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Sample records for belgian reactor 1

  1. Irradiation of Fuel Elements in the Belgian BR3 Reactor

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

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

    Science.gov (United States)

    Wagemans, J.

    2008-04-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  8. Pandemic A/H1N1v influenza 2009 in hospitalized children: a multicenter Belgian survey

    Directory of Open Access Journals (Sweden)

    Blumental Sophie

    2011-11-01

    Full Text Available Abstract Background During the 2009 influenza A/H1N1v pandemic, children were identified as a specific "at risk" group. We conducted a multicentric study to describe pattern of influenza A/H1N1v infection among hospitalized children in Brussels, Belgium. Methods From July 1, 2009, to January 31, 2010, we collected epidemiological and clinical data of all proven (positive H1N1v PCR and probable (positive influenza A antigen or culture pediatric cases of influenza A/H1N1v infections, hospitalized in four tertiary centers. Results During the epidemic period, an excess of 18% of pediatric outpatients and emergency department visits was registered. 215 children were hospitalized with proven/probable influenza A/H1N1v infection. Median age was 31 months. 47% had ≥ 1 comorbid conditions. Febrile respiratory illness was the most common presentation. 36% presented with initial gastrointestinal symptoms and 10% with neurological manifestations. 34% had pneumonia. Only 24% of the patients received oseltamivir but 57% received antibiotics. 10% of children were admitted to PICU, seven of whom with ARDS. Case fatality-rate was 5/215 (2%, concerning only children suffering from chronic neurological disorders. Children over 2 years of age showed a higher propensity to be admitted to PICU (16% vs 1%, p = 0.002 and a higher mortality rate (4% vs 0%, p = 0.06. Infants less than 3 months old showed a milder course of infection, with few respiratory and neurological complications. Conclusion Although influenza A/H1N1v infections were generally self-limited, pediatric burden of disease was significant. Compared to other countries experiencing different health care systems, our Belgian cohort was younger and received less frequently antiviral therapy; disease course and mortality were however similar.

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

    Science.gov (United States)

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

    2002-04-23

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

  10. Reactor Materials Research

    International Nuclear Information System (INIS)

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

  11. 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)

  12. Strategic groups in the Belgian fishing fleet

    OpenAIRE

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

    2011-01-01

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

  13. R and D on fuzzy control applications to the BR1 research reactor

    International Nuclear Information System (INIS)

    Fuzzy control applications in nuclear reactor operations present a tremendous challenge. The main reason for this is the public awareness of the risks of nuclear industry and the very strict safety regulations in force for nuclear power plants. The very same regulations prevent a researcher from quickly introducing novel fuzzy-logic methods into this field. On the other hand, the application of fuzzy logic has, despite the ominous sound of the word 'fuzzy' to nuclear engineers, a number of very desirable advantages over classical methods, e.g., its robustness and the capability to include human experience into the controller. In this paper we report an on-going R and D project for controlling the power level of the Belgian Nuclear Reactor 1 (BR1) at the Belgian Nuclear Research Centre (SCK.CEN). The project started in 1995 and aims to investigate the added value of fuzzy control for nuclear reactors. We first review some relevant literature on fuzzy logic control in nuclear reactors, then present the state-of-the-art of the BR1 project. After experimenting fuzzy logic control under off-line tests at the BR1 reactor, we now foresee a new development for a closed-loop fuzzy control as an on-line operation of the BR1 reactor. Finally, we present the new development for a closed-loop fuzzy logic control at BR1 with an understanding of the safety requirements for this real fuzzy logic control application in nuclear rectors. (author)

  14. MOLIERE PROJECT - Belgian National Report

    OpenAIRE

    Naedenoen, Frédéric

    2014-01-01

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

  15. The Belgian nuclear research centre

    International Nuclear Information System (INIS)

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

  16. 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.

  17. The utility of different reactor types for the research

    International Nuclear Information System (INIS)

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

  18. Fuel cycle transition - A Belgian implementation scenario

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    RUAN, Da

    2000-01-01

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

  20. TRR-1/M1 reactor pool refurbishment

    International Nuclear Information System (INIS)

    The pool refurbishment of the TRR-1/M1 is intended to maintain the pool and irradiation facilities in the operable condition prior to the next decade before making decision whether the reactor will be shutdown and decommissioning or used for other purposes. What ever reason the TRR-1/M1 will serve as a training tool for scientists or engineers, and isotopes production or other analytical works for a period of time until the new research reactor will be established. (orig.)

  1. The Belgian Nuclear Higher Education Network

    International Nuclear Information System (INIS)

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

  2. Reactor operations at SAFARI-1

    International Nuclear Information System (INIS)

    A vigorous commercial programme of isotope production and other radiation services has been followed by the SAFARI-1 research reactor over the past ten years - superimposed on the original purpose of the reactor to provide a basic tool for nuclear research, development and education to the country at an institutional level. A combination of the binding nature of the resulting contractual obligations and tighter regulatory control has demanded an equally vigorous programme of upgrading, replacement and renovation of many systems in order to improve the safety and reliability of the reactor. Not least among these changes is the more effective training and deployment of operations personnel that has been necessitated as the operational demands on the reactor evolved from five days per week to twenty four hours per day, seven days per week, with more than 300 days per year at full power. This paper briefly sketches the operational history of SAFARI-1 and then focuses on the training and structuring currently in place to meet the operational needs. There is a detailed step-by-step look at the operator?s career plan and pre-defined milestones. Shift work, especially the shift cycle, has a negative influence on the operator's career path development, especially due to his unavailability for training. Methods utilised to minimise this influence are presented. The increase of responsibilities regarding the operation of the reactor, ancillaries and experimental facilities as the operator progresses with his career are discussed. (author)

  3. Moving ring reactor 'Karin-1'

    International Nuclear Information System (INIS)

    The conceptual design of a moving ring reactor ''Karin-1'' has been carried out to advance fusion system design, to clarify the research and development problems, and to decide their priority. In order to attain these objectives, a D-T reactor with tritium breeding blanket is designed, a commercial reactor with net power output of 500 MWe is designed, the compatibility of plasma physics with fusion engineering is demonstrated, and some other guideline is indicated. A moving ring reactor is composed mainly of three parts. In the first formation section, a plasma ring is formed and heated up to ignition temperature. The plasma ring of compact torus is transported from the formation section through the next burning section to generate fusion power. Then the plasma ring moves into the last recovery section, and the energy and particles of the plasma ring are recovered. The outline of a moving ring reactor ''Karin-1'' is described. As a candidate material for the first wall, SiC was adopted to reduce the MHD effect and to minimize the interaction with neutrons and charged particles. The thin metal lining was applied to the SiC surface to solve the problem of the compatibility with lithium blanket. Plasma physics, the engineering aspect and the items of research and development are described. (Kako, I.)

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

    OpenAIRE

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  8. 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.

  9. Reactor pressure vessel steels

    International Nuclear Information System (INIS)

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

  10. Safety operation of training reactor VR-1

    International Nuclear Information System (INIS)

    There are three nuclear research reactors in the Czech Republic in operation now: light water reactor LVR-15, maximum reactor power 10 MWt, owner and operator Nuclear Research Institute Rez; light water zero power reactor LR-0, maximum reactor power 5 kWt, owner and operator Nuclear Research Institute Rez and training reactor VR-1 Sparrow, maximum reactor power 5 kWt, owner and operate Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. The training reactor VR-1 Vrabec 'Sparrow', operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly it is designed for training the students of Czech universities, preparing the experts for the Czech nuclear programme, as well as for certain research work, and for information programmes in the nuclear programme, as well as for certain research work, and for information programmes in sphere of using the nuclear energy (public relations). (author)

  11. Artificial intelligence in nuclear reactor operation

    International Nuclear Information System (INIS)

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

  12. Annual report on JEN-1 reactor

    International Nuclear Information System (INIS)

    In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

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

    International Nuclear Information System (INIS)

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

  14. Utilization of the VR-1 training reactor

    International Nuclear Information System (INIS)

    The Paper presents basic information about utilization of the VR-1 Training Reactor at FNSPE, CTU in Prague. The reactor has been used very efficiently especially for education of university students and specialists in favour of the Czech nuclear programme for more than 15 years. It is the only reactor of this type in the Czech Republic. Therefore, students from several Czech technical universities and also from universities in Central Europe participate on its use. The VR-1 Reactor is well equipped for education and training not only by the experimental facility itself but also by carefully developed training methods. The education experiments can be combined into training courses attended by students according to their study specialization. The training programme covers overall information on nuclear safety, radiation protection, emergency preparedness, and physical protection principles. Every year, approximately 250 university students undergo training at VR-1 Reactor. Their stay at reactor site means an enormous benefit for their study process. (author)

  15. Shielding calculations for ETRR-1 reactor

    International Nuclear Information System (INIS)

    The flux and dose through ETRR-1 reactor shielding are calculated using ANISN code. The neutron and gamma radiation sources in the reactor core are determined by using Madland Nix Model (MNM Model ). The results show that the flux in the core is in good agreement with the reactor flux. The dose in the radial and axial shields at outside boundary is less than the maximum allowable dose

  16. Training courses at VR-1 reactor

    International Nuclear Information System (INIS)

    This paper describes one of the main purposes of the VR-1 training reactor utilization - i.e. extensive educational program. The educational program is intended for the training of university students and selected nuclear power plant personnel. The training courses provide them experience in reactor and neutron physics, dosimetry, nuclear safety and operation of nuclear facilities. At present, the training course participants can go through more than 20 standard experimental exercises; particular exercises for special training can be prepared. Approximately 200 university students become familiar with the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. The VR-1 reactor takes also part in Eugene Wigner Course on Reactor Physics Experiments in the framework of European Nuclear Educational Network (ENEN) association. Recently, training courses for Bulgarian research reactor specialists supported by IAEA were carried out. An attractive program including demonstration of reactor operation is prepared also for high school students. Every year, more than 1500 high school students come to visit the reactor, as do many foreigner visitors. (author)

  17. Reactor physics

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Buyens, Dirk; Wouters, Karen

    2005-01-01

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

  19. BRAMS --- the Belgian RAdio Meteor Stations

    Science.gov (United States)

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

    2014-07-01

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

  20. Safety of Ghana Research Reactor (GHARR-1)

    International Nuclear Information System (INIS)

    The Ghana Research Reactor, GHARR-1 is a low power research rector with maximum thermal power lever of 30kW. The reactor is inherently safe and uses highly enriched uranium (HEU) as fuel, light water as moderator and beryllium as a reflector. The construction, commissioning and operation of this reactor have been subjected to the system of authorization and inspection developed by the Regulatory Authority, the Radiation Protection Board (RPB) with the assistance of the International Atomic Energy Agency. The reactor has been regulated by the preparation of an Interim Safety Analysis Report (SAR) based upon International Atomic Energy Agency standards. An International Safety Assessment peer review and safe inspections have confirmed a high level of operational safety of the reactor since it started operation in 1994. Since its operation there has been no significant reported incident/accidents. Several studies have validated the inherent safety of the reactor. The reactor has been used for neutron activation analysis of various samples, research and teaching. About 1000 samples are analysed annually. The final Safety Analysis Report (SAR) was submitted (after five years of extensive research on the operational reactor) to the Regulatory Authority for review in June 2000. (author)

  1. A Belgian Approach to Learning Disabilities.

    Science.gov (United States)

    Hayes, Cheryl W.

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

  2. Annual report on JEN-1 reactor; Informe periodico del Reactor JEN-1 correspondiente al ano 1971

    Energy Technology Data Exchange (ETDEWEB)

    Montes, J.

    1972-07-01

    In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

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

    OpenAIRE

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

    2010-01-01

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

  4. Extensive utilization of training reactor VR-1

    International Nuclear Information System (INIS)

    Full text: The training reactor VR-1 Vrabec ('Sparrow'), operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly, it is designed and operated for training of students from Czech universities, preparing of experts for the Czech nuclear programme, as well as for certain research and development work, and for information programmes in the sphere of non-military nuclear energy use (public relation). The VR-1 training reactor is a pool-type light-water reactor based on enriched uranium with maximum thermal power 1kWth and short time period up to 5kWth. The moderator of neutrons is light demineralized water (H2O) that is also used as a reflector, a biological shielding, and a coolant. Heat is removed from the core with natural convection. The reactor core contains 14 to 18 fuel assemblies IRT-3M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The core is accommodated in a cylindrical stainless steel vessel - pool, which is filled with water. UR-70 control rods serve the reactor control and safe shutdown. Training of the VR-1 reactor provides students with experience in reactor and neutron physics, dosimetry, nuclear safety, and nuclear installation operation. Students from technical universities and from natural sciences universities come to the reactor for training. Approximately 200 university students are introduced to the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. Practical Course on Reactor Physics in Framework of European Nuclear Engineering Network has been newly introduced. Currently, students can try out more than 20 experimental exercises. Further training courses have been included to

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

    Science.gov (United States)

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

    2007-01-01

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

  6. History of the Belgian nuclear power controversy

    International Nuclear Information System (INIS)

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

  7. Pakistan research reactor-1 and its upgradation

    International Nuclear Information System (INIS)

    In this article the author describes the procedure of renovation and upgradation of a swimming pool type Pakistan Research Reactor-1 (PARR-1) installed at PINSTECH. The reactor originally designed for a thermal power of 5 MW using highly enriched uranium as has been upgraded 10 MW with low enriched uranium as fuel. All the required safety precaution has been also modified with the new requirements. The cooling system of PARR-1 was modified to meet the requirements of upgraded power of 10 MW. In order to ensure safety for upgraded PARR-1 and to bring the reactor the current safety standards, some additional safety systems have been provided. An emergency core cooling system ECCS has been installed to remove core decay heat in case of loss of coolant accident (LOCA). (A.B.)

  8. Perspectives on reactor safety. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A. [Oak Ridge National Lab., TN (United States). Engineering Technology Div.

    1997-11-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  9. Perspectives on reactor safety. Revision 1

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course

  10. Integrated modelling of the Belgian coastal zone

    OpenAIRE

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

    2001-01-01

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

  11. An ecosystem approach towards Belgian coastal policy

    OpenAIRE

    Vanden Eede, S.

    2013-01-01

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

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

    OpenAIRE

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

    2002-01-01

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

  13. Environmental radiological monitoring at Pakistan research reactor - 1 (PARR-1)

    International Nuclear Information System (INIS)

    The radiological monitoring channels of Pakistan Research Reactor-1 (PARR)1 to monitor the release of radioactive materials into the environment. This paper presents the scope of the radiological monitoring in different areas of reactor facility and describes the detection of various probable hazards and remedial action taken which generally lead to scramming the reactor. This paper also describes a new radiological monitoring channel, which is locally developed and is in use for several years for measurement of nuclear radiation in the environment. (author)

  14. Estimation of radioactivity in structural materials of ETRR-1 reactor

    International Nuclear Information System (INIS)

    Precise knowledge of the thermal neutron flux in the different structural materials of a reactor is necessary to estimate the radioactive inventory in these materials that are needed in any decommissioning study of the reactor. ETRR-1 is a research reactor that went critical on 2/1691. In spite of this long age of the reactor, the effective operation time of this reactor is very short since the reactor was shutdown for long periods. Because of this long age one may think of reactor decommissioning. For this purpose, the radioactivity of the reactor structural materials was estimated. Apart from the reactor core, the important structural materials in the ETRR-1 are the reactor tank, shielding concrete, and the graphite thermal column. The thermal neutron flux was determined by the monte Carlo method in these materials and the isotope inventory and the radioactivity were calculated by the international code ORIGEN-JR. 1 fig

  15. Planning and implementation of the Belgian nuclear programme

    International Nuclear Information System (INIS)

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

  16. Determination flux in the Reactor JEN-1

    International Nuclear Information System (INIS)

    This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 μ gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs

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

    International Nuclear Information System (INIS)

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

  18. Simulation of the TR-1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dayday, N.; Alsan, S.; Erk, S.

    1978-01-01

    TR-1 is a 1 MW pool-type research reactor. A simulation of TR-1 was attempted in order to predict the values and the variations of principal parameters during the normal and accident conditions. A model based on point kinetics was developed and the variations of neutronics and thermal parameters were studied. A computer program was prepared and successfully run on a desktop calculator HP 9821. Thus it has been shown that a digital computer may be used in a simulation problem in contrast to an analog or hybrid type which are commonly used.

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

    OpenAIRE

    Garrone, Maria; Pieters, Hannah; Swinnen, Jo

    2015-01-01

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

  20. Reactor inventory monitoring system for Angra-1 reactor

    International Nuclear Information System (INIS)

    This work describes the project of Reactor Inventory Monitoring System, which will be installed in Angra I Nuclear Power Plant. The inventory information is important to the operators take corrective actions in case of an incident that may cause a failure in the core cooling. (author)

  1. Probabilistic risk analysis of Angra-1 reactor

    International Nuclear Information System (INIS)

    The first phase of probabilistic study for safety analysis and operational analysis of Angra-1 reactor is presented. The study objectives and uses are: to support decisions about safety problems; to identify operational and/or project failures; to amplify operator qualification tests to include accidents in addition to project base; to provide informations to be used in development and/or review of operation procedures in emergency, test and maintenance procedures; to obtain experience for data collection about abnormal accurences; utilization of study results for training operators; and training of evaluation and reliability techniques for the personnel of CNEN and FURNAS. (M.C.K.)

  2. FFTF reactor immersion heaters. Revision 1

    International Nuclear Information System (INIS)

    This specification establishes requirements for design, testing, and quality assurance for electric heaters that will be used to maintain primary Sodium temperature in the Fast Test Facility (FFTF) reactor vessel. The Test Specification (WHC-SD-FF-SDS-003) has been revised to Rev. 1. This change modifies the fabrication of approximately 25 feet of the subject heater using ceramic insulators over the heater lead wire rather than compressed magnesium oxide. Also, 304 or 316 stainless steel can be used for the heater sheath. This change should simplify fabrication and improve the heater operational reliability

  3. Prospects for Revitalising the Belgian University System.

    Science.gov (United States)

    Hecquet, Ignace

    1984-01-01

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

  4. Performance communication of the Belgian Railway

    NARCIS (Netherlands)

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

    2007-01-01

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

  5. Ceramic solvents for the RA-1 Reactor

    International Nuclear Information System (INIS)

    Full text: The new version of fuel for the RA-1 reactor will consist of Zry cladding tubes of 8.2 mm external diameter and, as fuel material, of pellets of 20% enriched U3O8 and Al powder mixture. Because significant internal temperatures are expected (higher than those obtained in plate fuel elements) and that Al oxide formation in the interfaces between U3O8 particles have been detected in experiences outside the reactor and in post-irradiation examinations, the possibility to use ceramic oxides as fissile material solvent, for instance Alumina or Zirconia is explored, since they are inert materials which would avoid fuel transformations and, therefore, possible dimensional and thermal changes. Besides, the oxidation resistance improvement is analyzed through the utilization of UO2 in the Al cermet, since it is more stable in the reaction with Al. The possibility of using the existing utilities to explore these alternatives is considered. This type of fuel is also of interest because it is potentially apt to be used with highly enriched uranium coming from nuclear weapons

  6. Utilization of the Thai Research Reactor (TRR-1/M-1)

    International Nuclear Information System (INIS)

    The Thai Research Reactor type TRIGA Mark III has reached 18th core configuration in 2010 after more than thirty years of service since 1977. The recent hexagonal core comprises mixed 107 fuel elements of 8.5-20% U wt.% and five control rods and with maximum neutron flux of 3×1013 cm-2s-1 at 1.2 MW. Core calculation is carried out by Neutronics Computer Codes (3D Deterministic method SRAC and 3D Monte Carlo method MVP) and Thermal Hydraulics Codes (Steady state Calculation COOLODN2 and Reactivity Insertion Analysis EUREKA2/RR). There are 10 in-core tubes (CT, C8, C12, F3, F12, F22, F29, G5, G22 and G33) and 12 out-core tubes and facilities (A1, A4, CA2, CA3, TA, TC, wet tube, SNIF, Rotary specimen rack, void tank and beam tubes and thermal column). The reactor serves research and development on neutron activation analysis, neutron radiography, plant mutation, as well as services on medical radioisotope production (I-131, P-32 and Sm-153), gems quality enhancement and elemental analysis with the total income around 0.45 Million US$ in 2009. The reactor also serves education, training and technology transfer including university reactor experiments and technical tours for the public. (author)

  7. Internal capital market efficiency of Belgian holding companies

    OpenAIRE

    Gautier, Axel; Malika HAMADI

    2004-01-01

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

  8. Flow visualization techniques in nuclear reactors, (1)

    International Nuclear Information System (INIS)

    Heat energy generated in nuclear reactors is transferred by coolants to utilizing systems such as electric power and process industries etc. Therefore, heat removal characteristics of nuclear reactors depends on flow conditions of coolants in a reactor core and in cooling systems. In order to make clear flow patterns of these coolants, the flow visualization method is often applied prior to actual measurements of pressure, velocity and so on. This paper describes basic techniques for flow visualization especially in nuclear reactor, and gives applied examples of this technique. (author)

  9. Thai Research Reactor (TRR-1/M1) Neutron Beam Measurements

    International Nuclear Information System (INIS)

    Full text: Neutron beam tube of neutron radiography facility at Thai Research Reactor (TRR-1/M1) Thailand Institute of Nuclear Technology (public organization) is a divergent beam. The rectangular open-end of the beam tube is 16 cm x 17 cm while the inner-end is closed to the reactor core. The neutron beam size was measured using 20 cm x 40 cm neutron imaging plate. The measurement at the position 100 cm from the end of the collimator has shown that the beam size was 18.2 cm x 19.0 cm. Gamma ray in neutron the beam was also measured by the identical position using industrial X ray film. The area of gamma ray was 27.8 cm x 31.1 cm with the highest intensity found to be along the neutron beam circumference

  10. Problems of nuclear reactor safety. Vol. 1

    International Nuclear Information System (INIS)

    Proceedings of the 9. Topical Meeting 'Problems of nuclear reactor safety' are presented. Papers include results of studies and developments associated with methods of calculation and complex computerized simulation for stationary and transient processes in nuclear power plants. Main problems of reactor safety are discussed as well as rector accidents on operating NPP's are analyzed

  11. Nuclear reactors Monitoring using neutrinos detectors1

    International Nuclear Information System (INIS)

    We study the feasibility to use antineutrinos detectors for monitoring of nuclear reactors. Using a simple model of fission shower with two components, we illustrate how the numbers of antineutrinos detected at a distance L from the reactor depend on the composition of the nuclear combustible and how it could be used for nuclear safeguards policy.

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

    Science.gov (United States)

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

  13. Focal epilepsy in the Belgian shepherd

    DEFF Research Database (Denmark)

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

    2009-01-01

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

  14. Optical remote sensing of Belgian coastal waters

    OpenAIRE

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

    2001-01-01

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

  15. Advertising budgeting practices of Belgian industrial marketers.

    OpenAIRE

    François, Pierre

    2003-01-01

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

  16. RA reactor operation in 1991, Part 1

    International Nuclear Information System (INIS)

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1991, three major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. Renewal of the reactor instrumentation was started but but it is behind the schedule in 1991 because the delivery of components from USSR is late. Production of this instruments is financed by the IAEA according to the contract signed in December 1988 with Russian Atomenergoexport. According to this contract, it has been planned that the RA reactor instrumentation should be delivered to the Vinca Institute by the end of 1990. Since then any delivery of components to Yugoslavia was stopped because of the temporary embargo imposed by the IAEA for political reasons. In 1991 most of the existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Training of the existing personnel was done regularly, but lack of financial support prevented employment of new workers that would be needed for operation in shifts and regular maintenance

  17. Burnup analysis of the power reactor, 1

    International Nuclear Information System (INIS)

    Several years of endeavors has been devoted to development of three-dimensional nuclear-thermal-hydro-dynamic simulators and research by basing the progress on the merits and demerits of the variational method, the functional approximation method, etc. As the result, the three-dimensional nuclear-thermal-hydro-dynamic code FLORA has been prepared. It has the following features. (1) The executive time is one third -- half as much as that by the convensional programs. (2) Numerical error is small when neutron spectrum mismatches. (3) In the fuels in which the distributions of Gd2O3 and enrichments are localized axially in the reactor core, three-dimensional nuclear-thermal-hydro-dynamic calculations are possible. (4) The transport kernel can be obtained by the coarse mesh method and the functional approximation method. (5) Albedo can be calculated by the two-group diffusion theory. (6) Power distribution can be obtained in the case of partial control rods inserted in the core. The course taken to the preparation, the theoretical background and example calculations with FLORA are described. The present report can be also used as a manual. (auth.)

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

    Science.gov (United States)

    Huybrechts, Bart; Callebaut, Alfons

    2015-01-01

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

  19. Use of the VR-1 ''Vrabec'' training reactor

    International Nuclear Information System (INIS)

    An overview is presented of the extent and ways of using the VR-1 training reactor, which is operated by the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague. A list and the characteristics of 16 problems developed for teaching purposes is given, and the 14 faculties and 2 research institutes participating in the teaching activities are listed. The reactor is used in the education and training of nuclear scientists and engineers. The instrumentation, experimental, handling and operating tools, as well as documentation and texts relating to the reactor are described. The following examples of the teaching activities are included: a guided visit to the operating reactor site, reactor dynamics study and delayed neutron measurement, training course, and the basic criticality experiment. Nuclear safety aspects (hypothetical accidents, quality control and system qualification demonstration, safety culture) are stressed during the education. The reactor department is involved in international cooperation projects. (J.B.). 3 refs

  20. core calculations for ETRR-1 research reactor upgrading

    International Nuclear Information System (INIS)

    nuclear research reactors play an important role in supporting the nuclear energy program for most countries. research reactors are categorized according to the type of fuel, fuel enrichment, type of moderator and reflector, the power of the reactor and its application. most reactors initially operated at low power then an era began to up-rate the power by changing the fuel type, improving the thermal-hydraulic system performance and modifying the control system to comply with the new trends in research reactors and its applications. in this thesis, we carried out static calculation for the egyptian first research reactor ETRR-1 to evaluate its power upgrade possibility. firstly, we carried out cell calculation using WIMSD/4 code to study the variation of the infinite multiplication factor with the variation of fuel enrichment, lattice pitch and adding heavy water by increasing percentage to the ordinary water coolant

  1. Annual report on JEN-1 and JEN-2 Reactors

    International Nuclear Information System (INIS)

    In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

  2. Upgrading of the research reactors FRG-1 and FRG-2

    International Nuclear Information System (INIS)

    In 1972 for the research reactor FRG-2 we applied for a license to increase the power from 15 MW to 21 MW. During this procedure a public laying out of the safety report and an upgrading procedure for both research reactors - FRG-1 (5 MW) and FRG-2 - were required by the licensing authorities. After discussing the legal background for licensing procedures in the Federal Republic of Germany the upgrading for both research reactors is described. The present status and future licensing aspects for changes of our research reactors are discussed, too. (orig.)

  3. Data base for analysis of RA-1 reactor decommissioning

    International Nuclear Information System (INIS)

    Full text: The RA-1 'Enrico Fermi' reactor is located at the Constituyentes Atomic Centre near Buenos Aires city. It reached criticality for the first time on January 17th, 1958. During 45 years, major modifications were introduced. The Decommissioning of RA-1 is not foreseen in a near future, but nevertheless CNEA (legally responsible of Dismantling and Decommissioning of all relevant nuclear facilities in Argentina) has started Dismantling and Decommissioning planning activities. As a part of these activities, a historical information data base of the RA-1 reactor was performed. This report contains a set of global and specific data of the RA-1 Research Reactor prepared as a data base for a future decommissioning analysis. It describes the whole installation, and specially the core, fuel type, present configuration, shielding and the operation devices. An exhaustive listing of materials and components located in the Reactor building is given. It also reconstructs the reactor operation history based on the available information. (author)

  4. RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

    The RA nuclear reactor stopped operation after March 1979 campaign due to appearance of aluminium oxyhydrates deposits on the surface of fuel element claddings. Relevant decisions of the Sanitary inspection body of the Ministry of health and the Director General of the 'Boris Kidric' Institute of nuclear sciences, Vinca, banned further reactor operation until reasons caused aluminium oxyhydrates deposition are investigated and removed to enable regular reactor operation. Until the end of 1979 and during 1980, after a series of analyses and findings that caused cease of reactor operation, all the preparatory actions needed for restart were performed. Due to the fact that there is no emergency cooling system and no appropriate filtering system at the reactor, and according to the new regulations about start up of nuclear facilities, the Sanitary inspection body made a decision about temporary licence for reactor start-up meaning performance of the 'zero experiment' limiting the operating power to 1% of the nominal power. Accordingly the reactor was restarted on January 21 1981. Criticality was reached with the core made of 80% enriched fuel elements only. After the experiment was finished by the end of March a permission was demanded for operation at higher power levels at full power. Taking into account the state of the reactor components the operating licence was issued limiting the power to 2 MW until reconstruction of the ventilation system and construction of the emergency cooling system are fulfilled. Program of testing operation started on September 15 1981 increasing gradually the operating power. Thus the reactor was operated at 2 MW power for 15 days during November and December. The total production achieved in 1981 was 1698 MWh. This enabled isotopes production at the reactor during last two months. Control and maintenance of the reactor components and systems was done regularly and efficiently within limits imposed by availability of spare parts. The

  5. Investigation of neutron distribution in training reactor VR-1

    International Nuclear Information System (INIS)

    The VR-1 training reactor is a pool-type light-water reactor with the low-enriched uranium and maximum thermal power of 1 kW. The reactor is mainly used for students' training. The training is aimed to areas such as the reactor physics, neutronics, dosimetry, nuclear safety and I and C systems. Since neutron flux in the VR-1 core is well measured, this work focuses on one part of the reactor - its Radial experimental Channel (RC). This paper deals with the measurement of the neutron distribution by means of gold-foil neutron-activation technique and continual measurement with 3He-filled detector. Obtained experimental results were verified with the simulation in the Monte-Carlo N-Particle Transport Code. Results and conclusions from this paper will be used for further investigation of neutrons and their spatial distribution inside the low-power training reactor. Also, the data obtained in this paper can be used as a basis for future detailed measurements of neutron flux and its distribution in other hard accessible areas inside the reactor. The paper gives a simple theoretical introduction concerning neutron measurement procedures and available techniques in this field, which is particularly important for improving training courses and a content of offered experiments in the VR-1 reactor. (author)

  6. Korea Research Reactor -1 and 2 Decommissioning Project in Korea

    International Nuclear Information System (INIS)

    Korea Research Reactor 1 (KRR-1), the first research reactor in Korea, has been operated since 1962, and the second one, Korea Research Reactor 2 (KRR-2) since 1972. The operation of both of them was phased out in 1995 due to their lifetime and operation of the new and more powerful research reactor, HANARO (High-flux Advanced Neutron Application Reactor; 30MW). Both are TRIGA Pool type reactors in which the cores are small self-contained units sitting in tanks filled with cooling water. The KRR-1 is a TRIGA Mark II, which could operate at a level of up to 250 kW. The second one, the KRR-2 is a TRIGA Mark III, which could operate at a level of up 2,000 kW. The decontamination and decommissioning (D and D) project of these two research reactors, the first D and D project in Korea, was started in January 1997 and will be completed to stage 3 by 2008. The aim of this decommissioning program is to decommission the KRR-1 and 2 reactors and to decontaminate the residual building structure s and the site to release them as unrestricted areas. KAERI (Korea Atomic Energy Research Institute) submitted the decommissioning plan and the environmental impact assessment reports to the Ministry of Science and Technology (MOST) for the license in December 1998, and was approved in November 2000

  7. Irradiation routine in the IPR-R1 Triga reactor

    International Nuclear Information System (INIS)

    Information about irradiations in the IPR-R1 TRIGA reactor and procedures necessary for radioisotope solicitation are presented All procedures necessary for asking irradiation in the reactor, shielding types, norms of terrestrial and aerial expeditions, payment conditions, and catalogue of disposable isotopes with their respective saturation activities are described. (M.C.K.)

  8. Korea Research Reactor -1 & 2 Decommissioning Project in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. K.; Chung, U. S.; Jung, K. J.; Park, J. H.

    2003-02-24

    Korea Research Reactor 1 (KRR-1), the first research reactor in Korea, has been operated since 1962, and the second one, Korea Research Reactor 2 (KRR-2) since 1972. The operation of both of them was phased out in 1995 due to their lifetime and operation of the new and more powerful research reactor, HANARO (High-flux Advanced Neutron Application Reactor; 30MW). Both are TRIGA Pool type reactors in which the cores are small self-contained units sitting in tanks filled with cooling water. The KRR-1 is a TRIGA Mark II, which could operate at a level of up to 250 kW. The second one, the KRR-2 is a TRIGA Mark III, which could operate at a level of up 2,000 kW. The decontamination and decommissioning (D & D) project of these two research reactors, the first D & D project in Korea, was started in January 1997 and will be completed to stage 3 by 2008. The aim of this decommissioning program is to decommission the KRR-1 & 2 reactors and to decontaminate the residual building structure s and the site to release them as unrestricted areas. KAERI (Korea Atomic Energy Research Institute) submitted the decommissioning plan and the environmental impact assessment reports to the Ministry of Science and Technology (MOST) for the license in December 1998, and was approved in November 2000.

