve; Rijpens, Nancy

    2004-11-01

    Listeria monocytogenes is a major concern to producers of ready-to-eat foods because of the high mortality rate associated with listeriosis and the widespread nature of the organism. To investigate the prevalence of this pathogen in different ready-to-eat food products on the Belgian market, a variety of 252 ready-to-eat food products, mainly fish and meat products, were analyzed. Overall, L. monocytogenes was detected in 23.4% of the samples. The highest prevalence of L. monocytogenes was found in prepared minced meat (42.1%) and smoked halibut (33.3%). Contamination levels were in most cases low (prevalence of Listeria innocua (15.8%) and Listeria welshimeri (36.8%) was detected in prepared minced meat. L. monocytogenes strains isolated from different contaminated products were subjected to repetitive element sequence-based PCR (REP-PCR) typing to determine possible associations with product type, producer, or market. REP-PCR patterns were analyzed using BioNumerics software, and seven different groups with at least 90% similarity were identified. The cluster analysis indicates that cross-contamination occurred at the producer and retail level. Serotype identification of the strains by PCR revealed that most belonged to the 1/2a(3a) serotype group.

  1. A high performance neutron powder diffractometer at 3 MW Triga Mark-II research reactor in Bangladesh

    Energy Technology Data Exchange (ETDEWEB)

    Kamal, I., E-mail: imtiaz-kamal26@yahoo.com; Yunus, S. M., E-mail: yunussm11@yahoo.com; Datta, T. K., E-mail: tk-datta4@yahoo.com; Zakaria, A. K. M.; Das, A. K.; Aktar, S.; Hossain, S. [Institute of Nuclear Science and Technology, Atomic Energy Research Establishment, Ganakbari, Savar, Dhaka (Bangladesh); Berliner, R., E-mail: RB@instrumentationAssociates.com [Instrumentation Associates, Durham, North Carolina (United States); Yelon, W. B., E-mail: yelonwb@hotmail.com [Oak Ridge National Laboratory, Oak Ridge, Tennessee (United States)

    2016-07-12

    A high performance neutron diffractometer called Savar Neutron Diffractometer (SAND) was built and installed at radial beam port-2 of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh. Structural studies of materials are being done by this technique to characterize materials crystallograpohically and magnetically. The micro-structural information obtainable by neutron scattering method is very essential for determining its technological applications. This technique is unique for understanding the magnetic behavior in magnetic materials. Ceramic, steel, electronic and electric industries can be benefited from this facility for improving their products and fabrication process. This instrument consists of a Popovicimonochromator with a large linear position sensitive detector array. The monochromator consists of nine blades of perfect single crystal of silicon with 6 mm thickness each. The monochromator design was optimized to provide maximum flux on 3 mm diameter cylindrical sample with a relatively flat angular dependence of resolution. Five different wave lengths can be selected by orienting the crystal at various angles. A sapphire filter was used before the primary collimator to minimize the first neutron. The detector assembly is composed of 15 linear position sensitive proportional counters placed at either 1.1 m or 1.6 m from the sample position and enclosed in a air pad supported high density polythene shield. Position sensing is obtained by charge division using 1-wide NIM position encoding modules (PEM). The PEMs communicate with the host computer via USB. The detector when placed at 1.1 m, subtends 30° (2θ) at each step and covers 120° in 4 steps. When the detector is placed at 1.6 m it subtends 20° at each step and covers 120° in 6 steps. The instrument supports both low and high temperature sample environment. The instrument supports both low and high temperature sample environment. The diffractometer is a state-of-the art technology

  2. A high performance neutron powder diffractometer at 3 MW Triga Mark-II research reactor in Bangladesh

    Science.gov (United States)

    Kamal, I.; Yunus, S. M.; Datta, T. K.; Zakaria, A. K. M.; Das, A. K.; Aktar, S.; Hossain, S.; Berliner, R.; Yelon, W. B.