  9. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    International Nuclear Information System (INIS)

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results

  10. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    Energy Technology Data Exchange (ETDEWEB)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

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

    Science.gov (United States)

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

    2012-01-01

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

  12. The fast breeder reactor. v. 1

    International Nuclear Information System (INIS)

    The Energy Committee's report was prepared after hearing evidence (the minutes of which are published in Volume II) from the Central Electricity Generating Board, the United Kingdom Atomic Energy Authority and the Department of Energy. Memoranda received from other interested bodies or individuals were also considered and members of the Committee visited fast breeder projects in France, West Germany and Japan. As well as the development of the fast reactors, the economics and timescale were reviewed. The particular case of the fast breeder reactor and proposed fuel reprocessing plant at Dounreay was considered. The main conclusion is that major expenditure on fast reactor programmes can only be justified if there is a potential economic case, i.e. if the fuel cycle costs are lower than for PWRs. This would only be the case if uranium costs increased greatly. It is not considered worthwhile to participate in the European Fast Reactor although this should be reviewed in 1993 and 1997. The Committee agree with the Government's decision to cease funding the PFR in 1994 and endorses the need to regenerate the local economy which will be affected by this decision. (UK)

  13. Coolant technology of water cooled reactors. V. 1: Chemistry of primary coolant in water cooled reactors

    International Nuclear Information System (INIS)

    This report is a summary of the work performed within the framework of the Coordinated Research Programme on Investigations on Water Chemistry Control and Coolant Interaction with Fuel and Primary Circuit Materials in Water Cooled Power Reactors organized by the IAEA and carried out from 1987 to 1991. It is the continuation of a programme entitled Reactor Water Chemistry Relevant to Coolant-Cladding Interaction (IAEA-TECDOC-429), which ran from 1981 to 1986. Subsequent meetings resulted in the title of the programme being changed to Coolant Technology of Water Cooled Reactors. The results of this Coordinated Research Programme are published in four volumes with an overview in the Technical Reports Series. The titles of the volumes are: Volume 1: Chemistry of Primary Coolant in Water Cooled Reactors; Volume 2: Corrosion in the Primary Coolant Systems of Water Cooled Reactors; Volume 3: Activity Transport Mechanisms in Water Cooled Reactors; Volume 4: Decontamination of Water Cooled Reactors. These publications should be of interest to experts in water chemistry at nuclear power plants, experts in engineering, fuel designers, research and development institutes active in the field and to consultants to these organizations. Refs, figs and tabs

  14. Fusion reactor materials

    International Nuclear Information System (INIS)

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

  15. Belgian Photography: Towards a Minor Photography.

    OpenAIRE

    Jan Baetens; Hilde Van Gelder; Mieke Bleyen

    2011-01-01

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

  16. Monte Carlo modelling of VR-1 reactor core

    International Nuclear Information System (INIS)

    The possibilities of reactor core analysis by precise Monte Carlo codes are gradually increasing along with the accessibility of computing power. In the case of zero power research reactors, where temperature and burn-up effects remain negligible, model can approximate the reality to a very high degree. In such a case, most of calculation uncertainty can be caused by uncertainties in technical specifications of fuel and reactor internals. Thus performance of the modelling and its predictive power can be significantly improved via comparison with a large set of experimental data that can be acquired during reactor operation and via subtle tuning and improving the calculation model. The paper describes the case for neutronics calculations of VR-1 zero power reactor core. (author)

  17. Ehlers-Danlos Syndrome, Hypermobility Type, Is Linked to Chromosome 8p22-8p21.1 in an Extended Belgian Family

    Directory of Open Access Journals (Sweden)

    Delfien Syx

    2015-01-01

    Full Text Available Joint hypermobility is a common, mostly benign, finding in the general population. In a subset of individuals, however, it causes a range of clinical problems, mainly affecting the musculoskeletal system. Joint hypermobility often appears as a familial trait and is shared by several heritable connective tissue disorders, including the hypermobility subtype of the Ehlers-Danlos syndrome (EDS-HT or benign joint hypermobility syndrome (BJHS. These hereditary conditions provide unique models for the study of the genetic basis of joint hypermobility. Nevertheless, these studies are largely hampered by the great variability in clinical presentation and the often vague mode of inheritance in many families. Here, we performed a genome-wide linkage scan in a unique three-generation family with an autosomal dominant EDS-HT phenotype and identified a linkage interval on chromosome 8p22-8p21.1, with a maximum two-point LOD score of 4.73. Subsequent whole exome sequencing revealed the presence of a unique missense variant in the LZTS1 gene, located within the candidate region. Subsequent analysis of 230 EDS-HT/BJHS patients resulted in the identification of three additional rare variants. This is the first reported genome-wide linkage analysis in an EDS-HT family, thereby providing an opportunity to identify a new disease gene for this condition.

  18. Ehlers-Danlos Syndrome, Hypermobility Type, Is Linked to Chromosome 8p22-8p21.1 in an Extended Belgian Family.

    Science.gov (United States)

    Syx, Delfien; Symoens, Sofie; Steyaert, Wouter; De Paepe, Anne; Coucke, Paul J; Malfait, Fransiska

    2015-01-01

    Joint hypermobility is a common, mostly benign, finding in the general population. In a subset of individuals, however, it causes a range of clinical problems, mainly affecting the musculoskeletal system. Joint hypermobility often appears as a familial trait and is shared by several heritable connective tissue disorders, including the hypermobility subtype of the Ehlers-Danlos syndrome (EDS-HT) or benign joint hypermobility syndrome (BJHS). These hereditary conditions provide unique models for the study of the genetic basis of joint hypermobility. Nevertheless, these studies are largely hampered by the great variability in clinical presentation and the often vague mode of inheritance in many families. Here, we performed a genome-wide linkage scan in a unique three-generation family with an autosomal dominant EDS-HT phenotype and identified a linkage interval on chromosome 8p22-8p21.1, with a maximum two-point LOD score of 4.73. Subsequent whole exome sequencing revealed the presence of a unique missense variant in the LZTS1 gene, located within the candidate region. Subsequent analysis of 230 EDS-HT/BJHS patients resulted in the identification of three additional rare variants. This is the first reported genome-wide linkage analysis in an EDS-HT family, thereby providing an opportunity to identify a new disease gene for this condition. PMID:26504261

  19. Significant aspects concerning RC-1 TRIGA reactor operations

    International Nuclear Information System (INIS)

    In the present work, a brief critical survey is made of the various nuclear configurations with which the reactor operated after completion of the works directed at obtaining a power increase, from 0.1 to 1 MW, and the reactor utilizations are hinted. Furthermore, through experimental measurements of the neutron flux in significant positions, the volumetric integral of the flux in the core is solved by defining, in conservative conditions of nuclear configuration and of reactor operation, the peak factor, referred to the most stressed element of the TRIGA core. (author)

  20. Army Gas-Cooled Reactor Systems Program. Operation of ML-1 reactor skid in GCRE: safety evaluation report

    Energy Technology Data Exchange (ETDEWEB)

    None

    1964-10-01

    The operation of the ML-1 reactor skid in the modified GCRE facility, utilizing the GCRE reactor coolant circulating and heat removal systems, is described. An evaluation of the safety considerations associated with this mode of operation indicates that the consequences of the maximum credible accident are less severe than those previously approved for operation of the ML-1 reactor at the ML-1 test site or for operation of the GCRE-I reactor in the GCRE facility.

  1. Licensing of the first reload of Angra-1 reactor

    International Nuclear Information System (INIS)

    The historical aspects related to the licensing of the first reload of Angra-1 reactor are presented. The dates, the institutions, the experts, as well as the documents generated during that process are presented. (M.I.)

  2. Belgian NPPS fit passive autocatalytic recombiners

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  4. 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.

  5. Reactor Section standard analytical methods. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Sowden, D.

    1954-07-01

    the Standard Analytical Methods manual was prepared for the purpose of consolidating and standardizing all current analytical methods and procedures used in the Reactor Section for routine chemical analyses. All procedures are established in accordance with accepted practice and the general analytical methods specified by the Engineering Department. These procedures are specifically adapted to the requirements of the water treatment process and related operations. The methods included in this manual are organized alphabetically within the following five sections which correspond to the various phases of the analytical control program in which these analyses are to be used: water analyses, essential material analyses, cotton plug analyses boiler water analyses, and miscellaneous control analyses.

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

    International Nuclear Information System (INIS)

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

  7. Compact tokamak reactors. Part 1 (analytic results)

    International Nuclear Information System (INIS)

    We discuss the possible use of tokamaks for thermonuclear power plants, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First we review and summarize the existing literature. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamak power plant, by including the power required to drive the toroidal field, and considering two extremes of plasma current drive efficiency. The analytic results will be augmented by a numerical calculation which permits arbitrary plasma current drive efficiency; the results of which will be presented in Part II. Third, a scaling from any given reference reactor design to a copper toroidal field coil device is discussed. Throughout the paper the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculating electric power. We conclude that the latest published reactor studies, which show little advantage in using low aspect ratio unless remarkably high efficiency plasma current drive and low safety factor are combined, can be reproduced with the analytic model

  8. New human machine interface for VR-1 training reactor

    International Nuclear Information System (INIS)

    The contribution describes a new human machine interface that was installed at the VR-1 training reactor. The human machine interface update was completed in the summer 2001. The human machine interface enables to operate the training reactor. The interface was designed with respect to functional, ergonomic and aesthetic requirements. The interface is based on a personal computer equipped with two displays. One display enables alphanumeric communication between a reactor operator and the control and safety system of the nuclear reactor. Messages appear from the control system, the operator can write commands and send them there. The second display is a graphical one. It is possible to represent there the status of the reactor, principle parameters (as power, period), control rods' positions, the course of the reactor power. Furthermore, it is possible to set parameters, to show the active core configuration, to perform reactivity calculations, etc. The software for the new human machine interface was produced in the InTouch developing environment of the WonderWare Company. It is possible to switch the language of the interface between Czech and English because of many foreign students and visitors at the reactor. The former operator's desk was completely removed and superseded with a new one. Besides of the computer and the two displays, there are control buttons, indicators and individual numerical displays of instrumentation there. Utilised components guarantee high quality of the new equipment. Microcomputer based communication units with proper software were developed to connect the contemporary control and safety system with the personal computer of the human machine interface and the individual displays. New human machine interface at the VR-1 training reactor improves the safety and comfort of the reactor utilisation, facilitates experiments and training, and provides better support of foreign visitors.(author)

  9. 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.

  10. Education of students and experts and personnel preparation on training reactor VR-1

    International Nuclear Information System (INIS)

    The paper describes main purposes in the training reactor VR-1 operation. It also gives a typical schedule for one week training course in reactor physics and kinetics, which is organised at the VR-1 reactor

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

    Science.gov (United States)

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

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

  13. Department of Reactor Technology annual progress report 1 January -31 December 1977

    International Nuclear Information System (INIS)

    The work of the Department of Reactor Technology within the following fields is described: reactor engineering, reactor operation, structural reliability, system reliability, reactor physics, fuel management, reactor accident analysis for LOCA and ECC, containment analysis, experimental heat transfer, reactor core dynamics and power plant simulators, experimental activation measurements and neutron radiography at the DR 1 reactor, underground storage of gas, solar heating and underground heat storage, wind power. (author)

  14. Operation characteristics and conditions of training reactor VR-1

    International Nuclear Information System (INIS)

    The first 3 years of operation of the VR-1 training reactor are reviewed. This period includes its physical start-up (preparation, implementation, results) and operation development as far as the current operating configuration of the reactor core. The physical start-up was commenced using a reactor core referred to as AZ A1, whose physical parameters had been verified by calculation and whose configuration was based on data tested experimentally on the SR-0 reactor at Vochov. The next operating core, labelled AZ A2, was already prepared during the test operation of the VR-1 reactor. Its configuration was such that both of the main horizontal channels, radial and tangential, could be employed. The configuration that followed, AZ A3, was an intermediate step before testing the graphite side reflector. The current reactor core, labelled AZ A3 G, was obtained by supplementing the previous core with a one-sided graphite side reflector. (Z.S.). 2 tabs., 11 figs., 2 refs

  15. Refurbishment of Pakistan research reactor (PARR-1) for stainless steel lining of the reactor pool

    International Nuclear Information System (INIS)

    Pakistan Research Reactor-1 (PARR-1) is a pool-type research reactor. Reactor aging has resulted in the increase of water seepage from the concrete walls of the reactor pool. To stop the seepage, it was decided to augment the existing pool walls with an inner lining of stainless steel. This could be achieved only if the pool walls could be accessed unhindered and without excessive radiation doses. For this purpose a partial decommissioning was done by removing all active core components including standard/control fuel elements, reflector elements, beam tubes, thermal shield, core support structure, grid plate and the pool's ceramic tiles, etc. An overall decommissioning program was devised which included procedures specific to each item. This led to the development of a fuel transport cask for transportation, and an interim fuel storage bay for temporary storage of fuel elements (until final disposal). The safety of workers and the environment was ensured by the use of specially designed remote handling tools, appropriate shielding and pre-planned exposure reduction procedures based on the ALARA principle. During the implementation of this program, liquid and solid wastes generated were legally disposed of. It is felt that the experience gained during the refurbishment of PARR-1 to install the stainless steel liner will prove useful and better planning and execution for the future decommissioning of PARR-1, in particular, and for other research reactors like PARR-2 (27 kW MNSR), in general. Furthermore, due to the worldwide activities on decommissioning, especially those communicated through the IAEA CRP on 'Decommissioning Techniques for Research Reactors', the importance of early planning has been well recognized. This has made possible the implementation of some early steps like better record keeping, rehiring of trained manpower, and creation of interim and final waste storage. (author)

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

    International Nuclear Information System (INIS)

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

  17. Electrical system regulations of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

  18. Pebble Bed Reactor Plant screening evaluation. Volume 1. Overall plant and reactor system

    International Nuclear Information System (INIS)

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW/sub 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. Core scoping studies were performed which evaluated the effects of annular and cyclindrical 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

  19. Research program plan: reactor vessels. Volume 1

    International Nuclear Information System (INIS)

    The ability of the licensing staff of the NRC to make decisions concerning the present and continuing safety of nuclear reactor pressure vessels under both normal and abnormal operating conditions is dependent upon the existence of verified analysis methods and a solid background of applicable experimental data. It is the role of this program to provide both the analytical methods and the experimental data needed. Specifically, this program develops fracture mechanics analysis methods and design criteria for predicting the stress levels and flaw sizes required for crack initiation, propagation, and arrest in LWR pressure vessels under all known and postulated operations conditions. To do this, not only must the methods be developed but they must be experimentally validated. Further, the materials data necessary for input to these analytical methods must be developed. Thus, in addition to methods development and large scale experimental verification this program also develops data to show that slow-load fracture toughness, rapid-load fracture toughness, and crack arrest toughness obtained from small laboratory specimens are truly representative of the toughness characteristics of the material behavior in pressure vessels in both the unirradiated and the irradiated conditions

  20. FiR 1 reactor and the spent fuel management

    International Nuclear Information System (INIS)

    The FiR 1-reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose to run the reactor has been lately the Boron Neutron Capture Therapy (BNCT). According to the current operating license of our reactor we have to achieve a binding agreement between our Research Centre and either the domestic Nuclear Power Companies about the possibility to use the Olkiluoto final disposal facility for our spent fuel or USDOE about the return of our spent fuel back to USA. If we want to continue the reactor operation beyond the year 2006, the domestic final disposal is the only possibility. At the moment the domestic final disposal is the primary alternative, but it seems still to be reasonable to be prepared to both possibilities: the domestic final disposal and the return to the USA offered by USDOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will decide, which of the spent fuel policies will be obeyed. In less than two years' time it will be reviewed, if the results of the BNC treatments are satisfactory for the continuation of the treatments. If the BNCT and other irradiations develop satisfactorily and seem to have positive future, the reactor can be kept in operation beyond the year 2006 and the domestic final disposal will be implemented. If, however, the development of the BNCT will not develop positively and there will be lack of money, there is no reason to continue the operation of the reactor and the choice of USDOE alternative is natural. (author)

  1. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities

  2. RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

    Reactor test operation started in September 1981 at 2 MW power with 80% enriched fuel continued during 1982 according to the previous plan. The initial reactor core was made of 44 fuel channel each containing 10 fuel slugs. The first half of 1982 was used for the needed measurements and analysis of operating parameters and functioning of reactor systems and equipment under operating conditions. Program concerned with the testing operation at higher power levels was started in the second half of this year. It was found that the inherent excess reactivity and control rod worths ensure safe operation according to the IAEA safety standards. Excess reactivity is high enough to enable higher power level of 4.7 MW during 4 monthly cycles each lasting 15-20 days. Favourable conditions for cooling exist for the initial core configuration. Effects of poisoning at startup on the reactivity and power density distribution were measured as well as initial spatial distribution of the neutron flux which was 3,9 1013 cm-2 s-1 at 2 MW power. Modification of the calibration coefficient in the system for automated power level control was determined. All the results show that all the safety criteria and limitations concerned with fuel utilization are fulfilled if reactor power would be 4.7 MW. Additional testing operation at 3, 4, and 4.7 MW power levels will be needed after obtaining the licence for operating at nominal power. Transition from the initial core with 44 fuel channels to the equilibrium lattice configuration with 72 fuel channels each containing 10 fuel slugs, would be done gradually. Reactor was not operated in September because of the secondary coolant pipes were exchanged between Danube and the horizontal sedimentary. Control and maintenance of the reactor equipment was done regularly and efficiently dependent on the availability of the spare parts. Difficulties in maintenance of the reactor instrumentation were caused by unavailability of the outdated spare parts

  3. New reactor concepts

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Lafleur, Jean-Michel

    2011-01-01

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

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

    OpenAIRE

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

    2001-01-01

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

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

    OpenAIRE

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

  8. Thickness measurement of A-1 reactor caisson tube walls

    International Nuclear Information System (INIS)

    The equipment is described of measuring the thickness of caisson pipes built in the Bohunice A-1 reactor. The pulse-type ultrasonic thickness gauge is based on the reflection method using the double probe. The measurement accuracy is 0.1 mm. (J.B.)

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

    OpenAIRE

    Dobbelaere, Sabien

    2002-01-01

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

  10. Current status of the Thai Research Reactor (TRR-1/M1)

    International Nuclear Information System (INIS)

    The first Thai Research Reactor, TRR-1 went critical on 27 October 1962 at the maximum power of 1 MW. It was located at Office of Atoms for Peace (OAP) in Bangkok. Since then, TRR-1 was continuously operated and eventually shut down in 1975. Plate type, high-enriched uranium (HEU) and U3O8A1 cladding were used as the reactor fuel. Light water was used as moderator and coolant as well. In 1975, because of the problem from fuel supplier and also to supporting the Treaty of Non Proliferation of Nuclear Weapon or NPT, TRR-1 was shut down for modification. The reactor core and control system were disassembled and replaced by TRIGA Mark III. A new core was a hexagonal core shape designed by General Atomics (GA). Afterwards, TRR-1 was officially renamed to the Thai Research Reactor-1/Modification 1 (TRR-1/M1). TRR-1/M1 is a multipurpose swimming pool type reactor with nominal power of 2 MW. The TRR-1/M1 uses uranium enriched at 20% in U-235 (LEU) and ZrH alloy as fuel. Light water is also used as coolant and moderator. At present, the reactor is operating with core No.14. The reactor has been serving for various kinds of utilization namely, radioisotope production, neutron activation analysis, beam experiments and reactor physics experiments. (author)

  11. Reactor physics recommissioning of Pickering NGS Units 1 and 2

    International Nuclear Information System (INIS)

    Investigations following the rupture of Pickering Unit 1 pressure tube G16 in 1983, led to the shutdown of Units 1 and 2 for pressure tube replacement and numerous other upgrades. They were recommissioned in 1987 and 1988 respectively. This paper surveys the procedures used during the reactor physics recommissioning of these two reactors and presents the results of these measurements. Special note is made of the differences between this recommissioning work, and the initial commissioning of new CANDU reactors. From a physics point of view, the restarted units differed substantially from the original design. The main difference in the core configuration involved the conversion of 10 of the original adjuster rods into shutoff rods. The reactivities of the remaining adjusters were increased. These substantial changes to the core, together with the full core of fresh fuel, necessitated a complete set of reactor physics recommissioning experiments. Some of our procedures differed from those used to commission a new reactor. This was due mainly to the high levels of tritium in the moderator D2O and to radiological hazards on the reactivity deck. Also, the high residual activities in the rebuilt cores lead to increased difficulties in neutron monitoring and higher subcritical neutron count rates (hence a higher than usual reactor power at first criticality). In general the results of our recommissioning measurements closely matched the results of presimulations using the OHRFSP and SMOKIN computer codes. Results for Unit 2 were generally better than those for Unit 1. This was due to improved procedures which resulted from our experiences with Unit 1. (author). 4 tabs., 9 figs

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. Distribution of doses resulting from cosmic rays exposure for Belgian airlines

    International Nuclear Information System (INIS)

    The Belgian Radiation Protection Act of 2001 requires air line companies registered in Belgium to evaluate the doses resulting from the exposure of their crew to cosmic radiation. If the annual dose of 1 mSv is exceeded, the crew must be informed about their individual doses and pregnant women have to be protected, among other measures. The Federal Agency for Nuclear Control (FANC, the competent Belgian radiation protection authority) has issued guidelines in order to help air line companies to fulfil their duties. Following the publication of these guidelines, all commercial Belgian air line companies have sent data on the exposure of their personnel. These data and, in particular, the distribution of the doses are presented. Except in the cases of small 'air taxi' companies, which fly only on very short distances and low altitudes, a significant number of air crew members gets doses of more than 1 mSv/y. The maximum value amounts to ∼ 4 mSv/y. The computer codes used by the companies in order to evaluate the individual doses are PCAIRE, CARI and IASON-FREE. The FANC imposed the concerned companies to reassess yearly the individual doses. (author)

  15. Core burnup characteristics of high conversion light water reactor, (1)

    International Nuclear Information System (INIS)

    In order to evaluate core burnup characteristics of a high conversion light water reactor (HCLWR) with tight pitched lattice, core burnup calculation was made using two dimensional diffusion method. The volume ratio of moderator to fuel is about 0.8 in the reactor (HCLWR-J1) under study. The burnup calculations were carried out under the assumption of three batch and out-in fuel loading from the first cycle to the equilibrium cycle. A detailed evaluation was made for discharge burnup, conversion ratio, power distribution, and reactivity coefficients and so on. (author)

  16. Life extension of the St. Lucie unit 1 reactor vessel

    International Nuclear Information System (INIS)

    In late 1989, Florida Power and Light Company (FP and L) established the policy that St. Lucie unit 1 should not be prevented from achieving a 60-yr operating life by reactor vessel embrittlement. A 60-yr operating life means that the plant would be allowed to operate until the year 2036, which is 20 years beyond the current license expiration date of 2016. Since modifications to the reactor vessel and its components are projected to be expensive, the desire of FP and L management was to achieve this lifetime extension through the use of fuel management and proven technology. The following limitations were placed on any acceptable method for achieving this lifetime extension capability: low fuel cycle cost; low impact on safety parameters; very little or no operations impact; and use of normal reactor materials. A task team was formed along with the Advanced Nuclear Fuels Company (ANF) to develop a vessel-life extension program

  17. Research performed at the ETRR-1 reactor using TOF technique

    International Nuclear Information System (INIS)

    This paper represents the results of studies of neutron transmission at several single crystals, performed at ETRR-1 reactor horizontal channels. The results of these studies starting in 1984 and continuing to date are discussed; the use of large single crystals as a band pass filter is also assessed

  18. ATMEA and medium power reactors. The ATMEA joint venture and the ATMEA1 medium power reactor

    International Nuclear Information System (INIS)

    This Power Point presentation presents the ATMEA company (a joint venture of Areva and Mitsubishi), the main features of its medium power reactor (ATMEA1) and its building arrangement, indicates the general safety objectives. It outlines the features of its robust design which aim at protecting, cooling down and containing. It indicates the regulatory and safety frameworks, comments the review of the safety options by the ASN and the results of this assessment

  19. Predicting the environmental risks of radioactive discharges from Belgian nuclear power plants

    International Nuclear Information System (INIS)

    An environmental risk assessment (ERA) was performed to evaluate the impact on non-human biota from liquid and atmospheric radioactive discharges by the Belgian Nuclear Power Plants (NPP) of Doel and Tihange. For both sites, characterisation of the source term and wildlife population around the NPPs was provided, whereupon the selection of reference organisms and the general approach taken for the environmental risk assessment was established. A deterministic risk assessment for aquatic and terrestrial ecosystems was performed using the ERICA assessment tool and applying the ERICA screening value of 10 μGy h−1. The study was performed for the radioactive discharge limits and for the actual releases (maxima and averages over the period 1999–2008 or 2000–2009). It is concluded that the current discharge limits for the Belgian NPPs considered do not result in significant risks to the aquatic and terrestrial environment and that the actual discharges, which are a fraction of the release limits, are unlikely to harm the environment. -- Highlights: • Impact of radioactive discharges by the Belgian NPPs of Doel and Tihange on wildlife was evaluated. • Deterministic risk assessment for aquatic and terrestrial ecosystems performed with the ERICA tool. • NPP discharge limits do not result in significant risks to the aquatic and terrestrial environment. • Actual discharges, a fraction of the release limits, are unlikely to harm the environment

  20. Belgian canine population and purebred study for forensics by improved mitochondrial DNA sequencing.

    Science.gov (United States)

    Desmyter, Stijn; Gijsbers, Leonie

    2012-01-01

    In canine population studies for forensics, the mitochondrial DNA is profiled by sequencing the two hyper variable regions, HV1 and HV2 of the control region. In a first effort to create a Belgian population database some samples showed partially poor sequence quality. We demonstrated that a nuclear pseudogene was co-amplified with the mtDNA control region. Using a new combination of primers this adverse result was no longer observed and sequencing quality was improved. All former samples with poor sequence data were reanalyzed. Furthermore, the forensic canine population study was extended to 208 breed and mixed dogs. In total, 58 haplotypes were identified, resulting in an exclusion capacity of 0.92. The profile distribution of the Belgian population sample was not significantly different from those observed in population studies of three other countries. In addition to the total population study 107 Belgian registered pedigree dogs of six breeds were profiled. Per breed, the obtained haplotypes were supplemented with those from population and purebred studies. The combined data revealed that some haplotypes were more or less prominent present in particular dog breeds. The statistically significant differences in haplotype distribution between breeds and population sample can have consequences on mtDNA databasing and matching probabilities in forensics. PMID:21489897

  1. Mechanical degradation processes: The Belgian experience

    International Nuclear Information System (INIS)

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

  2. RA reactor operation and maintenance in 1992, Part 1

    International Nuclear Information System (INIS)

    During 1992 Ra reactor was not in operation. All the activities were fulfilled according to the previously adopted plan. Basic activities were concerned with revitalisation of the RA reactor and maintenance of reactor components. All the reactor personnel was busy with reconstruction and renewal of the existing reactor systems and building of the new systems, maintenance of the reactor devices. Part of the staff was trained for relevant tasks and maintenance of reactor systems

  3. Thermal and hydraulic characteristics of the JEN-1 Reactor; Caracteristicas hidraulicas y termicas del Reactor JEN-1

    Energy Technology Data Exchange (ETDEWEB)

    Otra Otra, F.; Leira Rey, G.

    1971-07-01

    In this report an analysis is made of the thermal and hydraulic performances of the JEN-1 reactor operating steadily at 3 Mw of thermal power. The analysis is made separately for the core, main heat exchanger and cooling tower. A portion of the report is devoted to predict the performances of these three main components when and if the reactor was going to operate at a power higher than the maximum 3 Mw attainable today. Finally an study is made of the unsteady operation of the reactor, focusing the attention towards the pumping characteristics and the temperatures obtained in the fuel elements. Reference is made to several digital calculation programmes that nave been developed for such purpose. (Author) 21 refs.

  4. Analysis of reactivity induced accidents at Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    Analysis of reactivity induced accidents in Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel, has been carried out using standard computer code PARET. The present core comprises of 29 standard and five control fuel elements. Various modes of reactivity insertions have been considered. The events studied include: start-up accident; accidental drop of a fuel element on the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results reveal that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is concluded that the reactor, which is operated safely at a steady-state power level of 10 MW, with coolant flow rate of 950 m3/h, will also be safe against any possible reactivity induced accident and will not result in a fuel failure

  5. Analysis of reactivity induced accidents at Pakistan Research Reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Bokhari, I.H. E-mail: ishtiaq@pinstech.org.pk; Israr, M.; Pervez, S

    2002-12-01

    Analysis of reactivity induced accidents in Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel, has been carried out using standard computer code PARET. The present core comprises of 29 standard and five control fuel elements. Various modes of reactivity insertions have been considered. The events studied include: start-up accident; accidental drop of a fuel element on the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results reveal that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is concluded that the reactor, which is operated safely at a steady-state power level of 10 MW, with coolant flow rate of 950 m{sup 3}/h, will also be safe against any possible reactivity induced accident and will not result in a fuel failure.

  6. Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m3/h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

  7. An Ageing Management Programme for the SAFARI-1 Research Reactor

    International Nuclear Information System (INIS)

    The SAFARI-1 Research Reactor is a 20 MW high flux material test reactor and has been continuously operational for nearly 47 years. In this period, ageing of the facility has been addressed by means of various, mainly reactive, maintenance and upgrade initiatives to replace components that have become unmaintainable for various reasons such as wear, corrosion and obsolescence. With the facility now approaching 50 years of continuous operation, a programme has been implemented to assess, address and implement ageing management in a more formal and proactive manner. The programme conforms to the recently published IAEA Safety Guide SSG-10 Ageing Management for Research Reactors and makes extensive use of the guidelines set therein as well as other tools and methods developed in various international meetings and workshops at the IAEA for identifying ageing issues at the facility. The paper presents an outline of the methodology for implementing the ageing management programme at the SAFARI-1 research reactor. Identification of SSCs important to the safety and sustainability of the facility that are susceptible to ageing, the ageing mechanisms affecting them and the remedial actions identified to mitigate or remove the effects of ageing are discussed. A methodology for determining priorities is also elaborated. Remedial actions are divided into four groups: safety critical, mission critical, lifetime extension and organizational, and an implementation strategy is described. (author)

  8. Linear systems stability analysis of Ghana Research Reactor 1

    International Nuclear Information System (INIS)

    The thesis describes the sequential analytical and experimental investigations of reactor systems stability analysis related to kinetics and control of Ghana Research Reactor 1 (GHARR-1). First, the two-point, fifteen-group reactor point-kinetics equations and redrop method are used to derive the theoretical and experimental zero-power transfer function and frequency responses. A computer code is developed to compute theoretical and experimental gain and phase response values and plots for a selected frequency range. Next, the thermal feedback transfer function and frequency response are deduce by solving the coupled thermal-hydraulics equations. Reactivity feedback dynamics yield a feedback control law from which thermal feedback parameters are obtained. With the aid of computer codes, the frequency response values for the gain and phase and their plots are obtained for the selected frequency range. Finally, the closed-loop response is obtained for the frequency and time domains by the superposition of the zero-power and thermal feedback frequency response functions. Frequency and time domain specifications are obtained and then used to define the overall performance of the system. A computer code is developed to compute the gain and phase of the frequency and time response for selected frequency values. Comparison of the values obtained from this work compares very well with that obtained from the safety analysis report and other related work performed on the reactor. (author)

  9. Current activities at the FiR 1 TRIGA reactor

    International Nuclear Information System (INIS)

    The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The epithermal neutrons needed for the irradiation of brain tumor patients are produced from the fast fission neutrons by a moderator block consisting of Al+AlF3 (FLUENTAL), which showed to be the optimum material for this purpose. Twenty-one patients have been treated since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization. The treatment organization has a close connection to the Helsinki University Central Hospital. The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. In the near future the back end solutions of the spent fuel management will have a very important role in our activities. The Finnish Parliament ratified in May 2001 the Decision in Principle on the final disposal facility for spent fuel in Olkiluoto, on the western coast of Finland. There is a special condition in our operating license. We have now about two years' time to achieve a binding agreement between VTT and the Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel. If this will not happen, we have to make the agreement with the USDOE with the well-known time limits. At the moment it seems to be reasonable to prepare for both spent fuel management possibilities: the domestic final disposal and the return to the USA offered by USDOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will determine, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNC treatments will cover

  10. Reactor Engineering Department annual report (April 1, 1991-March 31, 1992)

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1991 (April 1, 1991-March 31, 1992). The major Department's programs promoted in the year are assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researchers on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative work to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  11. Reactor Engineering Department annual report (April 1, 1987 - March 31, 1988)

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1987 (April 1, 1987 - March 31, 1988). The major activities in the Department concerns the programs of the high temperature gas-cooled reactor, the high conversion light water reactor, the advanced fission reactor system and the fusion reactor at JAERI and the fast breeder reactor at PNC. The report contains the latest progress in nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control/diagnosis and robotics, as well as the new topics from this fiscal year on advanced reactors system design studies and technique developments related the facilities in the Department. Also described are the activities of the Research Committee on Reactor Physics. (author)

  12. Reactor Engineering Department annual report (April 1, 1988 - March 31, 1989)

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1988 (April 1, 1988 - March 31, 1989). The Department has promoted cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and also to PNC's fast reactor project. Other major Department's programs are the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Application of a high energy accelerator to the nuclear engineering is also preliminarily assessed. The report also contains the latest progress in various basic researches as nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/ diagnosis and technical developments related to the reactor physics facilities. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  13. Reactor Engineering Department annual report (April 1, 1990 - March 31, 1991)

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1990 (April 1, 1990 - March 31, 1991). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  14. The radiological impact of the Belgian phosphate industry

    International Nuclear Information System (INIS)

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

  15. Reactor core conversion studies of Ghana: Research Reactor-1 and proposal for addition of safety rod

    International Nuclear Information System (INIS)

    The inclusion of an additional safety rod in conjunction with a core conversion study of Ghana Research Reactor-1 (GHARR-1) was carried out using neutronics, thermal hydraulics and burnup codes. The study is based on a recommendation by Integrated Safety Assessment for Research Reactors (INSARP) mission to incorporate a safety rod to the reactor safety system as well as the need to replace the reactor fuel with LEU. Conversion from one fuel type to another requires a complete re-evaluation of the safety analysis. Changes to the reactivity worth, shutdown margin, power density and material properties must be taken into account, and appropriate modifications made. Neutronics analysis including burnup was studied followed by thermal hydraulics analyses which comprise steady state and transients. Four computer codes were used for the analysis; MCNP, REBUS, PLTEP and PARET. The neutronics analysis revealed that the LEU core must be operated at 34 Kw in order to attain the flux of 1.0E12 n/cm2.s as the nominal flux of the HEU core. The auxiliary safety rod placed at a modified irradiation site gives a better worth than the cadmium capsules. For core excess reactivity of 4 mk, 348 fuel pins would be appropriate for the GHARR-1 LEU core. Results indicate that flux level of 1.0E12 n/cm2.s in the inner irradiation channel will not be compromised, if the power of the LEU core is increased to 34 kW. The GHARR-1 core using LEU-U02-12.5% fuel can be operated for 23 shim cycles, with cycles length 2.5 years, for over 57 years at the 17 kW power level. All 23 LEU cycles meet the ∼ 4.0 mk excess reactivity required at the beginning of cycle . For comparison, the MNSR HEU reference core can also be operated for 23 shim cycles, but with a cycle length of 2.0 years for just over 46 years at 15.0kW power level. It is observed that the GHARR-1 core with LEU UO2 fuel enriched to 12.5% and a power level of 34 kW can be operated ∼25% longer than the current HEU core operated at 30 k

  16. Role of research reactors in training of NPP personnel with special focus on training reactor VR-1

    International Nuclear Information System (INIS)

    Research reactors play an important role in providing key personnel of nuclear power plants a hands-on experience from operation and experiments at nuclear facilities. Training of NPP (Nuclear Power Plant) staff is usually deeply theoretical with an extensive utilisation of simulators and computer visualisation. But a direct sensing of the reactor response to various actions can only improve the personnel awareness of important aspects of reactor operation. Training Reactor VR-1 and its utilization for training of NPP operators and other professionals from Czech Republic and Slovakia is described. Typical experimental exercises and good practices in organization of a training program are demonstrated. (authors)

  17. Reactor Core Internals Replacement of Ikata Units 1 and 2

    International Nuclear Information System (INIS)

    Ikata Units 1 and 2 have been in operation for a very long time. Unit 1, in particular, is one of the longest operating PWRs in Japan. In view of this history, preventive and proactive strategy has been adopted for the maintenance of major primary system components. Both units successfully completed the replacement of steam generators and reactor vessel heads approximately ten years ago. With regard to the reactor core internals, baffle former bolts (BFBs) were found to have been damaged by stress corrosion cracking (SCC) in 1989 at Bugey 2 in France. Since then, similar incidents have been reported in other European and U.S. plants, resulting in the replacement of failed BFBs. The BFB issue can be dealt with either by replacing bolts when damage is found or by replacing the entire core internals with those of a new design. Ikata Units 1 and 2 chose the latter and carried it out in 2004 and 2005, respectively.

  18. Equipment for neutron measurements at VR-1 Sparrow training reactor

    International Nuclear Information System (INIS)

    Full text: The VR-1 Sparrow training reactor is the experimental nuclear facility especially employed for education and teaching of students from different technical universities in the Czech Republic and other countries. Since 2005 the uniform all-purpose devices EMK310 have been used for measurement at reactor laboratory with different type of gas filled neutron detectors. The neutron detection system are employed for reactivity measurement, control rod calibration, critical experiment, study of delayed neutrons, study of nuclear reactor dynamics and study of detection systems dead time. The small dimension isotropic detectors are especially used for measurement of thermal neutron flux distribution inside the reactor core. The EMK-310 is a high performance, portable, three-channel fast amplitude analyzer designed for counting applications. It was developed for nuclear applications and made in close co-operation with firm TEMA Ltd. The precise rack eliminates electromagnetic disturbance and contains the control unit and four modules. The modules of high voltage supply and amplifier for gas filled detectors or scintillation probes are used in basic configuration. Software is tailored specifically to the reactor measurement and allows full online control. For applications involving the study of signals that may vary with the time, example study of delayed neutrons or nuclear reactor dynamics, the EMK-310 provides a Multichannel Scaling (MCS) acquisition mode. MCS dwell time can be set from 2 ms. Now, the new generation of digital multichannel analyzers DA310 is introduced. They have similarly attributes as EMK310 but the output information of unipolar signals from detector is more complete. The pipeline A/D converter with field programmable gate array (FPGA) is the hearth of the DA310 device. The resolution is 12 bits (4096 channels); the sample frequency is 80 MHz. The application for the neutron noise analysis is supposed. The correction method for non linearity

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

    NARCIS (Netherlands)

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

    2014-01-01

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

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

    Science.gov (United States)

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

    2011-01-01

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

  1. Inventory of nuclear liabilities - The Belgian perspective

    International Nuclear Information System (INIS)

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

  2. Refurbishment and power upgrade of Pakistan Research Reactor-1 (PARR-1)

    International Nuclear Information System (INIS)

    The Pakistan Research Reactor-1 (PARR-1) was commissioned in 1965 at a power level of 5 MW. The reactor was originally designed for HEU fuel with 93% enrichment in the form of UAlx-Al. The reactor equipment remained unchanged for the first 20 years of its operation. However, several factors, such as obsolescence of equipment and the non-availability of spare parts, forced a refurbishment programme of the facility. For this purpose, the old thermionic tube-based instrumentation and control system was replaced with new and modern instrumentation, the reactor pool was lined with stainless steel, and the old cooling system was removed and a new enhanced cooling system was installed for a power upgrade. In order to increase reactor safety, a new emergency core cooling system was installed, the reactor building was repaired, the HVAC system was improved, and a new compressor and additional diesel generators were installed. In order to meet international requirements, the reactor core was converted from HEU to LEU fuel. Due to changing experimental needs and demand for a higher neutron flux, reactor power was increased from 5 MW to 10 MW. The refurbished and upgraded PARR-1 achieved first criticality with LEU fuel on 31 October 1991 and full power of 9 MW on 7 May 1992. (author)

  3. Shielding assessment of the ETRR-1 Reactor Under power upgrading

    International Nuclear Information System (INIS)

    The assessment of existing shielding of the ETRR-1 reactor in case of power upgrading is presented and discussed. It was carried out using both the present EK-10 type fuel elements and some other types of fuel elements with different enrichments. The shielding requirements for the ETRR-1 when power is upgraded are also discussed. The optimization curves between the upgraded reactor power and the shield thickness are presented. The calculation have been made using the ANISN code with the DLC-75 data library. The results showed that the present shield necessitates an additional layer of steel with thickness of 10.20 and 25 cm. When its power is upgraded to 3, 6 and 10 MWt in order to cutoff all neutron energy groups to be adequately safe under normal operating conditions. 4 figs

  4. Fuel-failure detection system for Pakistan Research Reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Ayazuddin, S.K.; Hayat, Tariq; Qureshi, A.A.; Khan, H.A. [Pakistan Inst. of Nuclear Science and Technology, Nuclear Engineering Div., Islamabad (Pakistan)

    1997-12-01

    After the conversion and upgrading of Pakistan Research Reactor-1 (PARR-1), it was decided to install a fuel-failure detection system to confirm the performance and integrity of the new fuel elements. The fuel-failure detection is based on monitoring of delayed neutrons emitted from fission products leaking into the primary coolant loop from the fuel. For this purpose, two neutron detectors (BF{sub 3}) were replaced in the graphite moderator blocks that were installed at the outlet coolant pipe in the valve pit. The fuel-failure detection system was tested and calibrated at a miniature neutron source reactor (PARR-2) which provided the basis for alarm limits setting in the event of fuel failure. (author).

  5. Fuel-failure detection system for Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    After the conversion and upgrading of Pakistan Research Reactor-1 (PARR-1), it was decided to install a fuel-failure detection system to confirm the performance and integrity of the new fuel elements. The fuel-failure detection is based on monitoring of delayed neutrons emitted from fission products leaking into the primary coolant loop from the fuel. For this purpose, two neutron detectors (BF3) were replaced in the graphite moderator blocks that were installed at the outlet coolant pipe in the valve pit. The fuel-failure detection system was tested and calibrated at a miniature neutron source reactor (PARR-2) which provided the basis for alarm limits setting in the event of fuel failure. (author)

  6. Twenty-five years of the VR-1 reactor operation for nuclear education in Czech Republic

    International Nuclear Information System (INIS)

    The VR-1 reactor, operated by the Faculty of Nuclear Sciences and Physical Engineering, the Czech Technical University in Prague, was launched in 1989 and attained criticality in 1990. The history of the reactor is highlighted, the design is described in great detail with focus on the reactor vessels, fuel and reactor core, reactor control system, and equipment for experiments, and the uses of the reactor are outlined. The reactor is a key educational and training facility for students and experts from nuclear industry, power engineering and research. (orig.)

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

    International Nuclear Information System (INIS)

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

  8. 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.

  9. Coastal flooding risk calculations for the Belgian coast

    OpenAIRE

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

    2009-01-01

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

  10. RA reactor operation and maintenance in 1999, Part 1

    International Nuclear Information System (INIS)

    Activities at the RA reactor in 1999 were defined according to the needs of maintaining the reactor components and systems according to the existing funding. Basic activities during the past year were related to the maintenance of the reactor devices which must be in constant operation (special and regular ventilation power supply system, radioactivity and contamination control system, internal transportation system), reactor security system, and other systems that are useful independent of the future status of the reactor. (secondary cooling system, hot cells). maintenance of the reactor building was done on a limited scale due to lack of financial support. Possible solutions for the future status of the RA reactor discussed in this report are: renewal of reactor components for the reactor restart, conservation of the reactor (temporary shutdown) or permanent reactor shutdown

  11. RA reactor operation and maintenance in 1998, Part 1

    International Nuclear Information System (INIS)

    Activities at the RA reactor in 1998 were defined according to the needs of maintaining the reactor components and systems according to the existing funding. Basic activities during the past year were related to the maintenance of the reactor devices which must be in constant operation (special and regular ventilation power supply system, radioactivity and contamination control system, internal transportation system), reactor security system, and other systems that are useful independent of the future status of the reactor. (secondary cooling system, hot cells). maintenance of the reactor building was done on a limited scale due to lack of financial support. Possible solutions for the future status of the RA reactor discussed in this report are: renewal of reactor components for the reactor restart, conservation of the reactor (temporary shutdown) or permanent reactor shutdown

  12. Graphite stack corrosion of BUGEY-1 reactor (synthesis)

    International Nuclear Information System (INIS)

    The definitive shutdown date for the BUGEY-1 reactor was May 27th, 1994, after 12.18 full power equivalent years and this document briefly describes some of the feedback of experience from operation of this reactor. The radiolytic corrosion of graphite stack is the major problem for BUGEY-1 reactor, despite the inhibition of the reaction by small quantities of CH4 added to the coolant gas. The mechanical behaviour of the pile is predicted using the ''INCA'' code (stress calculation), which uses the results of graphite weight loss variation determined using the ''USURE'' code. The weight loss of graphite is determined by annually taking core samples from the channel walls. The results of the last test programme undertaken after the definitive shutdown of BUGEY-1 have enabled an experimental graph to be established showing the evolution of the compression resistance (perpendicular and parallel direction to the extrusion axis) as a function of the weight loss. The numerous analyses, made on the samples carried out in the most sensitive regions, have allowed to verify that no brutal degradation of the mechanical properties of graphite happens for the high value of weight loss up to 40% (maximum weight loss reached locally). (author). 10 refs, 3 figs, 4 tabs

  13. Welding of the A1 reactor pressure vessel

    International Nuclear Information System (INIS)

    As concerns welding, the A-1 reactor pressure vessel represents a geometrically complex unit containing 1492 welded joints. The length of welded sections varies between 10 and 620 mm. At an operating temperature of 120 degC and a pressure of 650 N/cm2 the welded joints in the reactor core are exposed to an integral dose of 3x1018 n/cm2. The chemical composition is shown for pressure vessel steel as specified by CSN 413090.9 modified by Ni, Ti and Al additions, and for the welding electrodes used. The requirements are also shown for the mechanical properties of the base and the weld metals. The technique and conditions of welding are described. No defects were found in ultrasonic testing of welded joints. (J.B.)

  14. Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 1: Reactor Design and Model Development

    KAUST Repository

    Zhao, Zhenlong

    2013-01-17

    Chemical-looping combustion (CLC) is a novel and promising technology for power generation with inherent CO2 capture. Currently, almost all of the research has been focused on developing CLC-based interconnected fluidized-bed reactors. In this two-part series, a new rotary reactor concept for gas-fueled CLC is proposed and analyzed. In part 1, the detailed configuration of the rotary reactor is described. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet and exit. Two purging sectors are used to avoid the mixing between the fuel stream and the air stream. The rotary wheel consists of a large number of channels with copper oxide coated on the inner surface of the channels. The support material is boron nitride, which has high specific heat and thermal conductivity. Gas flows through the reactor at elevated pressure, and it is heated to a high temperature by fuel combustion. Typical design parameters for a thermal capacity of 1 MW have been proposed, and a simplified model is developed to predict the performances of the reactor. The potential drawbacks of the rotary reactor are also discussed. © 2012 American Chemical Society.

  15. IPR-R1 TRIGA research reactor decommissioning plan

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) is concerning to establish or adopt standards of safety for the protection of health, life and property in the development and application of nuclear energy for peaceful purposes. In this way the IAEA recommends that decommissioning planning should be part of all radioactive installation licensing process. There are over 200 research reactors that have either not operated for a considerable period of time and may never return to operation or, are close to permanent shutdown. Many countries do not have a decommissioning policy, and like Brazil not all installations have their decommissioning plan as part of the licensing documentation. Brazil is signatory of Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management, but until now there is no decommissioning policy, and specifically for research reactor there is no decommissioning guidelines in the standards. The Nuclear Technology Development Centre (CDTN/CNEN) has a TRIGA Mark I Research Reactor IPR-R1 in operation for 47 years with 3.6% average fuel burn-up. The original power was 100 k W and it is being licensed for 250 k W, and it needs the decommissioning plan as part of the licensing requirements. In the paper it is presented the basis of decommissioning plan, an overview and the end state / final goal of decommissioning activities for the IPR-R1, and the Brazilian ongoing activities about this subject. (author)

  16. IPR-R1 TRIGA research reactor decommissioning plan

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) is concerning to establish or adopt standards of safety for the protection of health, life and property in the development and application of nuclear energy for peaceful purposes. In this way the IAEA recommends that decommissioning planning should be part of all radioactive installation licensing process. There are over 200 research reactors that have either not operated for a considerable period of time and may never return to operation or, are close to permanent shutdown. Many countries do not have a decommissioning policy, and like Brazil not all installations have their decommissioning plan as part of the licensing documentation. Brazil is signatory of Joint Convention on the safety of Spent Fuel Management and on the Safety of Radioactive Waste Management, but until now there is no decommissioning policy, and specifically for research reactor there is no decommissioning guidelines in the standards. The Nuclear Technology Development Centre (CDTN/CNEN) has a TRIGA Mark I Research Reactor IPR-R1 in operation for 47 years with 3.6% average fuel burn-up. The original power was 100 kW and it is being licensed for 250 kW, and it needs the decommissioning plan as part of the licensing requirements. In the paper it is presented the basis of decommissioning plan, an overview and the end state / final goal of decommissioning activities for the IPR-R1, and the Brazilian ongoing activities about this subject. (author)

  17. Monitoring of primary circuit and reactor of NPP A-1

    International Nuclear Information System (INIS)

    Nuclear Power Plant A-1 in Jaslovske Bohunice was commissioned in 1972. Heavy water moderated, carbon dioxide cooled channel type reactor was shut down after two accidents in 1977. During more serious second accident, the reduced coolant flow caused local overheating of the fuel and consequent damage/melting of the fuel channel. Both accidents had led to the damage of several fuel assemblies with extensive local damage of fuel claddings. As a consequence, the main cooling circuit was significantly contaminated by fission products and long-life alpha nuclides. The detailed monitoring of dose rates, smearable contamination and sampling of contamination was performed. Extended monitoring in reacto vessel, primary circuit pipes, turbo-compressors, steam generators, main valves, gas tanks and also heavy water system with collectors, coolers, distilling and purification station, pumps and valves was done. Appropriate devices and procedures for the monitoring and examination of the installations were prepared and applied. Obtained results will serve for the future planning of the decontamination and decommissioning works. The 3-D model of the reactor that had been developed as part of this Project proved invaluable for orientation, visualisation, planning and analysis of results. Dose rates were measured in the technological channels from the reactor hall floor to the bottom of the hot gas chamber in decrements of 1 m and 0.5 m. The highest absolute values of dose rates were found in channels located in the middle of the reactor (up to 1900 mGy/h in the active zone region). It is estimated that the total contaminated area of primary circuit equipment (pipework, steam generators and turbo-compressors) is some 48 000 m2. It follows that the total gamma contamination is of the order of 1014 to 1015 Bq and total alpha contamination 1011 to 1013 Bq. The total amount of deposits in the gas circuit is about 14.3 tons. (authors)

  18. Modification of the IAN-R1 reactor

    International Nuclear Information System (INIS)

    The IAN-R1 reactor is the only nuclear reactor operating in Colombia; it is installed at the Institute of Nuclear Affairs (AIN) in Bogota, which is an official body coming under the Ministry of Mining and Energy. This reactor started operation in January 1965 with a rated power of 10 kW and was modified a year later to operate at 20 kW, which has been its rated power up to the present. Given its importance for the application of nuclear technology in Columbia for various purposes, principally in the areas of neutron activation analysis, determination of uranium content in minerals using the delayed neutron counting method, production of certain radioisotopes such as 198Au and 82Br for engineering applications, and production of radioactive material for teaching and research purposes, research has been in progress for some years into ways of increasing its power. The study on experimental requirements and on the demand for locally produced radioisotopes came to the conclusion that its power should be increased to 1000 kW, which would allow the facility to remain on the same site. The modification includes conversion of the core to low-enriched fuel, operation up to 1 MW, modification of the shielding, renovation of instrumentation and installation of a radioisotope processing plant. When the reactor is modified we will be able to produce other radioisotopes for applications in nuclear medicine, industry and engineering; at the same time, the safety of the facility will be optimized and the experimental facilities improved

  19. Ageing management and refurbishment of Ghana Research Reactor-1 (GHARR-1)

    International Nuclear Information System (INIS)

    Ageing management is an essential component of the routine practices at the Ghana Research Reactor-1 (GHARR-1) Facility. The reactor is Miniature Neutron Source Reactor with a rated power of 30 kW. GHARR-1 was installed and attained criticality on December 17, 1994 and commissioned on 8th March, 1995. It has since been in operation. The routine practices and operational procedures have been set out with clear emphasis on ageing policy at the facility. Some electronic components are changed regularly during maintenance sessions and keeping to regular purification of the reactor and pool water to mitigate against corrosion. This paper outlines the ageing management programme, mitigation practices, strategies for ageing management, periodic safety reviews, consideration of ageing during designing, design features for components and unit replacement, top beryllium shim addition, and succession planning. Information sharing with other operating organization is one of the means considered by GHARR-1 to attain excellence

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

    International Nuclear Information System (INIS)

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

  1. Improving the radionuclide Inventory Determination of the Irradiated Graphite from BR1 in Mol

    OpenAIRE

    Nijst, Stefan

    2014-01-01

    The Belgian Reactor 1 (BR1) operational since 1956 at SCK'CEN in Mol, is the oldest research reactor in Belgium. It is a graphite-moderated and air-cooled reactor fuelled with natural metallic uranium. The active core consists of a 6.66 x 6.84 x 6.84 m3 graphite matrix, built by stacking squared-base prismatic graphite blocks (~14500), yielding a total mass of 492 tons. The BR1 is supposed to continue its operation for several decades; however it is necessary to already make studies abou...

  2. New burnup calculation of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

  3. New burnup calculation of TRIGA IPR-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z., E-mail: sinclercdtn@hotmail.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

  4. Validating the Serpent Model of FiR 1 Triga Mk-II Reactor by Means of Reactor Dosimetry

    OpenAIRE

    Viitanen Tuomas; Leppänen Jaakko

    2016-01-01

    A model of the FiR 1 Triga Mk-II reactor has been previously generated for the Serpent Monte Carlo reactor physics and burnup calculation code. In the current article, this model is validated by comparing the predicted reaction rates of nickel and manganese at 9 different positions in the reactor to measurements. In addition, track-length estimators are implemented in Serpent 2.1.18 to increase its performance in dosimetry calculations. The usage of the track-length estimators ...

  5. Renewal of instrumentations for Egypt's first research reactor (ETRR-1)

    International Nuclear Information System (INIS)

    The work reported in this paper presents the tasks completed or currently under completion for the renewal of the nuclear instrumentation and control system, radiation protection system and process instrumentation system for Egypt's first research reactor (ETRR-1). The mentioned tasks started in 1980. The work reported includes the procurement and installation procedures and gives also a historical background which introduces ETRR-1 and its operating history together with the need for and philosophy behind the renovation of the above mentioned systems which were first put in operation in 1961

  6. Stationary low power reactor No. 1 (SL-1) accident site decontamination ampersand dismantlement project

    International Nuclear Information System (INIS)

    The Army Reactor Area (ARA) II was constructed in the late 1950s as a test site for the Stationary Low Power Reactor No. 1 (SL-1). The SL-1 was a prototype power and heat source developed for use at remote military bases using a direct cycle, boiling water, natural circulation reactor designed to operate at a thermal power of 3,000 kW. The ARA II compound encompassed 3 acres and was comprised of (a) the SL-1 Reactor Building, (b) eight support facilities, (c) 50,000-gallon raw water storage tank, (d) electrical substation, (e) aboveground 1,400-gallon heating oil tank, (f) underground 1,000-gallon hazardous waste storage tank, and (g) belowground power, sewer, and water systems. The reactor building was a cylindrical, aboveground facility, 39 ft in diameter and 48 ft high. The lower portion of the building contained the reactor pressure vessel surrounded by gravel shielding. Above the pressure vessel, in the center portion of the building, was a turbine generator and plant support equipment. The upper section of the building contained an air cooled condenser and its circulation fan. The major support facilities included a 2,500 ft2 two story, cinder block administrative building; two 4,000 ft2 single story, steel frame office buildings; a 850 ft2 steel framed, metal sided PL condenser building, and a 550 ft2 steel framed decontamination and laydown building

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

    International Nuclear Information System (INIS)

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

  8. Belgian support programme to the IAEA for safeguards implementation

    International Nuclear Information System (INIS)

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

  9. The Dutch-Belgian beamline at the ESRF.

    Science.gov (United States)

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

    1998-05-01

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

  10. The attitudes of Belgian adolescents towards peers with disabilities

    NARCIS (Netherlands)

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

    2011-01-01

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

  11. Open Access to the Belgian Nuclear higher Education Network

    International Nuclear Information System (INIS)

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

  12. Reactor engineering department annual report. April 1, 1993-March 31, 1994

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1993 (April 1, 1993-March 31, 1994). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees organized by the Department are also summarized in this report. (author)

  13. RA Research nuclear reactor Part 1, RA Reactor operation and maintenance in 1987

    International Nuclear Information System (INIS)

    RA research reacto was not operated due to the prohibition issued in 1984 by the Government of Serbia. Three major tasks were finished in order to fulfill the licensing regulations about safety of nuclear facilities which is the condition for obtaining permanent operation licence. These projects involved construction of the emergency cooling system, reconstruction of the existing special ventilation system, and renewal of the system for electric power supply of the reactor systems. Renewal of the RA reactor instrumentation system was initiated. Design project was done by the Russian Atomenergoeksport, and is foreseen to be completed by the end of 1988. The RA reactor safety report was finished in 1987. This annual report includes 8 annexes concerning reactor operation, activities of services and financial issues, and three special annexes: report on testing the emergency cooling system, report on renewal of the RA reactor and design specifications for reactor renewal and reconstruction

  14. Verification of computer codes for dynamic processes in nuclear reactors against experiments at loop facility of IGR-1 pulse reactor

    International Nuclear Information System (INIS)

    Basic principles of PRISDG and PRISET computer codes structure to analyze dynamic processes in nuclear reactors are presented. The codes were verified against experimental studies of dynamic processes related with flow-stop and power surge. The experimental data were obtained at loop facility of IGR-1 pulse reactor using fuel assemblies of IVV-2M research reactor. Accuracy of the codes is the same as the accuracy achieved in the experiments. Analysis could be performed at PS-2-type personal computers. Running time is not longer than several tens of minutes. (author)

  15. Upgrade of Instrumentation for Purdue Reactor PUR-1

    Energy Technology Data Exchange (ETDEWEB)

    Revankar, S.T.; Merritt, E.; Bean, R.

    2000-08-28

    The major objective of this program was 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 two projects on instrument upgrade were implemented. The first one was to convert the vacuum tube control and safety amplifiers (CSA) to solid state electronics, and the other was to upgrade the electrical and electronic shielding. This report is the annual report and gives the efforts and progress achieved on these two projects from July 1999 to June 2000.

  16. Fuel management methodology upgrade of Thai Research Reactor (TRR-1/M1) using SRAC computer code

    International Nuclear Information System (INIS)

    Thailand Institute of Nuclear Technology (TINT) is currently responsible for the nuclear research reactor called TRR-1/M1 located in Bangkok. The existing fuel management tool for TRR-1/M1 is a computer program called TRIGAP, which was developed in Slovenia during the 80's. Although TRIGAP is capable of calculating reactor parameters such as core excess reactivity or neutron fluxes, this tool has several drawbacks. Since TRIGAP only models the spatial distribution of neutrons in cylindrical geometry, the TRR-1/M1 core, which is formed in hexagonal lattices, needs to be homogenized into cylindrical rings. As a result, TRIGAP is unable to provide pin-wise data such as normalized power distribution of the reactor. To overcome this, an upgrade to the existing methodology is proposed. The upgraded methodology is actually similar to the existing methodology that it is executed in 2 steps. However, both steps are performed by more advanced computer programs collectively packaged into one system called SRAC which has been developed in Japan since 1978. To validate the upgraded methodology, the group cross sections of different lattices needed for the reactor core calculation were generated by the PIJ module of the SRAC system and the reactor core calculations of TRR-1/M1 core number 1 and 2 were performed afterwards

  17. Reactors

    International Nuclear Information System (INIS)

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

  18. 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.

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

    International Nuclear Information System (INIS)

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

  20. Modernization of control instrumentation and security of reactor IAN - R1

    International Nuclear Information System (INIS)

    The program to modernize IAN-R1 research reactor control and safety instrumentation has been carried out considering two main aspects: updating safety philosophy requirements and acquiring the newest reactor control instrumentation controlled by computer, following the present criteria internationally recognized, for safety and reliable reactor operations and the latest developments of nuclear electronic technology. The new IAN-R1 reactor instrumentation consist of two wide range neutron monitoring channels, commanded by microprocessor a data acquisition system and reactor control, (controlled by computers). The reactor control desk is providing through two displays; all safety and control signals to the reactor operators; furthermore some signals like reactor power, safety and period signals are also showed on digital bar graphics, which are hard wired directly from the neutron monitoring channels

  1. Nordic study on reactor waste. Technical part 1 and 2

    International Nuclear Information System (INIS)

    An important part of the Nordic studies on system- and safety analysis of the management of low and medium level radioactive waste from nuclear power plants, is the safety analysis of a Reference System. This reference system was established within the study and is described in this Technical Part 1. The reference system covers waste management Schemes that are potential possibilities in either one of the four participating Nordic countries. The reference system is based on: a power reactor system consisting of 6 BWR's of 500 MWe each, operated simultaneously over the same 30 year period, and deep bed granular ion exchange resin wastes from the Reactor Water Clean-Up System (RWCS and powdered ion exchange resin from the Spent Fuel Pool Cleanup System (SFPCS)). Both waste types are supposed to be solidified by mixing with cement and bitumen. Two basic types of containers are considered. Standard 200 liter steel drums and specially made cubicreinforced concrete moulds with a net volume of 1 m3. The Nordic study assumes temporary storage of the solidified waste for a maximum of 50 years before the waste is transferred to the disposal site. Transportation of the waste from the storage facilitiy to the disposal site will be by road or sea. Three different disposal facilities are considered: Shallow land burial, near surface concrete bunker, and rock cavern with about 30 m granite cover. (EG)

  2. 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.

  3. Upgrade of VR-1 training reactor I and C

    International Nuclear Information System (INIS)

    The contribution describes the upgrade of the VR-1 training reactor I and C (Instrumentation and Control). The reactor was put into operation in the 1990, and its I and C seems to be obsolete now. The new I and C utilises the same digital technology as the old one. The upgrade has been done gradually during holidays in order not to disturb the reactor utilisation during teaching and training. The first stage consisted in the human-machine interface and the control room upgrade in 2001. A new operator's desk, displays, indicators and buttons were installed. Completely new software and communication interface to the present I and C were developed. During the second stage in 2002, new control rod drivers and safety circuits were installed. The rod motors were replaced and necessary mechanical changes on the control rod mechanism, induced by the utilisation of the new motor, were done. The new safety circuits utilise high quality relays with forced contacts to guarantee high reliability of their operation. The third stage, the control system upgrade is being carried out now. The new control system is based on an industrial PC mounted in a 19 inch crate. The operating system of the PC is the Microsoft Windows XP with the real time support RTX of the VentureCom Company. A large amount of work has been devoted to the software requirements to specify all dependencies, modes and permitted actions, safety measures, etc. The Department took an active part in the setting of software requirements and later in verification and validation of the software and the whole control system. Finally, a new protection system consisting of power measuring and power protection channels will be installed in 2004 or 2005. (author)

  4. Neutron Detection and Radiation Monitoring For The Egyptian Research Reactor (ETRR-1)

    International Nuclear Information System (INIS)

    The detection of neutrons inside the reactor core is very important for the control, utilization and safety of nuclear reactors. Also, the radiation monitoring system was designed to allow continuous monitoring of radiation levels outside the nuclear reactors. We will illustrate the various types and function of the neutron detection system and the modernized radiation monitoring system which are used in the Egyptian research reactor (ETRR-1)

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Stevens, Peter A. J.

    2008-01-01

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

  7. RA reactor operation and maintenance in 2000, Part 1

    International Nuclear Information System (INIS)

    Activities related to revitalisation of the RA reactor started in 1986 were fulfilled except the exchange of the complete reactor instrumentation. Since 1992, due to economic and political reasons, RA reactor is in a difficult situation. The old RA reactor instrumentation was dismantled. Decision about the future status of the reactor should be made because the aging of all the components is becoming dramatic. Control and maintenance of the reactor components was done regularly and efficiently. The most important activity and investment in 1998 was improvement of conditions for spent fuel storage in the existing pools at the RA reactor. Russian company ENTEK and IAEA are involved in this activity which was initiated 1997. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis

  8. Annual report on JEN-1 and JEN-2 Reactors; Informe periodico de Reactores JEN-1 y JEN-2 correpondiente al ano 1972

    Energy Technology Data Exchange (ETDEWEB)

    Montes Ponce de Leon, J.

    1974-07-01

    In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

  9. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1988

    International Nuclear Information System (INIS)

    According to the action plan for 1988, operation of the RA reactor should have been restarted in October, but the operating license was not obtained. Control and maintenance of the reactor components was done regularly and efficiently dependent on the availability of the spare parts. The major difficulty was maintenance of the reactor instrumentation. Period of the reactor shutdown was used for repair of the heavy water pumps in the primary coolant loop. With the aim to ensure future safe and reliable reactor operation, action were started concerning renewal of the reactor instrumentation. Design project was done by the soviet company Atomenergoeksport. The contract for constructing this equipment was signed, and it is planned that the equipment will be delivered by the end of 1990. In order to increase the space for storage of the irradiated fuel elements and its more efficient usage, projects were started concerned with reconstruction of the existing fuel handling equipment, increase of the storage space and purification of the water in the fuel storage pools. These projects are scheduled to be finished in mid 1989. This report includes 8 annexes concerning reactor operation, activities of services and financial issues

  10. Engineering development of slurry bubble column reactor (SBCR) technology. Quarterly report, January 1--March 31, 1996

    Energy Technology Data Exchange (ETDEWEB)

    Toseland, B.A.; Tischer, R.E.

    1997-12-31

    The major technical objectives of this program are threefold: (1) to develop the design tools and a fundamental understanding of the fluid dynamics of a slurry bubble column reactor to maximize reactor productivity; (2) to develop the mathematical reactor design models and gain an understanding of the hydrodynamic fundamentals under industrially relevant process conditions; and (3) to develop an understanding of the hydrodynamics and their interaction with the chemistries occurring in the bubble column reactor. Successful completion of these objectives will permit more efficient usage of the reactor column and tighter design criteria, increase overall reactor efficiency, and ensure a design that leads to stable reactor behavior when scaling up to large diameter reactors. The main part of this report describes tracer studies of slurry bubble column hydrodynamics during methanol synthesis.

  11. Radiation protection programme of the Ghana Research Reactor (GHARR-1)

    International Nuclear Information System (INIS)

    The Radiation Protection Programme is generally based on a prior risk assessment in which the locations and magnitudes of all radiation hazards are taken into account. This work has shown that the Ghana Research Reactor-1 which is a Miniature Neutron Source Reactor has ensured both technical and administrative protocols for an effective radiation protection programme. The key principle that has aided the technical inherent safety is the defence in depth and the adoption of multiple barriers for prevention of the escape of radioactive materials into the environment. Administrative procedures established include the classification of working areas and access control; local rules and supervision of work; monitoring of individuals and the workplace; work planning and work permits; application of the principle of optimization of protection; removal or reduction in intensity of sources of radiation, health surveillance and training. The MNSR passed various rigorous tests as the required quality assurance and control was adhered to. The control systems are in accordance with the Chinese national standards and guidelines and are compatible with those of IEC, IEEE and IAEA. (author)

  12. In service inspection of ETRR-1 reactor vessels

    International Nuclear Information System (INIS)

    Technical survey included in-service inspection are needed in order to investigate the structural integrity and to insure safe operation of ET-R R-1 reactor after thirty years aging. An intensive work for the inspection of the central tank, shield tank, horizontal channel, primary coolant circuit and spent fuel storage tank have been carried out. The inspection procedures were visual method using video camera and magnification optical device as well as thickness measurements using ultrasonic gauge meter and replica for determining defect depth. Water chemical analysis of the primary cooling circuit and spent fuel storage were helpful in results explanation. The results showed that the reactor vessel have good surface conditions. The observed pitting did not affect the structural integrity. The majority of the defects were pits having maximum surface area of about 50 mm. Their depth dose not exceed 2 mm. The pits depth rate penetration is of the order of 0.5% per year. Thickness measurements showed insignificant variation. Water status and its chemical properties are very important in controlling corrosion rate. 11 fig., 14 tab

  13. Temelin-1 reactor unit commissioning and start-up

    International Nuclear Information System (INIS)

    The Temelin-1 commissioning process was affected substantially by the change in the Czech political situation in 1989. The effects thereof were both favourable and unfavourable. Among favourable effects are the replacement of the original Instrumentation and Control System by a more advanced system and design changes which have brought about additional improvement of the Temelin NPP design safety, although on the other hand, this had an adverse impact on the time span and price of the power plant construction. Additional adverse effects included an unstable political and economic situation, associated with frequent changes in the management of the utility CEZ, a.s. (owner of the plant) as well as frequent replacement of persons in the position of the managing director of the Temelin plant itself. Despite all the difficulties encountered, Temelin-1 reactor unit could be ultimately put into trial operation in June 2002. (author)

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

    International Nuclear Information System (INIS)

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

  15. Ecological status evaluation of the quality element macro-invertebrates for the Belgian coast (2007-2009)

    OpenAIRE

    Van Hoey, G.; Derweduwen, J; Hillewaert, H.; Hostens, K.; Pecceu, E.; Wittoeck, J.