    2016-07-01

    A high performance neutron diffractometer called Savar Neutron Diffractometer (SAND) was built and installed at radial beam port-2 of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh. Structural studies of materials are being done by this technique to characterize materials crystallograpohically and magnetically. The micro-structural information obtainable by neutron scattering method is very essential for determining its technological applications. This technique is unique for understanding the magnetic behavior in magnetic materials. Ceramic, steel, electronic and electric industries can be benefited from this facility for improving their products and fabrication process. This instrument consists of a Popovicimonochromator with a large linear position sensitive detector array. The monochromator consists of nine blades of perfect single crystal of silicon with 6mm thickness each. The monochromator design was optimized to provide maximum flux on 3mm diameter cylindrical sample with a relatively flat angular dependence of resolution. Five different wave lengths can be selected by orienting the crystal at various angles. A sapphire filter was used before the primary collimator to minimize the first neutron. The detector assembly is composed of 15 linear position sensitive proportional counters placed at either 1.1 m or 1.6 m from the sample position and enclosed in a air pad supported high density polythene shield. Position sensing is obtained by charge division using 1-wide NIM position encoding modules (PEM). The PEMs communicate with the host computer via USB. The detector when placed at 1.1 m, subtends 30˚ (2θ) at each step and covers 120˚ in 4 steps. When the detector is placed at 1.6 m it subtends 20˚ at each step and covers 120˚ in 6 steps. The instrument supports both low and high temperature sample environment. The instrument supports both low and high temperature sample environment. The diffractometer is a state-of-the art technology

  3. A high performance neutron powder diffractometer at 3 MW Triga Mark-II research reactor in Bangladesh

    International Nuclear Information System (INIS)

    Kamal, I.; Yunus, S. M.; Datta, T. K.; Zakaria, A. K. M.; Das, A. K.; Aktar, S.; Hossain, S.; Berliner, R.; Yelon, W. B.

    2016-01-01

    A high performance neutron diffractometer called Savar Neutron Diffractometer (SAND) was built and installed at radial beam port-2 of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh. Structural studies of materials are being done by this technique to characterize materials crystallograpohically and magnetically. The micro-structural information obtainable by neutron scattering method is very essential for determining its technological applications. This technique is unique for understanding the magnetic behavior in magnetic materials. Ceramic, steel, electronic and electric industries can be benefited from this facility for improving their products and fabrication process. This instrument consists of a Popovicimonochromator with a large linear position sensitive detector array. The monochromator consists of nine blades of perfect single crystal of silicon with 6 mm thickness each. The monochromator design was optimized to provide maximum flux on 3 mm diameter cylindrical sample with a relatively flat angular dependence of resolution. Five different wave lengths can be selected by orienting the crystal at various angles. A sapphire filter was used before the primary collimator to minimize the first neutron. The detector assembly is composed of 15 linear position sensitive proportional counters placed at either 1.1 m or 1.6 m from the sample position and enclosed in a air pad supported high density polythene shield. Position sensing is obtained by charge division using 1-wide NIM position encoding modules (PEM). The PEMs communicate with the host computer via USB. The detector when placed at 1.1 m, subtends 30° (2θ) at each step and covers 120° in 4 steps. When the detector is placed at 1.6 m it subtends 20° at each step and covers 120° in 6 steps. The instrument supports both low and high temperature sample environment. The instrument supports both low and high temperature sample environment. The diffractometer is a state-of-the art technology

  4. Mirror hybrid reactor optimization studies

    International Nuclear Information System (INIS)

    Bender, D.J.

    1976-01-01

    A system model of the mirror hybrid reactor has been developed. The major components of the model include (1) the reactor description, (2) a capital cost analysis, (3) various fuel management schemes, and (4) an economic analysis that includes the hybrid plus its associated fission burner reactors. The results presented describe the optimization of the mirror hybrid reactor, the objective being to minimize the cost of electricity from the hybrid fission-burner reactor complex. We have examined hybrid reactors with two types of blankets, one containing natural uranium, the other thorium. The major difference between the two optimized reactors is that the uranium hybrid is a significant net electrical power producer, whereas the thorium hybrid just about breaks even on electrical power. Our projected costs for fissile fuel production are approximately 50 $/g for 239 Pu and approximately 125 $/g for 233 U

  5. Reactor Neutron Sources

    International Nuclear Information System (INIS)

    Aksenov, V.L.

    1994-01-01

    The present status and the prospects for development of reactor neutron sources for neutron scattering research in the world are considered. The fields of application of neutron scattering relative to synchrotron radiation, the creation stages of reactors (steady state and pulsed) and their position in comparison with spallation neutron sources at present and in the foreseen future are discussed. (author). 15 refs.; 8 figs.; 3 tabs

  6. Generation IV reactors: economics

    International Nuclear Information System (INIS)

    Dupraz, B.; Bertel, E.