    2010-01-01

    The Water Framework Directive (WFD; 2000/60/EG) aims to achieve a good ecological and chemical quality status of the European waters (Rivers, Lakes, coastal- and transitional waters). The quality status is determined based on the evaluation of different quality elements, e.g. macro-invertebrates. Macro-invertebrates are good indicators for detecting anthropogenic impacts and ecological degradation. The Belgian Coastal water body (< 1 nautical mile) is a small area, but this environment is hig...

  16. Ageing management of Pakistan Research Reactor-1 (PARR-1)

    International Nuclear Information System (INIS)

    The physical ageing of PARR-1, due to normal running of the plant and equipment, wear and tear, corrosion, vibration, stressing, thermal and mechanical fatigue and the general deterioration of the plant etc., has been manifested and dealt with almost from the beginning. The non physical ageing issues have been demonstrated in satisfying changing regulatory requirements, updating safety and administrative documentation, coping with technical obsolescence of facilities, and maintaining essential staff skills keeping in view the loss of safety knowledge that occurs with the loss of staff due to either retirement or organizational changes. Ageing issues are still there and need continuous attention

  17. Calculation and measurement of kinetic parameters of Pakistan Research Reactor-1 (PARR-1)

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad, Atta, E-mail: atta1974pk@yahoo.co [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences (PIEAS), P.O. Nilore, Islamabad 45650 (Pakistan); Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology (PINSTECH), P.O. Nilore, Islamabad 45650 (Pakistan); Iqbal, Masood, E-mail: masiqbal@hotmail.co [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology (PINSTECH), P.O. Nilore, Islamabad 45650 (Pakistan); Mahmood, Tyyab; Qadir, Javed [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology (PINSTECH), P.O. Nilore, Islamabad 45650 (Pakistan)

    2011-01-15

    Rossi-Alpha ({beta}{sub eff}/{Lambda}) for critical reactor measured experimentally by noise analysis technique at PARR-1 core at 35.26 full power days burn up. In noise analysis technique the inherent reactivity fluctuations are taken as input to reactor system and the neutron density population fluctuations are considered as output of the reactor system. The auto power spectral density of the linear channel is taken and used to find out the break frequency by non-linear least square fitting method, which leads to {beta}{sub eff}/{Lambda} = 161.45 s{sup -1}. Calculations were performed with the help of computer codes WIMSD/4 and CITATION. The calculated {beta}{sub eff}/{Lambda} = 161.07 s{sup -1} at 35.26 full power days burn up. The measured and calculated values for Rossi-Alpha are in good agreement within 0.235% of error.

  18. Calculation and measurement of kinetic parameters of Pakistan Research Reactor-1 (PARR-1)

    International Nuclear Information System (INIS)

    Rossi-Alpha (βeff/Λ) for critical reactor measured experimentally by noise analysis technique at PARR-1 core at 35.26 full power days burn up. In noise analysis technique the inherent reactivity fluctuations are taken as input to reactor system and the neutron density population fluctuations are considered as output of the reactor system. The auto power spectral density of the linear channel is taken and used to find out the break frequency by non-linear least square fitting method, which leads to βeff/Λ = 161.45 s-1. Calculations were performed with the help of computer codes WIMSD/4 and CITATION. The calculated βeff/Λ = 161.07 s-1 at 35.26 full power days burn up. The measured and calculated values for Rossi-Alpha are in good agreement within 0.235% of error.

  19. Status of prompt gamma neutron activation analysis (PGAA) at TRR-1/M1 (Thai Research Reactor-1/Modified 1)

    International Nuclear Information System (INIS)

    The first prompt gamma activation analysis (PGAA) was designed, constructed and installed at a 6 inch diameter neutron beam port of the Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1989. Beam characteristic were made by Gd foil irradiation, X-ray film exposing and densitometry scanning consequently. The thermal neutron flux at sample position was measured by Au foil activation, and was about 1 x 107 n.cm2.sec-1 at 700 kW operating power. The experiments have been conducted successfully. In 1998, the PGAA facility has been developed for the reactor operating power at 1.2 MW. The new PGAA system, e.g., beam shutter, gamma collimator and biological shields have been designed to reduce the leakage of neutrons and gamma radiation to the acceptance levels in accordance with the International Commission on Radiation Protection Publication 60 (ICRP 60). The construction and installation will be completed in April 1999. (author)

  20. The Mandate System for the Belgian Public Prosecution

    Directory of Open Access Journals (Sweden)

    Bruno BROUCKER

    2009-12-01

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

  1. Water cooled reactor technology: Safety research abstracts no. 1

    International Nuclear Information System (INIS)

    The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

  2. Management and Program Effectiveness in Belgian Sports Clubs

    OpenAIRE

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

    2009-01-01

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

  3. Characteristics and challenges of the modern Belgian veal industry

    OpenAIRE

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

    2014-01-01

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

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

    OpenAIRE

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

    2015-01-01

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

  5. Development path and capital structure of belgian biotechnology firms

    OpenAIRE

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

    2002-01-01

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

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

    OpenAIRE

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

    2011-01-01

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

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

    OpenAIRE

    Barran, Fernando; Peeters, Marga

    1998-01-01

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

  8. Dietary Intake of Artificial Sweeteners by the Belgian Population

    OpenAIRE

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

    2011-01-01

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

  9. The Attitudes of Belgian Adolescents towards Peers with Disabilities

    OpenAIRE

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

    2011-01-01

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

  10. Assessment of marine debris on the Belgian Continental Shelf

    OpenAIRE

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

    2013-01-01

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

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

    OpenAIRE

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

    2015-01-01

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

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

    OpenAIRE

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

    2009-01-01

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

  13. Operating reactors licensing actions summary. Volume 5, Number 1

    International Nuclear Information System (INIS)

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

  14. RA reactor operation and maintenance in 1994, Part 1

    International Nuclear Information System (INIS)

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. The spent fuel elements used from the very beginning of reactor operation are stored in the existing pools. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988 and was fulfilled in 1990. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor

  15. Major update of Safety Analysis Report for Thai Research Reactor-1/Modification 1

    International Nuclear Information System (INIS)

    Thai Research Reactor-1/Modification 1 (TRR-1/M1) was converted from a Material Testing Reactor in 1975 and it had been operated by Office of Atom for Peace (OAP) since 1977 until 2007. During the period, Office of Atom for Peace had two duties for the reactor, that is, to operate and to regulate the reactor. However, in 2007, there was governmental office reformation which resulted in the separation of the reactor operating organization from the regulatory body in order to comply with international standard. The new organization is called Thailand Institute of Nuclear Technology (TINT) which has the mission to promote peaceful utilization of nuclear technology while OAP remains essentially the regulatory body. After the separation, a new ministerial regulation was enforced reflecting a new licensing scheme in which TINT has to apply for a license to operate the reactor. The safety analysis report (SAR) shall be submitted as part of the license application. The ministerial regulation stipulates the outlines of the SAR almost equivalent to IAEA standard 35-G1. Comparing to the IAEA 35-G1 standard, there were several incomplete and missing chapters in the original SAR of TRR1/M1. The major update of the SAR was therefore conducted and took approximately one year. The update work included detail safety evaluation of core configuration which used two fuel element types, the classification of systems, structures and components (SSC), the compilation of detail descriptions of all SSCs and the review and evaluation of radiation protection program, emergency plan and emergency procedure. Additionally, the code of conduct and operating limits and conditions were revised and finalized in this work. A lot of new information was added to the SAR as well, for example, the description of commissioning program, information on environmental impact assessment, decommissioning program, quality assurance program and etc. Due to the complexity of this work, extensive knowledge was

  16. « Congobéton Léopoldville. Congés payés du 1/1/57 au 31/12/57 »: Postwar Architecture, Construction Work and Local Labor in a Belgian Colony

    OpenAIRE

    Lagae, Johan; Craenenbroeck, Ludwine Van

    2016-01-01

    Ever since the first article on colonial architecture in the former Belgian Congo, the territory known today as the Democratic Republic of the Congo, appeared in 1986, a substantial amount of research has been conducted on the topic, scrutinizing late nineteenth-century prefabricated metal structures, the introduction of modernist ideas in design and planning since the 1920s, and the emergence of 1950s tropical modernism. More recently, the built production of the post-independence era has al...

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

    OpenAIRE

    Faniel, Jean

    2012-01-01

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

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

    OpenAIRE

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

    2006-01-01

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

  19. Thermal facility for BNCT in RA-1 Argentine research reactor

    International Nuclear Information System (INIS)

    Full text: A thermal facility for BNCT experiments is being developed in an Argentine Research Reactor: RA-1 'Enrico Fermi'. RA-1 research nuclear reactor is working at Constituyentes Atomic Center, near Buenos Aires, and started operations in 1958. It worked at several power levels, up to 120 k W. Today, RA-1 is licensed to work at 40 k W. RA-1 was used to produce radioisotopes in the early 60's, and today gives irradiation services to test materials, to calibrate detectors and activation analysis. RA-1 users are CNEA researchers, Nuclear Regulatory Authority staff and private laboratories. Boron Neutron Capture Therapy (BNCT) is a method to fight against cancer. It consists to irradiate cancer tumors using thermal neutrons. The tumor tissue should include a dose of a boron solution. The Boron irradiation produces the following nuclear reactions: n + B10→ α + Li7 + γ. Being the α particle a radiation with short range, but high destructive energy, the tumor cells are destroyed. The neutron flux should be of 109 n/cm2seg, and the gamma dose lower than 0.48 s V/h. This method is oriented to treat brain tumors. Taking in account that the brain tumors usually are several centimeters deep in the head, to get thermal neutrons in the tumor is convenient to irradiate the patient using epithermal neutrons. moderation in the cells of the brain will permit to get more thermal neutrons in the tumor. In CNEA BNCT program there is in construction an epithermal clinical facility in the RA-6, a 500 k W research reactor that is at Bariloche Atomic Center. To perform some experiments for instance to test the boron compounds, RA-1 is used. In this experiments little animals like hamsters or bottles with cultivated cells are used, for that reasons thermal neutrons are used. The project in RA-1 consists in several stages. As the first stage a preliminary thermal facility was built. Irradiation times of 45-60 minutes were estimated, at power operation levels of 40 k W. Several

  20. Fast breeder reactor. The past, the present and the future. (6) History of fast reactor development in Japan - 1

    International Nuclear Information System (INIS)

    History and present state of fast breeder reactor was reviewed in series. As a history of fast reactor development in Japan - 1, this sixth lecture presented the start of FBR development, and construction and operation of the experimental FBR (JOYO). The JOYO began operation in 1977 and now is being operated at 140 MWt after two times of upgraded modification. The JOYO is aimed at (1) advancement of technology through and experiment, (2) conducting irradiation tests on fuels and materials and (3) validation of innovative technology for development of a future FBR. (T. Tanaka)

  1. Validating the Serpent Model of FiR 1 Triga Mk-II Reactor by Means of Reactor Dosimetry

    Science.gov (United States)

    Viitanen, Tuomas; Leppänen, Jaakko

    2016-02-01

    A model of the FiR 1 Triga Mk-II reactor has been previously generated for the Serpent Monte Carlo reactor physics and burnup calculation code. In the current article, this model is validated by comparing the predicted reaction rates of nickel and manganese at 9 different positions in the reactor to measurements. In addition, track-length estimators are implemented in Serpent 2.1.18 to increase its performance in dosimetry calculations. The usage of the track-length estimators is found to decrease the reaction rate calculation times by a factor of 7-8 compared to the standard estimator type in Serpent, the collision estimators. The differences in the reaction rates between the calculation and the measurement are below 20%.

  2. Dismantling of the reactor block of the FRJ-1 research reactor (MERLIN)

    International Nuclear Information System (INIS)

    By the end of 1998 the complete secondary cooling system and the major part of the primary cooling system were dismantled. Furthermore, the experimental devices, including a rabbit system conceived as an in-core irradiation device, were disassembled and disposed of. In total, approx. 65 t of contaminated and/or activated material as well as approx. 70 t of clearance-measured material were disposed of within the framework of these activities. The dismantling of the coolant loops and experimental devices was followed in 2000 by the removal of the reactor tank internals and the subsequent draining of the reactor tank water. The reactor tank internals were essentially the core support plate, the core box, the flow channel and the neutron flux bridges (s. Fig. 2, detailed reactor core). All components consisted of aluminium, the connecting elements such as bolts and nuts, however, of stainless steel. Due to the high activation of the core internals, disassembly had to be remotely controlled under water. All removal work was carried out from a tank intermediate floor (s. Fig. 2). These activities, which served for preparing the dismantling of the reactor block, were completed in summer 2001. The waste parts arising were transferred to the Service Department for Decontamination of the Research Centre. This included approx. 2.5 t of waste parts with a total activity of approx. 8 x 1011 Bq. (orig.)

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

    Directory of Open Access Journals (Sweden)

    Evelyne Verhaegen

    2015-11-01

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

  4. The role of Callionymus lyra (L.) and C. reticulatus in the life cycle of Lernaeocera lusci in Belgian coastal waters (Southern Bight of the North Sea)

    NARCIS (Netherlands)

    Van Damme, P.A.; Maertens, D.; Arrumm, A.; Hamerlynck, O.; Ollevier, F.

    1993-01-01

    A survey of the dragonet Callionymus lyra and the reticulated dragonet C. reticulatus from Belgian coastal waters (Southern Bight of the North Sea) in June 1991 revealed 34% of dragonets infected with 1–7 Lernaeocera lusci. This same parasite infected 9% of the reticulated dragonets (mean intensity

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

    Science.gov (United States)

    Penninckx, F; Danse, E

    2006-01-01

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

  6. The Belgian regulatory framework for NORM industries : test-case of a phosphate production plant

    International Nuclear Information System (INIS)

    According to the Belgian radiation protection regulatory framework, some categories of NORM industries must register to the competent Belgian radiation protection authority (FANC, Federal Agency for Nuclear Control). The facilities must provide a set of information which allows the authority to assess the radiological impact of the industrial activity on the workers, the population and the environment. If the dose exceeds or is likely to exceed 1 mSv/y, corrective measures have to be implemented. FANC has issued a methodology in order to define more precisely the data needed to evaluate the radiological impact. The methodology lists also the exposure pathways which have to be taken into account and the relevant parameters for the evaluation of the doses. The case of a phosphate production plant is discussed in more details: although the exposure of the workers in the production process as such is rather limited, the specific activity of some residues (calcium fluoride sludge) reaches around 10 kBq/kg of Ra-226. This sludge is disposed off in a specific landfill for which a program of radiological monitoring has been implemented. It includes periodic measurements of dose rate and radon concentration on the landfill. (author)

  7. Education and research at the VR-1 Vrabec training reactor facility

    International Nuclear Information System (INIS)

    The results of 12 years' efforts devoted to the construction of the VR-1 ''Vrabec'' training reactor at the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague and to establishing the training reactor department, as well as the contribution of the training reactor facility to the teaching and scientific activities of the Faculty are presented in a comprehensive manner. The thesis is divided into 2 parts: (i) preconditions, reactor construction and commissioning, and constituting the reactor department, and (ii) basic and comprehensive information concerning the current utilization of the reactor for the benefit of students from various university level institutions. The prospects of scientific activities of the department are also outlined. Attention is paid to selected nuclear safety aspects of the reactor during operation and teaching of students, as well as to its innovated digital control system whose implementation is planned. The results achieved are compared with the initial goals and with similar experience abroad. (P.A.)

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

    International Nuclear Information System (INIS)

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

  9. Argentina: Disposal aspects of RA-1 research reactor decommissioning waste

    International Nuclear Information System (INIS)

    The objective of the project is to analyze disposal aspects of waste from total dismantling of Argentinean research reactors, starting with the oldest one, 48 years old RA-1. In order to estimate decommissioning waste, data was collected from files, area monitoring, measurements, sampling to measure activity and composition, operational history and tracing of operational incidents. Measurements were complemented with neutron activation calculations. Decommissioning waste for RA-1 is estimated to be 71.5 metric tons, most of it concrete (57 tons), the rest being steels, lead and reflector graphite (4.8 tons). Due to their low specific activities, no disposal problems are foreseen in the case of metals and concrete. Disposal of aluminium, steel, lead and concrete is analyzed. On the contrary, as the country has no experience in managing graphite radioactive waste, work was concentrated on that material. Stored (Wigner) energy may exist in RA-1 graphite reflectors irradiated at room temperature. Evaluation of stored energy by calorimetric methods is proposed, and its annealing by inductive heating; HEPA filters should be used to deal with gaseous activity emissions, mainly Cl-36 and C-14. Galvanic corrosion, dust explosion, ignition and oxidation can be addressed and should not become disposal problems. Care must be taken with graphite dust generation and disposal, due to wetting and flotation problems. Lessons learned from the project are presented, and the benefits of sharing international experience are stressed. (author)

  10. Nuclear Reactor RA Safety Report, Vol. 1, Introduction

    International Nuclear Information System (INIS)

    Based on the agreement between governments of Socialist Federal Republic of Yugoslavia and USSR of January 28 1956, a contract was signed about construction of RA research reactor in the Boris Kidric Institute of nuclear sciences. Building of the RA reactor started in 1956, and has reached criticality in 1959. Since then it has been in almost permanent operation, except for five longer shutdown periods: in 1963 because of heavy water and primary coolant system contamination with cobalt; in 1970 because of transporting the heavy water to France for isotopic regeneration; in 1979/1980 and 1983 because of aluminium oxyhydrate deposition on the fuel element cladding in reactor active region; in 1985/1986 because of ventilation system reconstruction and construction of emergency core cooling system. RA reactor is a heavy water cooled and heavy water moderated research reactor. Since the beginning of its operation, 2% enriched metal uranium fuel was used. From 1976, 80% enriched uranium oxide fuel elements were used partially in some core regions and since 1981 the complete core was filled with this highly enriched fuel. RA reactor was designed to operate under normal conditions at 6.5 MW power and at 10 MW power under forced regime. As a powerful neutron source the reactor was meant to be used for research in the field of reactor and neutron physics, solid state physics, radiation chemistry, biology and radioactive isotopes production. RA reactor was build by Yugoslav companies based on USSR basic design project. Main components of the systems were produced in USSR

  11. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

  12. The Cairo Fourier Diffractometer Facility At The ETRR-1 Reactor

    International Nuclear Information System (INIS)

    The work presents the Cairo Fourier diffractometer facility (CFDF). The CFDF is based on the reverse time-of-flight (RTOF) concept and was recently installed, as IAEA TC Project, at one of the ET-RR-1 reactor horizontal channels. The CFDF performance is assessed and its main parameters are given. The facility applies a Fourier chopper system and 6Li-glass scintillators (NE-912) arranged according to the time focusing geometry in order to detect neutrons scattered from the sample at an angle 29= 90 diameter. The detector system has been optimized for studying the internal stresses in materials along with neutron diffraction measurements. Its angular aperture was found, from precise calculations, by a special program, to be equal to S.1 x 102 steradians. The neutron guide system attached to the CFDF provides a thermal neutron flux-l.lxl06n/cm2/sec at the sample position. It has been found, from measurements with different powder samples, that such value of the thermal neutron flux is adequate for neutron diffraction measurements, at scattering angle 2θ= 90ο and d-spacing values between 0.7 A and 2.5 A, within 0.45% resolution

  13. The Management of TRIGA Spent Fuel at ENEA RC-1 Research Reactor

    International Nuclear Information System (INIS)

    TRIGA Mark II reactor of ENEA's Casaccia research Center (in Italy named RC-1) reached first criticality in 1960. Reactor core was realized with 61 standard TRIGA fuel elements, aluminium clad. In this condition, the reactor was operated until August 1965 at a steady state power level of 100 kW. In the summer of 1965, a programme was established to increase the reactor power to 1 MW. After significant plant modifications (in order both to adapt the reactor to the new operative circumstances, including safety regulations, and to extend reactor flexibility in the widest research areas), the new criticality was reached in July 1967. The 1 MW reactor operative configuration was initially obtained with 76 standard TRIGA fuel elements, but stainless steel clad. The RC-1 Reactor is still operational and during these years, many fuel elements were used. In this paper we describe the facility, the infrastructure available for spent fuel storage, and the operative experience accumulated during these years in the management of RC-1 Spent Nuclear Fuel (SNF). The activities and the incumbencies during SNF shipment that was carried out in 1999, in the frame of the USA Return of Foreign Research Reactors Spent Fuel Programme, are also described. (author)

  14. Experimental possibilities of research reactors complex Bajkal-1 for the decision of the problems of atomic power

    International Nuclear Information System (INIS)

    Research reactors complex 'Bajkal' includes two research reactors IVG.1M and RA. The reactor IVG.1M is a research water-water heterogeneous tank type nuclear reactor on the thermal neutrons with light-water moderator and coolant and beryllium neutron reflector. At present time the experimental studies of processes of the fission yield, the precipitation, the filtration of fission products. The possibilities of this reactor and stand systems are allowed to begin the experimental studies of model fuel assemblies water-cooled reactors at the accidental regimes. The reactor RA is a research high temperature gas-cooled tank type nuclear reactor on the thermal neutrons with gas moderator, zirconium hydride coolant and beryllium neutron reflector. The reactor RA is used for studies of hard-working of fuel elements and fuel assemblies of gas-cooled reactors during long reactor irradiation and experimental study of yield processes, precipitation and filtration of fission products

  15. Fusion reactor design and technology 1986. V. 1

    International Nuclear Information System (INIS)

    The first volume of the Proceedings of the Fourth Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology organized by the IAEA (Yalta, 26 May - 6 June 1986) includes 36 papers devoted to the following topics: fusion programmes (3 papers), tokamaks (15 papers), non-tokamak reactors and open systems (9 papers), inertial confinement concepts (5 papers), fission-fusion hybrids (4 papers). Each of these papers has a separate abstract. Refs, figs and tabs

  16. An overview of the RECH-1 reactor conversion

    International Nuclear Information System (INIS)

    The RECH-l research reactor achieved the first criticality on October 13, 1974 using HEU MTR type fuel elements, which were fabricated by the UKAEA at Dounreay, Scotland. In 1979, the conversion of the reactor to use LEU fuel was decided; however, a rough estimate of the uranium density needed to convert the reactor gave 3.7 g/cm3. This density was not available, and to maintain the overall fuel element geometry it was necessary to convert the reactor to use 45% enriched uranium fuel. In 1985, the conversion of the reactor to use medium enriched uranium was achieved. Some years later, the Chilean Nuclear Energy Commission developed the capability to produce fuel elements based on U3Si2-Al dispersion fuel. Once the plant and the manufacturing and quality control procedures were commissioned to permit the production of fuel elements, a fabrication program starts to produce LEU fuel elements with a uranium density of 3.4 g/cm3. A fabrication qualification period that extended to the required fuel plates for the assembly of two fuel elements started. In November 1998, the first four LEU fuel elements manufactured by the Chilean Fuel Fabrication Plant were delivered to the reactor. When the first two fuel elements were introduced into the core a LEU fuel element qualification program began. While those fuel elements remain in the core, an evaluation program is being applied to observe its performance under irradiation condition. (author)

  17. Calculations of Changes in Reactivity during some regular periods of operation of JEN-1 MOD Reactor

    International Nuclear Information System (INIS)

    By a Point-Reactor model and Perturbation Theory, changes in reactivity during some regular operating periods of JEN-1 MOD Reactor have been calculated and compared with available measured values. they were in good agreement. Also changes in reactivity have been calculated during operations at higher power levels than the present one, concluding some practical consequences for the case of increasing the present power of this reactor. (Author)

  18. Approach to decommissioning planning of the Egyptian research reactor ET-RR-1

    International Nuclear Information System (INIS)

    Based on the broad inspection programme carried out for ET-RR-1, it has been clearly demonstrated that the whole reactor system, subsystem and components need urgent renewing, replacement, maintenance and testing; (in order to increase the plant lifetime for another 5-10 years). Intensive studies and evaluation of different strategies for future planning to update the current reactor systems, upgrading reactor power or decommissioning are under investigations with planning for the future decommissioning as part of the programme. (author)

  19. SWAN-PPL, Fusion Reactor 1-D Particle Transport Optimization

    International Nuclear Information System (INIS)

    1 - Description of problem or function: Given the material density profiles which describe a one-dimensional reference system with a neutron source, SWAN will calculate, and optionally implement, density changes so as to optimize a single functional parameter of the system. 2 - Method of solution: The one-dimensional discrete-ordinate transport code ANISN is used to calculate flux and adjoint distributions for specified sources. The code SWIF calculates first-order estimates of the effect of material density changes on a goal functional, and from these evaluates effectiveness functions for the substitution of one material for another. Density distribution changes are then calculated which would optimize the goal functional, optionally subject to a constraint of holding another functional constant (to first order). 3 - Restrictions on the complexity of the problem: SWAN is not designed to analyze critical systems; it assumes that there is a fixed source, as in shielding or fusion reactor applications. Otherwise it is compatible with ANISN. All arrays are variably-dimensioned, so that there are no restrictions on individual dimensions

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

    OpenAIRE

    F.BOSSIER; D. DUWEIN

    1981-01-01

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

  1. University of Florida training reactor. Annual progress report, September 1, 1984-August 31, 1985

    International Nuclear Information System (INIS)

    This annual progress report of the University of Florida Training Reactor discusses: reactor operation; personnel; modifications made to the reactors; reactor maintenance; and testing of reactor systems

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

    Directory of Open Access Journals (Sweden)

    Pierre Robert Smeesters

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

  3. Performance of HEU and LEU fuels in Pakistan Research Reactor-1 (PARR-1)

    International Nuclear Information System (INIS)

    Pakistan Research Reactor-1 (PARR-1) a swimming pool MTR type 5 MW research reactor went critical in 1965 with HEU fuel. The reactor was operated with HEU fuel for about 30,000 hours in 25 years and produced about 93,000 MWh energy. The reactor was then converted to LEU fuel and its power upgraded from 5 MW to 10 MW to meet the demand of higher neutron flux and compensate the penalty in neutron flux due to conversion from HEU to LEU. The reactor went critical with LEU fuel in 1991. The operation with LEU fuel in the last 14 years has been about 10,000 hours and the energy produced is about 66,000 MWh. The performance of both HEU and LEU fuels has been excellent during the long operation history. The average and maximum burn up with HEU fuel was 34 % and 49 % respectively whereas that with LEU fuel has been 42 % and 48 % respectively. No signs of fission product release in pool water have ever been observed thus establishing full integrity of the fuel. Post irradiation visual inspection of the fuel has revealed no abnormality. No signs of geometrical distortion, corrosion or any other damage to the fuel have ever been observed. A fuel element got damaged during fuel handling which was repaired and replaced in the core and it achieved 28 % burn up without causing any problem. To establish the quality of the new fuel, a fuel failure detection system has been installed. This system is based upon monitoring the delayed neutrons emitted from fission products leaking into the primary coolant as a result of any clad failure. No incident of fuel failure has even been recorded by this system. (author)

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

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Maes, Ivo; Buyst, Erik

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

  7. Design of a new research reactor : 1st year conceptual design

    International Nuclear Information System (INIS)

    A new research reactor model satisfying the strengthened regulatory environments and the changed circumstances around nuclear society should be prepared for the domestic and international demand of research reactor. This can also lead to the improvement of technologies and fostering manpower obtained during the construction and the operation of HANARO. In this aspect, this study has been launched and the 1st year conceptual design has been carried out in 2003. The major tasks performed at the first year of conceptual design stage are as follows; Establishments of general design requirements of research reactors and experimental facilities, Establishment of fuel and reactor core concepts, Preliminary analysis of reactor physics and thermal-hydraulics for conceptual core, Conceptual design of reactor structure and major systems, International cooperation to establish foundations for exporting

  8. FiR 1 Reactor in Service for Boron Neutron Capture Therapy (BNCT) and Isotope Production

    International Nuclear Information System (INIS)

    The FiR 1 reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). Although BNCT dominates the current utilization of the reactor, it also has an important national role in providing local enterprises and research institutions in the fields of industrial measurements, pharmaceuticals, electronics, etc. with isotope produc- tion and activation analysis services. The whole reactor building has been renovated, creating a dedicated clinical BNCT facility at the reactor. Close to 30 patients have been treated since May 1999, when the licence for patient treatment was granted to the responsible BNCT treatment organization. The treatment organization has a close connection to the Helsinki University Central Hospital. (author)

  9. Neutronic calculations for the new fuel configuration of the ETRR-1 research reactor

    International Nuclear Information System (INIS)

    Neutronic calculations were performed for the new loading configuration of the ETRR-1 research reactor. The MCNP three dimensions Monte Carlo code and the two dimensions CITATION code are used to model the reactor. The power and thermal flux distributions in the reactor core are calculated. The power peak factor and the effect of control rod insertion on both flux and power profiles in the reactor core are determined and analyzed. The partial and total control rods worth are calculated. It was found that the difference between MCNP and CITATION in power distributions is 4 to 8% and for thermal flux ranges between 3 to 14%. (orig.)

  10. Conversion of the IAN-R1 reactor from MTR fuel to TRIGA LEU fuel

    International Nuclear Information System (INIS)

    The Institute of Nuclear Sciences and Alternative Energies (INEA) in Bogota, Colombia, has operated since 1965, a small 10 kW(t) research reactor, known as the IAN-R1 reactor, which was upgraded to 30 kW(t) in 1980. This reactor was provided to the Republic of Colombia under the U.S. Atoms for Peace Program, and which has been fueled with MTR HEU fuel, enriched nominally to 93% U-235. With the cooperation of the International Atomic Energy Agency (IAEA), a gradual reactor upgrade program has been undertaken beginning in 1987. The first step in this program was the upgrade of reactor instrumentation and control systems. In December, 1994, the IAEA and INEA entered into a tripartite contract with General Atomics (GA) to prepare a new safety analysis report for performing an HEU to LEU conversion of the R-1 reactor, manufacture TRIGA type LEU (19.7% enriched) fuel to replace the original MTR-HEU fuel plate assemblies, upgrade the reactor power to 100 kW(t), carry out additional upgrades of auxiliary reactor systems and commission the reactor with TRIGA fuel. (author)

  11. The Osiris reactor. Descriptive report - Volume 1 - text

    International Nuclear Information System (INIS)

    Osiris is a pool type reactor with a 70 MW thermal power. Its main purpose is to irradiate under high flows of neutrons the materials of which future nuclear power stations are made. This report proposes a description of this pool reactor. A first part describes the functional aspects and general characteristics of all installations which are in principle definitely defined (premises, irradiation and experimentation equipment, water circuits, power supply, venting, controls). The second part addresses elements which are likely to be changed, and more particularly the reactor core: fuel elements and controls (uranium and boron load in different fuel element generations, experimental locations within the core), neutron transport aspects (calculation and experiment), and thermal aspects (power generation and removal) of the pile). The third part addresses the operation: operation cycles, stops, exploitation organisation

  12. Neutron Beam Characterization for Neutron Radiography Facility at the Thai Research Reactor TRR-1/M1

    International Nuclear Information System (INIS)

    The aim of this research is to characterize the present status of neutron beam coming out from the reactor core of Thai Research Reactor TRR-1/M1 through neutron radiography facility. In this study, the neutron beam profiles at different positions along the beam exit were recorded using digital imaging devices. In addition, thin foil activation technique, with and without cadmium cover, was employed to determine thermal neutron flux and Cd ratio. An acrylic step wedge was exposed to neutron at different time. In parallel to image construction, neutron detection was carried out using a BF3 gas-filled detector. Then, the image intensities at particular thicknesses were normalized by neutron counts from the BF3 detector to determine relative neutron intensity. The obtained information of neutron beam characterization will be useful not only for monitoring the present status of neutron radiography facility but also for determining the optimum exposure conditions for particular samples in the future.

  13. Extensive utilisation of VR-1 reactor for nuclear education and training

    International Nuclear Information System (INIS)

    The paper presents utilisation of the VR-1 reactor for nuclear education and training at national and international level. VR-1 reactor has been operating by the Czech Technical University since December 1990. The reactor is a pool-type light water reactor based on enriched uranium (19.7% 235U) with maximum thermal power 1kW and for short time period up to 5kW. The moderator of neutrons is light water, which is also used as a reflector, a biological shielding and a coolant. Heat is removed from the core by natural convection. The pool disposition of the reactor facilitates access to the core, setting and removing of various experimental samples and detectors, easy and safe handling of fuel assemblies. The reactor core can contain from 17 to 21 fuel assemblies IRT-4M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The reactor is equipped with several experimental devices; e.g. horizontal, radial and tangential channels used to take out a neutron beam, reactivity oscillator for dynamics study and bubble boiling simulator. The reactor has been used very efficiently especially for education and training of university students and NPP's specialists for more than 18 years. The VR-1 reactor is utilised within various national and international activities such as Czech Nuclear Education Network (CENEN), European Nuclear Education Network and also Eastern European Research Reactor Initiative (EERRI). The reactor is well equipped for education and training not only by the experimental facility itself but also by incessant development of training methods and improvement of education experiments. The education experiments can be combined into training courses attended by students according to their study specialization and knowledge level. The training programme is aimed to the reactor and neutron physics, dosimetry, nuclear safety, and control of nuclear installations. Every year, approximately 250 university students undergo

  14. Reactor safety research programs. Quarterly progress report, January 1--March 31, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Romano, A. J. [comp.

    1978-04-01

    Progress is summarized in the following areas: (1) gas reactor safety evaluation, (2) THOR code development, (3) foreign code review, (4) SSC code development, (5) LMFBR and LWR safety experiments, (6) fast reactor safety code validation, (7) stress corrosion cracking of PWR steam generator tubing, and (8) technical coordination of structural integrity.

  15. Quality control of pool water from IEA-R1 reactor

    International Nuclear Information System (INIS)

    This paper presents the results of the pool water monitoring program of the IEA-R1 reactor of IPEN/CNEN-SP in normal operation. The considered period was previous to the systems upgrade and modernization for the new reactor operation condition: a power of 5 MW and operation time of 100 hours weekly. (author)

  16. Magnetic Fustion Reactor Design Studies Program final report, 1 July 1986--30 September 1986

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-09-30

    This report presents progress reported during the period, 7/1/86 - 9/30/86 for the Technical Support Services (TSS) for the Magnetic Fusion Reactor Design Studies Program. Tasks reported include: systems studies work plan, normalization of reactor design studies, interpretation of design study activities, research and development plan, conference support, and reports generated.

  17. Modifications done in the IPR-R1 reactor and their auxiliary systems

    International Nuclear Information System (INIS)

    The improvements done in the IPR-R1 reactor for adequateness of operation conditions and increase of irradiation sample capability. The cooling systems, reactor pool, system of control rods were substituted. The optimization of transfer pneumatic system was done. (M.C.K.)

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

    International Nuclear Information System (INIS)

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

  19. A level 1+ probabilistic safety assessment for Dhruva reactor

    International Nuclear Information System (INIS)

    Full text: Probabilistic safety assessment of Dhruva research reactor has been carried out. The scope of this work extends beyond level 1 PSA study to include the limited scope level 2 PSA study to give the likelihood of releases, during the postulated LOCA scenario, to the member of public. The work involved under this project include the reliability analysis of safety systems and safety support systems, estimation of failure frequency of the initiating events and modeling of accident sequences towards giving the statement of core damage frequency (CDF) for the plant. The uncertainty analysis has been carried out at system level as well as at CDF level to account for possible data and modeling error. The sensitivity analysis has been performed to check the affect of major assumptions and critical system parameters on the result of this analysis. The results of this analysis include the statement of CDF and important accident sequences for the plant. The important accident initiators identified in this study which contribute significantly to the CDF value. Though LOCA contribution is small to the CDF the consequences of LOCA have potential of radioactivity release into the containment. There exists potential, however small it may be, for release of radioactivity outside the plant depending on the performance of containment isolation and emergency exhaust system. Based on the results and findings of the deterministic safety analysis of the plant, further study was carried out to estimate the frequency of release of radioactivity during LOCA scenario. It was found that the likelihood of release of radioactivity of the order of around one-tenth of the permissible limit, to the member of public is very low

  20. Level 1 Tornado PRA for the High Flux Beam Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bozoki, G.E.; Conrad, C.S.

    1994-05-01

    This report describes a risk analysis primarily directed at providing an estimate for the frequency of tornado induced damage to the core of the High Flux Beam Reactor (HFBR), and thus it constitutes a Level 1 Probabilistic Risk Assessment (PRA) covering tornado induced accident sequences. The basic methodology of the risk analysis was to develop a ``tornado specific`` plant logic model that integrates the internal random hardware failures with failures caused externally by the tornado strike and includes operator errors worsened by the tornado modified environment. The tornado hazard frequency, as well as earlier prepared structural and equipment fragility data, were used as input data to the model. To keep modeling/calculational complexity as simple as reasonable a ``bounding`` type, slightly conservative, approach was applied. By a thorough screening process a single dominant initiating event was selected as a representative initiator, defined as: ``Tornado Induced Loss of Offsite Power.`` The frequency of this initiator was determined to be 6.37E-5/year. The safety response of the HFBR facility resulted in a total Conditional Core Damage Probability of .621. Thus, the point estimate of the HFBR`s Tornado Induced Core Damage Frequency (CDF) was found to be: (CDF){sub Tornado} = 3.96E-5/year. This value represents only 7.8% of the internal CDF and thus is considered to be a small contribution to the overall facility risk expressed in terms of total Core Damage Frequency. In addition to providing the estimate of (CDF){sub Tornado}, the report documents, the relative importance of various tornado induced system, component, and operator failures that contribute most to (CDF){sub Tornado}.

  1. 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.

  2. Average daily nitrate and nitrite intake in the Belgian population older than 15 years.

    Science.gov (United States)

    Temme, E H M; Vandevijvere, S; Vinkx, C; Huybrechts, I; Goeyens, L; Van Oyen, H

    2011-09-01

    The aim of this study was to assess the dietary intake of nitrate and nitrite in Belgium. The nitrate content of processed vegetables, cheeses and meat products was analysed. These data were completed by data from non-targeted official control and from the literature. In addition, the nitrite content of meat products was measured. Concentration data for nitrate and nitrite were linked to food consumption data of the Belgian Food Consumption Survey. This study included 3245 respondents, aged 15 years and older. Food intakes were estimated by a repeated 24-h recall using EPIC-SOFT. Only respondents with two completed 24-h recalls (n=3083) were included in the analysis. For the intake assessment, average concentration data and individual consumption data were combined. Usual intake of nitrate/nitrite was calculated using the Nusser method. The mean usual daily intake of nitrate was 1.38 mg kg(-1) bodyweight (bw) day(-1) and the usual daily intake at the 97.5 percentile was 2.76 mg kg(-1) bw day(-1). Exposure of the Belgian population to nitrate at a mean intake corresponded to 38% of the ADI (while 76% at the 97.5 percentile). For the average consumer, half of the intake was derived from vegetables (especially lettuce) and 20% from water and water-based drinks. The average daily intake of nitrate and nitrite from cheese and meat products was low (0.2% and 6% of the ADI at average intake, respectively). Scenario analyses with a higher consumption of vegetables or a higher nitrate concentration in tap water showed a significant higher intake of nitrate. Whether this is beneficial or harmful must be further assessed. PMID:21728895

  3. Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1)

    International Nuclear Information System (INIS)

    The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity (ρex), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU)

  4. Atmospheric dispersion modeling for an accidental release from the Pakistan Research Reactor-1 (PARR-1)

    International Nuclear Information System (INIS)

    Atmospheric dispersion modeling and radiation dose calculations have been performed for a postulated accidental airborne radionuclide release from the Pakistan Research Reactor-1 (PARR-1) appropriate to a power upgrade to 10 MW. Estimates of releases for various radionuclide groups are based upon US-NRC regulatory guide 1.183. Committed Effective Doses (CEDs) to the public at various downwind distances were calculated using a health physics computer code 'HotSpot' developed at the Lawrence Livermore National Laboratory, University of California, USA. The doses were calculated for various atmospheric stability classes, viz., Pasquill categories A-F with site-specific averaged meteorological conditions. The meteorological data on atmospheric stability conditions, mean wind speed and the frequency distribution of wind direction based on data collected near the reactor site have also been analyzed and are presented here. The results indicate that a person located within a downwind distance of about 500 m from the reactor would receive more than the permissible CED under the analyzed severe accident scenario. Analysis of one typical year of wind data indicates that the predominant wind direction is East-North East (ENE), which occurs at the site for more than 15% of the time

  5. Atmospheric dispersion modeling for an accidental release from the Pakistan Research Reactor-1 (PARR-1)

    Energy Technology Data Exchange (ETDEWEB)

    Raza, S. Shoaib [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)]. E-mail: ssraza@msn.com; Iqbal, M. [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

    2005-07-15

    Atmospheric dispersion modeling and radiation dose calculations have been performed for a postulated accidental airborne radionuclide release from the Pakistan Research Reactor-1 (PARR-1) appropriate to a power upgrade to 10 MW. Estimates of releases for various radionuclide groups are based upon US-NRC regulatory guide 1.183. Committed Effective Doses (CEDs) to the public at various downwind distances were calculated using a health physics computer code 'HotSpot' developed at the Lawrence Livermore National Laboratory, University of California, USA. The doses were calculated for various atmospheric stability classes, viz., Pasquill categories A-F with site-specific averaged meteorological conditions. The meteorological data on atmospheric stability conditions, mean wind speed and the frequency distribution of wind direction based on data collected near the reactor site have also been analyzed and are presented here. The results indicate that a person located within a downwind distance of about 500 m from the reactor would receive more than the permissible CED under the analyzed severe accident scenario. Analysis of one typical year of wind data indicates that the predominant wind direction is East-North East (ENE), which occurs at the site for more than 15% of the time.

  6. Reactor engineering department annual report. April 1, 1994 - March 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author).

  7. Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department's programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

  8. Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-10-01

    This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department`s programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

  9. Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department's programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  10. Reactor engineering department annual report. April 1, 1995 - March 31, 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  11. Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1998-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department`s programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  12. Reactor engineering department annual report. April 1, 1994 - March 31, 1995

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  13. Reactor engineering department annual report. April 1, 1995 - March 31, 1996

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

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

    Science.gov (United States)

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

    2011-01-01

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

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

    Science.gov (United States)

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

    2009-05-01

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

  16. Status of special reactor process tube loadings, November 1, 1965

    Energy Technology Data Exchange (ETDEWEB)

    Bown, R.W.

    1965-11-10

    This report gives the status of production test control tube loadings in reactor process tubes containing significant amounts of SS materials. Data are given in table form. For further description of column headings and the current discharge goal exposure plan refer to Document RL-REA-837.

  17. Approach for Establishment of an Ageing Management Programme for Thai Research Reactor-1/Modification 1 (TRR- 1/M1)

    International Nuclear Information System (INIS)

    Thai Research Reactor-1/Mod. 1 is a TRIGA Mark III reactor and has been in operation for more than 30 years. As the reactor has become older, ageing issues have been more prominent. Therefore, the ageing management programme of TRR-1/M1 is being systematically formulated to manage ageing issues. The purpose of this paper is to discuss the approach being taken for the establishment of the TRR-1/M1 programme. Essentially, the approach is based on the new IAEA Safety Guide SSG-10 Ageing Management for Research Reactors, which was published in 2010. The key to success for the approach is to develop the ageing management programme and then integrate it into the current quality assurance programme. The formulation of the ageing management programme takes into account both major ageing categories, i.e. physical ageing and non-physical ageing. The physical ageing management begins by screening the SSCs for ageing management review. The SSCs in the SR-A class (SSCs performing safety functions) and the SR-B class (SSCs not performing safety functions but safety relevant) are evaluated whether to be included in the ageing review routine. The ageing mechanisms of these SSCs are then thoroughly studied to better understand the ageing degradation processes. The examples of ageing mechanisms of these SSCs are fatigue, corrosion and erosion, stress corrosion cracking and irradiation effects. Due to the wide variety of disciplines involved in the evaluation, external experts in each specific field are sought for consultation. The results from the study are to be reviewed for improvement of practices used in operation, maintenance, inspection and testing. The programme will also identify the measures to be taken for detection, monitoring and analysis of ageing degradation trends. The measures will be formulated and included in the routine inspection, maintenance and testing programme. The current conditions of the SSCs are to be factored into the programme. In addition, mitigating

  18. 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

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

    International Nuclear Information System (INIS)

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

  20. Enhancement of physical security at the IAN-R1 research reactor

    International Nuclear Information System (INIS)

    The IAN-R1 research reactor has undergone continuous substantial changes involving modifications to the instrumentation, power and fuel. The reactor group of the Institute of Nuclear Science and Alternative Energy (INEA) of the Ministry of Mines and Energy has endeavoured to improve the physical security of the reactor installations. Colombia has undertaken to maintain adequate physical protection measures with respect to the installations and the materials supplied, as well as any special fissionable material used, including subsequent generations of fissionable material produced. The paper gives details of the level of physical protection and the implementation of physical protection measures and the IAN-R1 research reactor and of the new project currently being developed under which the present system of security installed in the reactor will be upgraded and greater security will be applied to other sensitive installations of INEA. (author)

  1. Calculation of fissile nuclides and fission products inventory applied to ETRR-1 research reactor

    International Nuclear Information System (INIS)

    The study of the nuclear reactor fuel safety implies studying physical mechanical, thermal and chemical proportions of the fuel during normal operation and accident conditions. A model was developed to calculate the fissile nuclides and fission products inventory in an operating reactor. The model considers the production and removal of different radionuclides leaking into account the decay schemes of each. The mathematical formulas were treated without any approximations. A decay model was developed for the period after reactor shutdown. The amount of different nuclides was evaluated for a given cooling time. Egypt test and research reactor number 1, ETRR-1. Was chosen to apply the model. The amount of about 200 nuclides was calculated. A certain nuclides was chosen to be presented based on their poisoning ratios. Criticality calculations were carried out to investigate the criticality condition of the reactor at different operating times. 4 fig

  2. Application of MCNP for neutronic calculations at VR-1 training reactor

    International Nuclear Information System (INIS)

    The paper presents the utilization of the Monte Carlo MCNP transport code for neutronic calculations of the VR-1 training reactor. Zero power light water reactor VR-1 is used mainly for training and partially for research. The reactor core consists of IRT-4M fuel elements with enrichment below 20 % of 235U and other components (e.g. control rods, fuel dummies, dry channels etc.). The reactor offers large variety of the core configurations and every year at least one critical experiment with new original core configuration is performed. The results of the calculations are compared with measured data collected during the last critical experiments performed with various reactor core configurations. A very good agreement between calculations and measurements is observed

  3. Measures aimed at enhancing safe operation of the Nigeria Research Reactor-1 (NIRR-1)

    International Nuclear Information System (INIS)

    Safety culture has been defined as 'that assembly of characteristics and attitudes in organizations and individuals which establishes that as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. This paper briefly highlights efforts being made at the Centre for Energy Research and Training (CERT) towards realizing this broad objective as far as possible. To this end CERT realizes the need for instituted safety measures to reflect significant, site-specific peculiar characteristics of any generic reactor types. Consequently, standard procedures for pre-startup, startup and shutdown of NIRR-1 (a miniature neutron source reactor - MNSR) have been reviewed to reflect our local conditions and peculiarities. The review has revealed the need to incorporate important steps that impact on overall safety of the facility. For instance an interlocking system is being considered between NIRR-1 startup on the one hand and mandatory pre-startup measures on the other. Also a procedure has been put in place that would facilitate rapid response in the event of a rod-stuck-at-full-withdrawal incident. Furthermore, a program of automation of important analysis and design calculations of MNSRs is going on. Emphases are also placed, and deliberate efforts are being made, to ensure that a working atmosphere prevails that would foster the correct attitudinal approach to matters of reactor safety. A regime of constant dialogue and discussions amongst operating personnel has been factored into the overall operational program. (author)

  4. Proceedings of the topical meeting on reactor physics and safety: Sessions 1-10. Volume 1

    International Nuclear Information System (INIS)

    Technical papers and invited lectures presented at the International Topical Meeting on Reactor Physics and Safety are presented. The sessions include a general session on Challenges in Reactor Physics and Safety. Together with sessions on conventional reactor physics topics, there are sessions on safety in nuclear design, dynamic behavior of reactors, degraded cores, research reactors and pressure vessel embrittlement. This conference is broad in scope and brings together experts from all over the fee world to present papers and exchange ideas on the reactor physics and safety aspects of nuclear reactors

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

    Energy Technology Data Exchange (ETDEWEB)

    Zeevaert, T.

    1998-07-01

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

  6. Steady-state thermal hydraulic analysis of the equilibrium core of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Steady-state thermal hydraulic analysis of Pakistan Research Reactor-1 (PARR-1) has been carried out. RELAP5/Mod 3.4 (a best-estimate system code) was employed. PARR-1 is a swimming pool type research reactor using MTR (Material Testing Reactor) type fuel. It uses low enriched uranium (<20%) fuel with light water flowing from top to bottom under gravity. Standard correlations were employed to compute various parameters, which include: coolant velocity distribution in the core; critical velocity; pressure drop; saturation temperature; temperature distribution in the core, OFI (onset of flow instability) and DNB (departure from nucleate boiling)

  7. Thermal hydraulic analysis of the IPR-R1 TRIGA reactor

    International Nuclear Information System (INIS)

    The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

  8. Synthesis of the study performed by the IRSN regarding safety options of the ATMEA1 reactor

    International Nuclear Information System (INIS)

    The ATMEA1 reactor is a new concept for next generation 1000 MWe pressurized water reactors. This report first recalls the context of the IRSN study of safety options regarding this reactor: a first phase for the examination of these options, and a second phase for the examination of the safety option report and complementary reports on topics identified in phase 1. The main conclusions of phase 1 are recalled, as well as the topics addressed during the several meetings. Then, it lists the significant issues addressed in the different reports associated with these meetings

  9. Measurement of radionuclides in the cooling system of upgraded Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    During the operation of Pakistan Research Reactor-1 (PARR-1) a variety of radionuclides are produced by the neutron activation of the materials in the core which include fuel, cladding, structural material, coolant and impurities in the coolant water. The fission products remain within the cladding and are not released except in the event of its rupture. To ensure that concentration of various radionuclides in the cooling system in within acceptable limits and reactor operation is radiologically safe, coolant water samples are collected from PARR-1 and analyzed by gamma spectrometric technique. The radionuclides detected and their concentrations measured in the pool water during reactor operation are presented and discussed

  10. Steady-state thermal hydraulic analysis of the equilibrium core of Pakistan research reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Bokhari, I.H. [SDTP User Group, Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore 45650, Islamabad (Pakistan)], E-mail: ihbokhari@yahoo.co.uk; Mahmood, T.; Chaudri, K.S. [SDTP User Group, Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore 45650, Islamabad (Pakistan)

    2007-10-15

    Steady-state thermal hydraulic analysis of Pakistan Research Reactor-1 (PARR-1) has been carried out. RELAP5/Mod 3.4 (a best-estimate system code) was employed. PARR-1 is a swimming pool type research reactor using MTR (Material Testing Reactor) type fuel. It uses low enriched uranium (<20%) fuel with light water flowing from top to bottom under gravity. Standard correlations were employed to compute various parameters, which include: coolant velocity distribution in the core; critical velocity; pressure drop; saturation temperature; temperature distribution in the core, OFI (onset of flow instability) and DNB (departure from nucleate boiling)

  11. IEA-R1 research reactor: operational life extension and considerations regarding future decommissioning

    International Nuclear Information System (INIS)

    The IEA-R1 reactor is a pool type research reactor moderated and cooled by light water and uses graphite and beryllium reflectors. The reactor is located at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), in the city of Sao Paulo, Brazil. It is the oldest research reactor in the southern hemisphere and one of the oldest of this kind in the world. The first criticality of the reactor was obtained on September 16, 1957. Given the fact that Brazil does not have yet a definitive radioactive waste repository and a national policy establishing rules for the spent fuel storage, the institutions which operate the research reactors for more than 50 years in the country have searched internal solutions for continued operation. This paper describes the spent fuel assemblies and radioactive waste management process for the IEA-R1 reactor and the refurbishment and modernization program adopted to extend its lifetime. Some considerations about the future decommissioning of the reactor are also discussed which, in my opinion, might help the operating organization to make decisions about financial, legal and technical aspects of the decommissioning procedures in a time frame of 10-15 years(author)

  12. 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...

  13. Recent historical changes on the Belgian Meuse

    International Nuclear Information System (INIS)

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

  14. Organ procurement after euthanasia: Belgian experience.

    Science.gov (United States)

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

    2009-03-01

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

  15. Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 1

    International Nuclear Information System (INIS)

    FR0 is a fast zero power reactor built for experiments in reactor physics. It is a split table machine containing vertical fuel elements. 120 kg of U235 are available as fuel, which is fabricated into metallic plates of 20 % enrichment. The control system comprises 5 spring-loaded safety elements and 3 + 1 elements for startup operations and power control. The reactor went critical in February 1964. The first assemblies studied were made up of undiluted fuel into a cylindrical and a spherical core, respectively, surrounded by a reflector made of copper. The present report describes some experiments made on these systems. Primarily, critical mass determinations, flux distribution measurements and studies of the conversion ratio are dealt with. The measured quantities have been compared with theoretical predictions using various transport theory programmes (DSN, TDC) and cross section sets. The experimental results show that the neutron spectrum in the copper reflector is softer than predicted, but apart from this discrepancy agreement with theory has generally been obtained

  16. TA-2 water boiler reactor decommissioning (Phase 1)

    International Nuclear Information System (INIS)

    Removal of external structures and underground piping associated with the gaseous effluent (stack) line from the TA-2 Water Boiler Reactor was performed as Phase I of reactor decommissioning. Six concrete structures were dismantled and 435 ft of contaminated underground piping was removed. Extensive soil contamination by 137Cs was encountered around structure TA-2-48 and in a suspected leach field near the stream flowing through Los Alamos Canyon. Efforts to remove all contaminated soil were hampered by infiltrating ground water and heavy rains. Methods, cleanup guidelines, and ALARA decisions used to successfully restore the area are described. The cost of the project was approximately $320K; 970 m3 of low-level solid radioactive waste resulted from the cleanup operations

  17. Metal fire implications for advanced reactors. Part 1, literature review

    International Nuclear Information System (INIS)

    Public safety and acceptance is extremely important for the nuclear power renaissance to get started. The Advanced Burner Reactor and other potential designs utilize liquid sodium as a primary coolant which provides distinct challenges to the nuclear power industry. Fire is a dominant contributor to total nuclear plant risk events for current generation nuclear power plants. Utilizing past experience to develop suitable safety systems and procedures will minimize the chance of sodium leaks and the associated consequences in the next generation. An advanced understanding of metal fire behavior in regards to the new designs will benefit both science and industry. This report presents an extensive literature review that captures past experiences, new advanced reactor designs, and the current state-of-knowledge related to liquid sodium combustion behavior

  18. Metal fire implications for advanced reactors. Part 1, literature review.

    Energy Technology Data Exchange (ETDEWEB)

    Nowlen, Steven Patrick; Radel, Ross F.; Hewson, John C.; Olivier, Tara Jean; Blanchat, Thomas K.

    2007-10-01

    Public safety and acceptance is extremely important for the nuclear power renaissance to get started. The Advanced Burner Reactor and other potential designs utilize liquid sodium as a primary coolant which provides distinct challenges to the nuclear power industry. Fire is a dominant contributor to total nuclear plant risk events for current generation nuclear power plants. Utilizing past experience to develop suitable safety systems and procedures will minimize the chance of sodium leaks and the associated consequences in the next generation. An advanced understanding of metal fire behavior in regards to the new designs will benefit both science and industry. This report presents an extensive literature review that captures past experiences, new advanced reactor designs, and the current state-of-knowledge related to liquid sodium combustion behavior.

  19. Advanced technologies for water cooled reactors 1990. Pt. 1

    International Nuclear Information System (INIS)

    The meeting was attended by 20 participants from 12 countries who reviewed and discussed the status and progress of national programmes on advanced water-cooled reactors and recommended to the Scientific Secretary a comprehensive programme for 1991/1992 which would support technology development programmes in IWGATWR Member States. This summary report outlines the activities of IWGATWR since its Second Meeting in June 1988 and main results of the Third Meeting

  20. Use of self powered neutron detectors in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    A survey of self-powered neutron detectors, SPND, which are used as part of the in-core instrumentation of nuclear reactors is presented. Measurements with Co and Er SPND's were made in the IEA-R1 reactor for determining the neutron flux distribution and the integral reactor power. Due to the size of the available detectors, the neutron flux distribution could not be obtained with accuracy. The results obtained in the reactor power measurements demonstrate that the SPND have the linearity and the quick response necessary for a reactor power channel. This work also presents a proposed design of a SPND using Pt as wire emissor. This proposed design is based in the experience gained in building two prototypes. The greatest difficulties encountered include materials and technology to perform the delicate weldings. (author)

  1. Continuous thermal balance monitoring for IEA-R1 nuclear research reactor power determination

    International Nuclear Information System (INIS)

    This research deals with thermal balance calculation for real time power level determination of IEA-R1 nuclear research reactor. It is also shown the development of a supervision software (Visual Basic) of operation parameters. The assembled data acquisition system allows data analysis during reactor operation, giving a reliable measurement of reactor power, and the organization of a data base allows a back-up surveillance of reactor operation whenever necessary. Results obtained from temperature and primary flow are shown in a continuous form and also the Data Base implementation for further studies and analysis of energy balance behavior of the many reactor components. Besides it is planned to manage N-16 activity measurement channel (monitoring) for comparison of acquired data results for thermal calculations. The results of this acquisition and related thermal balance calculations are shown in a continuous shape (On-Line) by means of windows operational system using Visual Basic VB6 software for development. (author)

  2. Radiometric analysis of the spent fuel pool water and reactor coolant of ET-RR.1

    International Nuclear Information System (INIS)

    This work aims at analysis of radioactivity levels in the water of spent fuel pool and reactor core of the Egyptian 2MW research reactor (ET-RR.1 at Inshas). Gamma spectrometric and laser fluorimetric analysis have been used for carrying out this study. The fission product 137Cs and activation product 60Co are found with very high concentration in the spent fuel storage pool water. Thirteen isotopes; La-140, Cr-51, Ba-140, I-131, Cs-137, Ce-144, Nb-95, Ce-141, Zr-95, Ru-103, Cs-134, Nd-147 and Zn-65 are identified in the reactor core water. However no radiological hazard resulted because the fission products are contained within the shielded reactor pool. The radioactivity released into the reactor coolant water is mainly controlled by the diffusion mechanism. (orig.)

  3. Evaluation of the trial design studies for an advanced marine reactor, (1)

    International Nuclear Information System (INIS)

    The trial design of three type reactors, semi-integrated, integrated and integrated (self-pressurized) type, was carried out in order to clarify the reactor type for the advanced marine reactor that would be developed for its realization in future and in order to extract its research and development theme. The trial design was carried and finished as for the three type reactors in same specifications in order to improve the following characteristics, small in size, light in weight, high in safety and reliability, and economic. In this report, a comparison and review of the following items are described as for the above three type reactors, (1) specifications, (2) shielding, (3) refueling, (4) in-service inspection, (5) analysis of the transients and accidents, (6) piping systems, (7) control systems, (8) dynamic analysis, (9) overall comparison, (10) research and development theme and theme for study in future. (author)

  4. Nuclear reactor and materials science research: Technical report, May 1, 1985-September 30, 1986

    International Nuclear Information System (INIS)

    Throughout the 17-month period of its grant, May 1, 1985-September 30, 1986, the MIT Research Reactor (MITR-II) was operated in support of research and academic programs in the physical and life sciences and in related engineering fields. The reactor was operated 4115 hours during FY 1986 and for 6080 hours during the entire 17-month period, an average of 82 hours per week. Utilization of the reactor during that period may be classified as follows: neutron beam tube research; nuclear materials research and development; radiochemistry and trace analysis; nuclear medicine; radiation health physics; computer control of reactors; dose reduction in nuclear power reactors; reactor irradiations and services for groups outside MIT; MIT Research Reactor. Data on the above utilization for FY 1986 show that the MIT Nuclear Reactor Laboratory (NRL) engaged in joint activities with nine academic departments and interdepartmental laboratories at MIT, the Charles Stark Draper Laboratory in Cambridge, and 22 other universities and nonprofit research institutions, such as teaching hospitals

  5. Visual inspections of the neutron absorber control rods of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    The Fuel Engineering Division at IPEN/CNEN-SP developed facilities for visual inspection of the IEA-R1 fuel elements and neutron absorbing control rod assemblies inside the research reactor pool. This work presents the method of visual inspection performed at IEA-R1 research reactor. These inspections were adopted to evaluate and to follow the state of the Ag-In-Cd control assemblies fabricated at CERCA in 1972 that remain in use at the reactor core. In 1998, 2000 and 20001, visual inspections were performed in these control rod assemblies, which the general conditions were evaluated. (author)

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2012-08-01

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

  8. Decontamination of the reactor pressure vessel and further internals and auxiliary systems in the German boiling water reactor Isar-1

    International Nuclear Information System (INIS)

    The German nuclear power plant ISAR 1 (KKI 1), a 878 MWe boiling water reactor of KWU design, was shut down on March 17th, 2011. With the objective to minimize the plants activity inventory accompanied by the reduction of contact dose rates of systems and components the project 'decontamination of the RPV incl. steam dryer and water separator and the connected auxiliary systems' was implemented in the first quarter of 2015. One major focus within the project was the specific in-situ decontamination of the steam dryer.

  9. Design evaluation system for class 1 component of fast breeder reactor plants

    International Nuclear Information System (INIS)

    The development of a new type of nuclear power plant called Fast Breeder Reactor has been greatly promoted of late by the Power Reactor and Nuclear Fuel Development Corporation (PNC) and others in hopes of replacing Light Water Reactors so far prevailing in Japan. Fast Breeder Reactor, unlike Light Water Reactor, is subjected to elevated temperature within the creep temperature range for long duration, thus requiring higher structural standards for reliability as well as for safety. In this connection, PNC has been conducting many years' research and development to establish reliable design methods based on an advanced analysis taking into consideration elevated temperature properties of materials, and finally worked out Structural Design Guide for Class 1 Components of the prototype of Fast Breeder Reactor in elevated temperature service. The POST-DS system in this paper has been developed as an design evaluation system based on the above design guide, by Mitsui Engineering and Shipbuilding Co., Ltd. since 1979 in accordance with a commission given by PNC. Using the results of the heat transfer analysis and stress analysis for Class 1 Components of Fast Breeder Reactor, this system can evaluate the following factors. 1) Primary stress limit, 2) Strain limit, 3) Creep Fatigue damage. (author)

  10. Irradiation experience of IPEN fuel at IEA-R1 research reactor

    International Nuclear Information System (INIS)

    IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U3 O8-Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

  11. Lifetime Neutron Fluence Analysis of the Ringhals Unit 1 Boiling Water Reactor

    OpenAIRE

    Kulesza Joel A.; Roudén Jenny; Green Eva-Lena

    2016-01-01

    This paper describes a neutron fluence assessment considering the entire commercial operating history (35 cycles or ∼ 25 effective full power years) of the Ringhals Unit 1 reactor pressure vessel beltline region. In this assessment, neutron (E >1.0 MeV) fluence and iron atom displacement distributions were calculated on the moderator tank and reactor pressure vessel structures. To validate those calculations, five in-vessel surveillance chain dosimetry sets were evaluated as well as material ...

  12. Spent fuel management plans for the FiR 1 Reactor

    International Nuclear Information System (INIS)

    The FiR 1-reactor, a 250 kW TRIGA reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. The final disposal site is situated in Olkiluoto, on the western coast of Finland. Olkiluoto is also one of the two nuclear power plant sites in Finland. In the new operating license of our reactor there is a special condition. We have to achieve a binding agreement between our Research Centre and either the domestic Nuclear Power Companies about the possibility to use the Olkiluoto final disposal facility for our spent fuel or US DOE about the return of our spent fuel back to USA. If we want to continue the reactor operation beyond the year 2006. the domestic final disposal is the only possibility. At the moment it seems to be reasonable to prepare to both possibilities: the domestic final disposal and the return to the USA offered by US DOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will decide, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNCT treatments will cover the costs. If the BNCT and other irradiations develop satisfactorily, the reactor can be kept in operation beyond the year 2006 and the domestic final disposal will be implemented. If, however, there is still lack of money, there is no reason to continue the operation of the reactor and the choice of US DOE alternative is natural. (author)

  13. Digital radiographic equipment in the Belgian dental office.

    Science.gov (United States)

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

    2005-01-01

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

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

    Science.gov (United States)

    Debbaut, Kevin; Krysinska, Karolina; Andriessen, Karl

    2014-01-01

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

  15. Can Belgian nuclear power stations work for longer?

    International Nuclear Information System (INIS)

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

  16. Uncertainty analysis on thermal hydraulics parameter of the IPR-R1 TRIGA research nuclear reactor

    International Nuclear Information System (INIS)

    Experimental studies have been performed in the IPR-R1 TRIGA Mark 1 Research Nuclear Reactor of CDTN/CNEN at Belo Horizonte (Brazil) to find out the temperature distribution as a function of reactor power, under steady-state conditions. During these experiments the reactor was set in many different power levels. These experiments are part of the research program, that have the main objective of commissioning the IPR-R1 reactor for routine operation at 250 k W. This paper presents the uncertainty analysis of the thermal-hydraulic experiments performed. The methodology used to evaluate the uncertainty propagation on the results was done based on the pioneering article of Kline and McClintock (1953), with the propagation of uncertainties based on the specification of uncertainties in various primary measurements. (author)

  17. Training reactor VR 1 at the Czech Technical University in Prague is operational

    International Nuclear Information System (INIS)

    Detailed information on the construction work of the VR 1 VRABEC training reactor is presented (building and assembling work, inactive tests) and its technical description is given, with emphasis on the conception of the device and the crucial parts of its technology. The basic parameters of the reactor are as follows: rated power output 1 kW(thermal); fuel of the IRT-2M type (36% 235U enrichment, imported from the former USSR); reactor vessels (pools) made of stainless steel, 2300 mm in diameter, 4720 mm high, wall thickness 15 mm, bottom thickness 20 mm; reactor shielding: 3000 mm water layer above the reactor core, lateral shielding: about 850 mm of water plus a 950 mm layer of special heavy concrete; working temperature inside the reactor, which is affected by the ambient temperature, is about 20 degC; reactor core cooling proceeds by natural convection; pressure is atmospheric; control system consists of 5 to 7 UR-70 type control rods distributed as follows: 3 scram rods, 2 control rods, 0 to 2 experimental rods; neutron source: Am-Be, 185 GBq. (Z.S.). 10 figs

  18. Sector emergency procedures for Pakistan Research Reactor-1 (PARR-1), Pinstech

    Energy Technology Data Exchange (ETDEWEB)

    Aziz, A.; Aslam, M.; Hasan, S.; Faruq, M.U.; Ahmad, B.