    2003-01-01

    The operating nuclear reactors were built over a short period: no more than 10 years and today their average age rounds 18 years. EDF (French electricity company) plans to renew its reactor park over a far longer period : 30 years from 2020 to 2050. According to EDF this objective implies 3 constraints: 1) a service life of 50 to 60 years for a significant part of the present operating reactors, 2) to be ready to built a generation 3+ unit in 2020 which infers the third constraint: 3) to launch the construction of an EPR (European pressurized reactor) prototype as soon as possible in order to have it operating in 2010. In this scheme, generation 4 reactor will benefit the feedback experience of generation 3 and will take over in 2030. Economic analysis is an important tool that has been used by the generation 4 international forum to select the likely future reactor systems. This analysis is based on 4 independent criteria: the basic construction cost, the construction time, the operation and maintenance costs and the fuel cycle cost. This analysis leads to the evaluation of the global cost of electricity generation and of the total investment required for each of the reactor system. The former defines the economic competitiveness in a de-regulated energy market while the latter is linked to the financial risk taken by the investor. It appears, within the limits of the assumptions and models used, that generation 4 reactors will be characterized by a better competitiveness and an equivalent financial risk when compared with the previous generation. (A.C.)

  7. Comparison of finite element and finite difference methods for 2D and 3D calculations with Monte Carlo method results for idealized cases of a heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Grant, Carlos; Marconi, Javier; Serra, Oscar [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)]. E-mail: grant@cnea.gov.ar; Mollerach, Ricardo; Fink, Jose [Nucleoelectrica Argentina S.A., Buenos Aires (Argentina)]. E-mail: RMollerach@na-sa.com.ar; JFink@na-sa.com.ar

    2005-07-01

    Nowadays, the increased calculation capacity of modern computers allows us to evaluate the 2D and 3D flux and power distribution of nuclear reactor in a reasonable amount of time using a Monte Carlo method. This method gives results that can be considered the most reliable evaluation of flux and power distribution with a great amount of detail. This is the reason why these results can be considered as benchmark cases that can be used for the validation of other methods. For this purpose, idealized models were calculated using Monte Carlo (code MCNP5) for the ATUCHA I reactor. 2D and 3D cases with and without control rods and channels without fuel element were analyzed. All of them were modeled using a finite element code (DELFIN) and a finite difference code (PUMA). In both cases two energy groups were use. (author)

  8. Development of a general coupling interface for the fuel performance code transuranus tested with the reactor dynamic code DYN3D

    International Nuclear Information System (INIS)

    Holt, L.; Rohde, U.; Seidl, M.; Schubert, A.; Van Uffelen, P.

    2013-01-01

    Several institutions plan to couple the fuel performance code TRANSURANUS developed by the European Institute for Transuranium Elements with their own codes. One of these codes is the reactor dynamic code DYN3D maintained by the Helmholtz-Zentrum Dresden - Rossendorf. DYN3D was developed originally for VVER type reactors and was extended later to western type reactors. Usually, the fuel rod behavior is modeled in thermal hydraulics and neutronic codes in a simplified manner. The main idea of this coupling is to describe the fuel rod behavior in the frame of core safety analysis in a more detailed way, e.g. including the influence of the high burn-up structure, geometry changes and fission gas release. It allows to take benefit from the improved computational power and software achieved over the last two decades. The coupling interface was developed in a general way from the beginning. Thence it can be easily used also by other codes for a coupling with TRANSURANUS. The user can choose between a one-way as well as a two-way online coupling option. For a one-way online coupling, DYN3D provides only the time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, but the fuel performance code doesn’t transfer any variable back to DYN3D. In a two-way online coupling, TRANSURANUS in addition transfers parameters like fuel temperature and cladding temperature back to DYN3D. This list of variables can be extended easily by geometric and further variables of interest. First results of the code system DYN3D-TRANSURANUS will be presented for a control rod ejection transient in a modern western type reactor. Pre-analyses show already that a detailed fuel rod behavior modeling will influence the thermal hydraulics and thence also the neutronics due to the Doppler reactivity effect of the fuel temperature. The coupled code system has therefore a potential to improve the assessment of safety criteria. The developed code system DYN3D-TRANSURANUS can be used also

  9. RB Research nuclear reactor, Annual report for 2005

    International Nuclear Information System (INIS)

    Milosevic, M.; Dasic, N.; Ljubenov, V.; Pesic, M.; Nikolic, D; Jevremovic, M.; Minic, D.

    2006-01-01

    Report on RB reactor operation during 2005 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation during 2005

  10. RB research nuclear reactor - Annual report for 1986, I - III

    International Nuclear Information System (INIS)

    Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Jevremovic, M.; Ilic, I.