    1992-12-01

    Pakistan Research Reactor (PARR-1) has many operational safety features against any accidental failure of its control system. Nevertheless the risk, how-so-ever remote, of any accidental excursion that may lead to the release of radioactive material from the core resulting in the possible contamination of the premises and the environment and the exposure of the workers and the population cannot be ruled out. This report outlines the detailed procedures and individual and collective responsibilities and actions to be undertaken for meeting the emergency situation.

  19. FMDP reactor alternative summary report. Volume 1 - existing LWR alternative

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-07

    Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] are becoming surplus to national defense needs in both the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES&H) consequences if surplus fissile materials are not properly managed. This document summarizes the results of analysis concerned with existing light water reactor plutonium disposition alternatives.

  20. FMDP reactor alternative summary report. Volume 1 - existing LWR alternative

    International Nuclear Information System (INIS)

    Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] are becoming surplus to national defense needs in both the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES ampersand H) consequences if surplus fissile materials are not properly managed. This document summarizes the results of analysis concerned with existing light water reactor plutonium disposition alternatives

  1. Water treatment process in the JEN-1 Research Reactors

    International Nuclear Information System (INIS)

    The main characteristics and requirements which must be met with by waters to be used for nuclear reactors were studied paying attention separately both to those used in primary and secondary circuits as well as to the purification systems to be employed in each case. The experiments carried out for the initial pretreatment of water and the ion-exchange de ionization processes including a number of systems consisting of separated and mixed beds loaded with a variety of different commercially available resins are described. (Author) 24 refs

  2. Water treatment process in the JEN-1 Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Urgel, M.; Perez-Bustamante, J. A.; Batuecas, T.

    1965-07-01

    The main characteristics and requirements which must be met with by waters to be used for nuclear reactors were studied paying attention separately both to those used in primary and secondary circuits as well as to the purification systems to be employed in each case. The experiments carried out for the initial pretreatment of water and the ion-exchange de ionization processes including a number of systems consisting of separated and mixed beds loaded with a variety of different commercially available resins are described. (Author) 24 refs.

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

    International Nuclear Information System (INIS)

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

  4. Calculation of the main neutron parameters of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    The main neutron parameters of the research reactor IEA-R1 were calculated using computer programs to generate cross sections and criticality calculations. A calculation procedure based on the programs available in the Processing Center Data of IEA was established and centered in the HAMMER and CITATION system. A study was done in order to verify the validity and accuracy of the calculation method comparing the results with experimental data. Some operating parameters of the reactor, namely the distribution of neutron flux, the critical mass, the variation of the reactivity with the burning of fuel, and the dead time of the reactor were determined

  5. Thermal hydraulic analysis for upgradation of Pakistan Research Reactor-1 from 9 to 10 MW

    International Nuclear Information System (INIS)

    Thermal hydraulic aspects of Pakistan Research Reactor-1 have been studied to upgrade its power level from 9 to 10 MW. Standard computer codes and correlations were used to compute: pressure drop and flow through different channels of the core, coolant critical velocity beyond which fuel plates may collapse, temperature distribution in the core, heat fluxes at onset of nucleate boiling, onset of flow instability and departure from nucleate boiling. Natural convection cooling at low power was also analyzed. Results indicate that the cores have reasonably high safety margins and reactor power can be upgraded to 10 MW without compromising on reactor safety. (author)

  6. Fusion reactor technology studies. Final report for period August 1, 1972 - October 31, 1978

    International Nuclear Information System (INIS)

    Major accomplishments for the period August 1, 1972 - October 31, 1978 include the publishing of four comprehensive fusion reactor conceptual design studies; experimental studies in the areas of radiation damage, plasma-wall interactions, superconducting magnets and 14-MeV neutron cross sections; development of the concepts of carbon curtains and ISSEC's for use in fusion reactors; development of a neutron and gamma heating computer code, a radioactivity and afterheat computer code and a neutral transport computer code; and studies in the areas of RF heating for tokamaks and resource assessment for fusion reactors

  7. Auxiliary control system of the safety parameters for IPR-R1 reactor

    International Nuclear Information System (INIS)

    This paper deals with the description of the control of three cooling water parameters, as well as the percent power fraction of the nuclear research reactor IPR-R1. In order to keep the reactor in good operation conditions, a permanent and accurate control of the cooling water is needed. This is achieved through this system, which allows the simultaneous measurement of the water parameters such as: conductivity, temperature and the maximum and minimum water levels. The monitoring of a fourth parameter, the percent power fraction, is obtained through the control of the uncompensated ion chamber current and aims to avoid the operation of the reactor without running the cooling system. (Author)

  8. Auxiliary control system of the safety parameters for IPR-R1 reactor

    International Nuclear Information System (INIS)

    This paper deals with the description for the control of three cooling water parameters (conductivity, temperature and the maximum and minimum water levels) as well as the percent power fraction of the nuclear research reactor IPR-R1. In order to keep the reactor in good operation conditions, one permanent and accurate control of the cooling water is needed. The double monitoring of a fourth parameter, part of the original design, the percent power fraction, is obtained through the control of the uncompensated ion chamber current and aims to avoid the operation of the reactor without running the cooling system. (Author)

  9. Sensitivity analysis of the RELAP5 nodalization to IPR-R1 TRIGA research reactor

    International Nuclear Information System (INIS)

    The main aim of this work is to identify how much the code results are affected by code user in the choice of, for example, the number of thermal-hydraulic channels in a nuclear reactor nodalization. To perform this, two essential modifications were made on a previous validated nodalization for analysis of steady state and forced recirculation off transient in the IPR-R1 TRIGA research reactor. Experimental data were taken as reference to compare the behavior of the reactor for two different types of model. The results found highlight the necessity of sensitivity analysis to obtain the ideal simulation model of a system. (author)

  10. Fuel reactor modelling in chemical-looping combustion of coal: 1. model formulation

    OpenAIRE

    Abad Secades, Alberto; Gayán Sanz, Pilar; Diego Poza, Luis F. de; García Labiano, Francisco; Adánez Elorza, Juan

    2013-01-01

    A fundamental part of the reliability of the Chemical-Looping Combution system when a 13 solid fuel, such as coal, is fed to the reactor is based on the behaviour of the fuel reactor, which determines the conversion of the solid fuel. The objective of this work is to develop a model describing the fuel reactor in the Chemical–Looping Combustion with coal (CLCC) process. The model is used to simulate the performance of the 1 MWth CLCC rig built in the Technology University of Darmsta...

  11. IPR-R1 reactor power control by the gamma radiation of N16

    International Nuclear Information System (INIS)

    The IPR-R1 reactor power control is realized by the ion chambers use. The information that the chambers send to the control console are deformed due the control rods movements during the operation. With the purpose to eliminate these interferences, was installed close the reactor water cooling circuit, one power control auxiliary system using the detection of the N 16 formed in the water. this paper presents an analysis of the results and propose one complete project that permits the control of the radioactives nuclides localized in the reactor cooling water. (Author)

  12. Proposed design for the PGAA facility at the TRIGA IPR-R1 research reactor

    OpenAIRE

    Guerra, Bruno T; Jacimovic, Radojko; Menezes, Maria Angela BC; Leal, Alexandre S.

    2013-01-01

    Background This work presents an initial proposed design of a Prompt Gamma Activation Analysis (PGAA) facility to be installed at the TRIGA IPR-R1, a 60 years old research reactor of the Centre of Development of Nuclear Technology (CDTN) in Brazil. The basic characteristics of the facility and the results of the neutron flux are presented and discussed. Findings The proposed design is based on a quasi vertical tube as a neutron guide from the reactor core, inside the reactor pool, 6 m below t...

  13. Fast breeder reactors: experience and trends. V. 1

    International Nuclear Information System (INIS)

    The IAEA Symposium on ''Fast Breeder Reactors: Experience and Future Trends'' was held, at the invitation of the Government of France, in Lyons, France, on 22-26 July 1985. It was hosted by the French Commissariat a l'energie atomique and Electricite de France. The purpose of the Symposium was to review the experience gained so far in the field of LMFBRs, taking into account the constructional, operational, technological, economic and fuel cycle aspects, and to consider the developmental trends as well as the international co-operation in fast breeder reactor design and utilization. The Symposium presentations were divided into sessions devoted to the following topics: Experience of LMFBR construction and operation and resultant development strategies (6 papers); LMFBR plant startup and commissioning tests and general behaviour (8 papers); Core performance experience for high burnup and core design trends (8 papers); Experience and trends in the LMFBR fuel cycle (4 papers); Core design and behaviour (3 papers); Fuels and materials (7 papers). A separate abstract was prepared for each of these papers

  14. Application of stable adaptive schemes to nuclear reactor systems, (1)

    International Nuclear Information System (INIS)

    Parameter identification and adaptive control schemes are presented for a point reactor with internal feedbacks which lead to the nonlinearity of the overall system. Both are shown stable with new representation of the system, which corresponds to the nonminimal system representation, in the vein of the Model Reference Adaptive System (MRAS) via the Lyapunov's method. For the sake of the parameter identification, model parameters can be adjusted adaptively as soon as measurements start, while plant parameters can also adaptively be compensated through control input to reduce the output error between the model and the plant for the case of the adaptive control. In the case of the adaptive control, control schemes are presented for two cases, the case of the unknown decay constant of the delayed neutron and the case of the known constant. The adaptive control scheme for the latter case is shown extremely simpler than that for the former. Furthermore, when plant parameters vary slowly with time, computer simulations show that the proposed adaptive control scheme works satisfactorily enough to stabilize an unstable reactor and that it does even in the noise with small variance. (auth.)

  15. Fragmentation of suddenly heated liquids in ICF reactors. Revision 1

    International Nuclear Information System (INIS)

    Fragmentation of free liquids in Inertial Confinement Fusion reactors could determine the upper bound on reactor pulse rate because increased surface area will enhance the cooling and condensation of coolant ablated by the fusion x rays. Relaxation from the suddenly (neutron) heated state will move a liquid into the negative pressure region under the liquid-vapor P-V dome. The resulting expansion in a diverging geometry will hydrodynamically force the liquid to fragment, with vapor then forming from the new surfaces to fill the cavities. An energy minimization model is used to determine the fragment size that produces the least amount of non-fragment-center-of-mass energy; i.e., the sum of the surface and dilational kinetic energies. This model predicts fragmentation dependence on original system size and amount of isochoric heating as well as liquid density, Grueneisen parameter, surface tension, and sound speed. A two dimensional molecular dynamics code was developed to test the model at a microscopic scale for the Lennard-Jones fluid with its two adjustable constants chosen to represent lithium

  16. Summary of session 1: 'Advanced fuel cycles and reactor concepts'

    International Nuclear Information System (INIS)

    Full text: During the opening session of the Scientific Forum presentations were made by the IAEA Director General, Mohamed ElBaradei, Mr. Carlo Rubbia from Italy, and by the Chairman of the Scientific Forum, Mr. B. Bigot from France. Mr. A. Kakodkar from India was the moderator of the session The audience included some 180 participants. Four keynote speakers from the UK, Argentina, Japan and India as well as one panelist from the USA contributed to the first session. The highlights can be summarized as follows: Nuclear energy as an emission free energy source is indispensable for sustainable development. The importance of continuous R and D in support of innovative reactors and fuel cycles was stressed. The overall goal for these technologies includes better uranium resource utilization and improved waste management strategies. Moreover, the development of accelerator-driven systems for transmutation and energy production is regarded as an important long-term option. In this respect the closure of the fuel cycle with fast reactors is considered to be essential. In meeting these objectives, a focus on economics, proliferation resistance, and safety is paramount. The transition to innovative nuclear energy systems from current systems must be gradual via a combination of evolutionary and innovative technologies. (author)

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Wielemans, Willy

    1988-01-01

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

  19. Recuperation of the energy released in the G-1, an air-cooled graphite reactor core

    International Nuclear Information System (INIS)

    The CEA (in his five-year setting plan) has objective among others, the realization of the two first french reactors moderated with graphite. The construction of the G-1 reactor in Marcoule, first french plutonic core, is achieved so that it will diverge in the beginning of 1956 and reach its full power in the beginning of the second semester of the same year. In this report we will detail the specificities of the reactor and in particular its cooling and energy recuperation system. The G-1 reactor being essentially intended to allow the french technicians to study the behavior of an energy installation supply taking its heat in a nuclear source as early as possible. (M.B.)

  20. Report on safety related occurrences and reactor trips July 1, 1977 - December 31, 1977

    International Nuclear Information System (INIS)

    This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1977 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 48 safety related occurrences and 49 reactor trips have been reported to the Nuclear Power Inspectorate. Included is also one incident June 21 in Barsebaeck 2 which was not included in the last compilation of occurrences. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips have increased nearly 30% since the last period. Those occurred during power operation however, were less. More than 50% of the reactor trips happened in the shutdown condition. The fact that even small deviations from prescribed operation result in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences withou02068NRM 0000169 450

  1. Report on safety related occurrences and reactor trips July 1, 1979 - December 31, 1979

    International Nuclear Information System (INIS)

    This is a report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1979 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 76 safety related occurrences and 27 reactor trips have been reported to the Nuclear Power Inspectorate. It is to the greatest extent conventional components such as valves and pumps which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant system and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips are normal. The average value for these 6 months is 4.5 trips/unit. Approximetely one half of the reactor trips happened at zero or very low power operation. The fact that even small deviations from prescribed operation result in an automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences without safety significance. (author)

  2. New digital control system for the operation of the Colombian research reactor IAN-R1

    International Nuclear Information System (INIS)

    En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

  3. FiR 1 reactor in service for boron neutron capture therapy (BNCT) and isotope production

    International Nuclear Information System (INIS)

    The FiR 1 reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose for the existence of the reactor is now the Boron Neutron Capture Therapy (BNCT), but FiR 1 has also an important national role in providing local enterprises and research institutions in the fields of industrial measurements, pharmaceuticals, electronics etc. with isotope production and activation analysis services. In the 1990's a BNCT treatment facility was built at the FiR 1 reactor located at Technical Research Centre of Finland. A special new neutron moderator material FluentalTM (Al+AlF3+Li) developed at VTT ensures the superior quality of the neutron beam. Also the treatment environment is of world top quality after a major renovation of the whole reactor building in 1997. Recently the lithiated polyethylene neutron shielding of the beam aperture was modified to ease the positioning of the patient close to the beam aperture. Increasing the reactor power to 500 kW would allow positioning of the patient further away from the beam aperture. Possibilities to accomplish a safety analysis for this is currently under considerations. Over thirty patients have been treated at FiR 1 since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization, Boneca Corporation. Currently three clinical trial protocols for tumours in the brain as well as in the head and neck region are recruiting patients. (author)

  4. Assessment of a RELAP5 model for the IPR-R1 Triga research reactor

    International Nuclear Information System (INIS)

    RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants. However, several current investigations have shown that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research systems with good predictions. In this way, as a contribution to the assessment of RELAP5/3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed by a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open-pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data and also calculation data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code were considered in the process of the model validation. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual reactor behavior in good agreement with the available data. (author)

  5. ENEA TRIGA RC-1 reactor spent fuel elements shipment to the USA

    International Nuclear Information System (INIS)

    TRIGA Mark II reactor of ENEA's Casaccia research Center (in Italy named RC-1) reached first criticality in 1960. In more than thirty years of operation, 1 MW reactor core has been modified many times for fuel elements burn-up optimization. Till now, because of achieved maximum burn-up, 146 fuel elements have been definitively removed from reactor core and transferred to the hot storages in reactor pool (5 racks around reactor vessel) and in the reactor room (pits). The activities planning, the organizing aspect study, the analysis and valuations both nuclear safety and radioprotection have been suitable for the TRIGA RC-1 fuel element shipment. Infact, no operative anomaly is appeared respect the approved procedures. Personnel engagement has been as expectations and the personnel absorbed gamma dose resulted negligible. Finally, the NAC disposable narrow time (only one week at the end of July) has not produced heavy organization problems but it has been a strong goad per all operative structures involved in the TRIGA RC-1 elements shipment

  6. Project management plan for the 105-C Reactor interim safe storage project. Revision 1

    International Nuclear Information System (INIS)

    In 1942, the Hanford Site was commissioned by the US Government to produce plutonium. Between 1942 and 1955, eight water-cooled, graphite-moderated reactors were constructed along the Columbia River at the Hanford Site to support the production of plutonium. The reactors were deactivated from 1964 to 1971 and declared surplus. The Surplus Production Reactor Decommissioning Project (BHI 1994b) will decommission these reactors and has selected the 105-C Reactor to be used as a demonstration project for interim safe storage at the present location and final disposition of the entire reactor core in the 200 West Area. This project will result in lower costs, accelerated schedules, reduced worker exposure, and provide direct benefit to the US Department of Energy for decommissioning projects complex wide. This project sets forth plans, organizational responsibilities, control systems, and procedures to manage the execution of the Project Management Plan for the 105-C Reactor Interim Safe Storage Project (Project Management Plan) activities to meet programmatic requirements within authorized funding and approved schedules. The Project Management Plan is organized following the guidelines provided by US Department of Energy Order 4700.1, Project Management System and the Richland Environmental Restoration Project Plan (DOE-RL 1992b)

  7. Ageing Management Programme for the IEA-R1 Reactor in São Paulo, Brazil

    International Nuclear Information System (INIS)

    IEA-R1 is a swimming pool type reactor. It is moderated and cooled by light water and uses graphite and beryllium as reflector elements. First criticality was achieved on 16 September 1957, and the reactor is currently operating at 4.0 MW on a 64 h per week cycle. In 1996, a reactor ageing study was established to determine general deterioration of systems and components such as cooling towers, secondary cooling system, piping, pumps, specimen irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation, and safety system. The basic structure of the reactor from the original design has been maintained, but several improvements and modifications have been made over the years to various components, systems and structures. During the period 1996–2005 the reactor power was increased from 2 MW to 5 MW and the operational cycle from 8 h per day for 5 days a week to 120 h continuous per week, mainly to increase production of 99Mo. Prior to increasing reactor power, several modifications were made to the reactor system and its components. Simultaneously, a vigorous ageing management, inspection and modernization programme was put in place

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

    OpenAIRE

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

    2003-01-01

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

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

    OpenAIRE

    Adja Awa Sissoko; Bart Van den Cruyce

    2009-01-01

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

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

    OpenAIRE

    Sven H. De Cleyn

    2014-01-01

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

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

    OpenAIRE

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

    2010-01-01

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

  12. 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.

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

    OpenAIRE

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

    2015-01-01

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

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

    OpenAIRE

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

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

  16. RPV-1: a first virtual reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    The presented work was aimed at building a first VTR (virtual test reactor) to simulate irradiation effects in pressure vessel steels of nuclear reactor. It mainly consisted in: - modeling the formation of the irradiation induced damage in such steels, as well as their plasticity behavior - selecting codes and models to carry out the simulations of the involved mechanisms. Since the main focus was to build a first tool (rather than a perfect tool), it was decided to use, as much as possible, existing codes and models in spite of their imperfections. - developing and parameterizing two missing codes: INCAS and DUPAIR. - proposing an architecture to link the selected codes and models. - constructing and validating the tool. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or transmit data. A user friendly Python interface facilitates the running of the simulations and the visualization of the results. RPV-1 relies on many simplifications and approximations and has to be considered as a prototype aimed at clearing the way. According to the functionalities targeted for RPV-1, the main weakness is a bad Ni and Mn sensitivity. However, the tool can already be used for many applications (understanding of experimental results, assessment of effects of material and irradiation conditions,....). (O.M.)

  17. Determination of gamma dose and neutron fluence during start-up of the Greifswald-1 reactor

    International Nuclear Information System (INIS)

    During start-up of the Greifswald-1 reactor gamma and neutron radiation was measured using activation probes and thermoluminescent detectors which provided more accurate results than colorimetric dosemeters and solid state track detectors. A correlation was found between the n,γ field intensity and reactor power. The spatial distribution of the gamma dose and neutron fluence resulted in corresponding values. The spectral fluence distribution confirmed the existence of a soft neutron spectrum

  18. Summary Report of Commercial reactor Criticality Data for Three Mile Island Unit 1

    International Nuclear Information System (INIS)

    The objective of the ''Summary Report of Commercial Reactor Criticality Data for Three Mile Island Unit I'' is to present the CRC data for the TMI-1 reactor. Results from the CRC evaluations will support the development and validation of the neutronics models used for criticality analyses involving commercial spent nuclear fuel. These models and their validation are discussed in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000)

  19. Core calculations for the upgrading of the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Adimir dos; Perrotta, Jose A.; Bastos, Jose Luis F.; Yamaguchi, Mitsuo; Umbehaun, Pedro E. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: asantos@net.ipen.br; perrotta@net.ipen.br; mitsuo@net.ipen.br

    1998-07-01

    The IEA-R1 Research Reactor is a multipurpose reactor. It has been used for basic and applied research in the nuclear area, training and radioisotopes production since 1957. In 1995, the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) took the decision to modernize and upgrade the power from 2 to 5 MW and increase the operational cycle. This work presents the design requirements and the calculations effectuated to reach this goal. (author)

  20. Neutronic analysis for upgradation of Pakistan Research Reactor-1 from 9 to 10 MW

    International Nuclear Information System (INIS)

    In connection with upgradation of PARR-1 from 9 to 10 MW, the effects of enhanced power on various neutronic parameters have been investigated. These include: neutron energy spectrum, absolute and relative flux, control rod of worth and shutdown margin, fission product poisons, kinetic parameters, temperature coefficients of reactivity, fuel burnup and core excess reactivity. Results indicate that the reactor operation at 10 MW will improve neutron flux levels without compromising on reactor safety. (author)

  1. Determination of the theoretical and experimental zero-power frequency response of Ghana Research Reactor-1

    International Nuclear Information System (INIS)

    The frequency response measurements of a reactor at low power help in determining the kinetic parameters of a reactor and ultimately in investigating its stability with respect to small perturbations in reactivity. In this report, we present the results of the zero-power frequency response measurements of GHARR-1 by rod method and its analytical analogue. The comparison in calculated and measured values is reasonably good in the frequency range used (author)

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

    International Nuclear Information System (INIS)

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

  3. Non-Power Reactor Operator Licensing Examiner Standards. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-01

    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience.

  4. Non-Power Reactor Operator Licensing Examiner Standards. Revision 1

    International Nuclear Information System (INIS)

    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience

  5. Neutron spectrometric evaluations in the Argentine research reactor RA-1

    International Nuclear Information System (INIS)

    Full text: The determination of the quantities dose equivalent H*(10) and personal dose equivalent Hp(10) in mixed field (n,γ) needs the knowledge of the related spectrum. In order to fulfill this aim spectrometer system has been built based on the combination of polyethylene spheres of different diameters (Bonner Spheres System-BSS) and a He3 proportional counter detector sensitive to thermal neutrons. The detector is located in the geometrical centre of each of the spheres and has an associated electronics with a charge preamplifier, an amplifier and a multichannel system that allows the outgoing spectrum analysis. In order to determine the neutron spectrum a deconvolution method is applied based on the LOUHI82 code. In this work are shown the spectra and the related values of H*(10) that have been got in five places of the reactor and in the command room with the BSS. (author)

  6. 1st International School of Fusion Reactor Technology "Ettore Majorana"

    CERN Document Server

    Knoepfel, Heinz; Safety, Environmental Impact and Economic Prospects of Nuclear Fusion

    1990-01-01

    This book contains the lectures and the concluding discussion of the "Seminar on Safety, Environmental Impact, and Economic Prospects of Nuclear Fusion", which was held at Erice, August 6-12, 1989. In selecting the contributions to this 9th meeting held by the International School of Fusion Reactor Technology at the E. Majorana Center for Scientific Cul­ ture in Erice, we tried to provide a comprehensive coverage of the many interre­ lated and interdisciplinary aspects of what ultimately turns out to be the global acceptance criteria of our society with respect to controlled nuclear fusion. Consequently, this edited collection of the papers presented should provide an overview of these issues. We thus hope that this book, with its extensive subject index, will also be of interest and help to nonfusion specialists and, in general, to those who from curiosity or by assignment are required to be informed on these as­ pects of fusion energy.

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

    International Nuclear Information System (INIS)

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

  8. FIR 1 reactor in service for boron neutron capture therapy (BNCT) and isotope production

    International Nuclear Information System (INIS)

    Full text: The FIR 1-reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose for the existence of the reactor is now the Boron Neutron Capture Therapy (BNCT). The BNCT work dominates the current utilization of the reactor: three or four days per week are reserved for BNCT purposes and the rest for other purposes such as isotope production and neutron activation analysis. In the 1990's a BNCT treatment facility was build at the FiR1 reactor located at Technical Research Centre of Finland. A special new neutron moderator material FluentalTM (Al+AlF3+Li) developed at VTT ensures the superior quality of the neutron beam. Also the treatment environment is of world top quality. The ground floor of the reactor hall was provided with a new entrance, easily accessible by any patient vehicle, a radio therapy control room and rooms for patient preparation and laboratories. The top of the reactor tank was separated from the reactor hall in order to confine contamination in case of a leakage from irradiation samples or fuel elements. The ventilation of the building, emergency power supply system, heat exchangers and the secondary cooling circuit of the reactor including cooling towers were completely redesigned and rebuilt. The expenditure of designing and accomplishing the construction work described was about 4 million euros. The costs were partly financed with venture capital via Radtek Ltd., particularly established for this enterprise. Close to thirty patients have been treated at FiR 1 since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization, Boneca Corporation. VTT as the reactor operator has a long term contract with the Boneca Corp. to provide the facility and irradiation services for the patient treatments. The BNCT facility has been licensed for clinical use and is being surveyed by several national public health authorities including the Finnish Nuclear and Radiation Safety

  9. Fuel management methodology upgrade of Thai Research Reactor (TRR-1/M1) using SRAC computer code

    International Nuclear Information System (INIS)

    This paper presents the effort to upgrade the fuel management methodology of Thai Research Reactor-1/Modification 1 (TRR-1/M1) which is currently under responsibility of Thailand Institute of Nuclear Technology (TINT). The more advanced SRAC computer code is being introduced to replace the TRIGAP computer code for the fuel management calculation of TRR-1/M1. With the new methodology, the hexagonal lattices of TRR-1/M1 can be modeled without approximating the lattices into cylindrical rings as performed by the TRIGAP computer code. In addition, the SRAC computer code is able to provide pin-wise results such as normalized power distribution which is unable to obtain by the TRIGAP computer code. Also, the paper compares the core excess reactivity of core loading 1 and core loading 2 calculated by SRAC computer code with the measurement data from the operation log book. The comparison shows good agreement between the calculated values and measured values. With the promising result, the SRAC computer code is expected to be employed as the usual fuel management methodology for TRR-1/M1 in the near future. (author)

  10. Inspection and replacement of baffle assembly screws inside American reactor vessels

    International Nuclear Information System (INIS)

    The baffle assembly inside the vessel of a 900 MWe reactor designed by Framatome, is made up of 44 plates fixed on 8 horizontal supports by a system of about 1000 screws. These plates undergo high neutron flux and the problem of screw cracking appeared at the end of the eighties in the first-generation reactors. The first operation on a large scale concerning the screws of a Westinghouse type reactor, was performed on the Tihange-1 power plant where Framatome controlled 960 screws and replaced 91. In 1997 as a consequence of the Belgian and French feedback experience, American plant operators launched a vast program of preventive actions: material analysis, inspection of baffle plate screws and replacement of defective screws. This program was held in cooperation with EPRI (electric power research institute) and under the control of NRC (nuclear regulatory commission). Framatome Technologies Inc (FTI) was in charge of the in-situ inspection and replacement of the screws. FTI designed special tools and equipment adapted to the 2-loop American reactors but the basis ideas were those applied on the Tihange reactor. The successful experience of FTI has allowed the firm to be commissioned for 6 2-loops American reactors. (A.C.)

  11. Reactor Physics

    International Nuclear Information System (INIS)

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

  12. Reactor Physics

    International Nuclear Information System (INIS)

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

  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. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  15. Neutron field for activation experiments in horizontal channel of training reactor VR-1

    International Nuclear Information System (INIS)

    The experimental channels of nuclear reactors often serve for nuclear data measurement and validation. The dosimetry-foils activation technique was employed to measure neutron field parameters in the horizontal radial channel of the training reactor VR-1, and to test the possibility of using the reactor for scientific purposes. The reaction rates, energy spectral indexes, and neutron spectrum at several irradiation positions of the experimental channel were determined. The experimental results show the feasibility of the radial channel for irradiating experiments and open new possibilities for data validation by using this nuclear facility. - Highlights: • Neutron activation analysis of various samples. • Neutron spectrometry and gamma-spectrometry. • Study of keff for various types of reactor core

  16. Computer-aided testing and operational aids for PARR-1 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ansari, S.A. (Nuclear Engineering Div., Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (PK))

    1990-06-01

    The utilization of the plant computer of Pakistan Research Reactor (PARR-1) for automatic periodic testing of nuclear instrumentation in the reactor is described. Computer algorithms have been developed for on-line acquisition and real-time processing of nuclear channel signals. The mean value, standard deviation, and probability distributions of nuclear channel signals are obtained in real time, and the computer generates a warning message if the signal error exceeds the maximum permissible error. In this way a faulty channel is automatically identified. Other real-time algorithms are also described that assist the operator in safe reactor operation by automatically computing approach-to-criticality during reactor start-up and the control rod worth determination.

  17. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease. (author)

  18. Computer-aided testing and operational aids for PARR-1 nuclear reactor

    International Nuclear Information System (INIS)

    The utilization of the plant computer of Pakistan Research Reactor (PARR-1) for automatic periodic testing of nuclear instrumentation in the reactor is described. Computer algorithms have been developed for on-line acquisition and real-time processing of nuclear channel signals. The mean value, standard deviation, and probability distributions of nuclear channel signals are obtained in real time, and the computer generates a warning message if the signal error exceeds the maximum permissible error. In this way a faulty channel is automatically identified. Other real-time algorithms are also described that assist the operator in safe reactor operation by automatically computing approach-to-criticality during reactor start-up and the control rod worth determination

  19. Report on the safety related occurrences and reactor trips July 1, 1980 - December 31, 1980

    International Nuclear Information System (INIS)

    This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1981 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarhamn 1 and 2, Ringhals 1, 2 and 3 and Forsmark 1 and 2. During this period of 6 months 88 safety related occurrences and 51 reactor trips have been reported to the Nuclear Power Inspectorate. As earlier experiences have shown it is to the greatest extent conventional components such as valves and pumps which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant system and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips are normal. The average value for these 6 months is 2,3 trips/unit if one looks at the 6 units in commercial operation. As can be expected, Ringhals 3 and Forsmark 1 have had significantly more reactor trips. These units have been in the start up phase during this period, which includes different transient and trip tests. Forsmark 2, beeing in its hot functional test period, has not yet been the subject of many of these tests. This is reflected in very few incidents and reactor trips. The fact that even small deviations from prescribed operation result in an automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences without safety significance. (author)

  20. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    International Nuclear Information System (INIS)

    This report is a final culmination of activities funded through the Department of Energy's (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher's workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead

  1. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.