    1987-01-01

    This report includes data concerning the RB reactor operation in 1986, state of the reactor components, data about the employed personnel and the database of experimental and other reactor related devices. It is made of 3 parts: Engineering description and operation of the RB reactor including dosimetry, reactor staff data and financial report; Reactor facility components and maintenance; RB reactor operation and utilization in 1986 [sr

  11. Synthesis of the IRSN report on the control-command architecture of the Flamanville 3 EPR reactor and the associated platforms

    International Nuclear Information System (INIS)

    2009-06-01

    This document briefly reports the IRSN's comments on the robustness of the control-command architecture of the Flamanville 3 EPR (European pressurized reactor), that is of the equipment which, to various extent, contribute to the installation safety. This report identifies the equipment the robustness of which should be improved, outlines the problems created by the complexity of computerized systems. Notably, the authors comment the ability of the off-the-shelf computer platforms acquired by EDF to deal with the nuclear context

  12. Methanol oxidation in a flow reactor: Implications for the branching ratio of the CH3OH+OH reaction

    DEFF Research Database (Denmark)

    Rasmussen, Christian Lund; Wassard, K.H.; Dam-Johansen, Kim

    2008-01-01

    The oxidation of methanol in a flow reactor has been studied experimentally under diluted, fuel-lean conditions at 650-1350 K, over a wide range of O-2 concentrations (1%-16%), and with and without the presence of nitric oxide. The reaction is initiated above 900 K, with the oxidation rate...

  13. Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jae Jun; Chung, Bub Dong

    2005-09-15

    For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis.

  14. The Belgian nuclear higher education network: the evolution of an academic programme in nuclear engineering

    International Nuclear Information System (INIS)

    Berkvens, T.; Coeck, M.

    2014-01-01

    The master-after-master in nuclear engineering provided by the Belgian Nuclear higher Education Network (BNEN) is a one-year, 60 ECTS programme which combines the expertise of six Belgian universities and SCK.CEN, the Belgian Nuclear Research Centre, which participates through its Academy for Nuclear Science and Technology. It was created in close collaboration with representatives of academia, research centres, industry and other nuclear stakeholders. The BNEN consortium Due to its modular programme, BNEN is accessible for both full-time students (mainly young engineering graduates) as well as young professionals already employed in the nuclear industry. The programme is offered in English to facilitate the participation of international students. One of the important aspects of the BNEN programme is the fact that exercises and hands-on sessions in the specialised laboratories of SCK.CEN complement the theoretical classes to bring the students into contact with all facets of nuclear energy. Several of SCK.CEN's researchers provide valuable contributions to the programme through seminars and practical exercises. From their daily practices and responsibilities they give an expert view on the subjects that are being taught. In 2012, in the framework of an official accreditation process all aspects of the BNEN programme were audited by an international visitation panel. The most important outcome of this process is the current reform of the academic programme, which will be implemented in the academic year 2014-2015, taking into account the recommendations by the visitation panel. In this paper, the history of the BNEN programme will be discussed, the new BNEN programme will be presented as well as the process that has led to its implementation. (authors)

  15. Individual and environmental correlates of objectively measured sedentary time in Dutch and Belgian adults.

    Directory of Open Access Journals (Sweden)

    Femke van Nassau

    Full Text Available As the detrimental health effects of sedentary behaviour are well established, insight into the individual and environmental factors that influence adults' sedentary behaviour is needed. Most studies to date rely on self-reported measures of sedentary time. Therefore, the aim of the current study was to examine individual and environmental correlates of objectively measured sedentary time in Dutch and Belgian adults. Between March and August 2014, Belgian (n = 133 and Dutch (n = 223 adults, recruited as sub-sample of the SPOTLIGHT survey, wore an ActiGraph accelerometer to provide objectively measured sedentary and moderate to vigorous physical activity time. Participants completed a questionnaire assessing sociodemographic (country of residence, age, gender and educational level, lifestyle (sleep, smoking, sugar-containing beverage consumption, alcohol intake, health (body mass index, self-rated health, work (employment status and type of work, happiness, physical environmental (owning a car, number of screens, socioeconomic status and residential density and social environmental factors (social network, social cohesion. Univariate and multivariable regression analyses showed that Belgian participants had a lower odds of being sedentary for at least 9 hours per day compared to Dutch participants. Women, older participants and those meeting the WHO recommendation for physical activity were also less likely to sit for 9 hours or more per day. Participants doing (heavy manual work or being in education, homemaker, unemployed had lower odds of being sedentary for at least 9 hours per day compared to participants with a sitting job. Those with a higher self-reported social network also had lower odds for sedentary time. No associations between physical and other social environmental characteristics and sedentary time were found. Our findings add to the growing evidence of factors associated with prolonged sedentary time in adults. These findings may

  16. Influence of NH3 on secondary organic aerosols from the ozonolysis and photooxidation of α-pinene in a flow reactor

    Science.gov (United States)

    Babar, Zaeem Bin; Park, Jun-Hyun; Lim, Ho-Jin

    2017-09-01

    This study presents detailed characterizations of a newly-developed flow reactor including (1) residence time distribution measurements, (2) relative humidity (RH) and temperature control, and (3) OH radical exposure range (i.e., atmospheric aging time). Hydroxyl (OH) radical exposures ranged from 8.20 × 1010 to 7.22 × 1011 molecules cm-3 s (0.5-4.9 d of atmospheric aging). In this study, the effects of NH3 gas on the secondary organic aerosol (SOA) formation of α-pinene by dark ozonolysis and photooxidation were investigated using the newly-developed flow reactor. For both dark α-pinene ozonolysis and photooxidation, higher SOA yields were observed in the presence of NH3 than in the absence of NH3. At RH of ∼50%, the SOA yield for ozonolysis and photooxidation in the presence of NH3 increased by 23% and 15% relative to those in the absence of NH3. Similar effects were observed at lower and higher RH conditions. Fourier transform infrared spectroscopy analysis confirmed the presence of nitrogen-containing functional groups in SOA formed in the presence of NH3. The α-pinene SOA formed in the presence of NH3 showed higher absorption and fluorescence for UV-visible radiation than those formed in the absence of NH3.

  17. Hydrophilic interaction chromatography and evaporative light scattering detection for the determination of polar analytes in Belgian endive.

    Science.gov (United States)

    Annaratone, Carlo; De Roeck, Ans; Hertog, Maarten L A T M; Nicolaï, Bart M

    2017-08-15

    Belgian endive (Cichorium intybus L. var. foliosum Hegi), a popular produce in northern Europe, has been thoroughly studied in regard to its bitter sesquiterpene lactones content. Much less is known about on its sweetness and crunchiness, which are typically linked to the content of polar compounds such as sugars, organic acids and salts. Through HILIC-HPLC-MS, it was shown that simple sugars, amino acids, and potassium chloride are abundant in Belgian endive extracts. Subsequently, a HILIC-HPLC-ELSD method for the analysis of such compounds with run times below six minutes was developed. Recoveries varied between 80 and 110% and an average reproducibility was 7.5RSD%. Finally, the method was applied to the study of three difference Belgian endive varieties. Takine, a variety known for its sweet taste, was found to contain significantly higher levels of fructose, and lower levels of potassium and glutamine. Copyright © 2017 Elsevier Ltd. All rights reserved.

  18. TU Electric reactor physics model verification: Power reactor benchmark

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1988-01-01

    Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

  19. Backfitting of research reactors

    International Nuclear Information System (INIS)

    Delrue, R.; Noesen, T.

    1985-01-01

    The backfitting of research reactors covers a variety of activities. 1. Instrumentation and control: Control systems have developed rapidly and many reactor operators wish to replace obsolete equipment by new systems. 2. Pool liners: Some pools are lined internally with ceramic tiles. These may become pervious with time necessitating replacement, e.g. by a new stainless steel liner. 3. Heat removal system: Deficiencies can occur in one or more of the cooling system components. Upgrading may require modifications of the system such as addition of primary loops, introduction of deactivation tanks, pump replacement. Recent experience in such work has shown that renewal, backfitting and upgrading of an existing reactor is economically attractive since the related costs and delivery times are substantially lower than those required to install a new research reactor

  20. Nuclear research reactors activities in INVAP

    International Nuclear Information System (INIS)

    Ordonez, Juan Pablo

    2013-01-01

    This presentation describes the different activities in the research reactor field that are being carried out by INVAP. INVAP is presently involved in the design of three new research reactors in three different countries. The RA-10 is a multipurpose reactor, in Argentina, planned as a replacement for the RA-3 reactor. INVAP was contracted by CNEA for carrying out the preliminary engineering for this reactor, and has recently been contracted by CNEA for the detailed engineering. CNEA groups are strongly involved in the design of this reactor. The RMB is a multipurpose reactor, planned by CNEN from Brazil. CNEN, through REDETEC, has contracted INVAP to carry out the preliminary engineering for this reactor. As the user requirements for RA-10 and RMB are very similar, an agreement was signed between Argentina and Brasil governments to cooperate in these two projects. The agreement included that both reactors would use the OPAL reactor in Australia, design and built by INVAP, as a reference reactor. INVAP has also designed the LPRR reactor for KACST in Saudi Arabia. The LPRR is a 30 kw reactor for educational purposes. KACST initially contracted INVAP for the engineering for this reactor and has recently signed the contract with INVAP for building the reactor. General details of these three reactors will be presented