  2. A sipping test system to the Triga Mark I IPR-R1 reactor

    International Nuclear Information System (INIS)

    The IPR-R1 TRIGA MARK I Research Reactor of the Nuclear Technology Development Centre (CDTN/CNEN-MG) is a tank type reactor of General Atomic Company that has been operating since 1960 at a power of 100 kW. At present there are 63 fuel rods at the reactor core (58 aluminum cladding, 5 stainless steel and 1 stainless steel instrumented). The oldest fuel elements are made with aluminum alloy and the new ones from stainless steel. Some of the old fuel rods present some spots along their lateral fuel plates. These spots are originated by galvanic corrosion between the fuel cladding and the aluminum core grid. To provide an ageing program to the reactor, a sipping tests system will be performed with the reactor fuel. The system intends evaluate the possible presence of 137Cs leaking rate. This work presents the system, the procedure and methodology that will is used to perform the sipping tests with the fuel rods at the reactor core. The results obtained for the 137Cs sipping water activity for some fuel assembly, if any, will be evaluated with the system in operation. A correlation between the possible corrosion and the activity values measured will be realized. (author)

  3. Waste management for spent resin from Reactor Experimental Chileno number-sign 1 (RECH-1)

    International Nuclear Information System (INIS)

    A strategy is reported to find a waste form for temporary storage of spent resin arising from Reactor Experimental Chileno No. 1, RECH-1, according to radioactive waste management principles. As a first step, activity levels and contributions from long-lived fission products is obtained. The method that is developed is used to quantify the radioisotopes present in the resin to be allowed to follow them in the radioactive waste studies. As the activity is low, it is possible to find a way to dispose of large quantities of resin per drum. Two different processes were investigated: a mixtures technique and the double container immobilization method. Both were tested at laboratory and pilot plant scales. Results indicate that the double container method achieves great savings in economy of the management of spent resins

  4. Assessment of the radiological impact and associated risk to non-human biota from routine liquid discharges of the Belgian nuclear power plants

    International Nuclear Information System (INIS)

    We performed an Environmental Risk Assessment (ERA) to evaluate the impact on non-human biota from liquid radioactive effluents discharged by the Belgian Nuclear Power Plants (NPPs) of Doel and Tihange. A deterministic risk assessment for aquatic and terrestrial ecosystems was performed using the ERICA tool and applying the ERICA screening value of 10 μGy.h-1. The ERA was performed for the radioactive discharge limits and for the actual releases (maxima and averages over the last 10 years, 1999-2008). All ERICA reference organisms were considered and depending on the assessment situation, additional reference organisms were included in the analysis. It can be concluded that the current discharge limits for the Belgian NPPs do not result in significant risks to the aquatic and terrestrial environment and that the actual discharges, which are a fraction of the liquid discharge limits, are unlikely to harm the environment. (authors)

  5. IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

    International Nuclear Information System (INIS)

    Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN2 test, Source LH2-H2O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface

  6. FIR 1 reactor and the plans for the spent fuel management

    International Nuclear Information System (INIS)

    Full text: The FIR 1-reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The BNCT work dominates the current utilization of the reactor: The weekly schedule allows still one or two days for other purposes such as isotope production and neutron activation analysis. The nuclear waste management plan of our reactor describes among others the methods, the schedule and the cost estimate of the whole spent fuel management procedure starting from the removal of the fuel from the reactor core and ending to the final disposal in Olkiluoto, the final disposal site on the western coast of Finland. The Finnish Parliament ratified in May 2001 the decision in principle on the final disposal facility for spent fuel. The final disposal facility is supposed to be in operation in 2020. The operation license of our reactor will expire in 2011. It is very probable that there will be certain waiting time from the shut down of the reactor to the opening of the final disposal facility. Therefore there have to be a sufficient interim storage for the spent fuel before the transportation to the final disposal facility. In addition to the domestic final disposal solution there is still the USDOE alternative available until 2006. In the current operating license of the reactor there is a special condition. We have to achieve a binding agreement, latest in 2005, between our Research Centre and either the domestic Nuclear Power Companies about the possibility to use the Olkiluoto final disposal facility for our spent fuel or USDOE about the return of our spent fuel back to USA. If we want to continue the reactor operation beyond the year 2006, the domestic final disposal is the only possibility It is reasonable to be prepared to both possibilities: the domestic final disposal and the return to the USA offered by USDOE. Because the cost estimates of the both possibilities are on the same order of

  7. The management system evolution of research nuclear reactor IEA-R1

    International Nuclear Information System (INIS)

    The IEA-R1 is the first research nuclear reactor in Brazil and since its inauguration in 1957, has been regularly used mainly for R and D, teaching and production of some radioisotopes for medical and other purpose. Until 1999 the IEA-R1 reactor adopted a Quality Assurance Program based on the Brazilian regulatory body standard (Brazilian Nuclear Energy Commission - CNEN) CNEN NN 1.16 and the IAEA Guide SS 50 CQ to control quality and safety requirements, quality procedures and records. In 2001 the Research Reactor Center (CRPq) has began to implement a Quality Management System (QMS) focused on 'Operation and Maintenance of the IEA-R1 Research Reactor and Irradiation Services'. In 2002 this facility obtained its first NBR ISO 9001:2000 certification on this scope. The present work relates the stages involving the implementation of QMS of IEA-R1 reactor since it started operation until now, reporting the mainly difficulties and results obtained. (author)

  8. Thirtieth anniversary of reactor accident in A-1 Nuclear Power Plant Jaslovske Bohunice

    International Nuclear Information System (INIS)

    The facts about reactor accidents in A-1 Nuclear Power Plant Jaslovske Bohunice, Slovakia are presented. There was the reactor KS150 (HWGCR) cooled with carbon dioxide and moderated with heavy water. A-1 NPP was commissioned on December 25, 1972. The first reactor accident happened on January 5, 1976 during fuel loading. This accident has not been evaluated according to the INES scale up to the present time. The second serious accident in A-1 NPP occurred on February 22, 1977 also during fuel loading. This INES level 4 of reactor accident resulted in damaged fuel integrity with extensive corrosion damage of fuel cladding and release of radioactivity into the plant area. The A-1 NPP was consecutively shut down and is being decommissioned in the present time. Both reactor accidents are described briefly. Some radioecological and radiobiological consequences of accidents and contamination of area of A-1 NPP as well as of Manivier Canal and Dudvah River as result of flooding during the decommissioning are presented (authors)

  9. Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

    Energy Technology Data Exchange (ETDEWEB)

    Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

    2006-01-01

    The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

  10. Small and medium power reactors: project initiation study, Phase 1

    International Nuclear Information System (INIS)

    In conformity with the Agency's promotional role in the peaceful uses of nuclear energy, IAEA has provided, over the past 20 years, assistance to Member States, particularly developing countries, in planning for the introduction of nuclear power plants in the Small and Medium range (SMPR). However these efforts did not produce any significant results in the market introduction of these reactors, due to various factors. In 1983 the Agency launched a new SMPR Project Initiation Study with the objective of surveying the available designs, examining the major factors influencing the decision-making processes in Developing Countries and thereby arriving at an estimate of the potential market. Two questionnaires were used to obtain information from possible suppliers and prospective buyers. The Nuclear Energy Agency of OECD assisted in making a study of the potential market in industrialized countries. The information gained during the study and discussed during a Technical Committee Meeting on SMPRs held in Vienna in March 1985, along with the contribution by OECD-NEA is embodied in the present report

  11. SCALE-4 analysis of pressurized water reactor critical configurations. Volume 1: Summary

    International Nuclear Information System (INIS)

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original fresh composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized water reactors (PWR). The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Each of the five volumes comprising this report provides an overview of the methodology applied. Subsequent volumes also describe in detail the approach taken in performing criticality calculations for these PWR configurations: Volume 2 describes criticality calculations for the Tennessee Valley Authority's Sequoyah Unit 2 reactor for Cycle 3; Volume 3 documents the analysis of Virginia Power's Surry Unit 1 reactor for the Cycle 2 core; Volume 4 documents the calculations performed based on GPU Nuclear Corporation's Three Mile Island Unit 1 Cycle 5 core; and, lastly, Volume 5 describes the analysis of Virginia Power's North Anna Unit 1 Cycle 5 core. Each of the reactor-specific volumes provides the details of calculations performed to determine the effective multiplication factor for each reactor core for one or more critical configurations using the SCALE-4 system; these results are summarized in this volume. Differences between the core designs and their possible impact on the criticality calculations are also discussed. Finally, results are presented for additional analyses performed to verify that solutions were sufficiently converged

  12. Power upgrade and conversion of the Colombia R-1 reactor to TRIGA-Leu fuel

    International Nuclear Information System (INIS)

    The IAN-R1 reactor was furnished to the Government of Colombia under the 'Atoms for Peace' program by the U.S. Government. The reactor was constructed by Lockheed Aircraft Corporation and achieved initial critically at the Institute for Nuclear Studies in Bogota on 20 January, 1965. The reactor core consists of aluminum clad MTR-type plate fuel elements containing fully-enriched uranium (HEU). The reactor was initially designed to operate at a steady-state power level of 20-kW but has operated at 30-kW for the past several years. The principal applications of the reactor are radioisotope production, radiochemistry, neutron activation analysis, training, and neutron beam physics. Beginning in 1987, a program was initiated to upgrade and modernize the reactor facility. The modernization program has included replacement of the original instrumentation and control system with a new state-of-the-art microprocessor-driven digital system and addition of a new radiation monitoring system

  13. Expanding the storage capability at ET-RR-1 research reactor at Inshass

    International Nuclear Information System (INIS)

    Storing of spent fuel from Test Reactor in developing countries has become a big dilemma for the following reasons: The transportation of spent fuel is very expensive; There are no reprocessing plants in most developing countries; The expanding of existing storage facilities in reactor building require experience that most of developing countries lack; Some political motivations from Nuclear Developed countries intervene which makes the transportation procedures and logistics to those countries difficult. This paper gives the conceptual design of a new spent fuel storage now under construction at Inshass research reactor (ET-RR-1). The location of the new storage facility is chosen to be within the premises of the reactor facility so that both reactor and the new storage are one Material Balance Area. The paper also proposes some ideas that can enhance the transportation and storage of spent fuel of test reactors, such as: Intensifying the role of IAEA in helping countries to get rid of the spent fuel; The initiation of regional spent fuel storage facilities in some developing countries. (author)

  14. The future of the IPR-R1 TRIGA MARK I reactor after 48 years operation

    International Nuclear Information System (INIS)

    The TRIGA Mark I IPR-R1 Reactor operates in the Nuclear Technology Development Center/ Brazilian Committion for Nuclear Energy (CDTN/CNEN), originally Institute of Radioactive Researches, in Belo Horizonte, Minas Gerais, since November 6, 1960. Initially it operated for isotope production for different uses, being later used in wide scale for another purposes as analyses for activation with neutrons and training of nuclear power plants operators. Dozens of degree theses were also developed with the use of the reactor. Along the years, several improvements were introduced in the reactor and its auxiliary systems, with the purpose to provide better use of the facilities and with the objective to increase the safety in the operation. The reactor is ready right now to operate at 250 kW, and for sure the nuclear applications programmed will be improved. The Operation Manual and the Safety Analysis report were already modified, as well as the Emergency Plan and the relative procedures to the same. After the tests at the end of 2008, the reactor will already be operating in the new power. This work presents a description of the several accomplishments of the last years and comments about the possibility of new uses for the reactor in the several areas of nuclear applications and some of the experiments and tests results during the upgrading program. (authors)

  15. Measured and calculated effective delayed neutron fraction of the IPR-R1 Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Rose Mary G.P.; Dalle, Hugo M.; Campolina, Daniel A.M., E-mail: souzarm@cdtn.b, E-mail: dallehm@cdtn.b, E-mail: campolina@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The effective delayed neutron fraction, {beta}{sub eff}, one of the most important parameter in reactor kinetics, was measured for the 100 kW IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil. The current reactor core has 63 fuel elements, containing about 8.5% and 8% by weight of uranium enriched to 20% in U{sup 235}. The core has cylindrical configuration with an annular graphite reflector. Since the first criticality of the reactor in November 1960, the core configuration and the number of fuel elements have been changed several times. At that time, the reactor power was 30 kW, there were 56 fuel elements in the core, and the {beta}{sub eff} value for the reactor recommended by General Atomic (manufacturer of TRIGA) was 790 pcm. The current {beta}{sub eff} parameter was determined from experimental methods based on inhour equation and on the control rod drops. The estimated values obtained were (774 {+-} 38) pcm and (744 {+-} 20) pcm, respectively. The {beta}{sub eff} was calculated by Monte Carlo transport code MCNP5 and it was obtained 747 pcm. The calculated and measured values are in good agreement, and the relative percentage error is -3.6% for the first case, and 0.4% for the second one. (author)

  16. The IPR-R1 TRIGA Mark I Reactor in 39 years: Operations and general improvements

    International Nuclear Information System (INIS)

    The nuclear IPR-R1 TRIGA Mark I Reactor operating in the Nuclear Technology Development Center, originally Institute for Radioactive Research in Minas Gerais, Brazil, was dedicated in November 11, 1960. Initially operating for the production of radioisotopes for different uses, it started later to be used in large scale for neutron activation analysis and training of operators for nuclear power plants. Many improvements have been made throughout these years to provide a better performance in its operation and safety conditions. A new cooling system to operate until 300 kW, a new control rod mechanism, an aluminum tank for the reactor pool, an optimization in the pneumatic system, a new reactor control console and a general remodeling of the reactor laboratory were some of the improvements added. To prevent and mitigate the ageing effects, the reactor operation personnel is starting a program to minimize future operation problems. This paper describes the improvements made, the results obtained during the past 39 years, and the precautions taken to ensure future safe operation of the reactor to give operators better conditions of safe work. (author)

  17. Analysis of loss of flow accident at Pakistan research reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Bokhari, I.H. [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)]. E-mail: ishtiaq@pinstech.org.pk; Mahmood, T. [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

    2005-12-15

    The main objective of the reactor safety is to keep the reactor core in a condition, which does not permit any release of radioactivity into the environment. In order to ensure this, the reactor must have sufficient safety margins during all possible operational conditions (normal as well as accidental). To accomplish this, a study has been carried out, for the analysis of loss of flow accident (LOFA), which is one of the probable scenarios among other possible events such as reactivity-induced-accidents, loss of coolant accident, etc. The study has been carried out for Pakistan research reactor, PARR-1, which was initially converted from HEU to LEU fuel. It is a swimming pool type reactor using MTR type fuel. Presently, a new core is proposed to be assembled containing LEU and some of the used (less burnt) HEU fuel elements. The accident is assumed when the reactor is running at a steady-state power level of 9.8 MW. Computer code PARET and standard correlations were employed to compute various parameters. Results predict nucleate boiling in the core but the temperatures would remain far below the fuel clad melting point.

  18. Analysis of loss of flow accident at Pakistan research reactor-1

    International Nuclear Information System (INIS)

    The main objective of the reactor safety is to keep the reactor core in a condition, which does not permit any release of radioactivity into the environment. In order to ensure this, the reactor must have sufficient safety margins during all possible operational conditions (normal as well as accidental). To accomplish this, a study has been carried out, for the analysis of loss of flow accident (LOFA), which is one of the probable scenarios among other possible events such as reactivity-induced-accidents, loss of coolant accident, etc. The study has been carried out for Pakistan research reactor, PARR-1, which was initially converted from HEU to LEU fuel. It is a swimming pool type reactor using MTR type fuel. Presently, a new core is proposed to be assembled containing LEU and some of the used (less burnt) HEU fuel elements. The accident is assumed when the reactor is running at a steady-state power level of 9.8 MW. Computer code PARET and standard correlations were employed to compute various parameters. Results predict nucleate boiling in the core but the temperatures would remain far below the fuel clad melting point

  19. Report on safety related occurrences and reactor trips July 1, 1976-December 31, 1976

    International Nuclear Information System (INIS)

    This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1, 1976 to December 31, 1976 inclusive. The facilities involved are Oskarshamn 1 and 2, Ringhals 1 and 2 and Barsebaeck 1. During this period of the 6 months 37 safety related occurrences and 34 reactor trips have been reported to the Nuclear Power Inspectorate. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The fact that even small deviations from prescribed operation results in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The number of reactor trips are almost as low as during the last period, which is a drastic reduction compared to earlier time periods. The greatest outages are caused by occurrences without safety significance.(author)

  20. Dose measurements in controlled area of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    The workers doses in exposure areas to the radiation are so important for a Radioprotection Quality Program, as well as to guarantee the workers safety. For that it is necessary to raise the doses in the radiation areas, to obtain the accumulated dose in certain procedures for detailed studies. Several risings were accomplished to obtain the radiation levels in the areas where the workers are exposed due the operation of a research nuclear reactor and in the radioisotopes manipulation laboratories of a nuclear institute. The radiation levels and doses can be observed through graphs in the dependences of the Controlled Area 1 (AC-1) and the Reactor Laboratory. Those limits are in according of the CNEN-NE-3.01 work limits rules. The conclusion of the work allowed to demonstrate that the Laboratory of the Reactor and AC-1, have booth an effective radiological program with efficient operational practices that contributes with low doses to the workers

  1. Current utilization and long term strategy of the Finnish TRIGA research reactor FIR 1

    International Nuclear Information System (INIS)

    The Finnish TRIGA reactor, FiR 1, started operation in 1962. From early on the reactor created versatile research to support the national nuclear program as well as generally the industry and health care sector. Production of short-lived radioisotopes is still a basic service. Education and training play a role in the form of university courses and training of nuclear industry personnel in the Baltic region. In the 1990's a BNCT cancer treatment facility was build. Over 200 patient irradiations have been performed since May 1999. FiR 1 is one of the few facilities in the world providing these treatments. A long term strategy is being worked out for FiR 1 by VTT supported by an independent survey. The survey recommends operation of the reactor at least till 2016 to enable continuation of the promising development of BNCT in parallel of developing accelerator based neutron sources for this treatment. (author)

  2. Build-Up Of Actinides In Irradiated Fuel Rods Of The ETRR-1 Reactor

    International Nuclear Information System (INIS)

    The content concentrations of actinides are calculated as a function of operating reactor regime and cooling time at different percentage of fuel burn-up. The build-up transmutation equations of actinides content in an irradiated fuel are solved numerically .A computer code BAC was written to operate on PC computer to provide the required calculations. The fuel element of 10% 235U enrichment of ETRR-1 reactor was taken as an example for calculations using BAC code. The results are compared with other calculations for the ETRR-1 fuel rod. An estimation of fissile build-up content of a proposed new fuel of 20% '235U enrichment for ETRR-1 reactor is given. The sensitivity coefficients of build-up plutonium concentrations as a function of cross-section data uncertainties are also calculated

  3. Build up of actinides in burnt fuel rods of the ET-RR-1 reactor

    International Nuclear Information System (INIS)

    The content concentrations of actinides are calculated as a function of operating reactor regime and cooling time at different percentage of fuel burn-up. The build-up transmutation equations of actinides content in an irradiated fuel are solved numerically .A computer code BAC was written to operate on a PC computer to provide the required calculations. The fuel element of 10% 235U enrichment of ET-RR-1 reactor was taken as an example for calculations using the BAC code. The results are compared with other calculations for the ET-RR-1 fuel rod. An estimation of fissile build-up content of a proposed new fuel of 20% 235U enrichment for ET-RR-1 reactor is given. The sensitivity coefficients of build-up plutonium concentrations as a function of cross-section data uncertainties are also calculated. (author)

  4. Four-neutrino analysis of 1.5km-baseline reactor antineutrino oscillations

    CERN Document Server

    Kang, Sin Kyu; Ko, Young-Ju; Siyeon, Kim

    2014-01-01

    The masses of sterile neutrinos are not yet known, and depending on the orders of magnitudes, their existence may explain reactor anomalies or the spectral shape of reactor neutrino events at 1.5km-baseline detector. Here, we present four-neutrino analysis of the results announced by RENO and Daya Bay, which performed the definitive measurements of $\\theta_{13}$ based on the disappearance of reactor antineutrinos at km-order baselines. Our results using 3+1 scheme include the exclusion curve of $\\Delta m^2_{41}$ vs. $\\theta_{14}$ and the adjustment of $\\theta_{13}$ due to correlation with $\\theta_{14}$. The value of $\\theta_{13}$ obtained by RENO and Daya Bay with a three-neutrino oscillation analysis is included in the $1\\sigma$ interval of $\\theta_{13}$ allowed by our four-neutrino analysis.

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

    International Nuclear Information System (INIS)

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

  6. Measurement of the thermal flux distribution in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    The knowledge of the neutron flux distribution in research reactors is important because it gives the power distribution over the core, and it provides better conditions to perform experiments and sample irradiations. The measured neutron flux distribution can also be of interest as a means of comparison for the calculational methods of reactor analysis currently in use at this institute. The thermal neutron flux distribution of the IEA-R1 reactor has been measured with the miniature chamber WL-23292. For carrying out the measurements, it was buit a guide system that permit the insertion of the mini-chamber i between the fuel of the fuel elements. It can be introduced in two diferent positions a fuel element and in each it spans 26 axial positions. With this guide system the thermal neutron flux distribution of the IEA-R1 nuclear reactor can be obtained in a fast and efficient manner. The element measured flux distribution shows clearly the effects of control rods and reflectors in the IEA-R1 reactor. The difficulties encountered during the measurements are mentioned with detail as well as the procedures adopteed to overcome them. (Author)

  7. 235 U consumption and 239 Pu formation on IPR-R1 reactor

    International Nuclear Information System (INIS)

    On the present world conjuncture it is necessary a rigorous storage and utilization control of nuclear materials under safeguard, specially those considered as strategic and those that could be used on the nuclear armaments fabrications as 235U and 239Pu. Brazil have signed and ratified many international pacts related to the peaceful utilization of the nuclear materials, making the compromise to attend nuclear safeguard procedures accorded with international regulatory organizations. The main goal of this work is the development of a simplified method for the estimation of the 239Pu productivity during the operation of the IPR-R1, TRIGA Mark-1 type, nuclear research reactor, that can achieve the maximum power of 100 KWt. A computational algorithm have been made to calculate the consumption of 235U on the fuel elements used on the reactor and the subsequent 239Pu formation (by the neutronic capture process), using the power generated by the reactor since the first criticality. The particular geometry of this kind of reactor allowed the calculation of the burn factors of the fuel elements discriminated by nuclear concentric rings, based on the thermal neutrons field distribution observed on experiments accomplished during a time period in similar reactors and on the IPR-R1. The simplified block diagram of the process, based on the calculation described above, is presented

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

    Science.gov (United States)

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

    2015-01-01

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

  9. Welding electrode for peripheral welds of A-1 reactor pressure vessel

    International Nuclear Information System (INIS)

    The properties are outlined of the VUZ-AC1-52 welding electrode used in welding the Bohunice A-1 reactor pressure vessel. The mechanical properties of welded joints after the final thermal treatment are summed up. (J.K.)

  10. Modelling for great breaks accident analysis in the primary system of Angra 1 reactor

    International Nuclear Information System (INIS)

    An analysis is made for a break in the cold leg, of the guillotine type with discharge coefficient C sub(D)=1.0, for the Angra 1 reactor. The computer codes, geometrical models and options used are described. A comparison between the method used and the requirements in the Appendix K of 10 CRF 50 is done. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-11-01

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

  12. Natural uranium-graphite system. Critial experiments on the G1 reactor

    International Nuclear Information System (INIS)

    A number of experiments have been performed during the start up period of the G1 (1956) and G2 (1958) reactors in Marcoule, both on their lattices and on different lattices (hollow rods, clusters, under moderated lattices). The first chapter gives a thorough description of the two reactors. The second chapter deals with buckling measurements, both absolute (flux plots) and relative by the method of progressive substitution. The experimental results are summarised in Table VI. The third chapter contains a number of other measurements performed on G1. (author)

  13. Kinetic and reactor modelling of lipases catalyzed (R,S)-1-phenylethanol resolution

    OpenAIRE

    Chua Lee-Suana; Cheng Kian-Kaib; Lee Chew-Tinb; Mohamad-Roji Sarmidic; Ramlan Abdul-Azizc; Tang Boon-Sengd

    2010-01-01

    This study was focused on the development of a kinetic model and a reactor model for the enzymatic resolution of (R,S)-1-phenylethanol. The reaction progress curves catalyzed by immobilized lipases, ChiroCLEC-PC in batch stirred tank reactor were used to develop the kinetic model. The resolution followed Ping-Pong Bi-Bi mechanism with the inhibition of lauric acid, (R,S)-1-phenylethanol and water. The validity of the model was verified by fitting it to another experimental data catalyzed by i...

  14. Proposal to the United States Energy Research and Development Administration for continuation of fusion reactor technology studies. Progress report October 1, 1977--July 1, 1978

    International Nuclear Information System (INIS)

    Since the last progress report we have concentrated on three main areas of research: (1) the study of the NUWMAK reactor design, (2) the study of rf heating for tokamak reactors, and (3) the initiation of a tandem mirror reactor study. The initial work on the tandem mirror reactor is included as background in the technical proposal. Summaries of our work on recent assessments of lithium reserves and neutral transport codes are included

  15. Proposal to the United States Energy Research and Development Administration for continuation of fusion reactor technology studies. Progress report October 1, 1977--July 1, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Conn, R.W.; Kulcinski, G.L.; Maynard, C.W.

    1978-01-01

    Since the last progress report we have concentrated on three main areas of research: (1) the study of the NUWMAK reactor design, (2) the study of rf heating for tokamak reactors, and (3) the initiation of a tandem mirror reactor study. The initial work on the tandem mirror reactor is included as background in the technical proposal. Summaries of our work on recent assessments of lithium reserves and neutral transport codes are included.

  16. KTA 3401.1. Reactor safety vessel of steel

    International Nuclear Information System (INIS)

    The text of the standard has been prepared by order of the Nuclear Committee of the Working Group on Pressure Vessels with the ''Verein Deutscher Eisenhuettenleute (VDEhL)'' acting as main contractor. This standard replaces the standard KTA 3401.1, edition 6/80. As against edition 6/80 the text of the standard has been editorially treated, in particular for adaptation to the newly included annex A: ''Material characteristics''. Steels: 15MnNi63 (DIN-1.6210); 40NiCrMo84 (DIN-1,6562); 26NiCrMo146 (DIN-1.6958); 20NiCrMo145 (DIN-1.6772); 34CrMo4 (DIN-1.7220); 42CrMo4 (DIN-1.7225); C45 (DIN-1.0503). (orig./HP)

  17. SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configuration

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2000-03-16

    Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the Waste Package Design team at the Yucca Mountain Project in the United States, who performed the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The CRCs involved relatively low-cycle burnups, and therefore contained a relatively high gadolinium poison content in the reactor assemblies. This report summarizes the data and methods used in analyzing the critical configurations and assesses the sensitivity of the results to some of the modeling approximations used to represent the gadolinium poison distribution within the assemblies. The KENO V.a calculations, performed using the SCALE 44GROUPNDF5 ENDF/B-V cross-section library, yield predicted k{sub eff} values within about 1% {Delta}k/k relative to reactor measurements for the five CRCs using general 8-pin and 9-pin heterogeneous gadolinium poison pin assembly models.

  18. SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configurations

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2000-03-01

    Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the Waste Package Design team at the Yucca Mountain Project in the US, who performed the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The CRCs involved relatively low-cycle burnups, and therefore contained a relatively high gadolinium poison content in the reactor assemblies. This report summarizes the data and methods used in analyzing the critical configurations and assesses the sensitivity of the results to some of the modeling approximations used to represent the gadolinium poison distribution within the assemblies. The KENO V.a calculations, performed using the SCALE 44GROUPNDF5 ENDF/B-V cross-section library, yield predicted k{sub eff} values within about 1% {Delta}k/k relative to reactor measurements for the five CRCs using general 8-pin and 9-pin heterogeneous gadolinium poison pin assembly models.

  19. SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configurations

    International Nuclear Information System (INIS)

    Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the Waste Package Design team at the Yucca Mountain Project in the US, who performed the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The CRCs involved relatively low-cycle burnups, and therefore contained a relatively high gadolinium poison content in the reactor assemblies. This report summarizes the data and methods used in analyzing the critical configurations and assesses the sensitivity of the results to some of the modeling approximations used to represent the gadolinium poison distribution within the assemblies. The KENO V.a calculations, performed using the SCALE 44GROUPNDF5 ENDF/B-V cross-section library, yield predicted keff values within about 1% Δk/k relative to reactor measurements for the five CRCs using general 8-pin and 9-pin heterogeneous gadolinium poison pin assembly models

  20. Development of a training simulator to operators of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    This work reports the development of a Simulator for the IEA-R1 Research Reactor. The Simulator was developed with Visual C++ in two stages: construction of the mathematics models and development and configuration of graphics interfaces in a Windows XP executable. A simplified modeling was used for main physics phenomena, using a point kinetics model for the nuclear process and the energy and mass conservation laws in the average channel of the reactor for the thermal hydraulic process. The dynamics differential equations were solved by using finite differences through the 4th order Runge- Kutta method. The reactivity control, reactor cooling, and reactor protection systems were also modeled. The process variables are stored in ASCII files. The Simulator allows navigating by screens of the systems and monitoring tendencies of the operational transients, being an interactive tool for teaching and training of IEA-R1 operators. It also can be used by students, professors, and researchers in teaching activities in reactor and thermal hydraulics theory. The Simulator allows simulations of operations of start up, power maneuver, and shut down. (author)

  1. Main refurbishment activities on electronic and electrical equipment for the FRG-1 research reactor

    International Nuclear Information System (INIS)

    As GKSS intends to operate the research reactor FRG-1 safely and reliably for many years to come, the plant is constantly refurbished and upgraded both in the interests of safety and operational reasons. The following electronic and electrical systems have been replaced or improved since 1990: Information and signalling systems; Emergency power plant (permit applied for); External and internal lightning protection system; Reactor protection system (in part); Safety lighting; Alarm and staff locating system; Control room telephone system; Closed-circuit television system; Beam tube controls; Storage plant for radioactive liquid waste; Ambient dose rate measuring system; Meteorological measuring system; Control and measuring system for the primary cooling circuit; Control rod drives; Control rod control system; Soft start for the secondary pumps; Control and switching devices for the emergency power plant; Trailing cable installation for the reactor bridge; Main-voltage distribution systems/cable routes. (author). 13 figs, 1 tab

  2. A model reconstruction of riverine nutrient fluxes and eutrophication in the Belgian Coastal Zone since 1984

    Science.gov (United States)

    Passy, P.; Gypens, N.; Billen, G.; Garnier, J.; Thieu, V.; Rousseau, V.; Callens, J.; Parent, J.-Y.; Lancelot, C.

    2013-12-01

    The OSPAR convention signed in 1992 by 15 European states including Belgium and France pledged to reduce the nutrient (nitrogen N and phosphorus P) loads from land-based sources to the Channel and the North Sea to half of what they were in 1985. In this paper, we use a river basin-coastal sea chain model to describe the evolution of nutrient loads to the Belgian Costal Zone originating from the Seine, Somme and Scheldt watersheds from 1984 to 2007 in order to assess the N and P reduction with respect to the OSPAR goals and the resulting effect on coastal eutrophication, especially Phaeocystis blooms. Since the early 1990s, most nutrient reduction actions have been devoted to domestic and industrial wastewater treatment, resulting in a sharp P decrease between 1984 and 2007: from 260 to 90 kgP km- 2 for the Seine River and from 215 to 110 kgP km- 2 for the Scheldt River. In spite of improved N treatment of wastewater, there is no clear decrease of N loads, which mostly originate from leaching intensively cultivated arable lands. N fluxes at the outlet of the Seine and Scheldt rivers were, respectively, 1990 and 2210 kgN km- 2 in 1984 and 1830 and 1390 kgN km- 2 in 2007. However, this relatively low decrease appears to be more influenced by hydrological conditions than by better efficiency of N use in agriculture. We conclude from this analysis that the OSPAR objectives for P have been achieved, whereas for N radical changes in agricultural practices are still required. The P reduction achieved allows, for the period of concern, a 50% decrease of Phaeocystis colony blooms in the Belgian Coastal Zone, both in magnitude and duration. However, the simulated decrease, of maximum abundance, i.e., from 60 · 106 in 1984 to 30 · 106 cells l- 1 in 2007, is still insufficient when compared to the ecological-quality indicator of 4 · 106 cells l- 1. A further decrease of nutrients is still necessary to decrease undesirable blooms more satisfactorily.

  3. Gas-cooled fast breeder reactor. Quarterly progress report, February 1-April 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-01

    Information is presented concerning the reactor vessel; reactivity control mechanisms and instrumentation; reactor internals; primary coolant circuits;core auxiliary cooling system; reactor core; systems engineering; and reactor safety and reliability;

  4. Reactor Engineering Department annual report, April 1, 1985 - March 31, 1986

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1985 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor, High Conversion Light Water Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, reactor decommissioning technology, and activities of the Committee on Reactor Physics. (author)

  5. SCALE-4 analysis of pressurized water reactor critical configurations. Volume 3: Surry Unit 1 Cycle 2

    International Nuclear Information System (INIS)

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using selected critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations in this report is based on the codes and data provided in the SCALE-4 code system. This volume of the report documents the SCALE system analysis of two reactor critical configurations for Surry Unit 1 Cycle 2. This unit and cycle were chosen for a previous analysis using a different methodology because detailed isotopics from multidimensional reactor calculations were available from the Virginia Power Company. These data permitted a direct comparison of criticality calculations using the utility-calculated isotopics, with those using, the isotopics generated by the SCALE-4 SAS2H sequence. These reactor critical benchmarks have been reanalyzed using the methodology described above. The two benchmark critical calculations were the beginning-of-cycle (BOC) startup at hot, zero-power (HZP) and an end-of-cycle (EOC) critical at hot, full-power (HFP) critical conditions. These calculations were used to check for consistency in the calculated results for different burnup, downtime, temperature, xenon, and boron conditions. The keff results were 1.0014 and 1.0113, respectively, with a standard deviation of 0.0005

  6. Methanol steam reforming via internal recycle reactor. Paper no. IGEC-1-144

    International Nuclear Information System (INIS)

    Hydrogen generation for PEMFC by methanol steam reforming using a Caldwell internal recycle reactor (IRR) was studied. BASF K3-110 copper-based catalyst was used. The impeller speed and methanol retention time almost proportionally affected the recycle ratio, one of the most direct and important indices to show the gradientlessness of concentration and temperature. When the recycle ratio was greater than 20, internal recycle reactor could be considered as continuously stirred tank reactor (CSTR), one ideal reactor for kinetics studies with no appreciable concentration and temperature gradients. The experiment results via CSTR fit very well with the kinetics model developed using a differential reactor by Peppley et al.. This verified the accuracy of the Peppley model and vice versa. The pseudo first order reaction rate constant developed in the CSTR was found to be 0.1-0.15 mol/bar.kg.s, and the activation energy was 93 kJ/mol, which were in good accordance with Peppley model and other values reported in the literature. However, when the recycle ratio was too low, less than 20 for instance, either because of the high GHSV of reactants or low impeller speed, methanol conversion rate as well as CO2, H2 production rates were well below the values predicted by the Peppley model due to the existence of strong gradients of concentration and temperature. Regardless of the recycle ratio, CO producing rate in the IRR was lower than that via the plug flow reactor (PFR) in terms of Peppley model, which could be presumably ascribed to the strong inhibition effect of hydrogen on the reaction rate of methanol decomposition and reverse water gas shift (WGS) reaction over Cu based catalyst. This characteristic could be of benefit in reactor design to suppress CO yield which will be beneficial for producing PEMFC-grade reformate. (author)

  7. Determine Operating Reactor to Use for the 2016 PCI Level 1 Milestone

    International Nuclear Information System (INIS)

    The Consortium for Advanced Simulation of Light Water Reactors (LWRs) (CASL) Level 1 milestone to 'Assess the analysis capability for core-wide [pressurized water reactor] PWR Pellet- Clad Interaction (PCI) screening and demonstrate detailed 3-D analysis on selected sub-region' (L1:CASL.P13.03) requires a particular type of nuclear power plant for the assessment. This report documents the operating reactor and cycles chosen for this assessment in completion of the physics integration (PHI) milestone to ''Determine Operating Reactor to use for PCI L1 Milestone'' (L3:PHI.CMD.P12.02). Watts Bar Unit 1 experienced (at least) one fuel rod failure in each of cycles 6 and 7, and at least one was deemed to be duty related rather than being primarily related to a manufacturing defect or grid effects. This brief report documents that the data required to model cycles 1-12 of Watts Bar Unit 1 using VERA-CS contains sufficient data to model the PHI portion of the PCI challenge problem. A list of additional data needs is also provided that will be important for verification and validation of the BISON results.

  8. Determine Operating Reactor to Use for the 2016 PCI Level 1 Milestone

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-01-30

    The Consortium for Advanced Simulation of Light Water Reactors (LWRs) (CASL) Level 1 milestone to “Assess the analysis capability for core-wide [pressurized water reactor] PWR Pellet- Clad Interaction (PCI) screening and demonstrate detailed 3-D analysis on selected sub-region” (L1:CASL.P13.03) requires a particular type of nuclear power plant for the assessment. This report documents the operating reactor and cycles chosen for this assessment in completion of the physics integration (PHI) milestone to “Determine Operating Reactor to use for PCI L1 Milestone” (L3:PHI.CMD.P12.02). Watts Bar Unit 1 experienced (at least) one fuel rod failure in each of cycles 6 and 7, and at least one was deemed to be duty related rather than being primarily related to a manufacturing defect or grid effects. This brief report documents that the data required to model cycles 1–12 of Watts Bar Unit 1 using VERA-CS contains sufficient data to model the PHI portion of the PCI challenge problem. A list of additional data needs is also provided that will be important for verification and validation of the BISON results.

  9. 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.

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

    Science.gov (United States)

    Vandendriessche, Sofie; Vincx, Magda; Degraer, Steven

    2006-02-01

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

  11. Reflector modelling with multi-group nodal equivalence theory for the SAFARI-1 research reactor

    International Nuclear Information System (INIS)

    Normalised Generalised Equivalence Theory is used to model the ex-core reflector region of the SAFARI-1 research reactor. This method is a one-dimensional homogenisation technique based on Generalised Equivalence Theory, but with only one discontinuity factor defined per node, and divided into the nodal parameters. The SAFARI-1 reactor is modelled with the deterministic code system OSCAR-4. Cross-sections for the reflector model is generated with NEWT (part of the SCALE 6.1 package) and EQUIVA-1 (part of OSCAR-4), which calculates the NGET parameters. A period of three years in the operational history of the SAFARI-1 research reactor is modelled. Two models are used, one with traditional flux-volume weighted and the other with equivalent ex-core reflector cross-sections. The performance of the two models over the three year period is compared. Reactor parameters such as reactivity and fuel burnup are investigated. Comparisons to experimental data, in particular control rod calibrations, are also made. The model with equivalent reflector parameters shows improved accuracy for control rod calibrations, a power tilt of about 10% across the core, no noticeable change in reactivity or burnup, and significant improvement in calculational time (reduced by over 40%) due to a reduction in the size of the core model. (author)

  12. The reactor kinetics code tank: a validation against selected SPERT-1b experiments

    International Nuclear Information System (INIS)

    The two-dimensional space-time analysis code TANK is being developed for the simulation of transient behaviour in the MAPLE class of research reactors. MAPLE research reactor cores are compact, light-water-cooled and -moderated, with a high degree of forced subcooling. The SPERT-1B(24/32) reactor core had many similarities to MAPLE-X10, and the results of the SPERT transient experiments are well documented. As a validation of TANK, a series of simulations of certain SPERT reactor transients was undertaken. Special features were added to the TANK code to model reactors with plate-type fuel and to allow for the simulation of rapid void production. The results of a series of super-prompt-critical reactivity step-insertion transient simulations are presented. The selected SPERT transients were all initiated from low power, at ambient temperatures, and with negligible coolant flow. Th results of the TANK simulations are in good agreement with the trends in the experimental SPERT data

  13. Initial core calculation of 1 MW reactor TRIGA PUSPATI (RTP) using SRAC code system

    International Nuclear Information System (INIS)

    The 1 MWatt TRIGA PUSPATI Reactor (RTP) was located in Malaysian Institute for Nuclear Technology Research (MINT). This research reactor was from TRIGA MARK II type and was put into operation on 1983 and has reached its first criticality on 28 June 1982. Since then, this reactor has been used for various beam experiments, irradiation facilities, radioisotope production and education and training. The RTP uses three types of fuel elements, namely, 8.5wt%, 12wt% and 20wt% which enriched to about 20% of U-235 for all types. The RTP has four control rods which made up of boron carbide. It has cylindrical core but not in periodically in its lattice structure, which possibly locates 127 of fuel elements. Both of the coolant and moderator uses light water system and the reflector was made from high purity graphite. Because of this research reactor's power is relatively small compared to the power reactor; it uses natural convection for its cooling system. To ensure the integrity of the core, fuel shuffling have been made for several times. Until now, there are 11 configurations of the core and recently has achieved the 12th configuration. This paper will described the first core configuration calculation using SRAC code system which was first introduced in 2005 during the FNCA workshop. (author)

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

    International Nuclear Information System (INIS)

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

  15. Improving of safe operation and new open policy at VR-1 reactor

    International Nuclear Information System (INIS)

    Full text: The VR-1 Reactor is operated for training of university students and nuclear power plant personnel, R and D, and information services for non-military nuclear energy use. During last four years a large number of improvements in operation was achieved. Some of them are the most important from the safety-related point of view. In the beginning of 2003 new web-portal with on-line information from the operation of the reactor was launched. Web-portal brings new feature in the way to opening information about operation of the VR-1 Reactor to public audience. Operation documentation and neutronics calculations reviewing. New Czech Atomic Act issued in 1997 and updated in 2002 requests reviewing all safety and operation documentation within five years from the date of releasing the Atomic Act. Majority of the VR-1 Reactor documentation was reviewed and updated or new documentation was created. According to requirements of the Atomic Act, Czech regulatory body requirements and IAEA recommendations there were reviewed: Operational limits and conditions; Quality assurance programmes and procedures; Inner emergency plan; emergency preparedness and emergency exercises; Operation staff qualification and training procedure, Operating instructions and procedures; Radiation protection and environmental monitoring procedure, waste management procedure. Two of them (quality assurance programmes and procedures and emergency preparedness and emergency exercises) were significantly innovated. New procedure for decommissioning was created. This preliminary version provides aims and methodology for potential decommissioning of reactor only. Next safety analysis report will be elaborated 10 years after the last SAR and after full upgrade of Control and safety system (2005-2006). At the end of 2003 works on Safety analysis by PSA method will be finished. Operation documentation reviewing for the reactor was very useful and brought at the same time new aspects and views on

  16. Neutronic and thermal-hydraulic experimental program in the IPR-R1 TRIGA reactor at CDTN

    International Nuclear Information System (INIS)

    The IPR-R1 TRIGA reactor, located at CDTN (Belo Horizonte/Brazil), is a typical 100 kW Mark I light-water reactor cooled by assisted natural convection with an annular graphite reflector. In order to study the safety aspects connected with the increase of the maximum steady state power of the IPR-R1 TRIGA reactor, experimental measures were taken. This paper summarizes the experimental program and some recent results and procedures of the neutronic and thermalhydraulic experiments carried out in the IPR-R1 TRIGA reactor. (authors)

  17. Digital Systems Implemented at the IPEN Nuclear Research Reactor (IEA-R1): Results and Necessities

    International Nuclear Information System (INIS)

    (Nuclear and Energy Research Institute) was founded in 1956 with the main purpose of doing research and development in the field of nuclear energy and its applications. It is located at the campus of University of Sao Paulo (USP), in the city of Sao Paulo, in an area of nearly 500, 000 m2. It has over 1.000 employees and 40% of them have qualification at master or doctor level The institute is recognized as a national leader institution in research and development (R and D) in the areas of radiopharmaceuticals, industrial applications of radiation, basic nuclear research, nuclear reactor operation and nuclear applications, materials science and technology, laser technology and applications. Along with the R and D, it has a strong educational activity, having a graduate program in Nuclear Technology, in association with the University of Sao Paulo, ranked as the best university in the country. The Federal Government Evaluation institution CAPES, granted to this course grade 6, considering it a program of Excellence. This program started at 1976 and has awarded 458 Ph.D. degrees and 937 master degrees since them. The actual graduate enrollment is around 400 students. One of major nuclear installation at IPEN is the IEA-R1 research reactor; it is the only Brazilian research reactor with substantial power level suitable for its utilization in researches concerning physics, chemistry, biology and engineering as well as for producing some useful radioisotopes for medical and other applications. IEA-R1 reactor is a swimming pool type reactor moderated and cooled by light water and uses graphite and beryllium as reflectors. The first criticality was achieved on September 16, 1957. The reactor is currently operating at 4.5 MW power level with an operational schedule of continuous 64 hours a week. In 1996 a Modernization Program was started to establish recommendations in order to mitigate equipment and structures ageing effects in the reactor components, detect and evaluate

  18. Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Demonstration. CEDT Phase 1 Preliminary Design Documentation

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, Rene Gerardo [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hutchinson, Jesson D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mcclure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Myers, William L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-08-20

    The intent of the integral experiment request IER 299 (called KiloPower by NASA) is to assemble and evaluate the operational performance of a compact reactor configuration that closely resembles the flight unit to be used by NASA to execute a deep space exploration mission. The reactor design will include heat pipes coupled to Stirling engines to demonstrate how one can generate electricity when extracting energy from a “nuclear generated” heat source. This series of experiments is a larger scale follow up to the DUFF series of experiments1,2 that were performed using the Flat-Top assembly.

  19. Disassembly of the fusion-1 capsule after irradiation in the BOR-60 reactor

    International Nuclear Information System (INIS)

    A U.S./Russia (RF) collaborative irradiation experiment, Fusion-1, was completed in June 1996 after reaching a peak exposure of ∼17 dpa in the BOR-60 fast reactor at the Research Institute of Atomic Reactors (RIAR) in Russia. The specimens were vanadium alloys, mainly of recent heats from both countries. In this reporting period, the capsule was disassembled at the RIAR hot cells and all test specimens were successfully retrieved. For the disassembly, an innovative method of using a heated diffusion oil to melt and separate the lithium bond from the test specimens was adopted. This method proved highly successful

  20. Acoustic emission monitoring of preservice testing at Watts Bar Unit 1 Nuclear Reactor

    International Nuclear Information System (INIS)

    Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Plant in the US during hot functional preservice testing is described. Background, methodology, and results are included. The work discussed here is a major milestone in a program supported by the US NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. The subject work demonstrated that anticipated problem areas can be overcome. Work is continuing to AE monitoring during reactor operation. 3 refs., 6 figs

  1. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    OpenAIRE

    Muhammad Atta; Iqbal Masood; Mahmood Tayyab

    2011-01-01

    The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determin...

  2. UKHEDD-1. Radioactive heavy element decay data for reactor calculations

    International Nuclear Information System (INIS)

    This document includes the introductory pages of the report AEEW-R-1407 (Dec. 1981), which contains voluminous data tables. The data file UKHEDD-1 is available on magnetic tape from the IAEA Nuclear Data Section. (author)

  3. ISO-9001: An approach to accreditation for an MTR facility: SAFARI-1 research reactor

    International Nuclear Information System (INIS)

    The SAFARI-1 Research Reactor obtained ISO-9001 accreditation via the South African Bureau of Standards in September 1998. In view of the commercial applications of the reactor, the value of acquisition of the accreditation was considered against the cost of implementation of the Quality System. The criteria identified in the ISO-9001 standard were appraised and a superstructure derived for management of the generation and implementation of a suitable Quality Management System (QMS) for the fairly unique application of a nuclear research reactor. A Quality Policy was established, which formed the basis of the QMS against which the various requirements and/or standards were identified. In addition, since it was considered advantageous to incorporate the management controls of Conventional and Radiological Safety as well as Plant Maintenance and Environmental Management (ISO 14001), these aspects were included in the QMS. (author)

  4. TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Liles, D.R.; Mahaffy, J.H.

    1984-02-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided.

  5. Hazards review: N-Reactor 1.25% co-producer fuel element test

    Energy Technology Data Exchange (ETDEWEB)

    Miller, N.R.; Nechodom, W.S.

    1964-07-13

    The N-Reactor Hazard Summary Report examines the hazard from operating the N-Reactor with a uniform fuel loading enriched to 0.947% U{sup 235}. Incentives have been developed for reactor testing of a block of 49 tubes loaded with co-producer elements, i.e. elements capable of producing both weapons grade plutonium and tritium. The element utilizes an outer fuel tube enriched to 1.25% U{sup 235} with an inner target lithium-aluminum rod. Criteria have been developed to guide the evaluation of safety aspects of such tests. It is the purpose of this document to review the hazards associated with the proposed test and to set forth special precautions which will be necessary to maintain a high level of safety.

  6. Experimental thermal-hydraulic analysis of the IPR-R1 TRIGA nuclear reactor

    International Nuclear Information System (INIS)

    The heat generated by nuclear fission in the IPR-R1 nuclear reactor is transferred from fuel elements to the cooling system through the fuel/cladding (gap) and the cladding to coolant interfaces. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were evaluated experimentally. A correlation for the gap conductance between the fuel and the cladding was also presented. As the reactor core power increases, the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling. Results indicated that subcooled boiling occurs at the cladding surface in the reactor core central channels at power levels in excess of 60 k W. (author)

  7. TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

    International Nuclear Information System (INIS)

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided

  8. Utilization of SRAC Computer Code as a Nuclear Fuel Management Tool for Thai Research Reactor (TRR-1/M1)

    International Nuclear Information System (INIS)

    Full text: An upgrade of fuel management methodology for TRR-1/M1 is presented in this paper. The advanced code system called SRAC is applicable to hexagonal arrangement reactor core thus it is applicable for TRR-1/M1 as a fuel management tool. The methodology is executed in 2 steps, namely hyperfine group structure cross section generation step and reactor core calculation step by applied SRAC's capabilities to create group cross section in to four group structure. For verification purpose, the TRR-1/M1 core configuration number 1 is modeled by the new methodology in this paper. The result of the core excess reactivity of this core shows good agreement with the operation data from operation log-book. This methodology is expected to be employed as a fuel management tool for TRR-1/M1 in the near future

  9. Reactor pressure vessel steels ASME SA533B and SA508 C1.2

    International Nuclear Information System (INIS)

    The report presents the results of the microstructural studies of steels SA533B and SA508 C1.2 obtained in connection with a programme initiated to gather and create information needed for the assessment of structural integrity of the reactor pressure vessels. The as-quenched and variably tempered microstructures were studied with optical, scanning and transmission electron microscopes. (author)

  10. Experience of IEA-R1 research reactor spent fuel transportation back to United States

    International Nuclear Information System (INIS)

    IPEN/CNEN-SP is sending the IEA-R1 Research Reactor spent fuels from USA origin back to this country. This paper describes the experience in organizing the negotiations, documents and activities to perform the transport. Subjects as cask licensing, transport licensing and fuel failure criteria for transportation are presented. (author)

  11. Nuclear Engineering Computer Modules: Reactor Dynamics, RD-1 and RD-2.

    Science.gov (United States)

    Onega, Ronald J.

    The objective of the Reactor Dynamics Module, RD-1, is to obtain the kinetics equation without feedback and solve the kinetics equations numerically for one to six delayed neutron groups for time varying reactivity insertions. The computer code FUMOKI (Fundamental Mode Kinetics) will calculate the power as a function of time for either uranium or…

  12. Turbine generator modifications at the Temelin-1 reactor unit during the start-up phase

    International Nuclear Information System (INIS)

    The history of the prototype saturated steam-driven turbine generator unit is described and information about the basic problems which had to be addressed during the Temelin-1 reactor unit start-up is presented. All the modifications performed are outlined. The repair of the damaged LP rotor is also mentioned. (author)

  13. Burn up calculations for ETRR 1 and ETRR 2 reactors with wims and origen codes

    International Nuclear Information System (INIS)

    For ETRR -1 and ETRR - 2 research reactor, the 235 U depletion is determined with wims and origen codes the two calculated results show good agreement with each other. The buildup of different fission products (important from both the safety and protection point of view) is also calculated. The radioactivity and decay heat of the spent fuel is determined up to 30 years

  14. International Thermonuclear Experimental Reactor (ITER). Engineering Design Activities (EDA). Agreement and protocol 1

    International Nuclear Information System (INIS)

    This document contains protocol 1 to the agreement among the European Atomic Energy Community, the government of Japan, the Government of the Russian Federation, and the Government of the United States of America on cooperation in the engineering design activities for the International Thermonuclear Experimental Reactor, which activities shall be conducted under the auspices of the International Atomic Energy Agency

  15. Experimental study of the IPR-R1 TRIGA reactor power channels responses

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Henrique F.A.; Ferreira, Andrea V., E-mail: hfam@cdtn.br, E-mail: avf@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The IPR-R1 nuclear reactor installed at Centro de Desenvolvimento da Tecnologia Nuclear CDTN/CNEN, Belo Horizonte, Brazil, is a Mark I TRIGA reactor (Training, Research, Isotopes, General Atomics) and became operational on November of 1960. The reactor has four irradiation devices: a rotary specimen rack with 40 irradiation channels, the central tube, and two pneumatic transfer tubes. The nuclear reactor is operated in a power range between zero and 100 kW. The instrumentation for IPR-R1 operation is mainly composed of four neutronic channels for power measurements. The aim of this work is to investigate the responses of neutronic channels of IPR-R1, Linear, Log N and Percent Power channels, and to check their linearity. Gold foils were activated at low powers (0.125-1.000 kW), and cobalt foils were activated at high powers (10-100kW). For each sample irradiated at rotary specimen rack, another one was irradiated at the same time at the pneumatic transfer tube-2. The obtained results allowed evaluating the linearity of the neutronic channels responses. (author)

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

    International Nuclear Information System (INIS)

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

  17. Development of level-1 PSA method applicable to Japan Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    This paper describes a study to develop the level-1 probabilistic safety assessment (PSA) method that is applicable to the Japan Sodium-cooled Fast Reactor (JSFR). This study has been started since August 2010 and aims to provide a new evaluation method of (1) passive safety architectures related to internal events and (2) an advanced seismic isolation system related to a seismic event as a representative external event in Japan. Regarding the internal events evaluation, a quantitative analysis on the frequency of the core damage caused by reactor shutdown failure was conducted. A failure in passive reactor shutdown was taken into account in the event tree model. The failure rate of sodium-cooled fast reactor (SFR) specific components was evaluated based on the operating experience in existing SFRs by applying the Hierarchical Bayesian Method, which can consider a plant-to-plant variability. By conducting an uncertainty analysis, it was found that the assumption about the correlation of the probability parameters between the main and backup reactor shutdown systems (RSSs) is sensitive to the mean value of the frequency of the core damage caused by reactor shutdown failure. As for the seismic event evaluation, seismic response analysis and sensitivity analysis of a seismic isolation system were carried out. Rubber bearings have a hardening property in horizontal direction and a softening property in vertical direction in case of large deformation. Therefore the analyses considered nonlinearity of rubber bearings. Both horizontal and vertical nonlinear characteristics of rubber bearings were explained by multi-linear model. Mass point analytical models were applied. At first, seismic response analysis was executed in order to investigate influence of nonlinearity of rubber bearing upon response of building. Then sensitivity analysis was executed. Parameters of rubber bearings, oil dampers and the building were fluctuated, and influence of dispersion of these

  18. Theory of nuclear reactors. Vol. 1. Theorie der Kernreaktoren. Bd. 1. Der stationaere Reaktor

    Energy Technology Data Exchange (ETDEWEB)

    Emendoerfer, D.; Hoecker, K.H.

    1982-01-01

    An introduction is given to the elements of reactor physics and reactor calculation which refers to practice from the present point of view. It is demonstrated to the reader how the reactor characteristics relevant to construction can be calculated from atomic factors by means of neutron transport and diffusion theory; these reactor characteristics are: multiplication factor, power density distribution, burn-up, plutonium build-up, xenon vibrations, short-time behaviour. The interaction between thermo- and fluid-dynamic processes is important for this calculation. On grounds of didactics the crucial point of this book is the establishment and calculation of simple models which give a clear description of all important characteristics of the events. Attempts for more exact simulation by computer are dealt with including typical solutions.

  19. Reactor Engineering Department annual report (April 1, 1986 - March 31, 1987)

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in the fiscal year 1986 are described. The major activities of the Department are closely related to the reactor physics of very high temperature gas-cooled reactor, high conversion light water reactor and liquid metal fast breeder reactor and to blanket neutronics of fusion reactor. Contents of this report are divided into the activities on nuclear data and group constants, theoretical methods and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control, diagnosis and robotics. The activity of the Research Committee on Reactor Physics is also included. (author)

  20. Calculations of Changes in Reactivity during some regular periods of operation of JEN-1 MOD Reactor; Calculo de vairaciones de reactividad en algunos periodos regulares de operacion del reactor JEN-1 Mod.

    Energy Technology Data Exchange (ETDEWEB)

    Alcala Ruiz, F.

    1973-07-01

    By a Point-Reactor model and Perturbation Theory, changes in reactivity during some regular operating periods of JEN-1 MOD Reactor have been calculated and compared with available measured values. they were in good agreement. Also changes in reactivity have been calculated during operations at higher power levels than the present one, concluding some practical consequences for the case of increasing the present power of this reactor. (Author)

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

    Directory of Open Access Journals (Sweden)

    Sven H. De Cleyn

    2014-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  3. Measurements at the RA Reactor related to the VISA-2 project - Part 1, Start-up of the RA reactor and measurement of new RA reactor core parameters

    International Nuclear Information System (INIS)

    The objective of the measurements was determining the neutron flux in the RA reactor core. Since the number of fuel channels is increased from 56 to 68 within the VISA-2 project, it was necessary to attain criticality of the RA reactor and measure the neutron flux properties. The 'program of RA reactor start-up' has been prepared separately and it is included in this report. Measurements were divided in two phases. First phase was measuring of the neutron flux after the criticality was achieved but at zero power. During phase two measurements were repeated at several power levels, at equilibrium xenon poisoning. This report includes experimental data of flux distributions and absolute values of the thermal and fast neutron flux in the RA reactor experimental channels and values of cadmium ratio for determining the neutron epithermal flux. Data related to calibration of regulatory rods for cold un poisoned core are included

  4. Use of PC based data acquisition systems, connected to reactor shutdown system No. 1 and 2

    International Nuclear Information System (INIS)

    The intention of this material is to present the experience in use and future development of PC Based Data Acquisition Systems (DAS), connected to Reactor Shutdown System (SDS) number 1 and number 2, at Cernavoda Nuclear Power Plant (NPP) Unit 1 (U1). Two major aspects, regarding the purpose of using DAS, are the subject of the material: post-event analysis and system impairments evaluation; economic penalties reduced. (author)

  5. Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

    International Nuclear Information System (INIS)

    RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.

  6. Estimated long lived isotope activities in ET-RR-1 reactor structural materials for decommissioning study

    International Nuclear Information System (INIS)

    The first Egyptian research reactor, ET-RR-1 is tank type with light water as a moderator, coolant and reflector. Its nominal power is 2MWt and the average thermal neutron flux is 10 13 n/cm2 sec-1. Its criticality was on the fall of 1961. The reactor went through several modifications and updating and is still utilized for experimental research. A plan for decommissioning of ET-RR-1 reactor should include estimation of radioactivity in structural materials. The inventory will help in assessing the radiological consequences of decommissioning. This paper presents a conservative calculation to estimate the activity of the long lived isotopes which can be produced by neutron activation. The materials which are presented in significant quantities in the reactor structural materials are aluminum, cast iron, graphite, ordinary and iron shot concrete. The radioactivity of each component is dependent not only upon the major elements, but also on the concentration of the trace elements. The main radioactive inventory are expected to be from 60Co and 55Fe which are presented in aluminium as trace elements and in large quantities in other construction materials. (author)

  7. Measurements and calculation of reactivity in the IEA-R1 nuclear reactor

    International Nuclear Information System (INIS)

    Techniques and experimentals procedures utilized in the measurement of some nuclear parameters related to reactivity are presented. Measurements of reactivity coefficients, such as void, temperature and power, and control rod worth were made in the IEA-R1 Research Reactor. The techniques used to perform the measurements were: i) stable period (control rod calibration), ii) inverse kinetics (digital reactivity meter), iii) aluminium slab insertion in the fuel element coolant channels (void reactivity), iv) nuclear reactor core temperature changes by means of the changes in the coolant systems of reactor core (isothermal reactivity coefficient) and v) by making perturbation in the core through the control rod motions (power reactivity coefficient and control rod calibration). By using the computer codes HAMMER, HAMMER-TECHNION and CITATION, the experiments realized in the IEA-R1 reactor were simulated. From this simulation, the theoretical reactivity parameters were estimated and compared with the respective experimental results. Furthermore, in the second fuel load of Angra-1 Nuclear Power Station, the IPEN-CNEN/SP digital reactivity - meter were used in the lower power test with the aim to assess the equipment performance. Among several tests, the reacticity-meter were used in parallel with a Westinghouse analogic reativimeter-meter) to measure the heat additiona point, critical boron concentration, control rod calibration, isothermal and moderator reactivity coefficient. These tests, and the results obtained by the digital reactivity-meter are described. The results were compared with those obtained by Westinghouse analogic reactivity meter, showing excellent agreement. (author)

  8. Preliminary design calculation for the TR-1 reactor

    International Nuclear Information System (INIS)

    This study covers the preliminary design calculations of the TR-1 core with the LEU fuel that is chosen to be suitable for the TR-2 core. Different initial loadings of the core are investigated. The flux distributions, excess reactivities, power peaking factors have been calculated. The effect of beam tubes around the core and the insertion of two dry irradiation tubes in the core, on the above parameters are studied. (author)

  9. 1-D Two-phase Flow Investigation for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    During a severe accident, when a molten corium is relocated in a reactor vessel lower head, the RCF(Reactor Cavity Flooding) system for ERVC (External Reactor Vessel Cooling) is actuated and coolants are supplied into a reactor cavity to remove a decay heat from the molten corium. This severe accident mitigation strategy for maintaining a integrity of reactor vessel was adopted in the nuclear power plants of APR1400, AP600, and AP1000. Under the ERVC condition, the upward two-phase flow is driven by the amount of the decay heat from the molten corium. To achieve the ERVC strategy, the two-phase natural circulation in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. Also the natural circulation flow restriction has to be minimized. In this reason, it is needed to review the fundamental structure of insulation. In the existing power plants, the insulation design is aimed at minimizing heat losses under a normal operation. Under the ERVC condition, however, the ability to form the two-phase natural circulation is uncertain. Namely, some important factors, such as the coolant inlet/outlet areas, flow restriction, and steam vent etc. in the flow channel, should be considered for ERVC design. T-HEMES 1D study is launched to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The air injection method was used to simulate the boiling at the external reactor vessel and generate the natural circulation two-phase flow. From the experimental results, the natural circulation flow rate highly depended on inlet/outlet areas and the circulation flow rate increased as the outlet height as well as the supplied water head increased. On the other hand, the simple analysis using the drift

  10. Effect of the Lattice Pitch Variation on the Burn-up of the ETRR-1 Reactor Fuel

    International Nuclear Information System (INIS)

    The recent development in the field of reactor physics and nuclear safety calculation is providing a modern aspect to the fuel cycle economics through the reactor operations. For that, it is necessary to review the previous design of the old operating reactors. One of these reactors is the ETRR-1, which is operating since 1961. The fuel basket of this reactor contains 16 fuel rods arranged in 4 x 4 shape with a lattice pitch 17.5 mm. The lattice cell parameters and burn-up calculations were performed by using the WIMSD4 code for different lattice pitch configurations including the exist one. This study shows that, the present lattice pitch of the ETRR-1 reactor fuel element is not the best one and the 15 mm lattice pitch is more convenient from the fuel cycle economics point of view

  11. Enrichment reduction of the FRG-1 research reactor

    International Nuclear Information System (INIS)

    The GKSS research centre is participating since 1980 in the national and international efforts on enrichment reduction to clarify the needs on safety and licensing issues, to decide on fuel tests, to perform fuel tests and conversion calculations. These efforts are being made within the IAEA working groups, the German AF-program and the GKSS R and D program. There are successfully tested fuel elements with 45% and 20% enrichment: UAIx, U3O8 and U3Si2 fuel meat; U-density up to 3,7 g U/cm3 and fission densities up to 1,25 x 1021 fission/cc (ex. Pu) and five U3Si2 fuel plates (ρ(U) 4,75 g U/cm3). A safety report for the conversion of the FRG-1 to LEU fuel (U3Si2, ρ = 37 g U/cm3) has been submitted to the licensing authority. After examining the safety report by independent experts (TUeV) and the licensing authority the license for the conversion has been granted in May 1988. The FRG-1 has been totally converted in 1991 to LEU fuel elements. All necessary tests have been performed in a three- week period. The conversion procedure includes a reduction in core size by a factor of two and the replacement of old oval control rods by fork type absorbers. Therefore the thermal neutron flux in beam tube positions could be increased by more than a factor of two. (author)

  12. [The Belgian project for the prevention of cardiovascular diseases: a model of multifactorial prevention].

    Science.gov (United States)

    Kornitzer, M

    1989-01-01

    Résults are presented from the "belgian heart disease prevention project, part of the WHO european collaborative trial in the multifactorial prevention of coronary heart disease (CHD)". 19.409 men aged 40-59 yr took part; they were employed in thirty factories which formed the allocation units for a randomised controlled trial lasting 5-6 yr. The intervention package consisted largely of health education promoting a cholesterol-lowering diet, smoking cessation, weight control, physical activity, and treatment of arterial hypertension. A programme of information was supplemented by face-to-face counselling at the workplace by two physicians attached to the project. The coronary risk profile was reduced in the intervention group, compared with that in the control group, especially during the first 4 yr, by effects on serum cholesterol, number of cigarettes smoked daily, and arterial blood-pressure. Total mortality was 17.5% lower in the intervention group than in the control group (p = 0.038); coronary mortality was reduced by a non-significant 20.8% whereas CHD incidence (non-fatal myocardial infarction plus fatal myocardial infarction plus sudden deaths) was reduced by 24.5%, (P = 0.031). Non-fatal myocardial infarction (not a major end-point) was similarly reduced by 26.1% (p = 0.030). PMID:2679938

  13. Risk factors and effect of selective removal on retroviral infections prevalence in Belgian stray cats.

    Science.gov (United States)

    Garigliany, M; Jolly, S; Dive, M; Bayrou, C; Berthemin, S; Robin, P; Godenir, R; Petry, J; Dahout, S; Cassart, D; Thiry, E; Desmecht, D; Saegerman, C

    2016-01-01

    The aim of this study was to evaluate the effect of several risk/protective factors and predictors on the prevalence of feline immunodeficiency virus (FIV) and feline leukaemia virus (FeLV) infections in 302 stray cats captured during a trap-neuter-release programme in a mixed urban-rural area from Belgium, from 2010 to 2012. The impact of selective removal of FIV-positive cats on the apparent prevalence in the remaining population over this three-year period was also assessed. The seroprevalences over three years were 18.8 per cent for FIV and 0.7 per cent for FeLV. For FIV, the seroprevalence decreased significantly from the first year of the programme (2010; 30.5 per cent) to the last (2012; 13.1 per cent). Sex (male) and age (adult and old cats) were risk factors, while the year of sampling (years 2011 and 2012) was a protective factor. Age, sex and location were the most relevant predictors of FIV status. The data presented in this study revealed a very high FIV seroprevalence in Belgian stray cats, while FeLV was almost absent. The selective removal of positive cats had a drastic effect on the FIV seroprevalence in the remaining cat population. PMID:26744011

  14. Analysis of liquid radioactive wastes of Angra-1 reactor

    International Nuclear Information System (INIS)

    Any activity that produces or uses radioactive materials generates radioactive wastes. Normal operation of nuclear power plant produces radioactive waste that can be in gas, liquid or solid form and its level of radioactivity can vary. Gases and liquids wastes are treated and released into environment. The main source of radioactivity released to environment from Angra 1 are liquids from Waste Monitor Tanks. Those releases are under administrative control to meet the discharge limits established by Comissao Nacional de Energia Nuclear (CNEN). A representative sample of each batch is taken for analysis for principal gamma- emitting radionuclides and, if the analysis indicate that release can be made, the quantity of activity is recorded. Within the licensing process of Angra 1, monthly a proportional composite samples are made up with a aliquot of each batch and sent to Instituto de Radioprotecao e Dosimetria (IRD) to analyze and compare with the results reported. This comparative analyses showed that when the activity of that samples was very high, the activity measured on composite samples was higher than the sum of the activities measured on each batch. The operator was advised and requested to identify and solve the problem. This work presents the problem occurred and the solution found to improve the performance of measurements. (author)

  15. NBR ISO 9001 Certification for activities carried out in IEA-R1 reactor

    International Nuclear Information System (INIS)

    Since its inauguration in 1957, the IEA-R1 research reactor has been used mainly for research, development and teaching by scientific community. In the last years, with the increase of the commercial radiopharmaceutical production by Radiopharmacy Center of IPEN, the IEA-R1 reactor was recognized as a service supplier for that center and has received a treatment more commercial from IPEN Management. In 1999 the radiopharmaceutical production obtained the NBR ISO 9002 Certification, since that the IPEN Management considered convenient to invest in the certification of its internal suppliers. In this context, in 2001 the Research Reactor Center (CRPq) began the implantation of a Quality Management System (QMS) based on NBR 9001: 2000 standard, for activities related to the operation and maintenance of the IEA-R1 research reactor and irradiation services. This QMS was structured to incorporate tools already implemented in order to complain the requirements related to nuclear and radiological safe for a nuclear installation established by the regulatory organism. The QMS is supported by a documentation system composed of approximately 150 documents including quality manual, business and action plans, operational procedures and work instruction. Carlos Alberto Vanzolini Foundation (FCAV), an INMETRO certified organism, certified the 'Operation and Maintenance of the IEA-R1 Research Reactor and Irradiation Services' in December 2002. In 2003 and 2004, the QMS was audited by FCAV that determined the maintenance of the certification. This work presents the main steps of the QMS implementation, including the difficulties found and results obtained in the process. (author)

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

    OpenAIRE

    De Baerdemaeker J.; Quenon V.

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

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

    2005-01-01

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

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

    OpenAIRE

    Backeljau, T.

    1988-01-01

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

  20. Results of the Level 1 probabilistic risk assessment (PRA) of internal events for heavy water production reactors (U)

    International Nuclear Information System (INIS)

    This paper reports on a full-scope probabilistic risk assessment (PRA) performed for the Savannah River Site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident