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Sample records for belgian reactor 2

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

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

    International Nuclear Information System (INIS)

    Gubel, P.

    1996-01-01

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

  5. 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.

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

    International Nuclear Information System (INIS)

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

    1997-10-01

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

  7. Ageing management of the BR2 research reactor

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    Couplet, D.; Francoise, T.

    2001-01-01

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

  9. SAFIR-2 and the Belgian methodological R and D programme on deep disposal

    International Nuclear Information System (INIS)

    Preter, Peter de

    2002-01-01

    Peter de Preter (NIRAS/ONDRAF, Belgium) provided an overview of the Belgian programme of research and development on deep disposal. The planned submission in December 2001 of SAFIR 2 (the second Safety Assessment and Feasibility Interim Report) would mark an important milestone, as the report would inform a decision by the Belgian government on the nature of future research. An independent committee of scientists established by NIRAS/ONDRAF had reviewed a draft version of the report. The committee was generally in agreement with the technical R and D priorities proposed in the report but suggested that there should be more integration of technical and societal aspects. The committee also recommended that a future research programme should compare the option of deep disposal with other strategies for long-term management of radioactive wastes. It was suggested that a strategic environmental assessment might provide an appropriate mechanism for comparing alternative management strategies, and would enable societal dimensions also to be addressed

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

    CERN Multimedia

    2003-01-01

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

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

    CERN Multimedia

    2003-01-01

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

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

    CERN Multimedia

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  15. The Belgian nuclear research centre

    International Nuclear Information System (INIS)

    Moons, F.

    2001-01-01

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

  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. Material test reactor fuel research at the BR2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

    2012-03-15

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

  18. Reactor BR2

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2000-07-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported.

  19. Reactor BR2

    International Nuclear Information System (INIS)

    Gubel, P.

    2000-01-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

  20. Balancing selection on CDH2 may be related to the behavioral features of the Belgian Malinois.

    Directory of Open Access Journals (Sweden)

    Xue Cao

    Full Text Available The Belgian Malinois (BM is an excellent working dog that typically shows a circling behavior when placed in a confined space. Moreover, individuals showing moderate running in circles (one kind of obsessive compulsive behavior in confined spaces typically show better work performance compared to those without the circling behavior or to those with a serious circling behavior (which can be defined as an obsessive compulsive disorder (OCD. To determine whether the candidate gene CDH2, Cadherin 2, which is associated with OCD in the Doberman pinscher breed of dogs and in humans, was linked with this behavioral character in the BM, population genetic analyses were performed on a BM population and a natural population of the Chinese indigenous dog (CID. Many genetic signals of balancing selection were detected for one specific region of the CDH2 gene, which suggests that a genomic block, which is included in the CDH2 gene, experienced balancing selection in the BM, and that the CDH2 gene might be associated with the behavioral characteristics of the BM dog (a balance between circling behavior and work performance. Moreover one specific variant, G63913941A, which creates a predicted transcription factor-binding site, may be the key mutation in the CDH2 gene affecting the behavior of BMs by allowing the binding of a transcription factor and increasing CDH2 expression.

  1. Reactor BR2: Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2000-01-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. A safety audit was conduced by the IAEA, the conclusions of which demonstrated the excellent performance of the plant in terms of operational safety. In 1999, the CALLISTO facility was extensively used for various programmes involving LWR pressure vessel materials, IASCC of LWR structural materials, fusion reactor materials and martensic steels for use in ADS systems. In 1999, BR2's commercial programmes were further developed

  2. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2002-01-01

    The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system

  3. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2001-01-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given

  4. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2002-04-01

    The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

  5. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2001-04-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

  6. BR2 Reactor: Introduction

    International Nuclear Information System (INIS)

    Moons, F.

    2007-01-01

    The irradiations in the BR2 reactor are in collaboration with or at the request of third parties such as the European Commission, the IAEA, research centres and utilities, reactor vendors or fuel manufacturers. The reactor also contributes significantly to the production of radioisotopes for medical and industrial applications, to neutron silicon doping for the semiconductor industry and to scientific irradiations for universities. Along the ongoing programmes on fuel and materials development, several new irradiation devices are in use or in design. Amongst others a loop providing enhanced cooling for novel materials testing reactor fuel, a device for high temperature gas cooled fuel as well as a rig for the irradiation of metallurgical samples in a Pb-Bi environment. A full scale 3-D heterogeneous model of BR2 is available. The model describes the real hyperbolic arrangement of the reactor and includes the detailed 3-D space dependent distribution of the isotopic fuel depletion in the fuel elements. The model is validated on the reactivity measurements of several tens of BR2 operation cycles. The accurate calculations of the axial and radial distributions of the poisoning of the beryllium matrix by 3 He, 6 Li and 3T are verified on the measured reactivity losses used to predict the reactivity behavior for the coming decades. The model calculates the main functionals in reactor physics like: conventional thermal and equivalent fission neutron fluxes, number of displacements per atom, fission rate, thermal power characteristics as heat flux and linear power density, neutron/gamma heating, determination of the fission energy deposited in fuel plates/rods, neutron multiplication factor and fuel burn-up. For each reactor irradiation project, a detailed geometry model of the experimental device and of its neighborhood is developed. Neutron fluxes are predicted within approximately 10 percent in comparison with the dosimetry measurements. Fission rate, heat flux and

  7. Reactor Materials Research

    International Nuclear Information System (INIS)

    Van Walle, E.

    2001-01-01

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

  8. 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)

  9. Pressure Vessel Steel Research: Belgian Activities

    International Nuclear Information System (INIS)

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

    1994-03-01

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

  10. Pressure Vessel Steel Research: Belgian Activities

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-03-01

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

  11. Prevalence and co-prevalence of comorbidities in Belgian patients with type 2 diabetes mellitus: a transversal, descriptive study.

    Science.gov (United States)

    Hermans, Michel P; Dath, Nicolas

    2018-02-01

    Most patients with type 2 diabetes mellitus (T2DM) have at least one comorbid chronic disease. These comorbidities increase disease burden and costs and may impact the selection of glucose-lowering therapies. The aim of our study was to describe the prevalence and co-prevalence of comorbidities in patients with T2DM in Belgium. This was a single centre, transversal, descriptive study performed at Saint-Luc University Hospital, Brussels, Belgium. T2DM patients aged >18 years presenting to the outpatient diabetes clinic and regularly followed (≥1 year) between July 2011 and 2016 were included in the database. Information on descriptive characteristics was collected and several comorbidities were assessed. Overall, 778 patients were included. The median age was 68 years. The most frequently observed comorbidities were hypertension (85.5%), metabolic syndrome (MetS; 84.6%) and non-alcoholic hepatic steatosis (71.7%). 78.5% of the T2DM patients had ≥4 comorbidities. The highest co-prevalence of comorbidities was the combination of MetS and hypertension (75.8%), MetS and overweight (60.5%), and hypertension and overweight (56.6%). The observed prevalence and co-prevalence of comorbidities in T2DM Belgian patients was high, with 95.0% patients having at least two comorbid conditions.

  12. The post irradiation examination of a sphere-pac (UPu)C fuel pin irradiated in the BR-2 reactor (MFBS 7 experiment)

    International Nuclear Information System (INIS)

    Smith, L.; Aerne, E.T.; Buergisser, B.; Flueckiger, U.; Hofer, R.; Petrik, F.

    1979-09-01

    A pin fuelled with Swiss made (UPu)C microspheres has been successfully irradiated to a peak burn-up of 6% fima in the Belgian BR2 Reactor. The pin, rated up to 95 kW/m, was intact after irradiation and exhibited a peak strain of just over 0.5%. The results of the post irradiation examination are reported. (Auth.)

  13. Prevalence of the AMHR2 mutation in Miniature Schnauzers and genetic investigation of a Belgian Malinois with persistent Müllerian duct syndrome.

    Science.gov (United States)

    Smit, M M; Ekenstedt, K J; Minor, K M; Lim, C K; Leegwater, Paj; Furrow, E

    2018-04-01

    Persistent Müllerian duct syndrome (PMDS) is a sex-limited disorder in which males develop portions of the female reproductive tract. Important consequences of PMDS are cryptorchidism and its sequelae of infertility and increased risk of testicular cancer. Anti-Müllerian hormone (AMH) and its receptor (AMHR2) induce the regression of the Müllerian ducts in male embryos. In Miniature Schnauzer dogs, the genetic basis has been identified as an autosomal recessive nonsense mutation in AMHR2, but the allele frequency of the mutation is unknown. Thus, the primary objective of this study was to estimate the prevalence of the AMHR2 mutation in North American Miniature Schnauzers, in order to ascertain the value of genetic testing in this breed. An additional objective was to determine whether mutations in AMH or AMHR2 were responsible for PMDS in a Belgian Malinois; this would aid development of a genetic test for the Belgian Malinois breed. Genomic DNA from 216 Miniature Schnauzers (including one known PMDS case) was genotyped for the AMHR2 mutation, and DNA from a single PMDS-affected Belgian Malinois was sequenced for all coding exons of AMH and AMHR2. The Miniature Schnauzer cohort had an AMHR2 mutation allele frequency of 0.16 and a carrier genotypic frequency of 0.27. The genetic basis for PMDS in the Belgian Malinois was not determined, as no coding or splicing mutations were identified in either AMH or AMHR2. These findings support a benefit to AMHR2 mutation testing Miniature Schnauzers used for breeding or with cryptorchidism. © 2017 Blackwell Verlag GmbH.

  14. An attempt to reconstruct geodynamically the Belgian palaeogene basin from mineralogical clay data, from radioelement (U, Th, K2O) geochemistry and palynology

    International Nuclear Information System (INIS)

    Quinif, Yves; Mercier, Monique; Dupuis Christian; Roche, Emile

    1983-01-01

    By analysing clay mineralogy, U-Th-K 2 O geochemistry and palynology data, we find that the Belgian Palaeogene is dominated by the action of a biostasic period toward the end of the lower Eocene. At the middle Eocene starts a new phase, at least a rhexistasic one at the begining. In the Basin, the geochemical and mineralogical traces of this phase just start from the upper Eocene [fr

  15. The utility of different reactor types for the research

    International Nuclear Information System (INIS)

    Stiennon, G.

    1983-01-01

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

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

    Science.gov (United States)

    Hoste, J; Daci, E; Mathieu, C

    2014-06-01

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

  17. Two Belgian University Hospitals

    Directory of Open Access Journals (Sweden)

    M. Huylebrouck

    2012-01-01

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

  18. BR2 reactor neutron beams

    International Nuclear Information System (INIS)

    Neve de Mevergnies, M.

    1977-01-01

    The use of reactor neutron beams is becoming increasingly more widespread for the study of some properties of condensed matter. It is mainly due to the unique properties of the ''thermal'' neutrons as regards wavelength, energy, magnetic moment and overall favorable ratio of scattering to absorption cross-sections. Besides these fundamental reasons, the impetus for using neutrons is also due to the existence of powerful research reactors (such as BR2) built mainly for nuclear engineering programs, but where a number of intense neutron beams are available at marginal cost. A brief introduction to the production of suitable neutron beams from a reactor is given. (author)

  19. Belgian national report

    International Nuclear Information System (INIS)

    Berthe, J.

    1998-01-01

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

  20. Recent advances in the utilization and the irradiation technology of the refurbished BR2 reactor

    International Nuclear Information System (INIS)

    Dekeyser, J.; Benoit, P.; Decloedt, C.; Pouleur, Y.; Verwimp, A.; Weber, M.; Vankeerberghen, M.; Ponsard, B.

    1999-01-01

    Operation and utilization of the materials testing reactor BR2 at the Belgian Nuclear Research Centre (SCK·CEN) has since its start in 1963 always followed closely the needs and developments of nuclear technology. In particular, a multitude of irradiation experiments have been carried out for most types of nuclear power reactors, existing or under design. Since the early 1990s and increased focus was directed towards more specific irradiation testing needs for light water reactor fuels and materials, although other areas of utilization continued as well (e.g. fusion reactor materials, safety research, ...), including also the growing activities of radioisotope production and silicon doping. An important milestone was the decision in 1994 to implement a comprehensive refurbishment programme for the BR2 reactor and plant installations. The scope of this programme comprised very substantial studies and hardware interventions, which have been completed in early 1997 within planning and budget. Directly connected to this strategic decision for reactor refurbishment was the reinforcement of our efforts to requalify and upgrade the existing irradiation facilities and to develop advanced devices in BR2 to support emerging programs in the following fields: - LWR pressure vessel steel, - LWR irradiation assisted stress corrosion cracking (IASCC), - reliability and safety of high-burnup LWR fuel, - fusion reactor materials and blanket components, - fast neutron reactor fuels and actinide burning, - extension and diversification of radioisotope production. The paper highlights these advances in the areas of BR2 utilisation and the ongoing development activities for the required new generation of irradiations devices. (author)

  1. LEU fuel cycle analyses for the Belgian BR2 Research Reactor

    International Nuclear Information System (INIS)

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

    1988-01-01

    Equilibrium fuel cycle characteristics were calculated for reference HEU and two proposed LEU fuel cycles using an 11-group diffusion-theory neutron flux solution in hexagonal-Z geometry. The diffusion theory model was benchmarked with a detailed Monte Carlo core model. The two proposed LEU fuel designs increased the 235 U loading 20% and the fuel meat volume 51%. The first LEU design used 10 B as a burnable absorber. Either proposed LEU fuel element would provide equilibrium fuel cycle characteristics similar to those of the HEU fuel cycle. Irradiation rates of Co control followers and Ir disks in the center of the core were reduced 6 ± 1% in the LEU equilibrium core compared to reference HEU core. 11 refs., 4 figs., 5 tabs

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

    International Nuclear Information System (INIS)

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

    1986-01-01

    Fuel cycle calculations have been performed with reference HEU fuel and LEU fuel using Cd wires or boron as burnable absorbers. The 235 U content in the LEU element has increased 20% to 480g compared to the reference HEU element. The number of fuel plates has remained unchanged while the fuel meat thickness has increased to 0.76 mm from 0.51 mm. The LEU meat density is 5.1 Mg U/m 3 . The reference fuel cycle was a 31 element core operating at 56 MW with a 19.8 day cycle length and eight fresh elements loaded per cycle. Comparable fuel cycle characteristics can be achieved using the proposed LEU fuel element with either Cd wires or boron burnable absorbers. The neutron flux for E/sub n/ > 1 eV changes very little (<5%) in LEU relative to HEU cores. Thermal flux reductions are 5 to 10% in non-fueled positions, and 20 to 30% in fuel elements

  3. Refurbishment of BR2 (Phase 4 and 5)[1997 Scientific Report of the Belgian Nuclear Research Centre

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P.; Dekeyser, J.; Van der Auwera, J.

    1998-07-01

    The extensive refurbishment of the BR-2 materials testing reactor should allow another 10 to 15years of continued operation. The refurbishment programme is required in order to comply with modern safety standards, to enhance the reliability of operation, and to compensate for the ageing of the installations of a facility that has reached about 35 years of intensive service. The main objectives and achievements of phase 4 and 5 are described.

  4. PARR-2: reactor description and experiments

    International Nuclear Information System (INIS)

    Wyne, M.F.; Meghji, J.H.

    1990-12-01

    PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

  5. Cost effectiveness of atorvastatin in patients with type 2 diabetes mellitus: a pharmacoeconomic analysis of the collaborative atorvastatin diabetes study in the belgian population.

    Science.gov (United States)

    Annemans, L; Marbaix, S; Webb, K; Van Gaal, L; Scheen, A

    2010-01-01

    Patients with type 2 diabetes mellitus have a high risk of developing cardiovascular (CV) disease. The clinical benefit of use of statins in patients with type 2 diabetes has been demonstrated in several randomized, controlled trials, including the CARDS clinical trial. Based on the clinical CARDS data, the favourable cost effectiveness of atorvastatin 10 mg in patients with type 2 diabetes has been demonstrated in countries such as the UK and France. This study aimed to estimate the cost effectiveness in the Belgian setting of atorvastatin 10 mg compared with no treatment for the primary prevention of CV events in type 2 diabetes patients without a history of CV disease. A Markov model with 1-year cycles was developed to simulate the CV event and death risk according to the therapeutic approach initiated. The transition probabilities for CV events in the 'no statin treatment' group were derived from the risk equations reported from the large UKPDS. Risk reductions from the CARDS clinical trial were used to adjust these CV event probabilities in the atorvastatin 10 mg treatment group. The characteristics of type 2 diabetes patients without a CV history were derived from the Belgian OCAPI survey. The public healthcare payers' perspective was taken into account for costing. The direct medical costs of CV events were based on the Public Health Authorities' hospital database for acute care costs and on the literature for the follow-up costs. The impact on the reimbursement system of generic entry to the market was considered in the drug cost. Costs were valued as at year 2009; costs and outcomes were discounted at 3% and 1.5%, respectively. Based on a 5-year time horizon, atorvastatin was demonstrated to be cost effective with an incremental cost/quality-adjusted life-year (QALY) of euro 16,681. Over a lifetime horizon (25 years), atorvastatin was demonstrated to be a cost-saving therapeutic intervention. At a threshold of euro 30,000/QALY, atorvastatin had a 98

  6. Spent fuel strategy for the BR2 reactor

    International Nuclear Information System (INIS)

    Gubel, P.; Collard, G.

    1998-01-01

    The Belgian MTR reactor is fuelled with HEU UAl x elements and the fuel cycle was normally closed by reprocessing consecutively in Belgium (Eurochemic), France (Marcoule) and finally in the U.S.A. (Idaho Falls and Savannah River). When the acceptance of spent fuel by the U.S. was terminated, the facility was left with a huge backlog of used elements stored under water. After a few years, urgent and mandatory actions were required to maintain the BR2 facility operating. Later the accent was put on the evaluation of an optimum long term solution for the BR2 spent fuel during the projected 15 years life extension after the refurbishment executed between 1995 and 1997. The paper gives an overview of these successive actions taken during the last years as well as the handled various criteria for comparing and evaluating the available long-term alternatives. After commitment to reprocessing in existing facilities operated for aluminum fuels the focus of the BR2 fuel cycle strategy is now moving to the procurement of the necessary HEU fuel for securing the long-term operation of the facility. (author)

  7. The Belgian Nuclear Higher Education Network

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  8. Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

    Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

  9. Reactor handbook. 2. rev. ed.

    International Nuclear Information System (INIS)

    Lederer, B.J.; Wildberg, D.W.

    1992-01-01

    On the basis of the guidelines on expert knowledge, the book discusses the subjects of atomic physics, heat transfer, nuclear power plants, reactor materials, radiation protection, reactor safety, reactor instrumentation, and reactor operation, with special regard to nuclear power plants with LWR-type reactors. The book is intended for shift personnel, especially gang bosses, reactor operators, and control station operators: for this reason a practical and rather popular style has been chosen. However, the book will also be a manual for other operating personnel, personnel of producer companies, expert organisations, authorities, and students. It can be used as a textbook for staff training, a manual for the practice, and as accompanying book for teaching at nuclear engineering schools. (orig.) With 173 figs [de

  10. Reactor pressure vessel steels

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  11. TA-2 Water Boiler Reactor Decommissioning Project

    International Nuclear Information System (INIS)

    Durbin, M.E.; Montoya, G.M.

    1991-06-01

    This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m 3 of low-level solid radioactive waste and 35 m 3 of mixed waste. 15 refs., 25 figs., 3 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-05-01

    The BR2 reactor (Mol, Belgium) is a high-flux materials testing reactor. The fuel is 93% {sup 235}U enriched uranium. The nominal power ranges from 60 to 100 MW. The main features of the design are the following: 1) maximum neutron flux, thermal: 1.2 x 10{sup 15} n/cm{sup 2} s; fast (E > 0.1 MeV) : 8.4 x 10{sup 14} n /cm{sup 2} s; 2) great flexibility of utilization: the core configuration and operation mode can be adapted to the experimental loading; 3) neutron spectrum tailoring; 4) availability of five 200 mm diameter channels besides the standard channels (84 mm diameter); 5) access to the top and bottom covers of the reactor authorizing the irradiation of loops. The reactor is used to study the behaviour of fuel elements and structural materials intended for future nuclear power stations of several types (fission and fusion). Irradiations are carried out in connection with performance tests up to very high burn-up or neutron fluence as well as for safety experiments, power cycling experiments, and generally speaking, tests under off-normal conditions. Irradiations for nuclear transmutation (production of high specific activity radio-isotopes and transplutonium elements), neutron-radiography, use of beam tubes for physics studies, and gamma irradiations are also carried out. The BR2 is used in support of Belgian programs, at the request of utilities, industry and universities and in the framework of international agreements. The paper reviews the past and ongoing upgrading and enhancement of reactor capabilities as well as those under study or consideration, namely with regard to: reactor equipment, fuel elements, irradiation facilities, reactor operation conditions and long-term strategy. (author)

  13. Implementation of a program for type 2 diabetes based on the Chronic Care Model in a hospital-centered health care system: "the Belgian experience"

    Directory of Open Access Journals (Sweden)

    Van Royen Paul

    2009-08-01

    Full Text Available Abstract Background Most research publications on Chronic Care Model (CCM implementation originate from organizations or countries with a well-structured primary health care system. Information about efforts made in countries with a less well-organized primary health care system is scarce. In 2003, the Belgian National Institute for Health and Disability Insurance commissioned a pilot study to explore how care for type 2 diabetes patients could be organized in a more efficient way in the Belgian healthcare setting, a setting where the organisational framework for chronic care is mainly hospital-centered. Methods Process evaluation of an action research project (2003–2007 guided by the CCM in a well-defined geographical area with 76,826 inhabitants and an estimated number of 2,300 type 2 diabetes patients. In consultation with the region a program for type 2 diabetes patients was developed. The degree of implementation of the CCM in the region was assessed using the Assessment of Chronic Illness Care survey (ACIC. A multimethod approach was used to evaluate the implementation process. The resulting data were triangulated in order to identify the main facilitators and barriers encountered during the implementation process. Results The overall ACIC score improved from 1.45 (limited support at the start of the study to 5.5 (basic support at the end of the study. The establishment of a local steering group and the appointment of a program manager were crucial steps in strengthening primary care. The willingness of a group of well-trained and motivated care providers to invest in quality improvement was an important facilitator. Important barriers were the complexity of the intervention, the lack of quality data, inadequate information technology support, the lack of commitment procedures and the uncertainty about sustainable funding. Conclusion Guided by the CCM, this study highlights the opportunities and the bottlenecks for adapting chronic care

  14. Core monitoring at the WNP-2 reactor

    International Nuclear Information System (INIS)

    Skeen, D.R.; Torres, R.H.; Burke, W.J.; Jenkins, I.; Jones, S.W.

    1992-01-01

    The WNP-2 reactor is a 3,323-MW(thermal) boiling water reactor (BWR) that is operated by the Washington Public Power Supply System. The WNP-2 reactor began commercial operation in 1984 and is currently in its eighth cycle. The core monitoring system used for the first cycle of operation was supplied by the reactor vendor. Cycles 2 through 6 were monitored with the POWERPLEX Core Monitoring Software System (CMSS) using the XTGBWR simulation code. In 1991, the supply system upgraded the core monitoring system by installing the POWERPLEX 2 CMSS prior to the seventh cycle of operation for WNP-2. The POWERPLEX 2 CMSS was developed by Siemens Power Corporation (SPC) and is based on SPC's advanced state-of-the-art reactor simulator code MICROBURN-B. The improvements in the POWERPLEX 2 system are possible as a result of advances in minicomputer hardware

  15. Safety features of TR-2 reactor

    International Nuclear Information System (INIS)

    Tuerker, T.

    2001-01-01

    TR-2 is a swimming pool type research reactor with 5 MW thermal power and uses standard MTR plate type fuel elements. Each standard fuel element consist of 23 fuel plates with a meat + cladding thickness of 0.127 cm, coolant channel clearance is 0.21 cm. Originally TR-2 is designed for %93 enriched U-Al. Alloy fuel meat.This work is based on the preparation of the Final Safety Analyses Report (FSAR) of the TR-2 reactor. The main aspect is to investigate the behaviour of TR-2 reactor under the accident and abnormal operating conditions, which cowers the accident spectrum unique for the TR-2 reactor. This presentation covers some selected transient analyses which are important for the safety aspects of the TR-2 reactor like reactivity induced startup accidents, pump coast down (Loss of Flow Accident, LOFA) and other accidents which are charecteristic to the TR-2

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

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    1998-10-01

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

  17. 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.

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

    International Nuclear Information System (INIS)

    Decreton, M.

    1999-10-01

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

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

    International Nuclear Information System (INIS)

    Decreton, M.

    1998-10-01

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

  20. Belgian national report

    International Nuclear Information System (INIS)

    Berthe, J.

    1995-01-01

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

  1. Fission product release from SLOWPOKE-2 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Harnden-Gillis, A M.C. [Queen` s Univ., Kingston, ON (Canada). Dept. of Physics

    1994-12-31

    Increasing radiation fields at several SLOWPOKE-2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements. 19 refs., 4 tabs., 8 figs.

  2. Reactor physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.

    1998-01-01

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

  3. BR2 Reactor: Irradiation of fuels

    International Nuclear Information System (INIS)

    Verwimp, A.

    2005-01-01

    Safe, reliable and economical operation of reactor fuels, both UO 2 and MOX types, requires in-pile testing and qualification up to high target burn-up levels. In-pile testing of advanced fuels for improved performance is also mandatory. The objectives of research performed at SCK-CEN are to perform Neutron irradiation of LWR (Light Water Reactor) fuels in the BR2 reactor under relevant operating and monitoring conditions, as specified by the experimenter's requirements and to improve the on-line measurements on the fuel rods themselves

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-01

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

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

    Science.gov (United States)

    2005-10-01

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

  6. Refurbishing the BR2 materials testing reactor

    International Nuclear Information System (INIS)

    Baugnet, J.M.; Dekeyser, J.; Gubel, P.

    1995-01-01

    SCK/CEN is refurbishing its BR2 reactor to allow its further operation during the next 15 years; in doing so, it chooses to keep BR2 available for future scientific and technological irradiation programs within an international context. (author) 2 figs

  7. TMI-2 reactor vessel plenum final lift

    International Nuclear Information System (INIS)

    Wilson, D.C.

    1986-01-01

    Removal of the plenum assembly from the TMI-2 reactor vessel was necessary to gain access to the core region for defueling. The plenum was lifted from the reactor vessel by the polar crane using three specially designed pendant assemblies. It was then transferred in air to the flooded deep end of the refueling canal and lowered onto a storage stand where it will remain throughout the defueling effort. The lift and transfer were successfully accomplished on May 15, 1985 in just under three hours by a lift team located in a shielded area within the reactor building. The success of the program is attributed to extensive mockup and training activities plus thorough preparations to address potential problems. 54 refs

  8. Studsvik's R2 reactor - Review of activities

    Energy Technology Data Exchange (ETDEWEB)

    Grounes, Mikael; Tomani, Hans; Graeslund, Christian; Rundquist, Hans; Skoeld, Kurt [Studsvik Nuclear AB, Nykoeping (Sweden)

    1993-07-01

    A general description of the R2 reactor, its associated facilities and its history is given. The facilities and range of work are described for the following types of activities: fuel testing, materials testing, neutron transmutation doping of silicon, activation analysis, radioisotope production and basic research including thermal neutron scattering, nuclear chemistry and neutron capture radiography. (author)

  9. Irradiation techniques at BR2 reactor

    International Nuclear Information System (INIS)

    Hebel, W.

    1978-01-01

    Since 1963 the material testing reactor BR2 at Mol is operated for the realisation of numerous research programs and experiments on the behavior of materials under nuclear radiation and in particular under intensive neutron exposure. During this period special irradiation techniques and experimental devices were developed according to the desiderata of the different experiments and to the irradiation possibilities offered at BR2. The design and the operating characteristics of quite a number of those irradiation rigs of proven reliability may be used or can be made available for new irradiation experiments. A brief description is given of some typical irradiation devices designed and constructed by CEN/SCK, Technology and Energy Dpt. They are compiled according to their main use for the different research and development programs realized at BR2. Their eventual application however for different objectives could be possible. A final chapter summarizes the principal irradiation conditions offered by BR2 reactor. (author)

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  11. 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…

  12. EL-2 reactor: Thermal neutron flux distribution

    International Nuclear Information System (INIS)

    Rousseau, A.; Genthon, J.P.

    1958-01-01

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

  13. TMI-2 reactor vessel head removal

    International Nuclear Information System (INIS)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1985-09-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  15. Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

    International Nuclear Information System (INIS)

    Bussac, J.; Reuss, P.

    1985-01-01

    This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

  16. Reactor safety analysis

    International Nuclear Information System (INIS)

    Arien, B.

    1998-01-01

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

  17. TMI-2 reactor vessel head removal

    International Nuclear Information System (INIS)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1984-12-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

  18. Decommissioning of reactor facilities (2). Required technology

    International Nuclear Information System (INIS)

    Yanagihara, Satoshi

    2014-01-01

    Decommissioning of reactor facilities was planned to perform progressive dismantling, decontamination and radioactive waste disposal with combination of required technology in a safe and economic way. This article outlined required technology for decommissioning as follows: (1) evaluation of kinds and amounts of residual radioactivity of reactor facilities with calculation and measurement, (2) decontamination technology of metal components and concrete structures so as to reduce worker's exposure and production of radioactive wastes during dismantling, (3) dismantling technology of metal components and concrete structures such as plasma arc cutting, band saw cutting and controlled demolition with mostly remote control operation, (3) radioactive waste disposal for volume reduction and reuse, and (4) project management of decommissioning for safe and rational work to secure reduction of worker's exposure and prevent the spreading of contamination. (T. Tanaka)

  19. Refurbishment programme for the BR2-reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koonen, E [Centre d' Etude de l' Energie Nucleaire, Studiecentrum voor Kernenergie, BR2 Department, Boeretang, Mol (Belgium)

    1992-07-01

    BR2 is a high flux engineering test reactor, which differs from comparable material testing reactors by its specific core array (fig. 1). It is a heterogeneous, thermal, tank-in-pool type reactor, moderated by beryllium and light water, which serves also as coolant. The fuel elements consist of cylindrical assemblies loaded in channels materialized by hexagonal beryllium prisms. The central 200 mm channel is vertical, while all others are inclined and form a hyperbolical arrangement around the central one. This feature combines a very compact core with the requirement of sufficient space for individual access to all channels through penetrations in the top cover of the aluminium pressure vessel. Each channel may hold a fuel element, a control rod, an experiment, an irradiation device or a beryllium plug. The refurbishment Program According to the present programme of C.E.N./S.C.K., BR2 will be in operation until 1996. At that time, the beryllium matrix will reach its foreseen end-of-life. In order to continue operation beyond this point, a thorough refurbishment of the reactor is foreseen, in addition to the unavoidable replacement of the matrix, to ensure quality of the installation and compliance with modern standards. Some fundamental options have been taken as a starting point: BR2 will continue to be used as a classical MTR, i.e. fuel and material irradiations and safety experiments with some additional service-activities. The present configuration is optimized for that use and there is no specific experimental requirement to change the basic concepts and performance characteristics. From the customers viewpoint, it is desirable to go ahead with the well-known features of BR2, to maintain a high degree of availability and reliability and to minimize the duration of the long shutdown. It is also important to limit the amount of nuclear liabilities. So the objective of the refurbishment programme is the life extension of BR2 for about 15 years, corresponding to

  20. Refurbishment programme for the BR2-reactor

    International Nuclear Information System (INIS)

    Koonen, E.

    1992-01-01

    BR2 is a high flux engineering test reactor, which differs from comparable material testing reactors by its specific core array (fig. 1). It is a heterogeneous, thermal, tank-in-pool type reactor, moderated by beryllium and light water, which serves also as coolant. The fuel elements consist of cylindrical assemblies loaded in channels materialized by hexagonal beryllium prisms. The central 200 mm channel is vertical, while all others are inclined and form a hyperbolical arrangement around the central one. This feature combines a very compact core with the requirement of sufficient space for individual access to all channels through penetrations in the top cover of the aluminium pressure vessel. Each channel may hold a fuel element, a control rod, an experiment, an irradiation device or a beryllium plug. The refurbishment Program According to the present programme of C.E.N./S.C.K., BR2 will be in operation until 1996. At that time, the beryllium matrix will reach its foreseen end-of-life. In order to continue operation beyond this point, a thorough refurbishment of the reactor is foreseen, in addition to the unavoidable replacement of the matrix, to ensure quality of the installation and compliance with modern standards. Some fundamental options have been taken as a starting point: BR2 will continue to be used as a classical MTR, i.e. fuel and material irradiations and safety experiments with some additional service-activities. The present configuration is optimized for that use and there is no specific experimental requirement to change the basic concepts and performance characteristics. From the customers viewpoint, it is desirable to go ahead with the well-known features of BR2, to maintain a high degree of availability and reliability and to minimize the duration of the long shutdown. It is also important to limit the amount of nuclear liabilities. So the objective of the refurbishment programme is the life extension of BR2 for about 15 years, corresponding to

  1. History of the Belgian nuclear power controversy

    International Nuclear Information System (INIS)

    Laes, E.

    2009-01-01

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

  2. EBR-2 [Experimental Breeder Reactor-2], IFR [Integral Fast Reactor] prototype testing programs

    International Nuclear Information System (INIS)

    Lehto, W.K.; Sackett, J.I.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  3. Thermal reactor benchmark tests on JENDL-2

    International Nuclear Information System (INIS)

    Takano, Hideki; Tsuchihashi, Keichiro; Yamane, Tsuyoshi; Akino, Fujiyoshi; Ishiguro, Yukio; Ido, Masaru.

    1983-11-01

    A group constant library for the thermal reactor standard nuclear design code system SRAC was produced by using the evaluated nuclear data JENDL-2. Furthermore, the group constants for 235 U were calculated also from ENDF/B-V. Thermal reactor benchmark calculations were performed using the produced group constant library. The selected benchmark cores are two water-moderated lattices (TRX-1 and 2), two heavy water-moderated cores (DCA and ETA-1), two graphite-moderated cores (SHE-8 and 13) and eight critical experiments for critical safety. The effective multiplication factors and lattice cell parameters were calculated and compared with the experimental values. The results are summarized as follows. (1) Effective multiplication factors: The results by JENDL-2 are considerably improved in comparison with ones by ENDF/B-IV. The best agreement is obtained by using JENDL-2 and ENDF/B-V (only 235 U) data. (2) Lattice cell parameters: For the rho 28 (the ratio of epithermal to thermal 238 U captures) and C* (the ratio of 238 U captures to 235 U fissions), the values calculated by JENDL-2 are in good agreement with the experimental values. The rho 28 (the ratio of 238 U to 235 U fissions) are overestimated as found also for the fast reactor benchmarks. The rho 02 (the ratio of epithermal to thermal 232 Th captures) calculated by JENDL-2 or ENDF/B-IV are considerably underestimated. The functions of the SRAC system have been continued to be extended according to the needs of its users. A brief description will be given, in Appendix B, to the extended parts of the SRAC system together with the input specification. (author)

  4. EBR-2 [Experimental Breeder Reactor-2] test programs

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.; Hill, D.J.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  5. BR2 reactor: medical and industrial applications

    International Nuclear Information System (INIS)

    Ponsard, B.

    2005-01-01

    The radioisotopes are produced for various applications in the nuclear medicine (diagnostic, therapy, palliation of metastatic bone pain), industry (radiography of welds, ...), agriculture (radiotracers, ...) and basic research. Due to the availability of high neutron fluxes (thermal neutron flux up to 10 15 n/cm 2 .s), the BR2 reactor is considered as a major facility through its contribution for a continuous supply of products such 99 Mo ( 99 mTc), 131 I, 133 Xe, 192 Ir, 186 Re, 153 Sm, 90 Y, 32 P, 188 W ( 188 Re), 203 Hg, 82 Br, 41 Ar, 125 I, 177 Lu, 89 Sr, 60 Co, 169 Yb, 147 Nd, and others. Neutron Transmutation Doped (NTD) silicon is produced for the semiconductor industry in the SIDONIE (Silicon Doping by Neutron Irradiation Experiment) facility, which is designed to continuously rotate and traverse the silicon through the neutron flux. These combined movements produce exceptional dopant homogeneity in batches of silicon measuring 4 and 5-inches in diameter by up to 750 mm in length. The main objectives of work performed were to provide a reliable and qualitative supply of radioisotopes and NTD-silicon to the customers in accordance with a quality system that has been certified to the requirements of the EN ISO 9001: 2000. This new Quality System Certificate has been obtained in November 2003 for the Production of radioisotopes for medical and industrial applications and the Production of Neutron Transmutation Doped (NTD) Silicon in the BR2 reactor

  6. Advances in Reactor Physics, Mathematics and Computation. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, Volume 2, are divided into 7 sessions bearing on: - session 7: Deterministic transport methods 1 (7 conferences), - session 8: Interpretation and analysis of reactor instrumentation (6 conferences), - session 9: High speed computing applied to reactor operations (5 conferences), - session 10: Diffusion theory and kinetics (7 conferences), - session 11: Fast reactor design, validation and operating experience (8 conferences), - session 12: Deterministic transport methods 2 (7 conferences), - session 13: Application of expert systems to physical aspects of reactor design and operation.

  7. Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

    International Nuclear Information System (INIS)

    Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

    2014-01-01

    Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are

  8. Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

    Energy Technology Data Exchange (ETDEWEB)

    Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)

    2014-07-01

    Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the

  9. Radiation protection at the RA Reactor in 1998, RA reactor annual report, Part -2

    International Nuclear Information System (INIS)

    Ninkovic, M.; Pavlovic, R.; Mandic, M.; Pavlovic, S.; Grsic, Z.

    1998-01-01

    Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Collecting and treatment of fluid effluents; and (4) radioactive wastes, decontamination and actions. Each of the category is described as a separate annex of this report [sr

  10. Radiation protection at the RA Reactor in 1988, Part -2, RA reactor annual report

    International Nuclear Information System (INIS)

    Ninkovic, M.; Ajdacic, N.; Zaric, M.; Vukovic, Z.

    1988-01-01

    Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor and radiation protection; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Decontamination and relevant actions, collecting and treatment of fluid effluents; and and solid radioactive wastes [sr

  11. Problems of nuclear reactor safety. Vol. 2

    International Nuclear Information System (INIS)

    Goncharov, L.A.

    1995-01-01

    Theses of proceedings of the 9 Topical Meeting on problems of nuclear power plant safety are presented. Reports include results of neutron-physical experiments carried out for reactor safety justification. Concepts of advanced reactors with improved safety are considered. Results of researches on fuel cycles are given too

  12. Utilization of the SLOWPOKE-2 research reactor

    International Nuclear Information System (INIS)

    Lalor, G.C.

    2001-01-01

    SLOWPOKEs are typically low power research reactors that have a limited number of applications. However, a significant range of NAA can be performed with such reactors. This paper describes a SLOWPOKE-based NAA program that is performing a valuable series of studies in Jamaica, including geological mapping and pollution assessment. (author)

  13. Pressurized water reactor simulator. Workshop material. 2. ed

    International Nuclear Information System (INIS)

    2005-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

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

    OpenAIRE

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

    Krull, W.

    1981-01-01

    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.) [de

  16. The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

    International Nuclear Information System (INIS)

    Townes, B.M.; Hilborn, J.W.

    1985-06-01

    A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected

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

    CERN Document Server

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

    2018-05-03

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

  18. PSA Level 2 activities for RBMK reactors

    International Nuclear Information System (INIS)

    Gubler, R.

    1998-01-01

    Probabilistic safety analyses (PSAs) of the boiling water graphite moderated pressure tube reactors (RBMKs) have been developed only recently and they are limited to Level 1. Activities at the IAEA were first motivated because of the difficulties to characterize core damage for RBMK reactors. Core damage probability is used in documents of the IAEA as a convenient single valued measure, for example for probabilistic safety criteria. The limited number of PSAs that have been completed for the RBMK reactors have shown that several special features of these channel type reactors necessitate revisiting of the characterization of core damage for these reactors. Furthermore, it has become increasingly evident that detailed deterministic analysis of DBAs and beyond design basis accidents reveal considerable insights into RBMK response to various accident conditions. These analyses can also help in better characterizing the outstanding phenomenological uncertainties, improved EOPs and AM strategies, including potential risk-beneficial accident negative backfits. The deterministic efforts should be focused first on elucidating accident progression processes and phenomena, and second on finding, qualifying and implementing procedures to minimize the risk of severe accident states The IAEA PSA procedures were mainly developed in New of vessel type LWRs, and would therefore require extensions to make them directly applicable. to channel type reactors. (author) (author)

  19. Antimicrobial use in Belgian broiler production.

    Science.gov (United States)

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

    2012-08-01

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

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

    International Nuclear Information System (INIS)

    Montes Ponce de Leon, J.

    1974-01-01

    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)

  1. Equipment for thermal neutron flux measurements in reactor R2

    Energy Technology Data Exchange (ETDEWEB)

    Johansson, E; Nilsson, T; Claeson, S

    1960-04-15

    For most of the thermal neutron flux measurements in reactor R2 cobalt wires will be used. The loading and removal of these wires from the reactor core will be performed by means of a long aluminium tube and electromagnets. After irradiation the wires will be scanned in a semi-automatic device.

  2. Economic evaluation of reprocessing - Indicative Belgian position

    International Nuclear Information System (INIS)

    1979-05-01

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

  3. Performance communication of the Belgian railway

    NARCIS (Netherlands)

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

    2007-01-01

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

  4. Fast breeder reactor at Kalkar. Pt. 2

    International Nuclear Information System (INIS)

    Degen, G.

    1979-02-01

    After a brief description of the previous development of the case the legal decisions are documented and commented on. The concept of the then FDP-Minister of Economy of North Rhine Westphalia (Riemer, Pu-combustion plant) is presented and the prospects and risk for the fast breeder reactor after the 3. partial construction license are discussed. (orig./HP) [de

  5. The mud deposits and the high turbidity in the Belgian-Dutch coastal zone, Southern Bight of the North Sea

    OpenAIRE

    Fettweis, M.; Van den Eynde, D.

    2003-01-01

    The suspended sediment processes and the mudfields found in the Belgian/Dutch coastal area (Southern North Sea) are discussed by presenting an integrated data-modelling approach of the suspended sediment transport along the Belgian-Dutch coast, using a fine-grid coupled 2D hydrodynamic and sediment transport model and existing field and literature data. These mudfields and turbidity maxima are situated in a well-mixed, highly energetic hydrodynamic environment. In the past the occurrence of t...

  6. The 2nd reactor core of the NS Otto Hahn

    International Nuclear Information System (INIS)

    Manthey, H.J.; Kracht, H.

    1979-01-01

    Details of the design of the 2nd reactor core are given, followed by a brief report summarising the operating experience gained with this 2nd core, as well as by an evaluation of measured data and statements concerning the usefulness of the knowledge gained for the development of future reactor cores. Quite a number of these data have been used to improve the concept and thus the specifications for the fuel elements of the 3rd core of the reactor of the NS Otto Hahn. (orig./HP) [de

  7. Set of rules SOR 2 reactor site criteria

    International Nuclear Information System (INIS)

    1976-06-01

    The purpose of this set of rules is to describe criteria which guide the Director in his evaluation of the suitability of proposed sites for stationary power and testing reactors subject to SOR 2. (B.G.)

  8. Reactivity effects due to beryllium poisoning of BR2

    International Nuclear Information System (INIS)

    Kalcheva, S.; Ponsard, B.; Koonen, E.

    2004-01-01

    This paper illustrates the impact of the poisoning of the beryllium reflector on reactivity variations of the Belgian MTR BR2 in SCK.CEN. Detailed calculations by MCNP-4C of reactivity effects caused by strong neutron absorbers 3 He and 6 Li during reactor operation history are presented. The importance of beryllium poisoning for the accuracy of reactivity predictions is discussed. (authors)

  9. On-line fast flux measurements in the BR2 reactor

    International Nuclear Information System (INIS)

    Vermeeren, L.

    2009-01-01

    Since 2001, CEA-Cadarache and the Belgian Nuclear Research Centre SCK-CEN are collaborating on the development and in-pile qualification of subminiature fission chambers (diameter of 1.5 mm). Initially, efforts concentrated on fission chambers for the in-pile measurement of thermal fluxes (with 235 U as fissile material). Meanwhile successful long-term tests of the prototypes have been performed in various environments: in low temperature (40-100 degress Celsius) BR2 pool water (up to a thermal neutron fluence of 3 1 0 21 n/cm 2 ) and in the CALLISTO PWR loop (300 degrees Celsius, 155 bars). The long-term qualification of derived industrial detectors (Photonis CFUZ53) in CALLISTO is still ongoing. However, for various types of irradiations in research reactors, the knowledge of the evolution of the fast neutron flux is even of more interest than the thermal flux data. Therefore the collaboration program was extended to the development and the in-pile qualification of subminiature or miniature fission chambers (with 3 mm diameter) for fast neutron detection, for which 242 Pu was selected as the optimal fissile material. In order to achieve the on-line in-pile measurement of fast neutron flux, the fission chambers will be operated in the Campbelling mode (based on the mean square fluctuation of the detector current). In this mode the gamma induced contribution to the signal can be efficiently suppressed. Moreover, a data processing software will take into account the evolution of the fissile deposit in order to assess on-line the fast flux sensitivity and to correct for the low energy neutron contributions. The final objective is to qualify a Fast Neutron Detector System (FNDS) able to provide on-line data for local fast neutron fluxes in Material Testing Reactors. The on-line measurement of the fast neutron flux would contribute significantly to the characterization of the irradiation conditions during test experiments with materials and innovative fuel elements

  10. Operation of the SLOWPOKE-2 reactor in Jamaica

    Energy Technology Data Exchange (ETDEWEB)

    Grant, C.N.; Lalor, G.C.; Vuchkov, M.K. [University of the West Indies, Kingston (Jamaica)

    2001-07-01

    Over the past sixteen years lCENS has operated a SLOWPOKE 2 nuclear reactor almost exclusively for the purpose of neutron activation analysis. During this period we have adopted a strategy of minimum irradiation times while optimizing our output in an effort to increase the lifetime of the reactor core and to maintaining fuel integrity. An inter-comparison study with results obtained with a much larger reactor at IPEN has validated this approach. The parameters routinely monitored at ICENS are also discussed and the method used to predict the next shim adjustment. (author)

  11. Assessment of the French and US embrittlement trend curves applied to RPV materials irradiated in the BR2 materials test reactor

    International Nuclear Information System (INIS)

    Chaouadi, R.; Gerard, R.; Boagaerts, A.S.

    2011-01-01

    The irradiation embrittlement of reactor pressure vessels (RPVs) in monitored through the surveillance programs associated with predictive formulas, the so-called embrittlement trend curves. These formulas are generally empirically derived and contain the major embrittlement-inducing elements such as copper, nickel and phosphorus. There are a number of such trend curves used in various regulatory guides used in the US, France, Germany, Russia and Japan. These trend curves are often supported by surveillance data and regularly assessed in view of updated surveillance databases. With the recent worldwide move towards life extension of existing reactors above their initially-scheduled lifetime of 40 years, adequate and accurate modeling of irradiation embrittlement becomes a concern for long term operation. The aim of this work is to assess the performance of the embrittlement trend curves used in a regulatory perspective. The work presented here is limited to US and French trend curves because the reactor pressure vessels of the Belgian nuclear power plants are either Westinghouse or Framatome design. The chemical composition of the Belgian RPVs being very close to the one of the French 900 MW units, the French trend curve is used except for the Doel 1-2 units for which these curves are not applicable due to the higher copper content of the welds. In this case, the U.S. trend curves are used. The aim of this work is to evaluate the performance of the embrittlement trend curves used in a regulatory perspective to represent the experimental data obtained in the BR2 reactor. In particular, the French (FIM, FIS) and the US (Reg. Guide 1.99 Rev. 2, ASTM E900-02, EWO and EONY) formulas are of prime interest. The results obtained clearly show that the French trend curves tend to over-estimate the actual irradiation hardening while the US curves under-estimate it. Within the long term operation perspective, both over- and under-estimating are undesirable and therefore the

  12. Maximum credible accident analysis for TR-2 reactor conceptual design

    International Nuclear Information System (INIS)

    Manopulo, E.

    1981-01-01

    A new reactor, TR-2, of 5 MW, designed in cooperation with CEN/GRENOBLE is under construction in the open pool of TR-1 reactor of 1 MW set up by AMF atomics at the Cekmece Nuclear Research and Training Center. In this report the fission product inventory and doses released after the maximum credible accident have been studied. The diffusion of the gaseous fission products to the environment and the potential radiation risks to the population have been evaluated

  13. Belgian guidelines for economic evaluations: second edition.

    Science.gov (United States)

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

    2014-12-01

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

  14. Compact tokamak reactors part 2 (numerical results)

    International Nuclear Information System (INIS)

    Wiley, J.C.; Wootton, A.J.; Ross, D.W.

    1996-01-01

    The authors describe a numerical optimization scheme for fusion reactors. The particular application described is to find the smallest copper coil spherical tokamak, although the numerical scheme is sufficiently general to allow many other problems to be solved. The solution to the steady state energy balance is found by first selecting the fixed variables. The range of all remaining variables is then selected, except for the temperature. Within the specified ranges, the temperature which satisfies the power balance is then found. Tests are applied to determine that remaining constraints are satisfied, and the acceptable results then stored. Results are presented for a range of auxiliary current drive efficiencies and different scaling relationships; for the range of variables chosen the machine encompassing volume increases or remains approximately unchanged as the aspect ratio is reduced

  15. Dynamic simulation of the 2 MWt slowpoke heating reactor

    International Nuclear Information System (INIS)

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

    1982-04-01

    A 2 MWt SLOWPOKE reactor, intended for commercial space heating, is being developed at the Chalk River Nuclear Laboratories. A small-signal dynamic simulation of this reactor, without closed-loop control, was developed. Basic equations were used to describe the physical phenomena in each kf the eight reactor subsystems. These equations were then linearized about the normal operation conditions and rearranged in a dimensionless form for implementation. The overall simulation is non-linear. Slow transient responses (minutes to days) of the simulation to both reactivity and temperature perturbations were measured at full power. In all cases the system reached a new steady state in times varying from 12 h to 250 h. These results illustrate the benefits of the inherent negative reactivity feedback of this reactor concept. The addition of closed-loop control using core outlet temperature as the controlled variable to move a beryllium reflector is also examined

  16. IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-07-01

    The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1.

  17. IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

    International Nuclear Information System (INIS)

    1992-01-01

    The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1

  18. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    1966-01-01

    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

  20. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    International Nuclear Information System (INIS)

    Greene, S.R.; Spellman, D.J.; Bevard, B.B.

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative

  1. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Spellman, D.J.; Bevard, B.B. [and others

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

  2. Reactor

    International Nuclear Information System (INIS)

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

    1979-01-01

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

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

    NARCIS (Netherlands)

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

    2015-01-01

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

  4. Clonal expansion of the Belgian Phytophthora ramorum populations based on new microsatellite markers

    Science.gov (United States)

    A. Vercauteren; I. De Dobbelaere; N. J. Grünwald; P. Bonants; E. Van Bockstaele; M. Maes; K. Heungens

    2010-01-01

    Co-existence of both mating types A1 and A2 within the EU1 lineage of Phytophthora ramorum has only been observed in Belgium, which begs the question whether sexual reproduction is occurring. A collection of 411 Belgian P. ramorum isolates was established during a 7-year survey. Our main objectives were genetic characterization of this population to test for sexual...

  5. Digital, remote control system for a 2-MW research reactor

    International Nuclear Information System (INIS)

    Battle, R.E.; Corbett, G.K.

    1988-01-01

    A fault-tolerant programmable logic controller (PLC) and operator workstations have been programmed to replace the hard-wired relay control system in the 2-MW Bulk Shielding Reactor (BSR) at Oak Ridge National Laboratory. In addition to the PLC and remote and local operator workstations, auxiliary systems for remote operation include a video system, an intercom system, and a fiber optic communication system. The remote control station, located at the High Flux Isotope Reactor 2.5 km from the BSR, has the capability of rector startup and power control. The system was designed with reliability and fail-safe features as important considerations. 4 refs., 3 figs

  6. Loss of coolant analysis for the tower shielding reactor 2

    International Nuclear Information System (INIS)

    Radcliff, T.D.; Williams, P.T.

    1990-06-01

    The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs

  7. Shadow corrosion evaluation in the Studsvik R2 reactor

    International Nuclear Information System (INIS)

    Sanders, Ch.; Lysell, G.

    2000-01-01

    Post-irradiation examination has shown that increased corrosion occurs when zirconium alloys are in contact with or in proximity to other metallic objects. The observations indicate an influence of irradiation from the adjacent component as the enhanced corrosion occurs as a 'shadow' of the metallic object on the zirconium surface. This phenomenon could ultimately limit the lifetime of certain zirconium alloy components in the reactor. The Studsvik R2 materials test reactor has an In-Core Autoclave (INCA) test facility especially designed for water chemistry and materials research. The INCA facility has been evaluated and found suitable for shadow corrosion studies. The R2 reactor core containing the INCA facility was modeled with the Monte Carlo N-Particle (MCNP) code in order to evaluate the electron deposition in various materials and to develop a hypothesis of the shadow corrosion mechanism. (authors)

  8. Research reactor FR2 - 20 years chemical and radiochemical measurements

    International Nuclear Information System (INIS)

    Feuerstein, H.; Graebner, H.; Oschinski, J.; Hoffmann, W.; Beyer, J.

    1986-09-01

    The FR2 has been a D 2 O cooled and moderated research reactor with a thermal output of 44 MW. It was in operation from 1961 to 1981. Because of the operating conditions of the reactor, only a small number of routine measurements were performed. For these however special techniques had to be developed. During the 20 years of operation a number of special events occured or have been observed, sometimes with very amazing results, e.g. the 'aceton effect'. This report describes the chemical and radiochemical conditions of the reactor systems, as well as the results of the surveilance work. Not described are measurements for the many experiments. The last chapter gives in a short form a description of the most unusual events and observations. (orig.) [de

  9. New reactor concepts

    International Nuclear Information System (INIS)

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

    1998-11-01

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

  10. Measurement of the natural radiation of the Belgian territory

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  11. OTUS - Reactor inventory management system based on ORIGEN2

    Energy Technology Data Exchange (ETDEWEB)

    Poellaenen, R; Toivonen, H; Lahtinen, J; Ilander, T

    1995-10-01

    ORIGEN2 is a computer code that calculates nuclide composition and other characteristics of nuclear fuel. The use of ORIGEN2 requires good knowledge in reactor physics. However, once the input has been defined for a particular reactor type, the calculations can be easily repeated for any burnup and decay time. This procedure produces large output files that are difficult to handle manually. A new computer code, known as OTUS, was designed to facilitate the postprocessing of the data. OTUS makes use of the inventory files precalculated with ORIGEN2 in a way that enables their versatile treatment for different safety analysis purposes. A data base is created containing a comprehensive set of ORIGEN2 calculations as a function of fuel burnup and decay time. OTUS is a reactor inventory management system for a microcomputer with Windows interface. Four major data operations are available: (1) Build data modifies ORIGEN2 output data into a suitable format, (2) View data enables flexible presentation of the data as such, (3) Different calculations, such as nuclide ratios and hot particle characteristics, can be performed for severe accident analyses, consequence analyses and research purposes, (4) Summary files contain both burnup dependent and decay time dependent inventory information related to the nuclide and the reactor specified. These files can be used for safeguards, radiation monitoring and safety assessment. (orig.) (22 refs., 29 figs.).

  12. Risk-return of Belgian SRI funds

    OpenAIRE

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

    2007-01-01

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

  13. Reproduction of the PSBR reactor with Exterminator-2; Reproduccion del reactor PSBR con exterminador-2

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1983-08-15

    To reproduce the reactor PSBR reported in (1), with the available version of the Exterminator-II in the ININ, they took the dimensions, composition specifications, effective sections of the different compositions (excepting those of the central thimble and of the moderator), the K{sub eff} and the factors of power (FP) for the different burners. Based on the comparison of the K{sub eff} and of the FP obtained with those reported the precision it is determined before in the reproduction of the reactor mentioned. (Author)

  14. Application of MCNPX 2.7.D for reactor core management at the research reactor BR2

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, Edgar

    2011-01-01

    The paper discusses application of the Monte Carlo burn up code MCNPX 2.7.D for whole core criticality and depletion analysis of the Material Testing Research Reactor BR2 at SCK-CEN in Mol, Belgium. Two different approaches in the use of MCNPX 2.7.D are presented. The first methodology couples the evolution of fuel depletion, evaluated by MCNPX 2.7.D in an infinite lattice with a steady-state 3-D power distribution in the full core model. The second method represents fully automatic whole core depletion and criticality calculations in the detailed 3-D heterogeneous geometry model of the BR2 reactor. The accuracy of the method and computational time as function of the number of used unique burn up materials in the model are being studied. The depletion capabilities of MCNPX 2.7.D are compared vs. the developed at the BR2 reactor department MCNPX & ORIGEN-S combined method. Testing of MCNPX 2.7.D on the criticality measurements at the BR2 reactor is presented. (author)

  15. SIRIUS 2: A versatile medium power research reactor

    International Nuclear Information System (INIS)

    Rousselle, P.

    1992-01-01

    Most of the Research Reactors in the world have been critical in the Sixties and operated for twenty to thirty years. Some of them have been completely shut down, modified, or simply refurbished; the total number of RR in operation has decreased but there is still an important need for medium power research reactors in order: - to sustain a power program with fuel and material testing for NPP or fusion reactors; - to produce radioisotopes for industrial or medical purposes, doped silicon, NAA or neutron radiography; - to investigate further the condensed matter, with cold neutrons routed through neutron guides to improved equipment; - to develop new technologies and applications such as medical alphatherapy. Hence, taking advantage of nearly hundred reactor x years operation and backed up by the CEA experience, TECHNICATOME assisted by FRAMATOME has designed a new versatile multipurpose Research Reactor (20-30 Mw) SIRIUS 2 taking into account: - more stringent safety rules; - the lifetime; - the flexibility enabling a wide range of experiments and, - the future dismantling of the facility according to the ALARA criteria

  16. Optimized Control Rods of the BR2 Reactor

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, E.

    2007-01-01

    At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

  17. Optimized Control Rods of the BR2 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kalcheva, Silva; Koonen, E.

    2007-09-15

    At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

  18. analysis and implementation of reactor protection system circuits - case study Egypt's 2 nd research reactor-

    International Nuclear Information System (INIS)

    Elnokity, O.E.M.

    2006-01-01

    this work presents a way to design and implement the trip unit of a reactor protection system (RPS) using a field programmable gate arrays (FPGA). instead of the traditional embedded microprocessor based interface design method, a proposed tailor made FPGA based circuit is built to substitute the trip unit (TU), which is used in Egypt's 2 nd research reactor ETRR-2. the existing embedded system is built around the STD32 field computer bus which is used in industrial and process control applications. it is modular, rugged, reliable, and easy-to-use and is able to support a large mix of I/O cards and to easily change its configuration in the future. therefore, the same bus is still used in the proposed design. the state machine of this bus is designed based around its timing diagrams and implemented in VHDL to interface the designed TU circuit

  19. A conceptual design of LIB fusion reactor: UTLIF(2)

    International Nuclear Information System (INIS)

    Madarame, Haruki; Kondo, Shunsuke; Iwata, Shuichi; Oka, Yoshiaki; Miya, Kenzo.

    1984-01-01

    UTLIF(2) is a conceptual design study on a light ion beam driven fusion reactor based on a concept of rod-bundle blanket. Survivability and maintainability of the first wall and the blanket are regarded as of major importance in the design. The blanket rod is composed of a thick tube which has enough stiffness, a thin wrapping wall which receives high heat flux, and liquid lithium which breeds tritium and removes generated heat. The rod can be pulled out from the outside of the reactor vessel, hence the replacement is very easy. Nuclear and thermal analysis have been made and the performance of the reactor has been shown to be satisfactory. (author)

  20. Quality assurance in the project of RECH-2 research reactor

    International Nuclear Information System (INIS)

    Goycolea Donoso, C.; Nino de Zepeda Schele, A.

    1989-01-01

    The implantation of a Quality Assurance Program for the design, supply, construction, installation, and testing of the RECH-2 research reactor, is described in this paper. The obtained results, demonstrate that a Quality Assurance Program constitutes a suitable mean to assure that the installation complies with the safety and reliability requirements. (author)

  1. 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.

  2. 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.

  3. Cronos 2: a neutronic simulation software for reactor core calculations

    International Nuclear Information System (INIS)

    Lautard, J.J.; Magnaud, C.; Moreau, F.; Baudron, A.M.

    1999-01-01

    The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

  4. Planned Scientific programs around the Triga Mark 2 Reactor

    International Nuclear Information System (INIS)

    Majah, M Ibn.

    2007-01-01

    Full text: Nuclear techniques have been introduced to Morocco since the sixties. After the energy crisis of 1973, Morocco decides to create the National Center for Energy Sciences and Nuclear Techniques (CNESTEN) under the supervision of the Ministry of high Education and Research, with a research commercial and support vocation. CNESTEN is in charge of promoting nuclear application, to act as technical support for the authorities and to prepare the technological basis for nuclear power option. In 1998, CNESTEN started the construction of Nuclear Research Centre. The on going activities cover many sectors : earth and environmental sciences, high energy physics, safety and security, waste management. In 2001, CNESTEN started the construction of a 2MW TRiga Mark 2 Reactor, with the possibility to increase the power to 3 MW. The construction was achieved in January 2007. The operation of the reactor is expected for April 2007. The program of the utilization of the reactor was established with th contribution of the university and with the assistance of IAEA. Some of the experimental set-up installed around the reactor have been designed. CNESTEN has developed cooperation with Nuclear research centres from other countries and is receiving visitors and trainees mainly through the IAEA [fr

  5. Operation of the BR2 Reactor

    International Nuclear Information System (INIS)

    Gubel, P.

    2006-01-01

    The BR2 is still SCK-CEN's most important nuclear facility. After an extensive refurbishment of 22 months to compensate for the ageing of the installations, to enhance the reliability of operation and to comply with modern safety standards, it was restarted in April 1997. The facility is mainly used for the irradiation and testing of fuels and materials and for commercial productions - including radioisotopes for the medical and industrial uses, and NTD-Silicon. The article describes the main achievements and activities in 2005

  6. Operation of the BR2 Reactor

    International Nuclear Information System (INIS)

    Gubel, P.

    2005-01-01

    The BR2 is still SCK-CEN's most important nuclear facility. After an extensive refurbishment of 22 months to compensate for the ageing of the installations, to enhance the reliability of operation and to comply with modern safety standards, it was restarted in April 1997. The facility is mainly used for the irradiation and testing of fuels and materials and for commercial productions - including radioisotopes for the medical and industrial uses, and NTD-Silicon. The article describes the main activities and achievements in 2004

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

    Directory of Open Access Journals (Sweden)

    Decoo D

    2015-11-01

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

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

    International Nuclear Information System (INIS)

    2004-01-01

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

  9. Dalhousie SLOWPOKE-2 reactor: A nuclear analytical chemistry facility

    International Nuclear Information System (INIS)

    Chatt, A.; Holzbecher, J.

    1990-01-01

    SLOWPOKE is an acronym for Safe Low POwer Kritical Experiment. The SOWPOKE-2 is a compact, inherently safe, swimming-pool-type reactor designed by the Atomic Energy of Canada Limited for neutron activation analysis (NAA) and isotope production. The Dalhousie University SLOWPOKE-2 reactor (DUSR) has been operating since 1976; a large beryllium reflector was added in 1986 to extend its lifetime by another 8 to 10 yr. The DUSR is generally operated at half-power with a maximum thermal flux of 1.1 x 10 12 n/cm 2 ·s in the inner pneumatic sites and that of 5.4 x 10 11 n/cm 2 ·s in the outer sites. Despite this comparatively low flux, SLOWPOKE-2 reactors have many beneficial features that are continuously being exploited at the DUSR facility for developing nuclear analytical methods for fundamental as well as applied studies. Although NAA is a well-established analytical technique, much of the activation analysis being performed in most facilities has been limited to methods using fairly long-lived nuclides. The approach at the DUSR facility has been to utilize the highly homogeneous, stable, and reproducible neutron flux to develop NAA methods based on short-lived nuclides. SLOWPOKE reactors have a fairly high epithermal neutron flux, which is being advantageously used for determining several trace elements in complex matrices. Radiochemical NAA (RNAA) methods using coprecipitation, distillation, and ion-exchange separations have been used for the determination of very low levels of several elements in biological materials

  10. Characterization of the Three Mile Island Unit-2 reactor building atmosphere prior to the reactor building purge

    International Nuclear Information System (INIS)

    Hartwell, J.K.; Mandler, J.W.; Duce, S.W.; Motes, B.G.

    1981-05-01

    The Three Mile Island Unit-2 reactor building atmosphere was sampled prior to the reactor building purge. Samples of the containment atmosphere were obtained using specialized sampling equipment installed through penetration R-626 at the 358-foot (109-meter) level of the TMI-2 reactor building. The samples were subsequently analyzed for radionuclide concentration and for gaseous molecular components (O 2 , N 2 , etc.) by two independent laboratories at the Idaho National Engineering Laboratory (INEL). The sampling procedures, analysis methods, and results are summarized

  11. Focal epilepsy in the Belgian shepherd

    DEFF Research Database (Denmark)

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

    2009-01-01

    and deceased) were ascertained through a telephone interview using a standardised questionnaire regarding seizure history and phenomenology. Living dogs were invited to a detailed clinical evaluation. Litters more than five years of age, or where epilepsy was present in all offspring before the age of five......, were included in the calculations of inheritance. results: Out of 199 family members, 66 dogs suffered from epilepsy. The prevalence of epilepsy in the family was 33%. Fifty-five dogs experienced focal seizures with or without secondary generalisation, while four dogs experienced primary generalised...... seizures. In seven dogs, seizures could not be classified. The mode of inheritance of epilepsy was simple Mendelian. CLINICAL SIGNIFICANCE: This study identified that the Belgian shepherd suffers from genetically transmitted focal epilepsy. The seizure phenomenology expressed by family members have...

  12. Benchmark calculations for VENUS-2 MOX -fueled reactor dosimetry

    International Nuclear Information System (INIS)

    Kim, Jong Kung; Kim, Hong Chul; Shin, Chang Ho; Han, Chi Young; Na, Byung Chan

    2004-01-01

    As a part of a Nuclear Energy Agency (NEA) Project, it was pursued the benchmark for dosimetry calculation of the VENUS-2 MOX-fueled reactor. In this benchmark, the goal is to test the current state-of-the-art computational methods of calculating neutron flux to reactor components against the measured data of the VENUS-2 MOX-fuelled critical experiments. The measured data to be used for this benchmark are the equivalent fission fluxes which are the reaction rates divided by the U 235 fission spectrum averaged cross-section of the corresponding dosimeter. The present benchmark is, therefore, defined to calculate reaction rates and corresponding equivalent fission fluxes measured on the core-mid plane at specific positions outside the core of the VENUS-2 MOX-fuelled reactor. This is a follow-up exercise to the previously completed UO 2 -fuelled VENUS-1 two-dimensional and VENUS-3 three-dimensional exercises. The use of MOX fuel in LWRs presents different neutron characteristics and this is the main interest of the current benchmark compared to the previous ones

  13. Severe accident analysis for level 2 PSA of SMART reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jin Yong; Lee, Jeong Hun; Kim, Jong Uk; Yoo, Tae Geun; Chung, Soon Il; Kim, Min Gi [FNC Technology Co., Seoul (Korea, Republic of)

    2010-12-15

    The objectives of this study are to produce data for level 2 PSA and evaluation results of severe accident by analyzing severe accident sequence of transient events, producing fault tree of containment systems and evaluating direct containment heating of the SMART. In this project, severe accident analysis results were produced for general transient, loss of feedwater, station blackout, and steam line break events, and based on the results, design safety of SMART was verified. Also, direct containment heating phenomenon of the SMART was evaluated using TCE methodology. For level 2 PSA, fault tree of the containment isolation system, reactor cavity flooding system, plant chilled water system, and reactor containment building HVAC system was produced and analyzed

  14. TMI-2 reactor vessel and balance of plant status

    International Nuclear Information System (INIS)

    Kuehn, G.A.

    1990-01-01

    In the fall of 1988 a corporate decision was made which concentrated effort on support of reactor vessel defueling and minimized activity in balance-of-plant areas. The auxiliary and fuel handling building are in a safe/stable state but final preparations for monitored storage won't be pursued until defueling and fuel shipping are complete. In addition to dispositioning fuel, the project is actively preparing for disposal of the Accident Generated Water (2.3 million gallons) by evaporation

  15. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

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

  16. 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

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

    Science.gov (United States)

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

    2018-04-10

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

  18. Neutronic study using oxide and nitride fuels for the Super Phenix 2 reactor

    International Nuclear Information System (INIS)

    Batista, J.L.; Renke, C.A.C.

    1991-11-01

    This report presents a neutronic analysis and a description of the Super Phenix 2 reactor, taken as reference. We present the methodology and results for cell and global reactor calculations for oxide (U O 2 - Pu O 2 ) and nitride (U N - Pu N) fuels. To conclude we compare the performance of oxide and nitride fuels for the reference reactor. (author)

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    Brognon, T.

    1993-01-01

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

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

    Science.gov (United States)

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

    1997-05-01

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

  2. The Effect Of Beryllium Interaction With Fast Neutrons On the Reactivity Of ETRR-2 Research Reactor

    International Nuclear Information System (INIS)

    Aziz, M.; El Messiry, A.M.

    2000-01-01

    The effect of beryllium interactions with fast neutrons is studied for Etrr 2 research reactors. Isotope build up inside beryllium blocks is calculated under different irradiation times. a new model for the Etrr 2 research reactor is designed using MCNP code to calculate the reactivity and flux change of the reactor due to beryllium poison

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

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 2 of the Proceedings of 2. Yugoslav symposium on reactor physics includes eight papers dealing with the following topics: method for measuring high anti reactivities of a reactor system; integration method for thermal reaction rate calculation; Determination of initial core configuration for BHWR-200 MWe; safety shutdowns and failures of the RA reactor equipment; determining the reactivity of absorption rods; measurements of thermal and fast neutron fluxes at the TRIGA reactor and other measurements during operation of the TRIGA reactor; mathematical modelling of the reactor safety; review of problems and methods for radiation risk assessment in the environment of a nuclear power plant

  4. TPDWR2: thermal power determination for Westinghouse reactors, Version 2. User's guide

    International Nuclear Information System (INIS)

    Kaczynski, G.M.; Woodruff, R.W.

    1985-12-01

    TPDWR2 is a computer program which was developed to determine the amount of thermal power generated by any Westinghouse nuclear power plant. From system conditions, TPDWR2 calculates enthalpies of water and steam and the power transferred to or from various components in the reactor coolant system and to or from the chemical and volume control system. From these results and assuming that the reactor core is operating at constant power and is at thermal equilibrium, TPDWR2 calculates the thermal power generated by the reactor core. TPDWR2 runs on the IBM PC and XT computers when IBM Personal Computer DOS, Version 2.00 or 2.10, and IBM Personal Computer Basic, Version D2.00 or D2.10, are stored on the same diskette with TPDWR2

  5. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

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

  6. Once-through CANDU reactor models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    Croff, A.G.; Bjerke, M.A.

    1980-11-01

    Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % 235 U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given

  7. G 2 reactor project; Projet de pile a double fin: G 2

    Energy Technology Data Exchange (ETDEWEB)

    Ailleret, [Electricite de France (EDF), Dir. General des Etudes de Recherches, 75 - Paris (France); Taranger, P; Yvon, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    The CEA actually constructs the G-2 reactor core working with natural uranium, which will use graphite as moderator, and gas under pressure as cooling fluid. This report presents the specificity of the new reactor: - the different elements of the reactor core, - the control and the security of the reactor, - the renewal of the fuel, - the biologic surrounding wall, - and the cooling circuit. (M.B.) [French] le Commissariat a l'Energie Atomique construit actuellement la pile G-2 a Uranium naturel, qui utilisera le graphite comme moderateur, et le gaz sous pression comme fluide de refroidissement. Ce rapport presente les specificite du nouveau reacteur: - les differents elements de la pile, - le controle et la securite du reacteur, - le renouvellement du combustible, - l'enceinte biologique, - et le circuit de refroidissement. (M.B.)

  8. Mechanical degradation processes: The Belgian experience

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  9. Generic atorvastatin, the Belgian statin market and the cost-effectiveness of statin therapy.

    Science.gov (United States)

    Simoens, Steven; Sinnaeve, Peter R

    2013-02-01

    This study examines how the market entry of generic atorvastatin influences the Belgian statin market and the cost-effectiveness of statin therapy. Using IMS Health data, the Belgian 2000-2011 statin market was analyzed in terms of total expenditure, annual price of statin treatment, and patient numbers. A simulation analysis projected statin market shares from 2012 to 2015 following market entry of generic atorvastatin. This analysis was based on three scenarios regarding the number of patients taking specific statins. Savings associated with an atorvastatin price reduction of 50-70 % were calculated. A literature review of economic evaluations assessed the cost-effectiveness of generic atorvastatin. Statin expenditure increased from €113 million in 2000 to €285 million in 2011 due to higher expenditure on atorvastatin and rosuvastatin. Although the number of patients treated with simvastatin increased by nearly 800 %, the resulting increase in expenditure was partially offset by price reductions. Atorvastatin is projected to become the dominant product in the Belgian statin market (market share of 47-66 % by 2015). Annual savings would attain €108.6-€153.7 million for a 50 % reduction in the atorvastatin price and €152.0-€215.2 million for a 70 % price reduction. The literature suggests that generic atorvastatin is cost-effective as compared to simvastatin. The limited evidence about the cost-effectiveness of rosuvastatin as compared with generic atorvastatin is inconclusive. Generic atorvastatin is cost-effective as compared to simvastatin, is projected to become the dominant product in the Belgian statin market and is expected to generate substantial savings to health care payers.

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

    Science.gov (United States)

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

    2007-09-01

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

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

    International Nuclear Information System (INIS)

    1985-04-01

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

  12. 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)

    Markovic, H.

    1962-07-01

    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 [sr

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

    Science.gov (United States)

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

    2014-10-01

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

  14. Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1992-04-01

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  15. Comparison of the N Reactor and Ignalina Unit No. 2 Level 1 Probabilistic Safety Assessments

    International Nuclear Information System (INIS)

    Coles, G.A.; McKay, S.L.

    1995-06-01

    A multilateral team recently completed a full-scope Level 1 Probabilistic Safety Assessment (PSA) on the Ignalina Unit No. 2 reactor plant in Lithuania. This allows comparison of results to those of the PSA for the U.S. Department of Energy's (DOE) N Reactor. The N Reactor, although unique as a Western design, has similarities to Eastern European and Soviet graphite block reactors

  16. Microstructure in Zircaloy Creep Tested in the R2 Reactor

    International Nuclear Information System (INIS)

    Pettersson, Kjell

    2004-12-01

    Tubular specimens of Zircaloy-4 have been creep tested in bending in the R2 reactor in Studsvik. The creep deformation in the reactor core is accelerated in comparison with creep deformation outside the reactor core. The possible mechanisms behind this behaviour are described briefly. In order to determine which the actual mechanism is, the microstructure of the material creep tested in the R2 reactor has been examined by transmission electron microscopy. Due to the bending, material subjected to both tensile and compressive stress during creep was available. Since some of the proposed mechanisms might give microstructures which are different when the material is subjected to compressive or tensile stress it was assumed that examination of both types of material would give valuable information with regard to the operating mechanism. The result of the examination was that in the as-irradiated condition there were no obvious differences detected between materials which had been deformed in tension or compression. After a heat treatment to coarsen the irradiation induced microstructure there were still no significant differences between the two types of material. However it was now observed that in addition to dislocation loops the microstructure also contained network dislocations which presumably had been invisible in the electron microscope before heat treatment due to the high density of small dislocation loops in this state. It is therefore concluded that the most probable mechanism for irradiation creep in this case is climb and glide of the network dislocations. The role of irradiation is two-fold: It accelerates climb due to the production of point defects of which more interstitials than vacancies arrive to the network dislocations stopped at an obstacles. This leads to a net climb after which a dislocation is released from the obstacle and an amount of glide takes place. The second effect is the production of loops which serve as an increasing density of

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

    Science.gov (United States)

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

    2016-06-01

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

  18. The radiological impact of the Belgian phosphate industry

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  19. 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)

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

    International Nuclear Information System (INIS)

    Ikeda, K.; Ishikawa, T.; Miyoshi, T.; Takagi, T.

    2012-01-01

    This paper presents an overview of the reactor core internals replacement project carried out at the Ikata Nuclear Power Station in Japan, which was the first of its kind among PWRs in the world. Failure of baffle former bolts was first reported in 1989 at Bugey 2 in France. Since then, similar incidents have been reported in Belgium and in the U.S., but not in Japan. However, the possibility of these bolts failing in Japanese plants cannot be denied in the future as operating hours increase. Ageing degradation mechanisms for the reactor core internals include irradiation-assisted stress corrosion cracking of baffle former bolts and mechanical wear of control rod guide cards. Two different approaches can be taken to address these ageing issues: to inspect and repair whenever a problem is found; and to replace the entire core internals with those of a new design having advanced features to prevent ageing degradation problems. The choice of our company was the latter. This paper explains the reasons for the choice and summarizes the replacement project activities at Ikata Units 1 and 2 as well as the improvements incorporated in the new design. (author)

  1. Further study on parameterization of reactor NAA: Pt. 2

    International Nuclear Information System (INIS)

    Tian Weizhi; Zhang Shuxin

    1989-01-01

    In the last paper, Ik 0 method was proposed for fission interference corrections. Another important kind of interferences in reator NAA is due to threshold reaction induced by reactor fast neutrons. In view of the increasing importance of this kind of interferences, and difficulties encountered in using the relative comparison method, a parameterized method has been introduced. Typical channels in heavy water reflector and No.2 horizontal channel of Heavy Water Research Reactor in the Insitute of Atomic Energy have been shown to have fast neutron energy distributions (E>4 MeV) close to primary fission neutron spectrum, by using multi-threshold detectors. On this basis, Ti foil is used as an 'instant fast neutron flux monitor' in parameterized corrections for threshold reaction interferences in the long irradiations. Constant values of φ f /φ s = 0.70 ± 0.02% have been obtained for No.2 rabbit channel. This value can be directly used for threshold reaction inference correction in the short irradiations

  2. Active species in a large volume N2-O2 post-discharge reactor

    International Nuclear Information System (INIS)

    Kutasi, K; Pintassilgo, C D; Loureiro, J; Coelho, P J

    2007-01-01

    A large volume post-discharge reactor placed downstream from a flowing N 2 -O 2 microwave discharge is modelled using a three-dimensional hydrodynamic model. The density distributions of the most populated active species present in the reactor-O( 3 P), O 2 (a 1 Δ g ), O 2 (b 1 Σ g + ), NO(X 2 Π), NO(A 2 Σ + ), NO(B 2 Π), NO 2 (X), O 3 , O 2 (X 3 Σ g - ) and N( 4 S)-are calculated and the main source and loss processes for each species are identified for two discharge conditions: (i) p = 2 Torr, f = 2450 MHz, and (ii) p = 8 Torr, f = 915 MHz; in the case of a N 2 -2%O 2 mixture composition and gas flow rate of 2 x 10 3 sccm. The modification of the species relative densities by changing the oxygen percentage in the initial gas mixture composition, in the 0.2%-5% range, are presented. The possible tuning of the species concentrations in the reactor by changing the size of the connecting afterglow tube between the active discharge and the large post-discharge reactor is investigated as well

  3. Coolant radiolysis studies in the high temperature, fuelled U-2 loop in the NRU reactor

    International Nuclear Information System (INIS)

    Elliot, A.J.; Stuart, C.R.

    2008-06-01

    An understanding of the radiolysis-induced chemistry in the coolant water of nuclear reactors is an important key to the understanding of materials integrity issues in reactor coolant systems. Significant materials and chemistry issues have emerged in Pressurized Water Reactors (PWR), Boiling Water Reactors (BWR) and CANDU reactors that have required a detailed understanding of the radiation chemistry of the coolant. For each reactor type, specific computer radiolysis models have been developed to gain insight into radiolysis processes and to make chemistry control adjustments to address the particular issue. In this respect, modelling the radiolysis chemistry has been successful enough to allow progress to be made. This report contains a description of the water radiolysis tests performed in the U-2 loop, NRU reactor in 1995, which measured the CHC under different physical conditions of the loop such as temperature, reactor power and steam quality. (author)

  4. Sterilization of E. coli bacterium in a flowing N2-O2 post-discharge reactor

    International Nuclear Information System (INIS)

    Villeger, S; Cousty, S; Ricard, A; Sixou, M

    2003-01-01

    Effective destruction of Escherichia coli (E. coli) bacteria has been obtained in a flowing N 2 -O 2 microwave post-discharge reactor. The sterilizing agents are the O atoms and the UV emissions of NOβ which are produced by N and O atoms recombination in the reactor. In the following plasma conditions: pressure 5 Torr, flow rate 1 L n min -1 , microwave power of 100 W in a quartz tube of 5 mm, an O atom density of 2.5x10 15 cm -3 is measured by NO titration in the post-discharge reactor with UV emission in a N 2 -(5%)O 2 gas mixture. Full destruction of 10 13 cfu ml -1 E. coli is observed after a treatment time of 25 min. (rapid communication)

  5. The probability safety assessment impact on the BR2 refurbishment

    International Nuclear Information System (INIS)

    Pouleur, Yvan

    1995-01-01

    The probabilistic safety assessment (PSA) study has proven its worth by establishing a sensitive safety screening of the reactor. It has focused engineering forces to technically improve safety systems and to measure the influence of functional modifications. In the future, the project will be developed in a living way, to reinforce the present structure along with continuous safety monitoring of the reactor and to develop engineers and operators safety skills. This paper presents the PSA impact on the BR2 (Belgian Reactor Two) refurbishment. (author)

  6. Preparation of covariance data for the fast reactor. 2

    International Nuclear Information System (INIS)

    Shibata, Keiichi; Hasagawa, Akira

    1998-03-01

    For some isotopes important for core analysis of the fast reactor, covariance data of neutron nuclear data in the evaluated nuclear data library (JENDL-3.2) were presumed to file. Objected isotopes were 10-B, 11-B, 55-Mn, 240-Pu and 241-Pu. Physical amounts presumed on covariance were cross section, isolated and unisolated resonance parameters and first order Legendre coefficient of elastic scattering angle distribution. Presumption of the covariance was conducted in accordance with the data estimation method of JENDL-3.2 as possible. In other ward, when the estimated value was based on the experimental one, error of the experimental value was calculated, and when based on the calculated value, error of the calculated one was obtained. Their estimated results were prepared with ENDF-6 format. (G.K.)

  7. Inventory of nuclear liabilities - The Belgian perspective

    International Nuclear Information System (INIS)

    Minon, Jean-Paul

    2003-01-01

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

  8. Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide

    International Nuclear Information System (INIS)

    Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

    1983-02-01

    The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems

  9. Security and Privacy Improvements for the Belgian eID Technology

    Science.gov (United States)

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

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

    BNEN, the Belgian Nuclear Higher Education Network has been created in 2001 by five Belgian universities and the Belgian Nuclear Research Centre (SCK CEN) as a joint effort to maintain and further develop a high quality programme in nuclear engineering in Belgium. More information: http://www.sckcen.be/BNEN. (author)

  11. 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.

  12. VISA-2, Reactor Vessel Failure Probability Under Thermal Shock

    International Nuclear Information System (INIS)

    Simonen, F.; Johnson, K.

    1992-01-01

    1 - Description of program or function: VISA2 (Vessel Integrity Simulation Analysis) was developed to estimate the failure probability of nuclear reactor pressure vessels under pressurized thermal shock conditions. The deterministic portion of the code performs heat transfer, stress, and fracture mechanics calculations for a vessel subjected to a user-specified temperature and pressure transient. The probabilistic analysis performs a Monte Carlo simulation to estimate the probability of vessel failure. Parameters such as initial crack size and position, copper and nickel content, fluence, and the fracture toughness values for crack initiation and arrest are treated as random variables. Linear elastic fracture mechanics methods are used to model crack initiation and growth. This includes cladding effects in the heat transfer, stress, and fracture mechanics calculations. The simulation procedure treats an entire vessel and recognizes that more than one flaw can exist in a given vessel. The flaw model allows random positioning of the flaw within the vessel wall thickness, and the user can specify either flaw length or length-to-depth aspect ratio for crack initiation and arrest predictions. The flaw size distribution can be adjust on the basis of different inservice inspection techniques and inspection conditions. The toughness simulation model includes a menu of alternative equations for predicting the shift in the reference temperature of the nil-ductility transition. 2 - Method of solution: The solution method uses closed form equations for temperatures, stresses, and stress intensity factors. A polynomial fitting procedure approximates the specified pressure and temperature transient. Failure probabilities are calculated by a Monte Carlo simulation. 3 - Restrictions on the complexity of the problem: Maxima of 30 welds. VISA2 models only the belt-line (cylindrical) region of a reactor vessel. The stresses are a function of the radial (through-wall) coordinate only

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

    Science.gov (United States)

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

    2012-01-01

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

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

    Science.gov (United States)

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

    2013-06-01

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

  15. RHTF 2, a 1200 MWe high temperature reactor

    International Nuclear Information System (INIS)

    Brisbois, Jacques

    1978-01-01

    After having adapted to French conditions the 1160 MWe G.A.C. reactor, Commissariat a l'Energie Atomique and French Industry have decided to design an High Temperature Reactor 1200 MWe based on the G.A.C. technology and taking into account the point of view of Electricite de France and the experience of C.E.A. and industry on the gas cooled reactor technology. The main objective of this work is to produce a reactor design having a low technical risk, good operability, with an emphasis on the safety aspects easing the licensing problems

  16. GENP-2, Program System for Integral Reactor Perturbation

    International Nuclear Information System (INIS)

    Boioli, A.; Cecchini, G.P.

    1975-01-01

    1 - Description of problem or function: GENP-2 is a system of programs that use 'generalized perturbation theory' to calculate the perturbations of reactor integral characteristics which can be expressed by means of ratios between linear or bilinear functionals of the real and/or adjoint fluxes (e.g. reaction rate ratios), due to cross section perturbations. 2 - Method of solution: GENP-2 consists of the following codes: DDV, SORCI, CIAP-PMN and GLOBP-2D. DDV calculates the real or adjoint fluxes and power distribution using multigroup diffusion theory in 2-dimensions. SORCI uses the fluxes from DDV to calculate the real and/or adjoint general perturbation sources. CIAP-PMN reads the sources from SORCI and uses them in the real or adjoint generalised importance calculations (2 dimensions, multi- group diffusion). GLOBP-2D uses the importance calculated by CIAP-PMN, and the fluxes calculated by DDV, in generalised perturbation expressions to calculate the perturbation in the quantity of interest. 3 - Restrictions on the complexity of the problem: DDV although variably dimensioned has the following restrictions: - max. number of mesh points 6400; - max. number of mesh points in 1-dimension 81; - max. number of regions 6400; - max. number of energy groups 100; - if power distribution calculated, product of number of groups and number of regions 2500. The other programs have the same restrictions if applicable

  17. BR2 reactor core steady state transient modeling

    International Nuclear Information System (INIS)

    Makarenko, A.; Petrova, T.

    2000-01-01

    A coupled neutronics/hydraulics/heat-conduction model of the BR2 reactor core is under development at SCK-CEN. The neutron transport phenomenon has been implemented as steady state and time dependent nodal diffusion. The non-linear heat conduction equation in-side fuel elements is solved with a time dependent finite element method. To allow coupling between functional modules and to simulate subcooled regimes, a simple single-phase hydraulics has been introduced, while the two-phase hydraulics is under development. Multiple tests, general benchmark cases as well as calculation/experiment comparisons demonstrated a good accuracy of both neutronic and thermal hydraulic models, numerical reliability and full code portability. A refinement methodology has been developed and tested for better neutronic representation in hexagonal geometry. Much effort is still needed to complete the development of an extended cross section library with kinetic data and two-phase flow representation. (author)

  18. Consideration of LH2 and LD2 cold neutron sources in heavy water reactor reflector

    International Nuclear Information System (INIS)

    Potapov, I.A.; Serebrov, A.P.

    2001-01-01

    The reactor power, the required CNS dimensions and power of the cryogenic equipment define the CNS type with maximized cold neutron production. Cold neutron fluxes from liquid hydrogen (LH 2 ) and liquid deuterium (LD 2 ) cold neutron sources (CNS) are analyzed. Different CNS volumes, presents and absence of reentrant holes inside the CNS, different adjustment of beam tube and containment are considered. (orig.)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    Fucks, I.; Hardeman, F.

    2002-01-01

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

  1. Reactor limitation system improves the safety and availability of the Angra 2 nuclear power plant

    International Nuclear Information System (INIS)

    Souza Mendes, J.E. de

    1987-01-01

    Beyond the classic Reactor Protection System and Reactor Control System, nuclear plant Angra 2 has a third system called Reactor Limitation System which combines the intelligence features of the control systems with the high reliability of the protection systems. In determined events, which are not controlled by the control system (e.g.: load rejection, failure of one main reactor coolant pump), the Reactor Limitation System actuates automatically in order to lead the plant to a safe operating condition and so it avoids the actuation of the Reactor Protection System and consequently the reactor trip. This increases safety and availability of the plant and reduces component stresses. After the safe operating condition is reached, the process guidance automatically returns to the control systems. (Author) [pt

  2. Set of rules SOR 2 licensing of nuclear reactors

    International Nuclear Information System (INIS)

    1976-05-01

    This is the set of rules promulgated by the Israel Atomic Energy Commission pursuant to the Supervision of Supplies and Services Law 5718-1957, Order regarding Supervision of Nuclear Reactors (1974) Chapter 3: Permits, to provide for the Licensing of Nuclear Reactors. (B.G.)

  3. Operating reactors licensing actions summary. Vol. 4, No. 2

    International Nuclear Information System (INIS)

    1984-04-01

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

  4. An experimental investigation of fission product release in SLOWPOKE-2 reactors - Data report

    International Nuclear Information System (INIS)

    Harnden, A.M.C.

    1995-09-01

    The results of an investigation into the release of fission products from SLOWPOKE-2 reactors fuelled with a highly-enriched uranium alloy core are detailed in Volume 1. This data report (Volume 2) contains plots of the activity concentrations of the fission products observed in the reactor container at the University of Toronto, Ecole Polytechnique and the Kanata Isotope Production Facility. Release rates from the reactor container water to the gas headspace are also included. (author)

  5. Development of the versatile reactor analysis code system, MARBLE2

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Jin, Tomoyuki; Hazama, Taira; Hirai, Yasushi

    2015-07-01

    The second version of the versatile reactor analysis code system, MARBLE2, has been developed. A lot of new functions have been added in MARBLE2 by using the base technology developed in the first version (MARBLE1). Introducing the remaining functions of the conventional code system (JOINT-FR and SAGEP-FR), MARBLE2 enables one to execute almost all analysis functions of the conventional code system with the unified user interfaces of its subsystem, SCHEME. In particular, the sensitivity analysis functionality is available in MARBLE2. On the other hand, new built-in solvers have been developed, and existing ones have been upgraded. Furthermore, some other analysis codes and libraries developed in JAEA have been consolidated and prepared in SCHEME. In addition, several analysis codes developed in the other institutes have been additionally introduced as plug-in solvers. Consequently, gamma-ray transport calculation and heating evaluation become available. As for another subsystem, ORPHEUS, various functionality updates and speed-up techniques have been applied based on user experience of MARBLE1 to enhance its usability. (author)

  6. Strippable coating used for the TMI-2 reactor building decontamination

    International Nuclear Information System (INIS)

    Adams, J.W.; Dougherty, D.R.; Barletta, R.E.

    1984-01-01

    Strippable coating material used in the TMI-2 reactor building decontamination has been tested for Sr, Cs, and Co leachability, for radiation stability, thermal stability, and for resistance to biodegradation. It was also immersion tested in water, a water solution saturated with toluene and xylene, toluene, xylene, and liquid scintillation counting (LSC) cocktail. Leach testing resulted in all of the Cs and Co activity and most of the Sr activity being released from the coating in just a few days. Immersion resulted in swelling of the coating in all of the liquids tested. Gamma irradiation and heating of the coating did not produce any apparent physical changes in the coating to 1 x 10 8 rad and 100 0 C; however, gas generation of H 2 , CO, CO 2 was observed in both cases. Biodegradation of the coating occurred readily in soils as indicated by monitoring CO 2 produced from microbial respiration. These test results indicate that strippable coating radwaste would have to be stabilized to meet the requirements for Class B waste outlined in 10 CFR Part 61 and the NRC Draft Technical Position on Waste Form

  7. Turkey's regulatory plans for high enriched to low enriched conversion of TR-2 reactor core

    International Nuclear Information System (INIS)

    Guelol Oezdere, Oya

    2003-01-01

    Turkey is a developing country and has three nuclear facilities two of which are research reactors and one pilot fuel production plant. One of the two research reactors is TR-2 which is located in Cekmece site in Istanbul. TR-2 Reactor's core is composed of both high enriched and low enriched fuel and from high enriched to low enriched core conversion project will take place in year 2005. This paper presents the plans for drafting regulations on the safety analysis report updates for high enriched to low enriched core conversion of TR-2 reactor, the present regulatory structure of Turkey and licensing activities of nuclear facilities. (author)

  8. Artificial intelligence in nuclear reactor operation

    International Nuclear Information System (INIS)

    Da Ruan; Benitez-Read, J.S.

    2005-01-01

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

  9. The Oak Ridge Research Reactor: safety analysis: Volume 2, supplement 2

    International Nuclear Information System (INIS)

    Hurt, S.S.

    1986-11-01

    The Oak Ridge Research Reactor Safety Analysis was last updated via ORNL-4169, Vol. 2, Supplement 1, in May of 1978. Since that date, several changes have been effected through the change-memo system described below. While these changes have involved the cooling system, the electrical system, and the reactor instrumentation and controls, they have not, for the most part, presented new or unreviewed safety questions. However, some of the changes have been based on questions or recommendations stemming from safety reviews or from reactor events at other sites. This paper discusses those changes which were judged to be safety related and which include revisions to the syphon-break system and changes related to seismic considerations which were very recently completed. The maximum hypothetical accident postulated in the original safety analysis requires dynamic containment and filtered flow for compliance with 10CFR100 limits at the site boundary

  10. Microflow photochemistry: UVC-induced [2 + 2]-photoadditions to furanone in a microcapillary reactor

    Directory of Open Access Journals (Sweden)

    Sylvestre Bachollet

    2013-10-01

    Full Text Available [2 + 2]-Cycloadditions of cyclopentene and 2,3-dimethylbut-2-ene to furanone were investigated under continuous-flow conditions. Irradiations were conducted in a FEP-microcapillary module which was placed in a Rayonet chamber photoreactor equipped with low wattage UVC-lamps. Conversion rates and isolated yields were compared to analogue batch reactions in a quartz test tube. In all cases examined, the microcapillary reactor furnished faster conversions and improved product qualities.

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

    International Nuclear Information System (INIS)

    Fraix, J.

    1992-01-01

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

  12. Open Access to the Belgian Nuclear higher Education Network

    International Nuclear Information System (INIS)

    Simons, S.

    2005-01-01

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

  13. Decomposing dynamic profit inefficiency of Belgian dairy farms

    NARCIS (Netherlands)

    Ang, Frederic; Lansink, Alfons Oude

    2018-01-01

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

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

    NARCIS (Netherlands)

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

    2007-01-01

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

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

    Science.gov (United States)

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

    2010-01-01

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

  16. Neutronic modeling of pebble bed reactors in APOLLO2

    International Nuclear Information System (INIS)

    Grimod, M.

    2010-01-01

    In this thesis we develop a new iterative homogenization technique for pebble bed reactors, based on a 'macro-stochastic' transport approximation in the collision probability method. A model has been developed to deal with the stochastic distribution of pebbles with different burnup in the core, considering spectral differences in homogenization and depletion calculations. This is generally not done in the codes presently used for pebble bed analyses, where a pebble with average isotopic composition is considered to perform the cell calculation. Also an iterative core calculation scheme has been set up, where the low-order RZ S N full-core calculation computes the entering currents in the spectrum zones subdividing the core. These currents, together with the core k eff , are then used as surface source in the fine-group heterogeneous calculation of the multi-pebble geometries. The developed method has been verified using reference Monte Carlo simulations of a simplified PBMR- 400 model. The pebbles in this model are individually positioned and have different randomly assigned burnup values. The APOLLO2 developed method matches the reference core k eff within ± 100 pcm, with relative differences on the production shape factors within ± 4%, and maximum discrepancy of 3% at the hotspot. Moreover, the first criticality experiment of the HTR-10 reactor was used to perform a first validation of the developed model. The computed critical number of pebbles to be loaded in the core is very close to the experimental value of 16890, only 77 pebbles less. A method to calculate the equilibrium reactor state was also developed and applied to analyze the simplified PBMR-400 model loaded with different fuel types (UO 2 , Pu, Pu + MA). The potential of the APOLLO2 method to compute different fluxes for the different pebble types of a multi-pebble geometry was used to evaluate the bias committed by the average composition pebble approximation. Thanks to a 'compensation of error

  17. Safe dismantling of the SVAFO research reactors R2 and R2-0 in Sweden

    International Nuclear Information System (INIS)

    ARNOLD, Hans-Uwe; BROY, Yvonne; Dirk Schneider

    2017-01-01

    The R2 and R2-0 reactors were part of the Swedish government's research program on nuclear power from the early 1960's. Both reactors were shut down in 2005 following a decision by former operator Studsvik Nuclear AB. The decommissioning of the R2 and R2-0 reactors is divided into three phases. The first phase - awarded to AREVA - involved dismantling of the reactors and associated systems in the reactor pool, treatment of the disassembled components as well as draining, cleaning and emptying the pool. In the second phase, the pool structure itself will be dismantled, while removal of remaining reactor systems, treatment and disposal of materials and clean-up will be carried out in the third stage. The entire work is planned to be completed before the end of this decade. The paper describes the several steps of phase 1 - starting with the team building, followed by the dismantling operations and covers challenges encountered and lessons learned as well. The reactors consist of 5.400 kg aluminum, 6.000 kg stainless steel restraint structures as well as, connection elements of the mostly flanged components (1.000 kg). The most demanding - from a radiological point of view - was the R2-0 reactor that was limited to ∼ 1 m"3 construction volumes but with an extremely heterogeneous activation profile. Based on the calculated radiological entrance data and later sampling, nuclide vectors for both reactors depending on the real placement of the single component and on the material (aluminum and stainless steel) were created. Finally, for the highest activated component from R2 reactor, 85 Sv/h were measured. The dismantling principles - adopted on a safety point of view - were the following: The always protected base area of the ponds served as a flexible buffer area for waste components and packaging. Specific protections were also installed on the walls to protect them from mechanical stress which may occur during dismantling work. A specific work platform was

  18. Development and testing of the EDF-2 reactor fuel element

    International Nuclear Information System (INIS)

    Delpeyroux, P.

    1964-01-01

    This technical report reviews the work which has been necessary for defining the EDF-2 fuel element. After giving briefly the EDF-2 reactor characteristics and the preliminary choice of parameters which made it possible to draw up a draft plan for the fuel element, the authors consider the research proper: - Uranium studies: tests on the passage into the β phase of an internal crown of a tube, bending of the tube under the effect of a localized force, welding of the end-pellets and testing for leaks. The resistance of the tube to crushing and of the pellets to yielding under the external pressure have been studied in detail in another CEA report. - Can studies: conditions of production and leak proof testing of the can, resistance of the fins to creep due to the effect of the gas flow. - Studies of the extremities of the element: creep under compression and welding of the plugs to the can. - Cartridge studies: determination of the characteristics of the can fuel fixing grooves and of the canning conditions, verification of the resistance of the fuel element to thermal cycling, determination of the temperature drop at the can-fuel interface dealt with in more detail in another CEA report. - Studies of the whole assembly: this work which concerns the graphite jacket, the support and the cartridge vibrations has been carried out by the Mechanical and Thermal Study Service (Mechanics Section). In this field the Fuel Element Study Section has investigated the behaviour of the centering devices in a gas current. The outcome of this research is the defining of the plan of the element the production process and the production specifications. The validity of ail these out-of-pile tests will be confirmed by the in-pile tests already under way and by irradiation of the elements in the EDF-2 reactor itself. In conclusion the programme is given for improving the fuel element and for defining the fuel element for the second charge. (authors) [fr

  19. Nuclear powerplant standardization: light water reactors. Volume 2. Appendixes

    International Nuclear Information System (INIS)

    1981-06-01

    This volume contains working papers written for OTA to assist in preparation of the report, NUCLEAR POWERPLANT STANDARDIZATION: LIGHT WATER REACTORS. Included in the appendixes are the following: the current state of standardization, an application of the principles of the Naval Reactors Program to commercial reactors; the NRC and standardization, impacts of nuclear powerplant standardization on public health and safety, descriptions of current control room designs and Duke Power's letter, Admiral Rickover's testimony, a history of standardization in the NRC, and details on the impact of standardization on public health and safety

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

    Science.gov (United States)

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

    2012-08-01

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

  1. 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.

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  3. Use of a SLOWPOKE-2 reactor for nuclear forensics applications

    Energy Technology Data Exchange (ETDEWEB)

    Andrews, M.T.; Beames-Canivet, T.L.; Elliott, R.S.; Kelly, D.G.; Corcoran, E.C., E-mail: Emily.Corcoran@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)

    2014-07-01

    A low enriched uranium SLOWPOKE-2 reactor is used as a neutron interrogation source in support of the identification and characterization of Special Nuclear Materials (SNM) at the Royal Military College of Canada (RMCC). Small amounts of fissile uranium and plutonium are sent into a SLOWPOKE-2 irradiation site before their transport to RMCC’s delayed neutron and gamma counting (DNGC) system. The counting arrangement of the DNGC consists of an array of six {sup 3}He and a high purity germanium detector. These detectors record the delayed neutron and photon emissions as a function of count time, to verify MCNP6 simulations of delayed particle emissions, and to detect and quantify trace amounts of fissile content. This paper discusses MCNP analyses done in preparation for an upcoming nuclear forensics exercise in the fall of 2014. MCNP6 simulations of the DNGC system focussed on the identification of characteristic gamma lines from prominent fission products. The relative intensities of these gamma lines are dependent on the SNM content in the sample. Gamma line pairs useful for SNM identification in RMCC's DNGC system are presented. (author)

  4. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-07-01

    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.

  5. A nodal Grean's function method of reactor core fuel management code, NGCFM2D

    International Nuclear Information System (INIS)

    Li Dongsheng; Yao Dong.

    1987-01-01

    This paper presents the mathematical model and program structure of the nodal Green's function method of reactor core fuel management code, NGCFM2D. Computing results of some reactor cores by NGCFM2D are analysed and compared with other codes

  6. Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

    International Nuclear Information System (INIS)

    Liu Guisheng

    1995-11-01

    How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testing of CENDL-2 and ENDF/B-6. (author). 8 refs, 2 figs, 4 tabs

  7. Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

    International Nuclear Information System (INIS)

    Liu Guisheng

    1995-01-01

    How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testings of CENDL-2 and ENDF/B-6. (4 tabs., 2 figs.)

  8. The Chernobyl reactor accident. Pt. 1 and 2

    International Nuclear Information System (INIS)

    1986-06-01

    The report first summarizes the available information on the various incidents of the whole accident scenario, and combines the information to present a first general outline and a basis for appraisal. The most significant incidents reported, namely power excursion, core meltdown, and fire, are discussed with a view to the reactor design and safety of reactors installed in the FRG. The main differences and advantages of German reactor designs are shown, as e.g.: Power excursions are mastered by inherent physical conditions; far better redundancy of engineered safety systems; enclosure of the complete reactor cooling system in a pressure-retaining steel containment; reactor buildings being made of reinforced concrete. The second part of the report deals with the radiological effects to be expected for our country. Data are given on the varying radiological exposure of the different regions. The fate and uptake of radioactivity in the human body are discussed. The conclusion drawn from the data presented is that the individual exposure due to the reactor accident will remain within the variations and limits of natural radioactivity and effects. (orig./HP) [de

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

    International Nuclear Information System (INIS)

    1966-01-01

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

  10. The reactor accident in the Three Mile Island-2 reactor plant. (Harrisburg, USA)

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The presentation of the accident development is based on the bulletin of the Atomic Industrial Forum (AIF) from the 6th of April, 1979. In addition, there are some short pieces of information, No. 14 and 15 of the association for reactor security as well as written and verbal information of the firms Brown, Boveri and Co/Babcock-Brown Boveri Reaktor GmbH (BBC/BBR). (orig.) [de

  11. Ageing/obsolescence management at the ZED-2 Research Reactor

    International Nuclear Information System (INIS)

    Mon, D.; Trudeau, C.

    2014-01-01

    The Zero Energy Deuterium (ZED-2) Research Reactor first achieved criticality in 1960 September. Ageing of Systems, Structures and Components (SSCs) as well as the obsolescence of many original components had led to the Facility being in a reactive mode with respect to maintenance rather than a preventative maintenance mode. Through the implementation of an Ageing Management Plan to access the effects of Ageing Related Degradation Mechanisms (ARDMs) on SSCs and a Critical Spare Parts Obsolescence Plan (establishing new specifications for components, procurement of new components and building a replacement inventory) the Facility is returning to a preventative maintenance state rather than a reactive maintenance state. Upgrades and refurbishment of the original equipment and components is underway within the Facility. This paper describes the programs being carried out in the Facility and focuses on the upgraded nuclear electronics, renewed moderator level control system and spare parts inventory system. Implementation of the above mentioned programs will enable the Facility to continue to fulfill its mandate as an integral part of AECL's Vision and Strategic Outcome: 'To be a global partner in nuclear innovation' and that 'Canadians and the world receive energy, health, environmental and economic benefits from nuclear science and technology with confidence that nuclear safety and security are assured', respectively. (author)

  12. Analytical investigations at the IBR-2 reactor in Dubna

    International Nuclear Information System (INIS)

    Frontas'eva, M.V.; Pavlov, S.S.

    2000-01-01

    Experience in applying activation analysis with epithermal neutrons to environmental studies at the IBR-2 pulsed fast reactor in Dubna is reviewed. The principal units of the pneumatic system REGATA are described as well as the main present-day parameters of the irradiation channels. Examples from two challenging areas: analysis of airborne particulate matter and analysis of the biomonitors of atmospheric deposition, where the feasibility of ENAA related to non-nuclear multi-element techniques has been shown, are given. ENAA is presently being used in several projects carried out at FLNP JINR that involve the analysis of aerosol filters. For these studies, instrumental neutron activation analysis cannot feasibly be replaced by any non-nuclear analytical technique at the present state of art due to the fact that the total mass of the aerosol collected on filter samples is often rather small. This favours direct instrumental techniques rather than those depending on dissolving the sample prior to analysis. The dominant part of air pollution studies at FLNP JINR is based on the use of the moss biomonitoring technique resulting in a final product-maps of the atmospheric deposition of heavy metals, rare-earths, actinides (U and Th), etc., constructed by the GIS (geographical information system) technology

  13. Ageing/obsolescence management at the ZED-2 Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mon, D.; Trudeau, C. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2014-07-01

    The Zero Energy Deuterium (ZED-2) Research Reactor first achieved criticality in 1960 September. Ageing of Systems, Structures and Components (SSCs) as well as the obsolescence of many original components had led to the Facility being in a reactive mode with respect to maintenance rather than a preventative maintenance mode. Through the implementation of an Ageing Management Plan to access the effects of Ageing Related Degradation Mechanisms (ARDMs) on SSCs and a Critical Spare Parts Obsolescence Plan (establishing new specifications for components, procurement of new components and building a replacement inventory) the Facility is returning to a preventative maintenance state rather than a reactive maintenance state. Upgrades and refurbishment of the original equipment and components is underway within the Facility. This paper describes the programs being carried out in the Facility and focuses on the upgraded nuclear electronics, renewed moderator level control system and spare parts inventory system. Implementation of the above mentioned programs will enable the Facility to continue to fulfill its mandate as an integral part of AECL's Vision and Strategic Outcome: 'To be a global partner in nuclear innovation' and that 'Canadians and the world receive energy, health, environmental and economic benefits from nuclear science and technology with confidence that nuclear safety and security are assured', respectively. (author)

  14. Economics and utilization of thorium in nuclear reactors. Technical annexes 1 and 2

    International Nuclear Information System (INIS)

    1978-05-01

    An assessment of the impact of utilizing the 233 U/thorium fuel cycle in the U.S. nuclear economy is strongly dependent upon several decisions involving nuclear energy policy. These decisions include: (1) to recycle or not recycle fissile material; (2) if fissile material is recycled, to recycle plutonium, 233 U, or both; and (3) to deploy or not to deploy advanced reactor designs such as Fast Breeder Reactors (FBR's), High Temperature Gas Reactors (HTGR's), and Canadian Deuterium Uranium Reactors (CANDU's). This report examines the role of thorium in the context of the above policy decisions while focusing special attention on economics and resource utilization

  15. TMI-2 reactor-vessel head removal and damaged-core-removal planning

    International Nuclear Information System (INIS)

    Logan, J.A.; Hultman, C.W.; Lewis, T.J.

    1982-01-01

    A major milestone in the cleanup and recovery effort at TMI-2 will be the removal of the reactor vessel closure head, planum, and damaged core fuel material. The data collected during these operations will provide the nuclear power industry with valuable information on the effects of high-temperature-dissociated coolant on fuel cladding, fuel materials, fuel support structural materials, neutron absorber material, and other materials used in reactor structural support components and drive mechanisms. In addition, examination of these materials will also be used to determine accident time-temperature histories in various regions of the core. Procedures for removing the reactor vessel head and reactor core are presented

  16. WWER-1000 reactor simulator. Material for training courses and workshops. 2. ed

    International Nuclear Information System (INIS)

    2005-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No.12, Reactor Simulator Development (2001). Course material for workshops using a pressurized water reactor (PWR) simulator developed for the IAEA by Cassiopeia Technologies Inc. of Canada is presented in the IAEA publication, Training Course Series No. 22, 2nd edition, Pressurized Water Reactor Simulator (2005) and Training Course Series No.23, 2nd edition, Boiling Water Reactor Simulator (2005). This report consists of course material for workshops using the WWER-1000 Reactor Department Simulator from the Moscow Engineering and Physics Institute, Russian Federation

  17. EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques

    Energy Technology Data Exchange (ETDEWEB)

    Rousseau, A; Genthon, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

  18. 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.

  19. Fissile fuel production and usage of thermal reactor waste fueled with UO2 by means of hybrid reactor system

    International Nuclear Information System (INIS)

    Ipek, O.

    1997-01-01

    The use of Fast Breeder Reactors to produce fissile fuel from nuclear waste and the operation of these reactors with a new neutron source are becoming today' topic. In the thermonuclear reactors, it is possible to use 2.45-14.1 MeV - neutrons which can be obtained by D-T, D-D Semicatalyzed (D-D) and other fusion reactions. To be able to do these, Hybrid Reactor System, which still has experimental and theoretical studies, have to be taken into consideration.In this study, neutronic analysis of hybrid blanket with grafit reflector, is performed. D-T driven fusion reaction is surrounded by UO 2 fuel layer and the production of ''2''3''9Pu fissile fuel from waste ''2''3''8U is analyzed. It is also compared to the other possible fusion reactions. The results show that 815.8 kg/year ''2''3''8Pu with D-T reaction and 1431.6 kg/year ''2''3''8Pu with semicatalyzed (D-D) reaction can be produced for 1000 MW fusion power. This means production of 2.8/ year and 4.94/ year LWR respectively. In addition, 1000 MW fusion flower is is multiplicated to 3415 MW and 4274 MW for D-T and semicatalyzed (D-D) reactions respectively. The system works subcritical and these values are 0.4115 and 0.312 in order. The calculations, ANISN-ORNL code, S 16 -P 3 approach and DLC36 data library are used

  20. Application of stable adaptive schemes to nuclear reactor systems, (2)

    International Nuclear Information System (INIS)

    Kukuda, Toshio

    1979-01-01

    The parameter identification and adaptive control schemes applied in a previous study to a nonlinear point reactor are extended to the case of a loosely-coupled-core reactor with internal feedbacks, constituting a nonlinear overall system. Both schemes are shown to be stable, with the system newly represented on the pattern of the Model Reference Adaptive System (MRAS) with use made of the Lyapunov's method. For either parameter identification or adaptive control of a loosely-coupled-core reactor, there exists no canonical form of multiple input-multiple output system which can be directly applied for deriving the MRAS with the matrix version of the Kalman-Yakubovich lemma as it was in the case of the point reactor. This difficulty is circumvented by the practical assumption that the neutron density can be directly measured on each core as reactivity change is applied as input into the coupled core as a whole. For parameter identification, the model parameters are adaptively adjusted to those of each core, while for the adaptive control, plant parameters of each core can be adaptively compensated, again through control inputs, to asymptotically reduce the output error between the model and the plant. The point reactor is shown to correspond to a special case. (author)

  1. A theoretical analysis of methanol synthesis from CO2 and H2 in a ceramic membrane reactor

    NARCIS (Netherlands)

    Gallucci, F.; Basile, A.

    2007-01-01

    In this theoretical work the CO2 conversion into methanol in both a traditional reactor (TR) and a membrane reactor (MR) is considered. The purpose of this study was to investigate the possibility of increasing CO2 conversion into methanol with respect to a TR. A zeolite MR, able to combine

  2. How Belgium helped establish a surveillance programme for Argentina's Atucha-2

    Energy Technology Data Exchange (ETDEWEB)

    Mitev, Lubomir [NucNet, Brussels (Belgium)

    2015-02-15

    Collaboration between Belgian experts and Argentina on the commissioning of Argentine reactors helped overcome problems caused by delays with construction. Marc Scibetta, deputy manager for nuclear materials science, from the Belgian Nuclear Research Centre (SCK CEN) gave an interview in which he told some facts to Lubomir Mitev of NucNet. The cooperation between SCK CEN an Argentina's Comision Nacional de Energia Atomica (CNEA) started in 2002. The first project was a support for the safety evaluation of the Atucha-1 reactor pressure vessel. When Argentina resumed the construction of Atucha-2 in 2006 - originally, construction started in 1981 but was suspended in 1985 due to financial reasons -, SCK CEN was asked to develop and implement a surveillance programme for the unit.

  3. System Design of a Supercritical CO_2 cooled Micro Modular Reactor

    International Nuclear Information System (INIS)

    Kim, Seong Gu; Cho, Seongkuk; Yu, Hwanyeal; Kim, Yonghee; Jeong, Yong Hoon; Lee, Jeong Ik

    2014-01-01

    Small modular reactor (SMR) systems that have advantages of little initial capital cost and small restriction on construction site are being developed by many research organizations around the world. Existing SMR concepts have the same objective: to achieve compact size and a long life core. Most of small modular reactors have much smaller size than the large nuclear power plant. However, existing SMR concepts are not fully modularized. This paper suggests a complete modular reactor with an innovative concept for reactor cooling by using a supercritical carbon dioxide. The authors propose the supercritical CO_2 Brayton cycle (S-CO_2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the reactor core and PCU in one vessel. A conceptual design of the proposed small modular reactor was developed, which is named as KAIST Micro Modular Reactor (MMR). The supercritical CO_2 Brayton cycle for the S-CO_2 cooled reactor core was optimized and the size of turbomachinery and heat exchanger were estimated preliminary. The nuclear fuel composed with UN was proposed and the core lifetime was obtained from a burnup versus reactivity calculation. Furthermore, a system layout with fully passive safety systems for both normal operation and emergency operation was proposed. (author)

  4. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 2-2

    International Nuclear Information System (INIS)

    Okuda, Eiji; Ito, Hiromichi; Yoshihara, Shizuya

    2014-01-01

    An accident occurred in experimental fast reactor 'Joyo' in 2007 which is obstruction of fuel change equipment caused by contacting rotating plug and MARICO-2. In addition, we confirmed two happenings in the reactor vessel that (1) Deformation of MARICO-2 subassembly on the in vessel storage rack together with a transfer pot, (2) Deformation of the Upper core structure of 'Joyo' caused by contacting MARICO-2 subassembly and the UCS. We do the restoration work for restoring it. This time, we describe current status of Replacement work of the UCS. (author)

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

    International Nuclear Information System (INIS)

    Leclere, Robert; Van Landeghem, Yves

    2010-01-01

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

  6. Radionuclide distribution in TMI-2 reactor building basement liquids and solids

    International Nuclear Information System (INIS)

    Horan, J.T.; McIsaac, C.V.; Keefer, D.G.

    1984-01-01

    As a result of the TMI-2 accident, approximately 2.46 x 10 6 L of contaminated water were released to the Reactor Building basement. The principal fission product release pathway from the damaged core was through the reactor coolant system (RCS) to the pressurizer, through the pressure-operated relief valve (PORV) on the pressurizer to the Reactor Coolant Drain Tank (RCDT), and then through the RCDT rupture disk to the Reactor Building basement. Since August 1979, a number of efforts have been made to determine the location, quantity, and composition of fission products released to the Reactor Building basement. These efforts have included sampling of the basement water and solids, the basement sump pump recirculation line, the RCDT, and visual surveys using a closed circuit television (CCTV) system. The analysis of basement samples has provided data on the physical and radioisotopic characteristics of the liquids and solids. This paper describes the sample collection techniques and discusses radiochemical analyses results

  7. Venting krypton-85 from the Three Mile Island Unit 2 reactor building

    International Nuclear Information System (INIS)

    Burton, H.M.

    1981-01-01

    To permit the less restricted access to the reactor building necessary to maintain instrumentation and equipment, and to proceed towad the total decontamination of the facility, General Public Utilities, operators of the facility referred to hereafter as GPU, asked the United States Nuclear Regulatory Commission, or NRC, for permission to remove the 85 Kr from the reactor building by venting it to the environment. GPU supported their request with the Safety Analysis and Environmental Assessment Report on the proposed reactor building venting plan. On June 12, 1980, after seven months of licensing deliberations and numerous public hearings, the NRC granted GPU's request. The actual venting took place between June 28 and July 11, 1980. This report presents an overview of the detailed effort involved in the TMI-2 reactor building venting program. The findings reported here are condensed from a published report entitled TMI-2 Reactor Building Purge--Kr-85 Venting

  8. The Mandate System for the Belgian Public Prosecution

    Directory of Open Access Journals (Sweden)

    Bruno BROUCKER

    2009-12-01

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

  9. Solid methane cold moderator for the IBR-2 reactor

    International Nuclear Information System (INIS)

    Beliakov, A. A.; Tretiakov, I. T.; Shabalin, E. P.; Golikov, V. V.; Luschivkov, V I.

    1997-09-01

    The paper describes the research and design work carried out since 1986 at the Frank Laboratory of Neutron Physics of the Joint Institute for Nuclear Research in Dubna to create a cryogenic moderator for the IBR-22 reactor using solid methane as a moderating substance.

  10. Methodology for reactor core physics analysis - part 2

    International Nuclear Information System (INIS)

    Ponzoni Filho, P.; Fernandes, V.B.; Lima Bezerra, J. de; Santos, T.I.C.

    1992-12-01

    The computer codes used for reactor core physics analysis are described. The modifications introduced in the public codes and the technical basis for the codes developed by the FURNAS utility are justified. An evaluation of the impact of these modifications on the parameter involved in qualifying the methodology is included. (F.E.). 5 ref, 7 figs, 5 tabs

  11. 105-B Reactor museum feasibility assessment (Phase 2) project

    International Nuclear Information System (INIS)

    Heckel, R. P.

    2000-01-01

    This 105-B Reactor Museum feasibility assessment project report documents project activities that have been performed, including a review and assessment of previously existing information, a walk-through of the facility, an assessment of potential hazards, and selection of mitigative measures deemed to be appropriate to allow unescorted access by members of the public to a specified primary tour route

  12. Contribution from Belgium - Belgian contribution to the PWG1 Generic Study on Undetected Failures of Safety systems

    International Nuclear Information System (INIS)

    Vincke, Marc

    1997-01-01

    In the frame of its participation to the PWG 1 generic study on 'Undetected Failures of Safety Systems', AVN performed a search of such cases among the Belgian plants, using the proposed criterion: to find significant events where equipment remained inoperable, or would have been unable to fulfil correctly its safety function for an extended period of time until their condition was discovered. An extended period of time means one cycle duration or several test interval periods at least; if unknown, it has to be estimated w.r.t. plant lifetime. Note that non safety systems preventing safety systems to perform their function are to be included. As a first information source, a screening of AVN's DIANE (Domestic Information about Nuclear Events) database, for undetected failures of safety systems was performed. This database is used to store and retrieve information on a selection of events which have occurred in the Belgian NPPs since 1985. The sources of information are the incident reports which AVN receives from the utilities, completed with the reports of our inspectors on site. The coding system used within this database is based on the IRS Coding Manual. This coding system does not always allow for an easy retrieval of events related to a specific subject. In addition the DIANE-coding system does not allow for direct retrieval of undetected failures. In a first step, the following systems were scanned: reactor coolant system, reactor heat removal system, emergency core cooling system, chemical and volume control, containment spray, main and auxiliary feedwater, component cooling water, control rod drives. For each system, records were selected by examining their title. Careful reading of the 64 reports selected this way finally led to two cases compatible with the criteria. The decennial revision studies formed a second set of information sources. An inquiry to AVN's engineers responsible for the decennial revision projects allowed to

  13. Iron microbial communities in Belgian Frasnian carbonate mounds

    OpenAIRE

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

    2001-01-01

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

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

    OpenAIRE

    Ang, Frederic; Oude Lansink, Alfons

    2014-01-01

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

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

    OpenAIRE

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

    2011-01-01

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

  16. Estimation of power feedback parameters of pulse reactor IBR-2M on transients

    International Nuclear Information System (INIS)

    Pepyolyshev, Yu.N.; Popov, A.K.

    2013-01-01

    Parameters of the IBR-2M reactor power feedback (PFB) on a model of the reactor dynamics by mathematical treatment of two registered transients are estimated. Frequency characteristics and the pulse transient characteristics corresponding to these PFB parameters are calculated. PFB parameters received thus can be considered as their express tentative estimation as real measurements in this case occupy no more than 30 minutes. Total PFB is negative at 1 and 2 MW. At the received estimations of PFB parameters in a self-regulation mode it is possible to consider the stability margins of the IBR-2M reactor satisfactory

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

    Directory of Open Access Journals (Sweden)

    Cor Wagenaar

    2014-08-01

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

  18. MULTI-LOOP CONTROL DESIGN IN MULTIVARIABLE (2X2 CONTINUOUS STIRRED TANK REACTOR

    Directory of Open Access Journals (Sweden)

    Abdul Wahid

    2015-06-01

    Full Text Available With this study, the design and tuning of multi-loop for multivariable (2x2 CSTR will be made in order to achieve optimum CSTR control performance. This study used Bequette model reactor and MATLAB software and is expected to be able to cope with disturbances in the reactor so that the reactor system is able to stabilize quickly despite the distractions. In this study, the design will be made using multi-loop approach, along with PI controller as the next step. Then, BLT and auto-tune tuning method will be used in PI controller and given disturbances to both of tuning method. The controller performances are then compared. Results of the study are then analyzed for discussions and conclusions. Results from this study have shown that in terms of disturbance rejection, BLT is better than auto-tune based on comparison between both of controller performances. For IAE for the case of temperature, BLT is 30% better than auto-tune, but it is almost the same for the case of concentration. For settling time for the case of concentration, BLT is 30% better than auto-tune, and for the case of temperature, BLT is 18% better than auto-tune. For rise time for the case of concentration and temperature, BLT is 30% better than auto-tune.

  19. Development of prototype reactor maintenance. (2) Application to piping support of sodium-cooled reactor prototype

    International Nuclear Information System (INIS)

    Arai, Masanobu; Kunogi, Kosuke; Aizawa, Kosuke; Chikazawa, Yoshitaka; Takaya, Shigeru; Kubo, Shigenobu; Kotake, Shoji; Ito, Takaya; Yamaguchi, Akira

    2017-01-01

    A maintenance program on piping support of prototype fast breeder reactor Monju are studied. Based on degradation mechanism, snubbers in Monju primary cooling system showed lifetime more than the plant lifetime of 30 years by experiments conservatively. For the first step during construction, visual inspection on accessible all supports could be available. In that visual inspection, mounting conditions and damages of all accessible supports could be monitored. One of major features of the Monju primary piping system is large thermal expansion due to large temperature difference between maintenance and operation conditions. Thanks to that large thermal expansion, integrity of piping supports could be monitored by measuring piping displacement. When technologies of piping displacement monitoring are matured in Monju, visual inspection on piping support could be shifted to piping displacement monitoring. At that stage, the visual inspection could be limited only on representative supports. (author)

  20. A KINETIC MODEL FOR H2O2/UV PROCESS IN A COMPLETELY MIXED BATCH REACTOR. (R825370C076)

    Science.gov (United States)

    A dynamic kinetic model for the advanced oxidation process (AOP) using hydrogen peroxide and ultraviolet irradiation (H2O2/UV) in a completely mixed batch reactor (CMBR) is developed. The model includes the known elementary chemical and photochemical reac...

  1. Sophistication of burnup analysis system for fast reactor (2)

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hirai, Yasushi; Tatsumi, Masahiro

    2010-10-01

    Improvement on prediction accuracy for neutronics characteristics of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified cross-section set as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores improvement of not only static characteristics but also burnup characteristics is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup characteristics using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous study, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for power reactor analysis systems. In the present study, by extending the prototype system, features for handling of control rods and energy collapse of group constants have been designed and implemented. Computational results from the present analysis system are stored into restart files which can be accessible by

  2. Diffusion calculation's for the SLOWPOKE-2 reactor using DONJON

    International Nuclear Information System (INIS)

    Noceir, S.; El Hajjaji, O.; Varin, E.

    1997-01-01

    The SLOWPOKE reactor at Ecole Polytechnique will be refueled with a Low Enriched Uranium (LEU) fuel in place of a High Enriched Uranium (HEU) fuel used until now. The purpose of this study is to provide various models, using the reactor physics chain of codes DRAGON/DONJON, in order to predict the behavior of the new LEU Slowpoke. In particle, we will present some numerical results concerning the separate temperature effects of the main components of the core, the effect of a partial void appearing near the fuel pins and the axial and radial flux distributions. Finally the difference between the present HEU and the future LEU fuel power will be given. (author)

  3. Fast breeder reactors: Experience and trends. V. 2

    International Nuclear Information System (INIS)

    1986-01-01

    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 was attended by almost 400 participants (340 participants, 35 observers and 20 journalists) from 25 countries and five international organizations. More than 80 papers were presented and discussed during six regular sessions and four poster sessions. A separate abstract was prepared for each of these papers

  4. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

  5. Production of Sn-117m in the BR2 high-flux reactor.

    Science.gov (United States)

    Ponsard, B; Srivastava, S C; Mausner, L F; Russ Knapp, F F; Garland, M A; Mirzadeh, S

    2009-01-01

    The BR2 reactor is a 100MW(th) high-flux 'materials testing reactor', which produces a wide range of radioisotopes for various applications in nuclear medicine and industry. Tin-117m ((117m)Sn), a promising radionuclide for therapeutic applications, and its production have been validated in the BR2 reactor. In contrast to therapeutic beta emitters, (117m)Sn decays via isomeric transition with the emission of monoenergetic conversion electrons which are effective for metastatic bone pain palliation and radiosynovectomy with lesser damage to the bone marrow and the healthy tissues. Furthermore, the emitted gamma photons are ideal for imaging and dosimetry.

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

    International Nuclear Information System (INIS)

    Takano, Hideki

    1995-01-01

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

  7. Analysis of key hardware factors and countermeasure for restricting 49-2 swimming pool reactor lifetime

    International Nuclear Information System (INIS)

    Zhang Yadong; Guo Yue; Yang Xiao; Wang Yiwei; Wang Zhanwen

    2013-01-01

    Safe operation is the most important factor to determine the lifetime of aged 49-2 swimming pool reactor. In this paper, the hardware factors of lifetime were analyzed, such as the pool concrete aging, corrosion of aluminum container and primary coolant system, and graphite swelling etc., and then the corresponding measures such as surveillance, prevention and maintenance were purposed. The results show that 49-2 swimming pool reactor can continue to operate safely due to that container is safe under 8 degree earthquake, the reactor is safe on flood level of once per millennium, adding dam break, and the ageing condition of primary coolant system and container is acceptable. (authors)

  8. Advanced technologies for water cooled reactors 1990. Pt. 2

    International Nuclear Information System (INIS)

    1991-05-01

    The main purpose of the meeting was to review and discuss the status of national programmes, the progress achieved since the last meeting held in June 1988 in the field of advanced technologies and design trends for existing and future water cooled reactors. 24 specialists from 14 countries and the IAEA took part in the meeting and 12 papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  9. European supercritical water cooled reactor (HPLWR Phase 2 project)

    International Nuclear Information System (INIS)

    Schulenberg, Thomas; Starflinger, Joerg; Marsault, Philippe; Bittermann, Dietmar; Maraczy, Czaba; Laurien, Eckart; Lycklama, Jan Aiso; Anglart, Henryk; Andreani, Michele; Ruzickova, Mariana; Heikinheimo, Liisa

    2010-01-01

    The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 deg C maximum core outlet temperature. It is designed and analyzed by a European consortium of 13 partners from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small, housed fuel assemblies with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The innovative core design with upward and downward flow through its assemblies has been studied with neutronic, thermal-hydraulic and stress analyses and has been reviewed carefully in a mid-term assessment. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. An overview of results achieved up to now, given in this paper, is illustrating the latest scientific and technological advances. (author)

  10. Irradiation effects on Zr-2.5Nb in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

  11. A simulation Model of the Reactor Hall Ventilation and air Conditioning Systems of ETRR-2

    International Nuclear Information System (INIS)

    Abd El-Rahman, M.F.

    2004-01-01

    Although the conceptual design for any system differs from one designer to another. each of them aims to achieve the function of the system required. the ventilation and air conditioning system of reactors hall is one of those systems that really differs but always dose its function for which it is designed. thus, ventilation and air conditioning in some reactor hall constitute only one system whereas in some other ones, they are separate systems. the Egypt Research Reactor-2 (ETRR-2)represents the second type. most studies conducted on ventilation and air conditioning simulation models either in traditional building or for research rectors show that those models were not designed similarly to the model of the hall of ETRR-2 in which ventilation and air conditioning constitute two separate systems.besides, those studies experimented on ventilation and air conditioning simulation models of reactor building predict the temperature and humidity inside these buildings at certain outside condition and it is difficult to predict when the outside conditions are changed . also those studies do not discuss the influences of reactor power changes. therefore, the present work deals with a computational study backed by infield experimental measurements of the performance of the ventilation and air conditioning systems of reactor hall during normal operation at different outside conditions as well as at different levels of reactor power

  12. Novel, Regenerable Microlith Catalytic Reactor for CO2 Reduction via Bosch Process, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Precision Combustion, Inc. (PCI) proposes to develop an extremely compact, lightweight and regenerable MicrolithREG catalytic CO2 reduction reactor, capable of...

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  14. Benchmark testing of CENDL-2 for U-fuel thermal reactors

    International Nuclear Information System (INIS)

    Zhang Baocheng; Liu Guisheng; Liu Ping

    1995-01-01

    Based on CENDL-2, NJOY-WIMS code system was used to generate 69-group constants, and do benchmark testing for TRX-1,2; BAPL-UO-2-1,2,3; ZEEP-1,2,3. All the results proved that CENDL-2 is reliable for thermal reactor calculations. (3 tabs.)

  15. PCU arrangement of a supercritical CO{sub 2} cooled micro modular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seong Gu; Baik, Seungjoon; Cho, Seong Kuk; Oh, Bong Seong; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    As part of the SMR(Small Modular Reactor)s development effort, the authors propose a concept of supercritical CO{sub 2} (S-CO{sub 2}) cooled fast reactor combined with the S-CO{sub 2} Brayton cycle. The reactor concept is named as KAIST Micro Modular Reactor (MMR). The S-CO{sub 2} Brayton cycle has many strong points when it is used for SMR's power conversion unit. It occupies small footprints due to the compact cycle components and simple layout. Thus, a concept of one module containing the S-CO{sub 2} cooled fast reactor and power conversion system is possible. This module can be shipped via ground transportation (by trailer) or marine transportation. In this study, the authors propose a new conceptual layout for the S-CO{sub 2} cooled direct cycle while considering various issues for arranging cycle components. The new design has an improved cycle efficiency (from 31% to 34%) than the earlier version of MMR by reducing pressure drops in the heat exchangers. As a more efficient option, a recompression recuperated cycle was also designed. It improves 5% of thermal efficiency while 18tons of mass can be added in comparison to the simple recuperated cycle. Even if we adopt recompression cycle as a PCU, the weight of module (152tons) is less than the ground transportable limit (260tons)

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

    Science.gov (United States)

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

    2017-08-01

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

  17. Keeping research reactors relevant: A pro-active approach for SLOWPOKE-2

    International Nuclear Information System (INIS)

    Cosby, L.R.; Bennett, L.G.I.; Nielsen, K.; Weir, R.

    2010-01-01

    The SLOWPOKE is a small, inherently safe, pool-type research reactor that was engineered and marketed by Atomic Energy of Canada Limited (AECL) in the 1970s and 80s. The original reactor, SLOWPOKE-1, was moved from Chalk River to the University of Toronto in 1970 and was operated until upgraded to the SLOWPOKE-2 reactor in 1973. In all, eight reactors in the two versions were produced and five are still in operation today, three having been decommissioned. All of the remaining reactors are designated as SLOWPOKE-2 reactors. These research reactors are prone to two major issues: aging components and lack of relevance to a younger audience. In order to combat these problems, one SLOWPOKE -2 facility has embraced a strategy that involves modernizing their reactor in order to keep the reactor up to date and relevant. In 2001, this facility replaced its aging analogue reactor control system with a digital control system. The system was successfully commissioned and has provided a renewed platform for student learning and research. The digital control system provides a better interface and allows flexibility in data storage and retrieval that was never possible with the analogue control system. This facility has started work on another upgrade to the digital control and instrumentation system that will be installed in 2010. The upgrade includes new computer hardware, updated software and a web-based simulation and training system that will allow licensed operators, students and researchers to use an online simulation tool for training, education and research. The tool consists of: 1) A dynamic simulation for reactor kinetics (e.g., core flux, power, core temperatures, etc). This tool is useful for operator training and student education; 2) Dynamic mapping of the reactor and pool container gamma and neutron fluxes as well as the vertical neutron beam tube flux. This research planning tool is used for various researchers who wish to do irradiations (e.g., neutron

  18. Apollo-L2, an advanced fuel tokamak reactor utilizing direct conversion

    International Nuclear Information System (INIS)

    Emmert, G.A.; Kulcinski, G.L.; Blanchard, J.P.; El-Guebaly, L.A.; Khater, H.Y.; Santarius, J.F.; Sawan, M.E.; Sviatoslavsky, I.N.; Wittenberg, L.J.; Witt, R.J.

    1989-01-01

    A scoping study of a tokamak reactor fueled by a D- 3 He plasma is presented. The Apollo D- 3 He tokamak capitalizes on recent advances in high field magnets (20 T) and utilizes rectennas to convert the synchrotron radiation directly to electricity. The low neutron wall loading (0.1 MW/m 2 ) permits a first wall lasting the life of the plant and enables the reactor to be classified as inherently safe. The cost of electricity is less than that from a similar power level DT reactor. 10 refs., 1 fig., 4 tabs

  19. A method of reactor power decrease by 2DOF control system during BWR power oscillation

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Suzuki, Katsuo

    1998-09-01

    Occurrence of power oscillation events caused by void feedback effects in BWRs operated at low-flow and high-power condition has been reported. After thoroughly examining these events, BWRs have been equipped with the SRI (Selected Rod Insertion) system to avoid the power oscillation by decreasing the power under such reactor condition. This report presents a power control method for decreasing the reactor power stably by a two degree of freedom (2DOF) control. Performing a numerical simulation by utilizing a simple reactor dynamics model, it is found that the control system designed attains a satisfactory control performance of power decrease from a viewpoint of setting time and oscillation. (author)

  20. Status of IVO-FR2-Vg7 experiment for irradiation of fast reactor fuel rods

    International Nuclear Information System (INIS)

    Elbel, H.; Kummerer, K.; Bojarsky, K.; Lopez Jimenez, J.; Otero de la Gandara, J.L.

    1979-01-01

    Report on the Seminar celebrated in Madrid between KfK (Karlsruhe) and JEN (Madrid) concerning a Joint Irradiation Program of Fast Reactor Fuel Rods. The design of fuel rods in general is defined, and, in particular of those with a density 94% DT and diameter 7.6 mm up to a burn-up of 7% FIMA, to be irradiated in the FR2 Reactor (Karlsruhe). Together with the design of NaK and single-wall capsules used in this irradiation, other possibilities of irradiation in the reactor will also be described. (auth.)

  1. In search of new neutrinos and dark matter. The return of fundamental research to BR2

    International Nuclear Information System (INIS)

    2015-01-01

    A consortium of three French, two British, and four Flemish universities and research institutions, including the Belgian Nuclear Research Center SCK-CEN, started in 2014 on the construction of a neutrino experiment in the BR2 reactor. A reactor such as this is an extremely intense source of neutrinos: elementary particles that are generated as a by-product of nuclear beta decay. BR2 is particularly suitable with regard to carrying out this measurement because of the compact core, the high operating capacity, sufficient space for placing a fairly large detector, and the extremely low background radiation. The article discusses recent developments.

  2. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

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

  3. 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.

  4. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

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

  5. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

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

  6. AMNT 2014. Key topic: Reactor operation, safety - report. Pt. 2

    International Nuclear Information System (INIS)

    Fischer, Klaus-Christian; Willschuetz, Hans-Georg; Wortmann, Birgit

    2014-01-01

    Summary report on the following sessions of the Annual Conference on Nuclear Technology held in Frankfurt, 6 to 8 May 2014: - Thermo Dynamics and Fluid Dynamics: Experiments and Backfittings for the Improvement of Safety and Efficiency; - Safety of Nuclear Installations - Methods, Analyses, Results: In-Vessel Phenomena; Ex-Vessel Phenomena; - Standards and Regulations; Hazard and Safety Analysis; and Validation and Uncertainty Analysis. The other Sessions of the Key Topics 'Reactor Operation, Safety', 'Competence, Innovation, Regulation' and 'Fuel, Decommissioning and Disposal' have been covered in atw 10 (2014) and will be covered in further issues of atw.

  7. Reactor safety research and safety technology. Pt. 2

    International Nuclear Information System (INIS)

    Theenhaus, R.; Wolters, J.

    1987-01-01

    The state of HTR safety research work reached permits a comprehensive and reliable answer to be given to questions which have been raised by the reactor accident at Chernobyl, regarding HTR safety. Together with the probability safety analyses, the way to a safety concept suitable for an HTR is cleared; instructions are given for design optimisation with regard to safety technique and economy. The consequences of a graphite fire, the neutron physics design and the consequenes of the lack of a safety containment are briefly described. (DG) [de

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

    International Nuclear Information System (INIS)

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

    2005-01-01

    G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

  9. COSTANZA, 1-D 2 Group Space-Dependent Reactor Dynamics of Spatial Reactor with 1 Group Delayed Neutrons

    International Nuclear Information System (INIS)

    Agazzi, A.; Gavazzi, C.; Vincenti, E.; Monterosso, R.

    1964-01-01

    1 - Nature of physical problem solved: The programme studies the spatial dynamics of reactor TESI, in the two group and one space dimension approximation. Only one group of delayed neutrons is considered. The programme simulates the vertical movement of the control rods according to any given movement law. The programme calculates the evolution of the fluxes and temperature and precursor concentration in space and time during the power excursion. 2 - Restrictions on the complexity of the problem: The maximum number of lattice points is 100

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  11. Development of Zr-2.5Nb pressure tubes for Advanced CANDU Reactor

    International Nuclear Information System (INIS)

    Bickel, G.A.; Griffiths, M.; Douchant, A.; Douglas, S.; Woo, O.T.; Buyers, A.

    2010-01-01

    In an Advanced CANDU Reactor (ACR), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure and flux than that in a CANDU reactor. In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a manufacturing process has been developed to produce 6.5 mm thick ACR pressure tubes with optimized chemical composition, improved mechanical properties and in-reactor behaviour. The test and examination results show that, when compared with current in-service pressure tubes, the mechanical properties of ACR pressure tubes are significantly improved. Based on previous experience with CANDU reactor pressure tubes an assessment of the grain structure and texture indicates that the in-reactor creep deformation will be improved also. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor. (author)

  12. Distribution of energy of impulses of the modernized IBR-2 REACTOR

    International Nuclear Information System (INIS)

    Tayibov, L.A; Mehtiyeva, R.N.; )

    2011-01-01

    Full text: For the modernized IBR-2 reactor there are two main reasons causing fluctuations of energy of impulses [1,3] on low power of stochastic fluctuations, on the nominal - giving rise to fluctuations of external reactance. The fluctuations of pulse energy is quite significant (20%). They affect the dynamics of the reactor, the process of regulation, starting, as well as the work of the experimental apparatus, etc. It is clear that research of fluctuation of energy of impulses has special value for the IBR-2 type reactor. Sufficient information about the statistical properties of the reactor noise gives the density distribution of the energy pulse power. We used the usual procedure of statistical analysis of time series. Calculated pulse energy of density and the parameters of this distribution.

  13. Estimation of power feedback parameters of the IBR-2M reactor by square wave reactivity

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Popov, A.K.; Sumkhuu, D.

    2016-01-01

    Parameters of the IBR-2M reactor power feedback (PFB) are estimated based on the analysis of power transients caused by deliberate square wave reactivity when the pulsed reactor operates in the self-regulation mode. The PFB of the IBR-2M is described by three linear first-order differential equations. Two components of the PFB are responsible for the negative feedback and one, for the positive. The overall feedback is negative, i.e., it has a stabilizing effect for the operation of the reactor. The slowest negative component of the PFB is probably caused by heating of the fuel. Periodically repeated in the process of exploitation, estimation of the PFB parameters is one of the methods to ensure safety operation of the reactor. [ru

  14. Groundwater Monitoring Plan for the Reactor Technology Complex Operable Unit 2-13

    International Nuclear Information System (INIS)

    Richard P. Wells

    2007-01-01

    This Groundwater Monitoring Plan describes the objectives, activities, and assessments that will be performed to support the on-going groundwater monitoring requirements at the Reactor Technology Complex, formerly the Test Reactor Area (TRA). The requirements for groundwater monitoring were stipulated in the Final Record of Decision for Test Reactor Area, Operable Unit 2-13, signed in December 1997. The monitoring requirements were modified by the First Five-Year Review Report for the Test Reactor Area, Operable Unit 2-13, at the Idaho National Engineering and Environmental Laboratory to focus on those contaminants of concern that warrant continued surveillance, including chromium, tritium, strontium-90, and cobalt-60. Based upon recommendations provided in the Annual Groundwater Monitoring Status Report for 2006, the groundwater monitoring frequency was reduced to annually from twice a year

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

    International Nuclear Information System (INIS)

    1981-08-01

    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 m 3 . 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)

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  17. An improved thermal-hydraulic modeling of the Jules Horowitz Reactor using the CATHARE2 system code

    Energy Technology Data Exchange (ETDEWEB)

    Pegonen, R., E-mail: pegonen@kth.se [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden); Bourdon, S.; Gonnier, C. [CEA, DEN, DER, SRJH, CEA Cadarache, 13108 Saint-Paul-lez-Durance Cedex (France); Anglart, H. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden)

    2017-01-15

    Highlights: • An improved thermal-hydraulic modeling of the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during loss of flow accident. • The heat exchanger approach gives more realistic and less conservative results. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support current and future nuclear reactor designs. The reactor is under construction at the CEA Cadarache research center in southern France and is expected to achieve first criticality at the end of this decade. This paper presents an improved thermal-hydraulic modeling of the reactor using solely CATHARE2 system code. Up to now, the CATHARE2 code was simulating the full reactor with a simplified approach for the core and the boundary conditions were transferred into the three-dimensional FLICA4 core simulation. A new more realistic methodology is utilized to analyze the thermal-hydraulic simulation of the reactor during a loss of flow accident.

  18. TOKMINA, Toroidal Magnetic Field Minimization for Tokamak Fusion Reactor. TOKMINA-2, Total Power for Tokamak Fusion Reactor

    International Nuclear Information System (INIS)

    Hatch, A.J.

    1975-01-01

    1 - Description of problem or function: TOKMINA finds the minimum magnetic field, Bm, required at the toroidal coil of a Tokamak type fusion reactor when the input is beta(ratio of plasma pressure to magnetic pressure), q(Kruskal-Shafranov plasma stability factor), and y(ratio of plasma radius to vacuum wall radius: rp/rw) and arrays of PT (total thermal power from both d-t and tritium breeding reactions), Pw (wall loading or power flux) and TB (thickness of blanket), following the method of Golovin, et al. TOKMINA2 finds the total power, PT, of such a fusion reactor, given a specified magnetic field, Bm, at the toroidal coil. 2 - Method of solution: TOKMINA: the aspect ratio(a) is minimized, giving a minimum value for Bm. TOKMINA2: a search is made for PT; the value of PT which minimizes Bm to the required value within 50 Gauss is chosen. 3 - Restrictions on the complexity of the problem: Input arrays presently are dimensioned at 20. This restriction can be overcome by changing a dimension card

  19. CO2 Energy Reactor - Integrated Mineral Carbonation: Perspectives on Lab-Scale Investigation and Products Valorization

    OpenAIRE

    Rafael M Santos; Pol CM Knops; Keesjan L Rijnsburger; Yi Wai eChiang

    2016-01-01

    To overcome the challenges of mineral CO2 sequestration, Innovation Concepts B.V. is developing a unique proprietary gravity pressure vessel (GPV) reactor technology and has focussed on generating reaction products of high economic value. The GPV provides intense process conditions through hydrostatic pressurization and heat exchange integration that harvests exothermic reaction energy, thereby reducing energy demand of conventional reactor designs, in addition to offering other benefits. In ...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1958-07-01

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

  1. Development of a TiO2-coated optical fiber reactor for water decontamination

    International Nuclear Information System (INIS)

    Danion, A.

    2004-09-01

    The objective of this study was to built and to study a photo-reactor composed by TiO 2 -coated optical fibers for water decontamination. The physico-chemical characteristics and the optical properties of the TiO 2 coating were first studied. Then, the influences of different parameters as the coating thickness, the coating length and the coating volume were investigated both on the light transmission in the TiO 2 - coated fiber and on the photo-catalytic activity of the fiber for a model compound (malic acid). The photo-catalytic degradation of malic acid was optimized using the experimental design methodology allowing to build a multi-fiber reactor comprising 57 optical fibers. The photo-degradation of malic acid was conducted in the multi-fiber reactor and it was demonstrated that the multi-fiber reactor was more efficient than the single-fiber reactor at the same fibers density. Finally, the multi-fiber reactor was applied to the photo-degradation of a fungicide, called fenamidone, and a degradation pathway was proposed. (author)

  2. Power noise spectrum classification in the problem of the IBR-2 reactor

    International Nuclear Information System (INIS)

    Bargel, M.; Kitowski, J.; Pepelyshev, Yu.N.

    1988-01-01

    The classification spectrum results of random fluctuations in the IBR-2 energy pulse are presented. The work is performed for the application of the obtained results to the reactor diagnostics and the study of its noise uncontrolled states. For classification of the spectra the method of pattern recognition based upon the ISODATA heuristic algorithm is used. It is shown that a set of noise uncontrolled reactor states, registered during the reactor operation period at power of 0.4-2 MVt with the first variant of moving reflector (1983-1986) is formed into 4(5) most typical states. Each of the states corresponds to the general conditions of the reactor core cooling and provides the normal work of the moving reflector. However, these states differ in coolant flow, power level and peculiarities of the moving reflector rotation regime. One type of anomal power noise, connected with some disorder in the moving reflctor work, is isolated. This work also presents the possibility of control over the state of moving reflectors according to the change in the amplitude of power oscillations at some frequences. The reactor noise classification results can be used as the data bank for the IBR-2 reactor diagnostic system

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

    NARCIS (Netherlands)

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

    2008-01-01

    This book contains the proceedings of the 20th edition of the Belgian-Netherlands Conference on Artificial Intelligence. The conference was organized by the Human Media Interaction group of the University of Twente. As usual, the conference was under the auspices of the Belgian-Dutch Association for

  4. 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.

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

    Science.gov (United States)

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

    2018-05-21

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-02-15

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

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

    International Nuclear Information System (INIS)

    Merny, R.

    1986-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-02-01

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

  9. Materials characterization for advanced pressurized water reactors: Pt. 2

    International Nuclear Information System (INIS)

    Little, E.A.; Gage, G.

    1994-01-01

    A compilation and overview is presented of the experimental techniques available for characterization of the microstructural changes induced by neutron irradiation of PWR pressure vessel steels, and directed towards monitoring of embrittlement processes by examination of surveillance samples from advanced reactor systems. The microstructural features of significance include copper precipitation, dislocation loop and/or microvoid matrix damage and grain boundary solute segregation. The techniques of transmission electron microscopy, field-emission gun scanning transmission electron microscopy, small angle neutron scattering, positron annihilation and field-ion microscopy have all developed to a degree of sophistication such that they are capable of providing detailed microstructural information in these areas, and afford considerable insight into embrittlement processes when used in combination. (author)

  10. Decommissioning plan for Tammuz-2 research reactor in Iraq

    International Nuclear Information System (INIS)

    Ahmed, A. A.; Jasim, H. I.

    2012-12-01

    For nuclear facilities, decommissioning is the final phase in the life cycle after sitting, design, construction, commissioning and operation. It is a process involving operations such as decontamination, dismantling of plant equipment of result in materials. All these activities take into account health and safety requirements for operating personnel and the general public, and any implications for the environment. (1) In several projects to decommission various type of nuclear facilities, it has been shown that technical methods and equipment are available today to dismantle safely nuclear facilities, of whatever type or size. Much experience in the use of these techniques has the decommissioning of prototype, demonstration, and small power reactors or other nuclear facilities. In Iraq these activities will be done by the cooperation with (IAEA) International Atomic Energy Agency and the other national regulatory bodies such as (IRSRA) Iraqi Radioactive Sources Regulatory Authority, and (RPC/MoEn) Radiation Protection Center/ Ministry of Environment in Iraq. (Author)

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

    International Nuclear Information System (INIS)

    Fujii, Yoshio; Usui, Hozumi; Shinohara, Yoshikuni

    1991-01-01

    This paper describes the prototype heavy-duty telerobotic system constructed as a cold test facility for the development of robotic remote handling system technology in reactor decommissioning. The total system is built up with a multi-functional electrical manipulator system, a manipulator transporter system equipped with a tripedal support mechanism, a monitoring system comprising a 3-D TV monitor, and a computer control system for overall system operation. The manipulator system consists of two master manipulators and two corresponding amphibious slave manipulators with load capacities of 100 and 25 daN, respectively. Valuable engineering experiences for developing more advanced heavy-duty telerobotic system have been gained through designing, constructing and testing the system. (author)

  12. Atomization of U3Si2 for research reactor fuel

    International Nuclear Information System (INIS)

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

    1995-01-01

    Rotating disk atomization technique is applied to KMRR (Korea Multi-purpose Research Reactor) fuel fabrication. A rotating disk atomizer is designed and manufactured locally and U-4.0 wt. % Si alloy powders are produced. The atomized powders are heat-treated to transform into U 3 Si and the mixture of U 3 Si and Al are extruded to fuel meat. Most of the atomized powders are spherical in shape. The microstructure of the powder is fine due to the rapid solidification. The time required for peritectoid reaction is reduced due to the fine microstructures and the resultant U 3 Si grain size is finer than ever obtained from ingot process. The mechanical properties of the fuel meat are improved: yield strength about 30 %, tensile strength 10% and elongation 250 % increased. (author)

  13. Alteration in reactor installations (Unit 1 and 2 reactor facilities) in the Hamaoka Nuclear Power Station of The Chubu Electric Power Co., Inc. (report)

    International Nuclear Information System (INIS)

    1982-01-01

    A report by the Nuclear Safety Commission to the Ministry of International Trade and Industry concerning the alteration in Unit 1 and 2 reactor facilities in the Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., was presented. The technical capabilities for the alteration of reactor facilities in Chubu Electric Power Co., Inc., were confirmed to be adequate. The safety of the reactor facilities after the alteration was confirmed to be adequate. The items of examination made for the confirmation of the safety are as follows: reactor core design (nuclear design, mechanical design, mixed reactor core), the analysis of abnormal transients in operation, the analysis of various accidents, the analysis of credible accidents for site evaluation. (Mori, K.)

  14. CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

    International Nuclear Information System (INIS)

    Kotas, J.F.; Stroh, K.R.

    1983-01-01

    The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident that simulates a control-rod withdrawal at full power

  15. Experimental Irradiations of Materials and Fuels in the BR2 Reactor: An Overview of Current Programmes

    International Nuclear Information System (INIS)

    Van Dyck, S.; Koonen, E.; Verwerft, M.; Wéber, M.

    2013-01-01

    The BR2 material test reactor offers a variety of experimental irradiation possibilities for testing of materials, fuels and instruments. The current paper gives an overview of the recent and ongoing programmes in order to illustrate the experimental potential of the reactor. Three domains of applications are reviewed: Irradiation of materials and fuels for pressurised water reactors (PWR); irradiation of materials for accelerator driven systems (ADS), cooled by liquid lead alloys; and irradiation of fuel for Material Test Reactors (MTR). For PWR relevant tests, a dedicated loop is available, providing a full simulation of the thermo hydraulic conditions of a PWR. ADS related tests require particular control of the irradiation environment and the necessary safety precautions in order to avoid 210 Po contamination. In-core mechanical testing of materials is done in comparison and complimentarily to post-irradiation examinations in order to assess flux related effects on the deformation behaviour of materials. (author)

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

    International Nuclear Information System (INIS)

    Neal, K.; Chaumont, J.C.

    1999-01-01

    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.)

  17. TMI-2 [Three Mile Island Unit 2] reactor building dose reduction task force

    International Nuclear Information System (INIS)

    Daniels, R.S.

    1988-01-01

    In late October 1982, the director of Three Mile Island Unit 2 (TMI-2) created the dose reduction task force with the objective of identifying the principal radiological sources in the reactor building and recommending actions to minimize the dose to workers on labor-intensive projects. Members of the task force were drawn form various groups at TMI. Findings and recommendations were presented to the US Nuclear Regulatory Commission in a briefing on November 18, 1982. The task force developed a three-step approach toward dose reduction. Step 1 identified the radiological sources. Step 2 modeled the source and estimated its contribution to the general area dose rates. Step 3 recommended actions to achieve dose reductions consistent with general exposure rate goals

  18. CYLFUX, Fast Reactor Reactivity Transients Simulation in LWR by 2-D 2 Group Diffusion

    International Nuclear Information System (INIS)

    Schmidt, A.

    1973-01-01

    1 - Nature of physical problem solved: A 2-dimensional calculation of the 2-group, space-dependent neutron diffusion equations is performed in r-z geometry using an arbitrary number of groups of delayed neutron precursors. The program is designed to simulate fast reactivity excursions in light water reactors taking into account Doppler feedback via adiabatic heatup of fuel. Axial motions of control rods may be considered including scram action on option. 2 - Method of solution: The differential equations are solved at each time step by an explicit finite difference method using two time levels. The stationary distributions are obtained by using the same algorithm. 3 - Restrictions on the complexity of the problem: No restriction to the number of space points and delayed neutron energy groups besides the computer size

  19. DRAGON and SERPENT 2-D modelling of the SLOWPOKE-2 reactor at Ecole Polytechnique Montreal

    International Nuclear Information System (INIS)

    Raouafi, H.; Marleau, G.

    2012-01-01

    DRAGON is a deterministic code that can be used to perform lattice cell calculations based on numerical solutions of neutron transport equation. DRAGON can also be used for full core 2-D and 3-D simulations in transport. One alternative to the use of such a deterministic code consist in following the history of neutrons in the core based on statistical Monte Carlo simulation with codes like MCNP and SERPENT. This second calculation approach has been used successfully for SLOWPOKE-2 simulation in the past. Here we present a comparison between DRAGON and SERPENT calculations for the SLOWPOKE-2 reactor. We also compare the flux distribution obtained using both codes for a copper sample placed inside a small irradiation site. (author)

  20. CO_2 capture with solid sorbent: CFD model of an innovative reactor concept

    International Nuclear Information System (INIS)

    Barelli, L.; Bidini, G.; Gallorini, F.

    2016-01-01

    Highlights: • A new reactor solution based on rotating fixed beds was presented. • The preliminary design of the reactor was approached. • A CFD model of the reactor, including CO_2 capture kinetic, was developed. • The CFD model is validated with experimental results. • Sorbent exploitation increasing is possible thanks to the new reactor. - Abstract: In future decarbonization scenarios, CCS with particular reference to post-combustion technologies will be an important option also for energy intensive industries. Nevertheless, today CCS systems are rarely installed due to high energy and cost penalties of current technology based on chemical scrubbing with amine solvent. Therefore, innovative solutions based on new/optimized solvents, sorbents, membranes and new process designs, are R&D priorities. Regarding the CO_2 capture through solid sorbents, a new reactor solution based on rotating fixed beds is presented in this paper. In order to design the innovative system, a suitable CFD model was developed considering also the kinetic capture process. The model was validated with experimental results obtained by the authors in previous research activities, showing a potential reduction of energy penalties respect to current technologies. In the future, the model will be used to identify the control logic of the innovative reactor in order to verify improvements in terms of sorbent exploitation and reduction of system energy consumption.

  1. Oxygen suppression in boiling water reactors. Phase 2. Annual report 1981, December 2, 1980-December 31, 1981

    International Nuclear Information System (INIS)

    Burley, E.L.

    1982-07-01

    A hydrogen addition test will be performed in the Dresden-2 reactor of Commonwealth Edison Company during 1982. Up to 2 ppM hydrogen will be added to and dissolved in the reactor feedwater to reverse the radiolysis reaction in the reactor core and suppress oxgen concentration in the primary coolant. At low oxygen levels the propensity of stressed and sensitized 304 stainless steel toward intergranular stress corrosion cracking is greatly reduced. The test will answer outstanding questions and uncertainties in the areas of water chemistry, equipment design and materials performance. Nine special sample facilities will be prepared in the primary coolant, main stream, feedwater/condensate, and offgas systems. Instrumentation will be available to measure hydrogen, oxygen, conductivity, pH, soluble and insoluble corrosion products, and electrochemical potentials. In addition, an autoclave in which confirming constant extension rate tests can be conducted in reactor water will be provided

  2. Reactor physics measurements with 19-element ThOsub(2)-sup(235)UOsub(2) cluster fuel in heavy water moderator

    International Nuclear Information System (INIS)

    French, P.M.

    1985-02-01

    Low power lattice physics measurements have been performed with a single rod of 19-element thorium oxide fuel enriched with 1.45 wt. percent sub(235)UOsub(2) (93 percent enriched) in a simulated heavy water moderated and cooled power reactor core. The experiments were designed to provide data relevant to a power reactor irradiation and to obtain some basic information on the physics of uranium-thorium fuel material. Some theoretical flux calculations are summarized and show reasonable agreement with experiment

  3. Research on economics and CO2 emission of magnetic and inertial fusion reactors

    International Nuclear Information System (INIS)

    Mori, Kenjiro; Yamazaki, Kozo; Oishi, Tetsutarou; Arimoto, Hideki; Shoji, Tatsuo

    2011-01-01

    An economical and environment-friendly fusion reactor system is needed for the realization of attractive power plants. Comparative system studies have been done for magnetic fusion energy (MFE) reactors, and been extended to include inertial fusion energy (IFE) reactors by Physics Engineering Cost (PEC) system code. In this study, we have evaluated both tokamak reactor (TR) and IFE reactor (IR). We clarify new scaling formulas for cost of electricity (COE) and CO 2 emission rate with respect to key design parameters. By the scaling formulas, it is clarified that the plant availability and operation year dependences are especially dominant for COE. On the other hand, the parameter dependences of CO 2 emission rate is rather weak than that of COE. This is because CO 2 emission percentage from manufacturing the fusion island is lower than COE percentage from that. Furthermore, the parameters dependences for IR are rather weak than those for TR. Because the CO 2 emission rate from manufacturing the laser system to be exchanged is very large in comparison with CO 2 emission rate from TR blanket exchanges. (author)

  4. Metal fire implications for advanced reactors. Part 2, PIRT results

    International Nuclear Information System (INIS)

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

    2008-01-01

    This report documents the results of a Phenomena Identification and Ranking Table (PIRT) exercise performed at Sandia National Laboratories (SNL) as well as the experimental and modeling program that have been designed based on the PIRT results. A PIRT exercise is a structured and facilitated expert elicitation process. In this case, the expert panel was comprised of nine recognized fire science and aerosol experts. The objective of a PIRT exercise is to identify phenomena associated with the intended application and to then rank the current state of knowledge relative to each identified phenomenon. In this particular PIRT exercise the intended application was sodium fire modeling related to sodium-cooled advanced reactors. The panel was presented with two specific fire scenarios, each based on a hypothetical sodium leak in an Advanced Breeder Test Reactor (ABTR) design. For both scenarios the figure of merit was the ability to predict the thermal and aerosol insult to nearby equipment (i.e. heat exchangers and other electrical equipment). When identifying phenomena of interest, and in particular when ranking phenomena importance and the adequacy of existing modeling tools and data, the panel was asked to subjectively weigh these factors in the context of the specified figure of merit. Given each scenario, the panel identified all those related phenomena that are of potential interest to an assessment of the scenario using fire modeling tools to evaluate the figure of merit. Each phenomenon is then ranked relative to its importance in predicting the figure of merit. Each phenomenon is then further ranked for the existing state of knowledge with respect to the ability of existing modeling tools to predict that phenomena, the underlying base of data associated with the phenomena, and the potential for developing new data to support improvements to the existing modeling tools. For this PIRT two hypothetical sodium leak scenarios were evaluated for the ABTR design. The

  5. VISIT OF BELGIAN FIRMS AT CERN

    CERN Multimedia

    2001-01-01

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

  6. Power Reactor Fuel Reprocessing Plant-2, Tarapur: a benchmark in Indian PHWR spent fuel reprocessing

    International Nuclear Information System (INIS)

    Pradhan, Sanjay; Dubey, K.; Qureshi, F.T.; Lokeswar, S.P.

    2017-01-01

    Power Reactor Fuel Reprocessing Plant-2 (PREFRE-2) is latest operating spent nuclear fuel reprocessing plant in India. This plant has improved design based on latest technology and feedback provided by the earlier plants. The design of PREFRE-2 plant is in five cycles of solvent extraction using TBP as extractant. The plant is commissioned in year 2011 after regulatory clearances

  7. 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.

  8. 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.

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

    Science.gov (United States)

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

    2014-12-01

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

  10. VENUS-2, Reactor Kinetics with Feedback, 2-D LMFBR Disassembly Excursions

    International Nuclear Information System (INIS)

    Jackson, J.F.; Nicholson, R.B.; Weber, D.P.

    1980-01-01

    1 - Description of problem or function: VENUS-2 is an improved edition of the VENUS fast-reactor disassembly program. It is a two- dimensional (r-z) coupled neutronics-hydrodynamics code that calculates the dynamic behavior of an LMFBR during a prompt-critical disassembly excursion. It calculates the power history and fission energy release as well as the space-time histories of the fuel temperatures, core material pressures, and core material motions. Reactivity feedback effects due to Doppler broadening and reactor material motion are taken into account. 2 - Method of solution: The power and energy release are calculated using a point-kinetics formulation with up to six delayed neutron groups. The reactivity is a combination of an input driving function and feedback effects due to Doppler broadening and material motion. An adiabatic model is used to calculate the temperature increase throughout the reactor based on an initial temperature distribution and power profile provided as input data. These temperatures are, in turn, converted to fuel pressures through one of several equation of state options provided. The material motion that results from the pressure buildup is calculated by a direct finite difference solution of a set of two-dimensional (r-z) hydrodynamics equations. This is done in Lagrangian coordinates. The reactivity change associated with this motion is calculated by first-order perturbation theory. The displacements are also used to adjust the fuel densities as required for the density dependent equation-of- state option. An automatic time-step-size selection scheme is provided. 3 - Restrictions on the complexity of the problem: VENUS-2 is written so that the dimensions of the storage arrays can be readily changed to accommodate a broad range of problem sizes. In the base version, the total number of mesh intervals is restricted such that (NR+3)*(NZ+3) is less than 700, where NR and NZ are the total number of mesh intervals in the r and z

  11. Belgian experience with radiation technologies for sterilization

    International Nuclear Information System (INIS)

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

    1998-01-01

    Belgium pioneered in non energetic applications of nuclear science. In 1970, the National Institute for Radioisotopes (IRE) was founded on the ground of the experience acquired at the CEN/SCK, for developing nuclear techniques oriented to the well-being of population. In 1978 IRE started operation of 2 γ-irradiation units with a 2.25 million Ci Co 60 , source having a capacity for sterilization of more than 100 m 3 of product per day. this installation is currently operated by Griffith-Mediris (Group Griffith Micro Science) and has accumulated 20 years of experience for irradiation of foodstuffs, medical appliances and pharmaceuticals. In 1986, IRE was producing radioisotopes from accelerators and joined UCL for founding Ion Beam Applications (IBA) company which shortly became the world leader for design and delivery of cyclotrons. More recently, on the basis of French CEA patent, IBA developed the Rhodotron, an e-beam accelerator, with an X-ray option, which is now facing commercial success i area of ionization of food and medical appliances. IRE developed also expertise for the back-end of these activities i.e. final disposal of radioactive sealed source and dismantling of the installations. (author)

  12. Physics design of advanced steady-state tokamak reactor A-SSTR2

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Ushigusa, Kenkichi

    2000-10-01

    Based on design studies on the fusion power reactor such as the DEMO reactor SSTR, the compact power reactor A-SSTR and the DREAM reactor with a high environmental safety and high availability, a new concept of compact and economic fusion power reactor (A-SSTR2) with high safety and high availability is proposed. Employing high temperature superconductor, the toroidal filed coils supplies the maximum field of 23T on conductor which corresponds to 11T at the magnetic axis. A-SSTR2 (R p =6.2m, a p =1.5m, I p =12MA) has a fusion power of 4GW with β N =4. For an easy maintenance and for an enough support against a strong electromagnetic force on coils, a poloidal coils system has no center solenoid coils and consists of 6 coils located on top and bottom of the machine. Physics studies on the plasma equilibrium, controllability of the configuration, the plasma initiation and non-inductive current ramp-up, fusion power controllability and the diverter have shown the validity of the A-SSTR2 concept. (author)

  13. 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.

  14. Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

    International Nuclear Information System (INIS)

    Bissani, M; O'Kelly, D S

    2006-01-01

    A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to provide color-enhanced gemstones but is

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

    International Nuclear Information System (INIS)

    Bombaerts, Gunter; Bovy, Michel; Laes, Erik

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-09-15

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

  17. Degradation of gas-phase trichloroethylene over thin-film TiO2 photocatalyst in multi-modules reactor

    International Nuclear Information System (INIS)

    Kim, Sang Bum; Lee, Jun Yub; Kim, Gyung Soo; Hong, Sung Chang

    2009-01-01

    The present paper examined the photocatalytic degradation (PCD) of gas-phase trichloroethylene (TCE) over thin-film TiO 2 . A large-scale treatment of TCE was carried out using scale-up continuous flow photo-reactor in which nine reactors were arranged in parallel and series. The parallel or serial arrangement is a significant factor to determine the special arrangement of whole reactor module as well as to compact the multi-modules in a continuous flow reactor. The conversion of TCE according to the space time was nearly same for parallel and serial connection of the reactors.

  18. 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...

  19. Design and computational analysis of passive siphon breaker for 49-2 swimming pool reactor

    International Nuclear Information System (INIS)

    Yue Zhiting; Song Yunpeng; Liu Xingmin; Zou Yao; Wu Yuanyuan

    2014-01-01

    Based on safety considerations, a passive siphon breaker will be added to the primary cooling system of 49-2 Swimming Pool Reactor (SPR). With the breaker location determined, the capability of siphon breakers with diameters of 1.5 cm and 2.0 cm was calculated and analyzed respectively by RELAP5/MOD3.3 code. The results show that in the condition of large break loss of coolant accident these two sizes of siphon breakers are able to break the siphon phenomena, and maintain the pool water level above the reactor core when the reactor and the pump are shutdown. In the end, to be conservative, the siphon breaker with diameter of 2.0 cm is adopted. (authors)

  20. Power Quality Problems Mitigation using Dynamic Voltage Restorer in Egypt Thermal Research Reactor (ETRR-2)

    International Nuclear Information System (INIS)

    Kandil, T.; Ayad, N.M.; Abdel Haleam, A.; Mahmoud, M.

    2013-01-01

    Egypt thermal research reactor (ETRR-2) was subjected to several Power Quality Problems such as voltage sags/swells, harmonics distortion, and short interruption. ETRR-2 encompasses a wide range of loads which are very sensitive to voltage variations and this leads to several unplanned shutdowns of the reactor due to trigger of the Reactor Protection System (RPS). The Dynamic Voltage Restorer (DVR) has recently been introduced to protect sensitive loads from voltage sags and other voltage disturbances. It is considered as one of the most efficient and effective solution. Its appeal includes smaller size and fast dynamic response to the disturbance. This paper describes a proposal of a DVR to improve power quality in ETRR-2 electrical distribution systems . The control of the compensation voltage is based on d-q-o algorithm. Simulation is carried out by Matlab/Simulink to verify the performance of the proposed method

  1. General outline of the operation and utilization of the BR2 reactor

    International Nuclear Information System (INIS)

    Baugnet, J.M.; Leonard, F.; Gandolfo, J.M.; Lenders, H.

    1978-01-01

    The BR2 reactor is a high-flux material testing reactor of the thermal heterogeneous type. The fuel is 93% 235 U enriched uranium in the form of plates clad in aluminium. The moderator consists of beryllium and light water, the water being pressurized (12.5kg/cm 2 )and acting also as coolant. The pressure vessel is of aluminium, and is placed in a pool of demineralized water. One should stress the following main features of the design: the experimental channels are skew, the tube bundle presenting the form of a hyperboloid of revolution (see figure 1)-this gives easy access at the top and bottom reactor covers allowing complex instrumented devices, while maintaining a very high neutron flux at the core; great flexibilty of utilization, due to the fact that it is possible to adapt the core configuration to the experimental loading as the fissile charge can be centred on different experimental channels; although BR2 is a thermal reactor, it is possible to achieve neutron spectra very similar to those obtained in a fast reactor, either by the use of absorbing screens or by the use of fissile material within the experimental device; five 200mm diameter channels are available for loading large experimental irradiation devices, as in-pile sodium, gas or water loops. (author)

  2. Shadow corrosion testing in the INCA facility in the Studsvik R2 reactor

    International Nuclear Information System (INIS)

    Nystrand, A.C.; Lassing, A.

    1999-01-01

    Shadow corrosion is a phenomenon which occurs when zirconium alloys are in contact with or in proximity to other metallic objects in a boiling water reactor environment (BWR, RBMK, SGHWR etc.). An enhanced corrosion occurs on the zirconium alloy with the appearance of a 'shadow' of the metallic object. The magnitude of the shadow corrosion can be significant, and is potentially limiting for the lifetime of certain zirconium alloy components in BWRs and other reactors with a similar water chemistry. In order to evaluate the suitability of the In-Core Autoclave (INCA) in the Studsvik R2 materials testing reactor as an experimental facility for studying shadow corrosion, a demonstration test has been performed. A number of test specimens consisting of Zircaloy-2 tubing in contact with Inconel were exposed in an oxidising water chemistry. Some of the specimens were placed within the reactor core and some above the core. The conclusion of this experiment after post irradiation examination is that it is possible to use the INCA facility in the Studsvik R2 reactor to develop a significant level of shadow corrosion after only 800 hours of irradiation. (author)

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

    International Nuclear Information System (INIS)

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  5. Reciprocity and depressive symptoms in Belgian workers: a cross-sectional multilevel analysis.

    Science.gov (United States)

    De Clercq, Bart; Clays, Els; Janssens, Heidi; De Bacquer, Dirk; Casini, Annalisa; Kittel, France; Braeckman, Lutgart

    2013-07-01

    This study examines the multidimensional association between reciprocity at work and depressive symptoms. Data from the Belgian BELSTRESS survey (32 companies; N = 24,402) were analyzed. Multilevel statistical procedures were used to account for company-level associations while controlling for individual-level associations. Different dimensions of individual reciprocity were negatively associated with depressive symptoms. On the company level, only vertical emotional reciprocity was negatively associated (β = -4.660; SE = 1.117) independently from individual reciprocity (β = -0.557; SE = 0.042). Complex interactions were found such that workplace reciprocity (1) may not uniformly benefit individuals and (2) related differently to depressive symptoms, depending on occupational group. This study extends the existing literature with evidence on the multidimensional, contextual, and cross-level interaction associations of reciprocity as a key aspect of social capital on depressive symptoms.

  6. TiO2 Solar Photocatalytic Reactor Systems: Selection of Reactor Design for Scale-up and Commercialization—Analytical Review

    Directory of Open Access Journals (Sweden)

    Yasmine Abdel-Maksoud

    2016-09-01

    Full Text Available For the last four decades, viability of photocatalytic degradation of organic compounds in water streams has been demonstrated. Different configurations for solar TiO2 photocatalytic reactors have been used, however pilot and demonstration plants are still countable. Degradation efficiency reported as a function of treatment time does not answer the question: which of these reactor configurations is the most suitable for photocatalytic process and optimum for scale-up and commercialization? Degradation efficiency expressed as a function of the reactor throughput and ease of catalyst removal from treated effluent are used for comparing performance of different reactor configurations to select the optimum for scale-up. Comparison included parabolic trough, flat plate, double skin sheet, shallow ponds, shallow tanks, thin-film fixed-bed, thin film cascade, step, compound parabolic concentrators, fountain, slurry bubble column, pebble bed and packed bed reactors. Degradation efficiency as a function of system throughput is a powerful indicator for comparing the performance of photocatalytic reactors of different types and geometries, at different development scales. Shallow ponds, shallow tanks and fountain reactors have the potential of meeting all the process requirements and a relatively high throughput are suitable for developing into continuous industrial-scale treatment units given that an efficient immobilized or supported photocatalyst is used.

  7. Safety assessments relating to the use of new fuels in research reactors: application to the case of FRM 2 reactor fuel

    International Nuclear Information System (INIS)

    Abou Yehia, H.; Bars, G.; Tran Dai

    2001-01-01

    After giving a brief reminder of the procedure applied in France for the licensing of the use of a new fuel type or design in a research reactor, we outline the main safety aspects associated with such a modification. Finally, by way of an example, we focus on the safety assessment relating to the IRIS irradiation device used in SILOE reactor, in particular for the qualification of the fuel dedicated to FRM II reactor of the Technical University of Munich. This qualification was carried out on a U 3 Si 2 fuel plate enriched to about 90 % in weight of 235 U and containing 1.5 g of uranium per cm 3 . The evaluation performed by the IPSN for GRS did not call into question the choice of U 3 Si 2 fuel plates for the FRM-II reactor. (authors)

  8. N2O Catalytic Decomposition – from Laboratory Experiment to Industry Reactor

    Czech Academy of Sciences Publication Activity Database

    Obalová, L.; Jirátová, Květa; Karásková, K.; Chromčáková, Ž.

    2012-01-01

    Roč. 191, č. 1 (2012), s. 116-120 ISSN 0920-5861 R&D Projects: GA TA ČR TA01020336 Institutional support: RVO:67985858 Keywords : N2O * catalytic decomposition * fixed bed reactor Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 2.980, year: 2012

  9. Analysis of SBO accident and natural circulation of 49-2 swimming pool reactor

    International Nuclear Information System (INIS)

    Wu Yuanyuan; Liu Tiancai; Sun Wei

    2012-01-01

    The transient thermal hydraulic characteristics of 49-2 Swimming Pool Reactor (SPR) were analyzed by RELAP5/MOD3.3 code to verify the capability of natural circulation and minus reactivity feedback for accident mitigation under the condition of station blackout (SBO). Then, the effects on accident consequence and sequence for core channels and primary pumps were briefly discussed. The calculation results show that the reactor can be shutdown by the effect of minus reactivity feedback, and the residual heat can be removed through the stable natural circulation. Therefore, it demonstrates that the 49-2 SPR is safe during the accident of SBO. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-07-01

    This Volume 2 of the Proceedings of 2. Yugoslav symposium on reactor physics includes eight papers dealing with the following topics: method for measuring high anti reactivities of a reactor system; integration method for thermal reaction rate calculation; Determination of initial core configuration for BHWR-200 MWe; safety shutdowns and failures of the RA reactor equipment; determining the reactivity of absorption rods; measurements of thermal and fast neutron fluxes at the TRIGA reactor and other measurements during operation of the TRIGA reactor; mathematical modelling of the reactor safety; review of problems and methods for radiation risk assessment in the environment of a nuclear power plant.

  11. Radiation protection at the RA Reactor in 1986, Part -2, Annex 2a, Radioactivity control of the RA reactor environment (atmospheric precipitations, dust, water, soil, plants, fruit...)

    International Nuclear Information System (INIS)

    Ajdacic, N.; Martic, M.; Jovanovic, J.

    1986-01-01

    Control of radioactivity in the biosphere in the vicinity of the RA reactor is part of the radioactivity control done regularly for the whole territory of the Vinca institute (report by the same authors included in this Annex). During 1986 control was conducted according to the plan until May 1, 1986 when a dramatic increase of the precipitations and all other samples from the biosphere was recorded. According to the measured data no significant changes have been found in the surroundings of the RA reactor, until April 29 1986. Since then more detailed control was conducted, the number of samples was increased, apart from standard measuring procedure of total beta activity measurements, gamma spectrometry of all samples was applied. High activity level of the following nuclides was found: Iodine, cerium,cesium, tellurium, ruthenium, barium, lanthanum, etc. As an example activity of ?1?3?1 I in one sample was 564±5 kBq/m 2 [sr

  12. Tests of Neutron Spectrum Calculations with the Help of Foil Measurements in a D{sub 2}O and in an H{sub 2}O-Moderated Reactor and in Reactor Shields of Concrete an Iron

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, R; Aalto, E

    1964-09-15

    Foil measurements covering the fast, epithermal and thermal neutron energy regions have been made in the centre of the Swedish D{sub 2}O-moderated reactor R1, in the pool reactor R2-0, and in different positions in reactor shields of iron, magnetite concrete and ordinary concrete. Neutron spectra have also been calculated for most of these positions, often with the help of a numerical integration of the Boltzmann equation. The measurements and the calculated spectra are presented.

  13. UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

    Energy Technology Data Exchange (ETDEWEB)

    Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Novitrian,; Pramuditya, Syeilendra; Su’ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Aji, Indarta K. [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia)

    2016-03-11

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.

  14. 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).

  15. Preliminary Design of S-CO2 Brayton Cycle for KAIST Micro Modular Reactor

    International Nuclear Information System (INIS)

    Kim, Seong Gu; Kim, Min Gil; Bae, Seong Jun; Lee, Jeong Ik

    2013-01-01

    This paper suggests a complete modular reactor with an innovative concept of reactor cooling by using a supercritical carbon dioxide directly. Authors propose the supercritical CO 2 Brayton cycle (S-CO 2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the core and PCU in one vessel for the full modularization. This study suggests a conceptual design of small modular reactor including PCU which is named as KAIST Micro Modular Reactor (MMR). As a part of ongoing research of conceptual design of KAIST MMR, preliminary design of power generation cycle was performed in this study. Since the targets of MMR are full modularization of a reactor system with S-CO 2 coolant, authors selected a simple recuperated S-CO 2 Brayton cycle as a power conversion system for KAIST MMR. The size of components of the S-CO 2 cycle is much smaller than existing helium Brayton cycle and steam Rankine cycle, and whole power conversion system can be contained with core and safety system in one containment vessel. From the investigation of the power conversion cycle, recompressing recuperated cycle showed higher efficiency than the simple recuperated cycle. However the volume of heat exchanger for recompressing cycle is too large so more space will be occupied by heat exchanger in the recompressing cycle than the simple recuperated cycle. Thus, authors consider that the simple recuperated cycle is more suitable for MMR. More research for the KAIST MMR will be followed in the future and detailed information of reactor core and safety system will be developed down the road. More refined cycle layout and design of turbomachinery and heat exchanger will be performed in the future study

  16. FORE-2, Thermohydraulics and Space-Independent Reactor Kinetics for Transients

    International Nuclear Information System (INIS)

    Fox, J.N.; Lawler, B.E.; Butz, H.R.; Heames, T.J.

    1984-01-01

    1 - Description of problem or function: FORE2 is a coupled thermal hydraulics-point kinetics digital computer code designed to calculate significant reactor parameters under steady-state conditions, or as functions of time during transients. The transients may result from a programmed reactivity insertion or a power change. Variable inlet coolant flow rate and temperature are considered. The code calculates the reactor power, the individual reactivity feedbacks, and the temperature of coolant, cladding, fuel, structure, and additional material for up to seven axial positions in three channel types which represent radial zones of the reactor. The heat of fusion, accompanying fuel melting, the liquid metal voiding reactivity, and the spatial and the time variation of the fuel cladding gap coefficient due to changes in gap size are considered. 2 - Method of solution: FORE2 input consists of property data, geometry, power and flow distribution factors, external time varying functions, experimental coefficients, and termination data. The differential equations for fluid flow, heat transfer, and point neutronics are solved by explicit finite-difference procedures. 3 - Restrictions on the complexity of the problem: Reactor excursions which can be calculated are restricted to those transients in which the reactor is not substantially destroyed. As a general rule, changes in reactor geometry and composition during an excursion are limited to those cases in which the reactivity effects of the changes may be considered as small perturbations of the initial system. Thus, accidents involving large-scale disassembly and bulk meltdown of a core are not covered by FORE2. FORE2 is valid only while the core retains its initial geometry

  17. Thermal neutron flux distribution in ET-RR-2 reactor thermal column

    Directory of Open Access Journals (Sweden)

    Imam Mahmoud M.

    2002-01-01

    Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

  18. Rapid data acquisition from the safety system of the FRJ-2 reactor

    International Nuclear Information System (INIS)

    Inhoven, H.

    1980-06-01

    The central department for research reactors (ZFR) of the Juelich Nuclear Research Centre (KFA) is operating the reactors FRJ-1 (MERLIN) and FRJ-2 (DIDO) since 1962. In 1976, a Siemens 330 computer has been put into operation especially for the processing of data from the DIDO reactor, followed by another computer of the same type for the purpose of processing data from the ZFR department in general. The present report is a result of the work investigating 'Data acquisition and data processing in the FRJ-2' and primarily discusses the complex of 'fast analog and binary signals'. The activities in this field of work have been and still are mainly concerned with general problems encountered in adapting a currently 14-year-old reactor system to a digital computer, namely problems such as data decoupling in the safety system of the reactor, data acquisition using the CAMAC system, data transfer via an 'extended branch', data acquisition software as core-resident programs, temporary storage as common data, interpreting software as peripheral - storage - resident programs. (orig./WB) [de

  19. An experimental investigation of fission product release in SLOWPOKE-2 reactors

    International Nuclear Information System (INIS)

    Harnden, A.M.C.

    1995-09-01

    Increasing radiation fields due to a release of fission products in the reactor container of several SLOWPOKE-2 reactors fuelled with a highly-enriched uranium (HEU) alloy core have been observed. It is believed that these increases are associated with the fuel fabrication where a small amount of uranium-bearing material is exposed to the coolant at the end-welds of the fuel element. To investigate this phenomenon samples of reactor water and gas from the headspace above the water have been obtained and examined by gamma spectrometry methods for reactors of various burnups at the University of Toronto, Ecole Polytechnique and Kanata Isotope Production Facility. An underwater visual examination of the fuel core at Ecole Polytechnique has also provided information on the condition of the core. This report (Volume 1) summarizes the equipment, analysis techniques and results of tests conducted at the various reactor sites. The data report is published as Volume 2. (author). 30 refs., 9 tabs., 20 figs

  20. Continuous backfitting measures for the FRG-1 and FRG-2 research reactors

    International Nuclear Information System (INIS)

    Blom, K.H.; Falck, K.; Krull, W.

    1990-01-01

    The GKSS-Research Centre Geesthacht GmbH has been operating the research reactors FRG-1 and FRG-2 with power levels of 5 MW and 15 MW for 31 and 26 years respectively. Safe operation at full power levels over so many years with an average utilization between 180 d to 250 d per year is possible only with great efforts in modernization and upgrading of the research reactors. Emphasis has been placed on backfitting since around 1975. At that time within the Federal Republic of Germany many new guidelines, rules, ordinances, and standards in the field of (power) reactor safety were published. Much work has been done on the modernization of the FRG-1 and FRG-2 research reactors therefore within the last ten years. Work done within the last two years and presently underway includes: measures against water leakage through the concrete and along the beam tubes; repair of both cooling towers; modernization of the ventilation system; measures for fire protection; activities in water chemistry and water quality; installation of a double tubing for parts of the primary piping of the FRG-1; replacement of instrumentation, process control systems (operation and monitoring system) and alarm system; renewal of the emergency power supply; installation of internal lightning protection; installation of a cold neutron source; enrichment reduction for FRG-1. These efforts will continue to allow safe operation of our research reactors over their whole operational life

  1. Experimental estimations of the kinetics parameters of the IBR-2M reactor by stochastic noises

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Tajybov, L.A.; Garibov, A.A.; Mekhtieva, R.N.

    2012-01-01

    Experimental investigations of stochastic fluctuations of pulse energy of the IBR-2M reactor have been carried out which allowed us to obtain some of the parameters of the reactor kinetics. At different levels of average power a sequence of values of pulse energy was recorded with the calculation of the distribution parameters. An ionization chamber with boron installed near the active zone was used as a neutron detector. The research results allowed us to estimate the average lifetime of prompt neutrons τ = (6.53±0.2)·10 -8 s, absolute power of the reactor and intensity of the source of spontaneous neutrons S sp ≤(6.72±0.12)·10 6 s -1 . It was shown that the experimental results are close to the calculated ones

  2. Investigation of hydrogen-burn damage in the Three Mile Island Unit 2 reactor building

    International Nuclear Information System (INIS)

    Alvares, N.J.; Beason, D.G.; Eidem, G.R.

    1982-06-01

    About 10 hours after the March 28, 1979 Loss-of-Coolant Accident began at Three Mile Island Unit 2, a hydrogen deflagration of undetermined extent occurred inside the reactor building. Examinations of photographic evidence, available from the first fifteen entries into the reactor building, yielded preliminary data on the possible extent and range of hydrogen burn damage. These data, although sparse, contributed to development of a possible damage path and to an estimate of the extent of damage to susceptible reactor building items. Further information gathered from analysis of additional photographs and samples can provide the means for estimating hydrogen source and production rate data crucial to developing a complete understanding of the TMI-2 hydrogen deflagration. 34 figures

  3. A review of the probabilistic safety assessment application to the TR-2 research reactor

    International Nuclear Information System (INIS)

    Goektepe, G.; Adalioglu, U.; Anac, H.; Sevdik, B.; Menteseoglu, S.

    2001-01-01

    A review of the Probabilistic Safety Assessment (PSA) to the TR-2 Research Reactor is presented. The level 1 PSA application involved: selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantification, fault tree constructions and quantification, human reliability, component failure data base development, dependent failure analysis. Each of the steps of the analysis given above is reviewed briefly with highlights from the selected results. PSA application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. Insights gained from the application of PSA methodology to the TR-2 research reactor led to a significant safety review of the system

  4. Reactor building integrity testing: A novel approach at Gentilly 2 - principles and methodology

    International Nuclear Information System (INIS)

    Collins, N.; Lafreniere, P.

    1991-01-01

    In 1987, Hydro-Quebec embarked on an ambitious development program to provide the Gentilly 2 nuclear power station with an effective, yet practical reactor building Integrity Test. The Gentilly 2 Integrity Test employs an innovative approach based on the reference volume concept. It is identified as the Temperature Compensation Method (TCM) System. This configuration has been demonstrated at both high and low test pressure and has achieved extraordinary precision in the leak rate measurement. The Gentilly 2 design allows the Integrity Test to be performed at a nominal 3 kPa(g) test pressure during an (11) hour period with the reactor at full power. The reactor building Pressure Test by comparison, is typically performed at high pressure 124 kPa(g)) in a 7 day window during an annual outage. The Integrity Test was developed with the goal of demonstrating containment availability. Specifically it was purported to detect a leak or hole in the 'bottled-up' reactor building greater in magnitude than an equivalent pipe of 25 mm diameter. However it is considered feasible that the high precision of the Gentilly 2 TCM System Integrity Test and a stable reactor building leak characteristic will constitute sufficient grounds for the reduction of the Pressure Test frequency. It is noted that only the TCM System has, to this date, allowed a relevant determination of the reactor building leak rate at a nominal test pressure of 3 kPa(g). Classical method tests at low pressure have lead to inconclusive results due to the high lack of precision

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

    Science.gov (United States)

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

    2016-07-01

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

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

    International Nuclear Information System (INIS)

    Liu Wenjin; Zeng Zhongxiu; Ye Xianhui; Wu Wanjun

    2013-01-01

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

  7. Possible future roles for fast breeder reactors Part 1 and 2

    International Nuclear Information System (INIS)

    1978-06-01

    Part 1. The Fast Breeder Reactor (in particular in its sodium cooled version) has been steadily developed in the Community. This report attempts to quantify the advantages of this system in terms of fossil energy and uranium savings in the medium/long term as well as to examine some long term economic implications. The methodology of comparing scenarios, not individual reactor systems is followed. These scenarios have been chosen taking into account a range of assumptions concerning Community energy demand growth, fossil energy and uranium availability and technological capabilities. Part 2. The fast breeder reactor (FBR), particularly its sodium-cooled form (LMFBR) has been under development in the Community for many years. Industrial enterprises dedicated to its commercialisation have been formed and long range plans for its industrial utilisation are being formulated. The value of breeder reactors from the point of view of minimising Community fuel requirements has been discussed in Part I of this report (1). In Part II the consequences of delaying their introduction, and the demands placed upon the recycle industry by the introduction of fast reactors of different characteristics, using the Community electricity demand scenarios developed for Part I, are discussed. In addition comments are provided upon the effect of FBR introduction on the size of plutonium stocks

  8. Addition of soluble and insoluble neutron absorbers to the reactor coolant system of TMI-2

    International Nuclear Information System (INIS)

    Hansen, R.F.; Silverman, J.; Queen, S.P.; Ryan, R.F.; Austin, W.E.

    1984-07-01

    The physical and chemical properties of six elements were studied and combined with cost estimates to determine the feasibility of adding them to the TMI-2 reactor coolant to depress k/sub eff/ to less than or equal to 0.95. Both soluble and insoluble forms of the elements B, Cd, Gd, Li, Sm, and Eu were examined. Criticality calculations were performed by Oak Ridge National Laboratory to determine the absorber concentration required to meet the 0.95 k/sub eff/ criterion. The conclusion reached is that all elements with the exception of boron have overriding disadvantages which preclude their use in this reactor. Solubility experiments in the reactor coolant show that boron solubility is the same as that of boron in pure aqueous solutions of sodium hydroxide and boric acid; consequently, solubility is not a limiting factor in reaching the k/sub eff/ criterion. Examination of the effect of pH on sodium requirements and costs for processing to remove radionuclides revealed a sharp dependence; small decreases in pH lead to a large decrease in both sodium requirements and processing costs. Boron addition to meet any contemplated reactor safety requirements can be accomplished with existing equipment; however, this addition must be made with the reactor coolant system filled and pressurized to ensure uniform boron concentration

  9. Enhancements to the SLOWPOKE-2 nuclear research reactor at the Royal Military College of Canada

    Energy Technology Data Exchange (ETDEWEB)

    Hungler, P.C.; Andrews, M.T.; Weir, R.D.; Nielson, K.S.; Chan, P.K.; Bennett, L.G.I., E-mail: paul.hungler@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)

    2014-07-01

    In 1985 a Safe Low Power C(K)ritical Experiment (SLOWPOKE) nuclear research reactor was installed at the Royal Military College of Canada (RMCC). The reactor at nominally 20 kW thermal was named SLOWPOKE-2 and the core was designed to have a total of 198 fuel pins with Low Enriched Uranium (LEU) fuel (19.89% U-235). Installation of the reactor was intended to provide an education tool for members of the Canadian Armed Forces (CAF) and an affordable neutron source for the application of neutron activation analysis (NAA) and radioisotope production. Today, the SLOWPOKE-2 at RMCC continues to be a key education tool for undergraduate and post-graduate students and successfully conducts NAA and isotope production as per its original design intent. RMCC has significantly upgraded the facility and instruments to develop capabilities such as delayed neutron and gamma counting (DNGC) and neutron imaging, including 2D thermal neutron radiography and 3D thermal neutron tomography. These unique nuclear capabilities have been applied to relevant issues in the CAF. The analog control system originally installed in 1985 has been removed and replaced in 2001 by the SLOWPOKE Integrated Reactor Control and Instrumentation System (SIRCIS) which is a digital controller. This control system continues to evolve with SIRCIS V2 currently in operation. The continual enhancement of the facility, instruments and systems at the SLOWPOKE-2 at RMCC will be discussed, including an update on RMCC's refueling plan. (author)

  10. Enhancements to the SLOWPOKE-2 nuclear research reactor at the Royal Military College of Canada

    International Nuclear Information System (INIS)

    Hungler, P.C.; Andrews, M.T.; Weir, R.D.; Nielson, K.S.; Chan, P.K.; Bennett, L.G.I.

    2014-01-01

    In 1985 a Safe Low Power C(K)ritical Experiment (SLOWPOKE) nuclear research reactor was installed at the Royal Military College of Canada (RMCC). The reactor at nominally 20 kW thermal was named SLOWPOKE-2 and the core was designed to have a total of 198 fuel pins with Low Enriched Uranium (LEU) fuel (19.89% U-235). Installation of the reactor was intended to provide an education tool for members of the Canadian Armed Forces (CAF) and an affordable neutron source for the application of neutron activation analysis (NAA) and radioisotope production. Today, the SLOWPOKE-2 at RMCC continues to be a key education tool for undergraduate and post-graduate students and successfully conducts NAA and isotope production as per its original design intent. RMCC has significantly upgraded the facility and instruments to develop capabilities such as delayed neutron and gamma counting (DNGC) and neutron imaging, including 2D thermal neutron radiography and 3D thermal neutron tomography. These unique nuclear capabilities have been applied to relevant issues in the CAF. The analog control system originally installed in 1985 has been removed and replaced in 2001 by the SLOWPOKE Integrated Reactor Control and Instrumentation System (SIRCIS) which is a digital controller. This control system continues to evolve with SIRCIS V2 currently in operation. The continual enhancement of the facility, instruments and systems at the SLOWPOKE-2 at RMCC will be discussed, including an update on RMCC's refueling plan. (author)

  11. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 2. User's manual

    International Nuclear Information System (INIS)

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

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear energy reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 2: User's Manual) describes the input requirements of VIPRE and its auxiliary programs, SPECSET, ASP and DECCON, and lists the input instructions for each code

  12. Calculation of the effectiveness of manual control rods for the reactor of Ignalina NPP Unit 2

    International Nuclear Information System (INIS)

    Bubelis, E.; Pabarcius, R.

    2001-01-01

    On the basis of one of the recent databases of the reactor of Ignalina NPP Unit 2, calculations of the effectiveness of separate manual control rods, groups of manual control rods and axial characteristic of effectiveness of separate manual control rods were performed. The results of the calculations indicated, that all analyzed separate manual control rods have approximately the same effectiveness, which doesn't depend on the location of a control rod in the reactor core layout Manual control rod of the new design has about 10% greater effectiveness than manual control rod of the old design. (author)

  13. Fusion reactor materials program plan. Section 2. Damage analysis and fundamental studies

    International Nuclear Information System (INIS)

    1978-07-01

    The scope of this program includes: (1) Development of procedures for characterizing neutron environments of test facilities and fusion reactors, (2) Theoretical and experimental investigations of the influence of irradiation environment on damage production, damage microstructure evolution, and mechanical and physical property changes, (3) Identification and, where appropriate, development of essential nuclear and materials data, and (4) Development of a methodology, based on damage mechanisms, for correlating the mechanical behavior of materials exposed to diverse test environments and projecting this behavior to magnetic fusion reactor (MFR) environments. Some major problem areas are addressed

  14. The performance of ENDF/B-V.2 nuclear data for fast reactor calculations

    International Nuclear Information System (INIS)

    Atkinson, C.A.; Collins, P.J.

    1987-01-01

    Calculations with ENDF/B-V.2 data have been made for twenty-five fast-spectrum integral assemblies covering a wide range of sizes and compositions. Analysis was done by transport codes with refined cross section processing methods and detailed reactor modelling. The predictions of fission rate distributions and control rod worths were emphasized for the more prototypic benchmark cores. The results show considerable improvements in agreement with experiment compared with analysis using ENDF/B-IV data, but it is apparent that significant errors remain for fast reactor design calculations

  15. Review of the accident source terms for aluminide fuel: Application to the BR2 reactor

    International Nuclear Information System (INIS)

    Joppen, F.

    2005-01-01

    A major safety review of the BR2, a material test reactor, is to be conducted for the year 2006. One of the subjects selected for the safety review is the definition of source terms for emergency planning and in particular the development of accident scenarios. For nuclear power plants the behaviour of fuel under accident conditions is a well studied object. In case of non-power reactors this basic knowledge is rather scarce. The usefulness of information from power plant fuels is limited due to the differences in fuel type, power level and thermohydraulical conditions. First investigation indicates that using data from power plant fuel leads to an overestimation of the source terms. Further research on this subject could be very useful for the research reactor community, in order to define more realistic source terms and to improve the emergency preparedness. (author)

  16. RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

    International Nuclear Information System (INIS)

    Kawai, H.

    2001-01-01

    JAPC purchased RETRAN, a program for transient thermal hydraulic analysis of complex fluid flow system, from the U.S. Electric Power Research Institute in 1992. Since then, JAPC has been utilizing RETRAN to evaluate safety margins of actual plant operation, in coping with troubles (investigating trouble causes and establishing countermeasures), and supporting reactor operation (reviewing operational procedures etc.). In this paper, a result of plant analysis performed on a CVCS reactor primary coolant leakage incident which occurred at JAPC's Tsuruga-2 plant (4-loop PWR, 3423 MWt, 1160 MW) on July 12 of 1999 and, based on the result, we made a plan to modify our operational procedure for reactor primary coolant leakage events in order to make earlier plant shutdown and this reduced primary coolant leakage. (author)

  17. Outline of construction planning on No. 2 Reactor of the Shika Nuclear Power Plant

    International Nuclear Information System (INIS)

    Nakagawa, Tetsuro; Kadoki, Shuichi; Kubo, Tetsuji

    1999-01-01

    The Hokuriku Electric Co., Ltd. carries out the expansion of the Shika Nuclear Power Plant No.2 (ABWR) to start its in March 2006. It is situated in north neighboring side of No. 1 reactor under operation at present, and its main buildings are planned to position a reactor building at mountain side and a turbine building at sea side as well as those in the No. 1 reactor. And, cooling water for steam condenser was taken in from an intake opening built at north side of the lifting space situated at the front of the power plant, and discharged into seawater from a flashing opening positioned about 600 m offing. Here were described on outline of main civil engineering such as base excavation engineering, concrete caisson production, oceanic establishment engineering, and facility for steam condenser, and characteristics of the engineering. (G.K.)

  18. Modernization of turbine control system and reactor control system in Almaraz 1 and 2

    International Nuclear Information System (INIS)

    Pulido, C.; Diez, J.; Carrasco, J. A.; Lopez, L.

    2005-01-01

    The replacement of the Turbine Control System and Reactor Control System are part of the Almaraz modernization program for the Instrumentation and Control. For these upgrades Almaraz has selected the Ovation Platform that provides open architecture and easy expansion to other systems, these platforms is highly used in many nuclear and thermal plants around the world. One of the main objective for this project were to minimize the impact on the installation and operation of the plant, for that reason the project is implemented in two phases, Turbine Control upgrade and Reactor Control upgrade. Another important objective was to increase the reliability of the control system making them fully fault tolerant to single failures. The turbine Control System has been installed in Units 1 and 2 while the Reactor Control System will be installed in 2006 and 2007 outages. (Author)

  19. Decontamination and decommissioning project of the TRIGA Mark-2 and 3 research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jung, K J; Baik, S T; Chung, U S; Jung, K H; Park, S K; Lee, B J; Kim, J K; Yang, S H

    2000-01-01

    During the review on the decommissioning plan and environmental impact assessment report by the KINS, the number of the inquired items were two hundred and fifty one, and the answers were made and sent until September 10, 1999, as the screened review results were reported to Ministry of Science and Technology(MOST) in December 14, 1999, all the reviews on the licence were over. Radioactive liquid wastes of 400 tons generated during the operation of the research reactors including reactor vessels are stored in the facility of the research reactor 1 and 2. Those liquid wastes have the low-level-radioactivity which can be discharged to the surroundings, but was wholly treated to be vaporized naturally by means of the increased numbers of the natural vaporization disposal facilities with the annual capacity of 200 tons for the purpose of the minimized environmental contamination.

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

    Science.gov (United States)

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

    2017-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Sarah DURY

    2010-01-01

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

  2. 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.

  3. 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.

  4. Irradiation experiments of recycled PuO2-UO2 fuels by SAXTON reactor, (1)

    International Nuclear Information System (INIS)

    Yumoto, Ryozo; Akutsu, Hideo

    1975-01-01

    Seventy two mixed oxide fuel rods made by PNC were irradiated in Saxton Core 3. This paper generally describes the fuel specifications, the power history of the fuel and the post-irradiation examination of the PNC fuel. The specifications of the 4.0 w/o and 5.0 w/o enriched PuO 2 fuel rods with zircaloy-4 cladding are presented in a table and a figure. The positions of PNC fuel rods in the Saxton reactor are shown in a figure. Sixty eight 5.0 w/o PuO 2 -UO 2 fuel rods were assembled in a 9 x 9 rod array together with zircaloy-4 bars, a flux thimble, and a Sb-Be source. The power history of the Saxton Core 3 and the irradiation history of the PNC fuel rods are summarized in tables. The peak power and burnup of each fuel rod and the axial power profile are also presented. The maximum linear power rate and burnup attained were 512W/cm and 8700 MWD/T, respectively. As for the post irradiation examination, the items of nondestructive test, destructive test, and cladding test are presented together with the working flow diagram of the examination. It is concluded that the performance of all fuel rods was safe and satisfactory throughout the power history. (Aoki, K.)

  5. The 5th surveillance testing for Kori unit 2 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules V, R, P, T and N are 2.837E+18, 1.105E+19, 2.110E+19, 3.705E+19 and 4.831E+19n/cm{sup 2}, respectively. The bias factor, the ratio of measurement/calculation, was 0.918 for the 1st through 5th testing and the calculational uncertainty, 11.6% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.898E+19n/cm{sup 2} based on the end of 15th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 4.203E+19, 5.232E+19, 6.262E+19 and 7.291E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 49 refs., 35 figs., 48 tabs. (Author)

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

    Science.gov (United States)

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

    2014-01-01

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

  7. Thermal limits validation of gamma thermometer power adaption in CFE Laguna Verde 2 reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas V, G.; Banfield, J. [GE-Hitachi Nuclear Energy Americas LLC, Global Nuclear Fuel, Americas LLC, 3901 Castle Hayne Road, Wilmingtonm, North Carolina (United States); Avila N, A., E-mail: Gabriel.Cuevas-Vivas@ge.com [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla Km 42.5, Alto Lucero, Veracruz (Mexico)

    2016-09-15

    This paper presents the status of GEH work on Gamma Thermometer (GT) validation using the signals of the instruments installed in the Laguna Verde Unit 2 reactor core. The long-standing technical collaboration between Comision Federal de Electricidad (CFE), Global Nuclear Fuel - Americas LLC (GNF) and GE-Hitachi Nuclear Energy Americas LLC (GEH) is moving forward with solid steps to a final implementation of GTs in a nuclear reactor core. Each GT is integrated into a slightly modified Local Power Range Monitor (LPRM) assembly. Six instrumentation strings are equipped with two gamma field detectors for a total of twenty-four bundles whose calculated powers are adapted to the instrumentation readings in addition to their use as calibration instruments for LPRMs. Since November 2007, the six GT instrumentation strings have been operable with almost no degradation by the strong neutron and gamma fluxes in the Laguna Verde Unit 2 reactor core. In this paper, the thermal limits, Critical Power Ratio (CPR) and maximum Linear Heat Generation Rate (LHGR), of bundles directly monitored by either Traverse In-core Probes (TIPs) or GTs are used to establish validation results that confirm the viability of TIP system replacement with automatic fixed in-core probes (AFIPs, GTs, in a Boiling Water Reactor. The new GNF steady-state reactor core simulator AETNA02 is used to obtain power and exposure distribution. Using this code with an updated methodology for GT power adaption, a reduced value of the GT interpolation uncertainty is obtained that is fed into the LHGR calculation. This new method achieves margin recovery for the adapted thermal limits for use in the Economic Simplified Boiling Water Reactor (ESBWR) or any other BWR in the future that employs a GT based AFIP system for local power measurements. (Author)

  8. Thermal limits validation of gamma thermometer power adaption in CFE Laguna Verde 2 reactor core

    International Nuclear Information System (INIS)

    Cuevas V, G.; Banfield, J.; Avila N, A.

    2016-09-01

    This paper presents the status of GEH work on Gamma Thermometer (GT) validation using the signals of the instruments installed in the Laguna Verde Unit 2 reactor core. The long-standing technical collaboration between Comision Federal de Electricidad (CFE), Global Nuclear Fuel - Americas LLC (GNF) and GE-Hitachi Nuclear Energy Americas LLC (GEH) is moving forward with solid steps to a final implementation of GTs in a nuclear reactor core. Each GT is integrated into a slightly modified Local Power Range Monitor (LPRM) assembly. Six instrumentation strings are equipped with two gamma field detectors for a total of twenty-four bundles whose calculated powers are adapted to the instrumentation readings in addition to their use as calibration instruments for LPRMs. Since November 2007, the six GT instrumentation strings have been operable with almost no degradation by the strong neutron and gamma fluxes in the Laguna Verde Unit 2 reactor core. In this paper, the thermal limits, Critical Power Ratio (CPR) and maximum Linear Heat Generation Rate (LHGR), of bundles directly monitored by either Traverse In-core Probes (TIPs) or GTs are used to establish validation results that confirm the viability of TIP system replacement with automatic fixed in-core probes (AFIPs, GTs, in a Boiling Water Reactor. The new GNF steady-state reactor core simulator AETNA02 is used to obtain power and exposure distribution. Using this code with an updated methodology for GT power adaption, a reduced value of the GT interpolation uncertainty is obtained that is fed into the LHGR calculation. This new method achieves margin recovery for the adapted thermal limits for use in the Economic Simplified Boiling Water Reactor (ESBWR) or any other BWR in the future that employs a GT based AFIP system for local power measurements. (Author)

  9. Pulsed irradiation of enriched UO{sub 2} in the Annular Core Pulse Reactor (ACPR)

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, T R; Lucoff, D M; Reil, K O; Croucher, D W [Sandia Laboratories (United States)

    1974-07-01

    A series of experiments have been conducted in the Annular Core Pulse Reactor (ACPR) to determine the energy deposition and behavior of enriched UO{sub 2} under pulse conditions. In the experiment single unirradiated pellets with enrichments up to 25 percent were pulse heated to melt temperatures. Temperature and fission product inventory measurements were made and compared with neutron transport calculations. (author)

  10. Ethanol production by immobilized yeast and its CO2 gas effects on a packed bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cho, G M; Choi, C Y; Choi, Y D; Han, M H

    1982-10-01

    Immobilised yeast trapped in an alginate matrix demonstrated maximum activity at 30 degrees C and showed no pH effect between 3 and 7. Substrate inhibition was observed at glucose concentrations above 8% but the immobilised cells retained 70% of their maximum activity at 20% glucose concentration. The operation stability of immobilised cells was lower in simple glucose solution than in the activation medium in which only 20% of the activity was lost after 10 days operation. Inactivated immobilised yeast beads were reactivated by incubation in activation medium without a significant increase in cell numbers in a bead. During the operation of the immobilised yeast in a packed bed reactor, CO/sub 2/ gas accumulation adversely affected the reactor performance. An ideal plus flow reactor, not taking into account the formation of CO/sub 2/ gas bubbles and the presence of mass trasnfer resistance, was simulated using a kinetic model for the production of ethanol and the simulation results were compared with the actual reactor performance to determine the CO/sub 2/ gas effect, quantitatively. Up to 45% of the substrate conversion was lost due to the accumulation of CO/sub 2/ gas bubbles in all cases. (Refs. 21).

  11. Formation of the high-spin Hf-179m2 isomer in reactor irradiations

    Czech Academy of Sciences Publication Activity Database

    Karamian, S. A.; Carroll, J. J.; Adam, Jindřich; Kulagin, EN.; Shabalin, EP.

    2004-01-01

    Roč. 14, č. 4 (2004), s. 438-441 ISSN 1054-660X R&D Projects: GA MŠk(CZ) ME 134 Keywords : reactor irradiation * high-spin Hf-179m2 Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 0.836, year: 2004

  12. The new fourier diffractometer at the IBR-2 reactor: design and first results

    International Nuclear Information System (INIS)

    Aksenov, V.L.; Balagurov, A.M.; Simkin, V.G.

    1992-01-01

    A high-resolution neutron powder Fourier diffractometer has been constructed at the pulsed reactor IBR-2 in Dubna. Preliminary results have confirmed resolution predictions indicating that Δd/d can be as small as 5x10 -4 over a wide d range with a good data accumulation rate. 3 refs.; 10 figs.; 1 tab

  13. Material properties of Bohunice 1 and 2 reactor pressure vessel materials before and after annealing

    International Nuclear Information System (INIS)

    Brumovsky, M.; Novosad, P.; Vacek, M.

    1994-01-01

    Six types of experimental RPV materials were studied before and after irradiation in host nuclear power and research reactors. Recovery of RPV materials from radiation hardening and embrittlement after annealing was evaluated including a rate of radiation damage after reirradiation used. (author). 3 refs, 4 figs, 2 tabs

  14. Recent Development of Radioanalytical Methods at the IBR-2 Pulsed Fast Reactor

    International Nuclear Information System (INIS)

    Nazarov, V.M.; Peresedov, V.F.

    1994-01-01

    Experience in the application of radioanalytical methods, including NAA, at the IBR-2 pulsed fast reactor is reviewed. Details of the instruments dedicated to neutron activation analysis and radiography studies are reported. Applications of resonance neutrons to environmental monitoring and to the investigation of high-purity materials, are examplified. 15 refs. 9 figs., 9 tabs

  15. Experimental determination of lattice parameters for 2% enriched uranium heavy water reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N; Takac, S; Markovic, H; Bosevski, T [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1963-04-15

    Systematic measurements of the buckling, infinite multiplication factor and the thermal utilization factor were made on a series of lattices for 2% enriched uranium tubular fuel elements in heavy water. This work represents the first phase of experimental verification of standard theoretical methods used for the determination of reactor parameters.

  16. Characterization of fuel distributions in the Three-Mile Island Unit 2 (TMI-2) reactor system by neutron and gamma-ray dosimetry

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McNeece, J.P.; Kaiser, B.J.; McElroy, W.N.

    1984-04-01

    The resolution of technical issues generated by the accident at Three-Mile Island Unit 2 (TMI-2) will inevitably be of long range benefit. Determination of the fuel debris dispersal in the TMI-2 reactor system represents a major technical issue. In reactor recovery operations, such as for the safe handling and final disposal of TMI-2 waste, quantitative fuel assessments are being conducted throughout the reactor core and primary coolant system

  17. Estimation of chemical carcass composition from 8th rib characteristics with Belgian blue double-muscled bulls.

    Science.gov (United States)

    De Campeneere, S; Fiems, L O; Van de Voorde, G; Vanacker, J M; Boucque, C V; Demeyer, D I

    1999-01-01

    Characteristics from the 8th rib cut: chemical composition, tissue composition after dissection, specific gravity (SG) and m. longissimus thoracis (LT) composition, collected on 17 Belgian Blue double-muscled fattening bulls were used to generate equations for predicting chemical carcass composition. Carcass composition was best predicted from chemical analysis of the 8th rib cut and the empty body weight (EBW) of the bull. Carcass chemical fat content (CCF, kg) was predicted from the 8th rib cut fat content (ether extract, 8RF, kg) by the following regression: CCF=1.94+27.37 8RF (R(2)=0.957, RSD =9.89%). A higher coefficient was found for carcass water (CCW, kg) predicted from 8RF and EBW: CCW=-2.26+0.28 EBW-34.28 8RF (R(2)=0.997, RSD=1.48%). No parameter was found to improve the prediction of CCP from EBW solely: CCP=-0.86+0.08 EBW (R(2) =0.992, RSD=2.61%). Prediction equations based solely on LT composition had low R(2) values of between 0.38 and 0.67, whereas no significant equations were found using SG. However, equations based on EBW had R(2) values between 0.78 and 0.99. Chemical components of the 8th rib cut in combination with EBW are most useful in predicting the chemical composition of the carcass of Belgian-Blue double-muscled bulls.

  18. Micro-scale H2-CO2 dynamics in a hydrogenotrophic methanogenic membrane reactor

    Directory of Open Access Journals (Sweden)

    Emilio Garcia-Robledo

    2016-08-01

    Full Text Available Biogas production is a key factor in a sustainable energy supply. It is possible to get biogas with very high methane content if the biogas reactors are supplied with exogenous hydrogen, and one of the technologies for supplying hydrogen is through gas permeable membranes. In this study the activity and stratification of hydrogen consumption above such a membrane was investigated by use of microsensors for hydrogen and pH. A hydrogenotrophic methanogenic community that was able to consume the hydrogen flux within 0.5 mm of the membrane with specific rates of up to 30 m3 H2 m-3 day-1 developed within 3 days in fresh manure and was already established at time zero when analyzing slurry from a biogas plant. The hydrogen consumption was dependent on a simultaneous carbon dioxide supply and was inhibited when carbon dioxide depletion elevated the pH to 9.2. The activity was only partially restored when the carbon dioxide supply was resumed. Bioreactors supplied with hydrogen gas should thus be carefully monitored and either have the hydrogen supply disrupted or be supplemented with carbon dioxide when the pH rises to values about 9.

  19. Passive Decay Heat Removal System Options for S-CO2 Cooled Micro Modular Reactor

    International Nuclear Information System (INIS)

    Moon, Jangsik; Jeong, Yong Hoon; Lee, Jeong Ik

    2014-01-01

    To achieve modularization of whole reactor system, Micro Modular Reactor (MMR) which has been being developed in KAIST took S-CO 2 Brayton power cycle. The S-CO 2 power cycle is suitable for SMR due to high cycle efficiency, simple layout, small turbine and small heat exchanger. These characteristics of S-CO 2 power cycle enable modular reactor system and make reduced system size. The reduced size and modular system motived MMR to have mobility by large trailer. Due to minimized on-site construction by modular system, MMR can be deployed in any electricity demand, even in isolated area. To achieve the objective, fully passive safety systems of MMR were designed to have high reliability when any offsite power is unavailable. In this research, the basic concept about MMR and Passive Decay Heat Removal (PDHR) system options for MMR are presented. LOCA, LOFA, LOHS and SBO are considered as DBAs of MMR. To cope with the DBAs, passive decay heat removal system is designed. Water cooled PDHR system shows simple layout, but has CCF with reactor systems and cannot cover all DBAs. On the other hand, air cooled PDHR system with two-phase closed thermosyphon shows high reliability due to minimized CCF and is able to cope with all DBAs. Therefore, the PDHR system of MMR will follows the air-cooled PDHR system and the air cooled system will be explored

  20. On-line reactor building integrity testing at Gentilly-2 (summary of results 1987-1994)

    International Nuclear Information System (INIS)

    Collins, N.; Lafreniere, P.

    1994-01-01

    In 1987, Hydro-0uebec embarked on an ambitious development program to provide the Gentilly-2 Nuclear Power Station with an effective and practical Reactor Building Containment integrity Test (CIT). In October 1992, the inaugural low pressure (3 kPa(g) nominal) CIT at 100% F.P was performed. The test was conclusive and the CIT was declared In-Service for containment integrity verification on-line. Five subsequent CITs performed in 1993 and 1994 have demonstrated the expected leak rate results and good reliability. The outstanding feature of the CITs is the demonstrated accurary of better than 5% of the measured leak rate. The CIT was developed with the primary goal of demonstrating 'overall' containment availability. Specifically it was designed to detect a 25 mm. diameter leak or hole in the Reactor Building. However, the remarkable CIT accuracy allows reliable detection of a 2 mm. hole. The Gentilly-2 CIT is an innovative approach based on the Temperature Compensation Method (TCM) which uses a reference volume composed of an extensive tubular network of several different diameters. This eliminates the need to track numerous temperature points. A second independent tubular network includes numerous humidity sampling points, thereby enabling the mearurernent of minute pressure variations inside the Reactor Building, independant of the spatial and temporal humidity behaviour. This Gentilly-2 TOM System has been demonstrated to work at both high and low test pressures. The GentiIly-2 design allows the CIT to be performed at a nominal 3 kPa(g) test pressure during a 12-hour period (28 hours total with alignment time) with the reactor at full power. The traditional Reactor Building Pressure Test (RBPT) is typically performed at high pressure (124 kPa(g) in a 5-day critical path window (7 days total with alignment time) during an annual shutdown

  1. Performance analysis of photocatalytic CO2 reduction in optical fiber monolith reactor with multiple inverse lights

    International Nuclear Information System (INIS)

    Yuan, Kai; Yang, Lijun; Du, Xiaoze; Yang, Yongping

    2014-01-01

    Highlights: • A new optical fiber monolith reactor model for CO 2 reduction was developed. • Methanol concentration versus fiber location and operation parameters was obtained. • Reaction efficiency increases by 31.1% due to the four fibers and inverse layout. • With increasing space of fiber and channel center, methanol concentration increases. • Methanol concentration increases as the vapor ratio and light intensity increase. - Abstract: Photocatalytic CO 2 reduction seems potential to mitigate greenhouse gas emissions and produce renewable energy. A new model of photocatalytic CO 2 reduction in optical fiber monolith reactor with multiple inverse lights was developed in this study to improve the conversion of CO 2 to CH 3 OH. The new light distribution equation was derived, by which the light distribution was modeled and analyzed. The variations of CH 3 OH concentration with the fiber location and operation parameters were obtained by means of numerical simulation. The results show that the outlet CH 3 OH concentration is 31.1% higher than the previous model, which is attributed to the four fibers and inverse layout. With the increase of the distance between the fiber and the monolith center, the average CH 3 OH concentration increases. The average CH 3 OH concentration also rises as the light input and water vapor percentage increase, but declines with increasing the inlet velocity. The maximum conversion rate and quantum efficiency in the model are 0.235 μmol g −1 h −1 and 0.0177% respectively, both higher than previous internally illuminated monolith reactor (0.16 μmol g −1 h −1 and 0.012%). The optical fiber monolith reactor layout with multiple inverse lights is recommended in the design of photocatalytic reactor of CO 2 reduction

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  3. Microspheres of UO2, ThO2 and PuO2 for the high temperature reactor

    International Nuclear Information System (INIS)

    Brandau, E.

    2002-01-01

    The production of high temperature reactor fuel, so called pebble fuel, was done in the eighties by a special vibrational dropping process to obtain as sintered UO 2 - or ThO 2 -microspheres, so called 'Kernels', with a diameter size of about 300 μm. These microspheres have been coated and embedded in carbon balls to get the pebble fuel. Since the early nineties BRACE is developing the processings of microspheres starting with sols and suspensions to produce Al 2 O 3 , ZrO 2 , HfO 2 and Actinide oxide microspheres. Two main developments have been made: 1) the preparation of the feed solution (sol, suspension) and the solidification processing, and 2) the equipment, design, and electronic control have been completely changed. A newly developed suspension process for actinide oxides and for metal oxides e.g. Al 2 O 3 , TiO 2 , SiO 2 , ZrO 2 , HfO 2 , CeO 2 , ThO 2 , UO 2 , PuO 2 leads to cheaper production of as sintered microspheres. The processing and the installations will be described and the experience of production will be shown. (author)

  4. Modifications to ORIGEN2 for generating N Reactor source terms. Volume 3: ORIGEN2 N-Reactor output files

    International Nuclear Information System (INIS)

    1997-01-01

    This text is intended to be a brief outline of the ORIGEN2 computer code which is a revised and updated version of the ORIGEN documented in report ORNL-4628 (May 1973). Included here are: a brief description of the functions of ORIGEN2; a listing of the major data sources; a listing of the published documentation concerning ORIGEN2; and an outline of the ORIGEN2 output organization. ORIGEN2 is available from the ORNL Radiation Shielding Information Center (RSIC). Past experience has indicated that many users encounter considerable difficulty in finding the desired information in a ORIGEN2 output which is sometimes rather massive. This section is intended as a brief outline of the organization of ORIGEN2 output

  5. A novel condensation reactor for efficient CO2 to methanol conversion for storage of renewable electric energy

    NARCIS (Netherlands)

    Bos, Martin Johan; Brilman, Derk Willem Frederik

    2015-01-01

    A novel reactor design for the conversion of CO2 and H2 to methanol is developed. The conversion limitations because of thermodynamic equilibrium are bypassed via in situ condensation of a water/methanol mixture. Two temperatures zones inside the reactor ensure optimal catalyst activity (high

  6. Independent CO2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, Task 2.50.05

    International Nuclear Information System (INIS)

    Stojic, M.; Pavicevic, M.

    1964-01-01

    This report contains the following volumes V and VI of the Project 'Independent CO 2 loop for cooling the samples irradiated in RA reactor vertical experimental channels': Design project of the dosimetry control system in the independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, and Safety report for the Independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels [sr

  7. Status and perspective of development of cold moderators at the IBR-2 reactor

    International Nuclear Information System (INIS)

    Kulikov, S; Shabalin, E

    2012-01-01

    The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams nos. 7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams nos. 2-3 and for beams nos. 1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (∼3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

  8. Status and perspective of development of cold moderators at the IBR-2 reactor

    Science.gov (United States)

    Kulikov, S.; Shabalin, E.

    2012-03-01

    The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams #7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams #2-3 and for beams #1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (~3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

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

    International Nuclear Information System (INIS)

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

    1996-09-01

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

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

    OpenAIRE

    Neyens, Evy; Smits, Tim

    2015-01-01

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

  11. 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.

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

    OpenAIRE

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

    2015-01-01

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

  13. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

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

  14. Report on the operation in 1973 of the FR 2 research reactor

    International Nuclear Information System (INIS)

    Moeller, I.; Steiger, W.

    1975-04-01

    Also in 1973, the heavy-water moderated research and testing reactor FR 2 was operated to schedule at 44 MW nominal power. Again, the availability of the plant was slightly improved. Experimental utilization through instrumented irradiation capsules strongly increased as compared to the previous year. Up to 16 capsule test rigs at a time were inserted in the reactor. As to the beam tube experiments, up to 13 experiments covering a total of 18 test rigs were conducted simultaneously at the 12 reasonably usable beam holes. At the beginning of the year all of the positions available were occupied by 5 loop experiments. Isotope production reached its highest value with a total of 2,372 irradiated capsules (1.3% more than the year before). Some remarkable figures characterized the year 1973: On August 16, 1973 ten years of full power operation at a nominal power of 12 and 44 MW, respectively, had been reached. On July 24, 1973 the 50,000th isotope irradiation was performed in the reactor and on December 26, 1973 a total energy release of 100,000 MWd was recorded. Moreover, the 125,000th visitor of the reactor was welcomed on December 6, 1973. (orig./UA) [de

  15. Increased SRP reactor power

    International Nuclear Information System (INIS)

    MacAfee, I.M.

    1983-01-01

    Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

  16. Gas reactor in-pile safety test project (GRIST-2)

    International Nuclear Information System (INIS)

    Kelley, A.P. Jr.; Arbtin, E.; St Pierre, R.

    1979-01-01

    Although out-of-pile tests may be expected to confirm individual phenomena models in core disruptive accident analysis codes, only in-pile tests are capable of verifying the extremely complex integrated model effects within the appropriate time phase for these accidents. For this reason, the GRIST-2 project, the purpose of which is to design and construct an in-pile helium loop capable of transient safety testing in the TREAT facility in Idaho, forms a cornerstone of the US GCFR safety program. The project organization, experiment program, facility, helium system design, and schedule which have been selected to meet the objectives are described

  17. Sintering of Kernel UO2 for High Temperature Reactor Fuel

    International Nuclear Information System (INIS)

    Sukarsono; Dwi-Heru-Sucahyo; Hidayati; Evi-Hertiviana; Bambang-Sugeng

    2000-01-01

    Sintering investigation of UO 2 gel has been done. The gel was preparedthrough two ways. The first, gel was produced using PVA as additive agent.The second gel was produced using HMTA and Urea as additive agent. From thepreparation of gel, the PVA method better than the urea - HMTA method,because was not necessary the cold temperature for sol preparation and alsowas not necessary the hot temperature for gelation process. After nextprocessing, the sintered gel of gel through PVA, also better than HMTAprocess. (author)

  18. Kinetics of Ar+*(2G9/2) metastable ions and transport of argon ions in ICP reactor

    NARCIS (Netherlands)

    Sadeghi, N.; Derouard, J.; Grift, van de M.; Kroesen, G.M.W.; Hoog, de F.J.; Tachibana, K.; Watanabe, Y.

    1997-01-01

    The decay time of the argon Ar~~(2G912) metastable ions was measured in the afterglow of a low pressure pulsed helicon reactor. From the argon pressure and electron density dependence of this decay time, rate coefficients for quenching of these ions by argon atoms and by plasma electrons have been

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

    Science.gov (United States)

    Parys, Nathalie

    2013-12-01

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

  20. Application of 2DOF controller for reactor power control. Verification by numerical simulation

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Suzuki, Katsuo

    1996-09-01

    In this report the usefulness of the two degree of freedom (2DOF) control is discussed to improve the reference response characteristics and robustness for reactor power control system. The 2DOF controller consists of feedforward and feedback elements. The feedforward element was designed by model matching method and the feedback element by solving the mixed sensitivity problem of H ∞ control. The 2DOF control gives good performance in both reference response and robustness to disturbance and plant perturbation. The simulation of reactor power control was performed by digitizing the 2DOF controller with the digital control periods of 10[msec]. It is found that the control period of 10[msec] is enough not to make degradation of the control performance by digitizing. (author)

  1. Exxon nuclear neutronics design methods for pressurized water reactors. Supplement 2

    International Nuclear Information System (INIS)

    Skogen, F.B.; Stout, R.B.

    1977-01-01

    Modifications to the Exxon Nuclear PWR neutronic design calculational methods are presented as well as the results obtained when these improved methods are compared to reactor measurements. The basic PWR design tools remain unchanged; i.e., the XPOSE code is used for generating the basic nuclear parameters, the PDQ-7 code is used for calculating reactivity and x-y power distributions, and the XTG code is used for three-dimensional analysis. The recent start-up experiences at D. C. Cook Unit 1 and H. B. Robinson Unit 2 have provided a significant increase in the data base supporting the current ENC PWR neutronic methods. The verification comparisons contained in the supplement include reactor measurements from D. C. Cook Unit 1, Cycle 2; H. B. Robinson Unit 2, Cycles 4 and 5; Palisades Cycle 2, and R. E. Ginna, Cycle 7

  2. Core design calculations of IRIS reactor using modified CORD-2 code package

    International Nuclear Information System (INIS)

    Pevec, D.; Grgic, D.; Jecmenica, R.; Petrovic, B.

    2002-01-01

    Core design calculations, with thermal-hydraulic feedback, for the first cycle of the IRIS reactor were performed using the modified CORD-2 code package. WIMSD-5B code is applied for cell and cluster calculations with two different 69-group data libraries (ENDF/BVI rev. 5 and JEF-2.2), while the nodal code GNOMER is used for diffusion calculations. The objective of the calculation was to address basic core design problems for innovative reactors with long fuel cycle. The results were compared to our results obtained with CORD-2 before the modification and to preliminary results obtained with CASMO code for a similar problem without thermal-hydraulic feedback.(author)

  3. Mirror Advanced Reactor Study (MARS). Final report. Volume 2. Commercial fusion synfuels plant

    International Nuclear Information System (INIS)

    Donohue, M.L.; Price, M.E.

    1984-07-01

    Volume 2 contains the following chapters: (1) synfuels; (2) physics base and parameters for TMR; (3) high-temperature two-temperature-zone blanket system for synfuel application; (4) thermochemical hydrogen processes; (5) interfacing the sulfur-iodine cycle; (6) interfacing the reactor with the thermochemical process; (7) tritium control in the blanket system; (8) the sulfur trioxide fluidized-bed composer; (9) preliminary cost estimates; and (10) fuels beyond hydrogen

  4. Optimization of binary breeder reactor. 2. Preliminary base for control analysis and fuel management

    International Nuclear Information System (INIS)

    Dias, A.F.; Nascimento, J.A. do; Ishiguro, Y.

    1985-01-01

    Neutronic calculations to verify the reactivity effects, of sodium voids and Doppler, with the variation of the composition of parasitic absorbers were done. A LMFBR type reactor loaded with mixed fuel, (U 233 -Th 232 )O 2 in the internal core and (U 238 -Pu 239 )O 2 in external core, was considered. In reactivity calculations the EXPANDA and CITATION computer codes were utilized. Buckling effects and importance of determination of the spatial selfshielding factors were analysed. (M.C.K.) [pt

  5. Investigation of a weld defect, reactor vessel head Ringhals 2

    International Nuclear Information System (INIS)

    Embring, G.; Pers-Anderson, E.B.

    1994-01-01

    During the summer-outage 1993 Ringhals unit 2 vessel head was inspected at weld-area of Alloy 182. One major defect was found Two plus two ''boat-samples'' were taken out from the zone between the weld and the stainless cladding. All samples were sent to Studsviks laboratories for detailed investigations. The metallographic and fractographic investigations revealed that the major weld-defect had been there from manufacturing. The defect was located between the Alloy 182-buttering and the pressure vessel steel SA 533 grB cl 1. No indications of PWSCC or IDSCC were found. An inspection programme was defined. Different types of reference blocks were provided by Ringhals in cooperation with ABB TRC. Reference reflectors of type flat bottom hole (FBH) and eroded notches (EDM), with different sizes and separation were manufactured. One weld sample with manufacturing defects -lack of fusion and slag was inclusions- was present. ABB TRC performed UT inspection in the gap between the penetration and the thermal sleeve. Inspection results like defect identification, defect separation and defect sizing accuracy were compared with result from the destructive inspection. No relevant additional defects were found. An analysing and repair program was performed. A special designed disc sealed off the defect area. (authors). 5 figs., 3 refs

  6. Assessment of United States industry structural codes and standards for application to advanced nuclear power reactors: Appendices. Volume 2

    International Nuclear Information System (INIS)

    Adams, T.M.; Stevenson, J.D.

    1995-10-01

    Throughout its history, the USNRC has remained committed to the use of industry consensus standards for the design, construction, and licensing of commercial nuclear power facilities. The existing industry standards are based on the current class of light water reactors and as such may not adequately address design and construction features of the next generation of Advanced Light Water Reactors and other types of Advanced Reactors. As part of their on-going commitment to industry standards, the USNRC commissioned this study to evaluate US industry structural standards for application to Advanced Light Water Reactors and Advanced Reactors. The initial review effort included (1) the review and study of the relevant reactor design basis documentation for eight Advanced Light Water Reactors and Advanced Reactor Designs, (2) the review of the USNRCs design requirements for advanced reactors, (3) the review of the latest revisions of the relevant industry consensus structural standards, and (4) the identification of the need for changes to these standards. The results of these studies were used to develop recommended changes to industry consensus structural standards which will be used in the construction of Advanced Light Water Reactors and Advanced Reactors. Over seventy sets of proposed standard changes were recommended and the need for the development of four new structural standards was identified. In addition to the recommended standard changes, several other sets of information and data were extracted for use by USNRC in other on-going programs. This information included (1) detailed observations on the response of structures and distribution system supports to the recent Northridge, California (1994) and Kobe, Japan (1995) earthquakes, (2) comparison of versions of certain standards cited in the standard review plan to the most current versions, and (3) comparison of the seismic and wind design basis for all the subject reactor designs

  7. Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor building decontamination. Summary status report. Volume 2

    International Nuclear Information System (INIS)

    Doerge, D.H.; Miller, R.L.; Scotti, K.S.

    1986-05-01

    This document summarizes information relating to decontamination of the Three Mile Island Unit 2 (TMI-2) reactor building. The report covers activities for the period of June 1, 1979 through March 29, 1985. The data collected from activity reports, reactor containment entry records, and other sources were entered into a computerized data system which permits extraction/manipulation of specific information which can be used in planning for recovery from an accident similar to that experienced at TMI-2 on March 28, 1979. This report contains summaries of man-hours, manpower, and radiation exposures incurred during decontamination of the reactor building. Support activities conducted outside of radiation areas are excluded from the scope of this report. Computerized reports included in this document are: a chronological summary listing work performed relating to reactor building decontamination for the period specified; and summary reports for each major task during the period. Each task summary is listed in chronological order for zone entry and subtotaled for the number of personnel entries, exposures, and man-hours. Manually-assembled table summaries are included for: labor and exposures by department and labor and exposures by major activity

  8. In-situ stripping of H{sub 2}S in gasoil hydrodesulphurization - reactor design considerations

    Energy Technology Data Exchange (ETDEWEB)

    Nava, J.A.O.; Krishna, R. [Amsterdam Univ., Dept. of Chemical Engineering, Amsterdam (Netherlands)

    2004-02-01

    In order to meet future diesel specifications the sulphur content of diesel would need to be reduced to below 50 ppm. This requirement would require improved reactor configurations. In this study we examine the benefits of counter-current contacting of gas oil with H{sub 2}, over conventional co-current contacting in a trickle bed hydrodesulphurization (HDS) reactor. In counter-current contacting, we achieve in-situ stripping of H{sub 2}S from the liquid phase; this is beneficial to the HDS kinetics. A comparison simulation study shows that counter-current contacting would require about 20% lower catalyst load than co-current contacting. However, counter-current contacting of gas and liquid phases in conventionally used HDS catalysts, of 1.5 mm sizes, is not possible due to flooding limitations. The catalysts need to be housed in special wire gauze envelopes as in the catalytic bales or KATAPAK-S configurations. A preliminary hardware design of a counter-current HDS reactor using catalytic bales was carried out in order to determine the technical feasibility. Using a realistic sulphur containing feedstock, the target of 50 ppm S content of desulphurized oil could be met in a reactor of reasonable dimensions. The study also underlines the need for accurate modelling of thermal effects during desulphurization. Our study also shows that interphase mass transfer is unlikely to be a limiting factor and there is a need to develop improved reactor configurations allowing for increased catalyst loading, at the expense of gas-liquid interfacial area. (Author)

  9. Energy Multiplier Module (EM{sup 2}) - advanced small modular reactor for electricity generation

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, T.; Schleicher, R.; Choi, H.; Rawls, J., E-mail: timothy.bertch@ga.com [General Atomics, San Diego, California (United States)

    2013-07-01

    In order to provide cost effective nuclear energy in other than large reactor, large grid applications, fission technology needs to make further advances. 'Convert and burn' fast reactors offer long life cores, improved fuel utilization, reduced waste and other benefits while achieving cost effective energy production in a smaller reactor. General Atomics' Energy Multiplier Module (EM{sup 2}), a helium-cooled compact fast reactor that augments its fissile fuel load with either depleted uranium (DU) or used nuclear fuel (UNF). The convert and burn in-situ provides 250 MWe with a 30 year core life. High temperature provides a simple, high efficiency direct cycle gas turbine which along with modular construction, fewer systems, road shipment and minimum on site construction support cost effectiveness. Additional advantages in fuel cycle, non-proliferation and siting flexibility and its ability to meet all safety requirements make for an attractive power source, especially in remote and small grid regions. (author)

  10. HERESY, 2-D Few-Group Static Eigenvalues Calculation for Thermal Reactor

    International Nuclear Information System (INIS)

    Finch, D.R.

    1965-01-01

    1 - Description of problem or function: HERESY3 solves the two- dimensional, few-group, static reactor eigenvalue problem using the heterogeneous (source-sink or Feinburg-Galanin) formalism. The solution yields the reactor k-effective and absorption reaction rates for each rod normalized to the most absorptive rod in the thermal level. Epithermal fissions are allowed at each resonance level, and lattice-averaged values of thermal utilization, resonance escape probability, thermal and resonance eta values, and the fast fission factor are calculated. Kernels in the calculation are based on age-diffusion theory. Both finite reactor lattices and infinitely repeating reactor super-cells may be calculated. Rod parameters may be calculated by several internal options, and a direct interface is provided to a HAMMER system (NESC Abstract 277) lattice library tape to obtain cell parameters. Criticality searches are provided on thermal utilization, thermal eta, and axial leakage buckling. 2 - Method of solution: Direct power iteration on matrix form of the heterogeneous critical equation is used. 3 - Restrictions on the complexity of the problem: Maxima of - 50 flux/geometry symmetry positions; 20 physically different assemblies; 9 resonance levels; 5000 rod coordinate positions

  11. In-pile modelling of nuclear fuel element for the MTR type reactors. Pt. 2

    Energy Technology Data Exchange (ETDEWEB)

    Farhadi, Kazem [AEOI, Tehran (Iran, Islamic Republic of). Radiations Application Research School

    2014-06-15

    In part two of the present paper, neutronic properties of the pool-type research reactor core are used to assess the similitude laws derived for out-of-pile modelling of the fuel element. The benchmark reactor used for this purpose is an IAEA 5 MW thermal pool-type research reactor currently in operation. The neutronic properties analysis are based on typical 2 200 m/sec and neutrons having 0.025 eV energy. The non-leakage capability of the system is estimated in terms of diffusion length. Also the slowing down power and the moderating ratio of the modelled methanol coolant are calculated in terms of lethargy of the diffusing medium. It is shown that the Iron which is substituted for Aluminium cladding is a relatively low absorber of neutrons but has a high neutron leakage. Methanol which replaced ordinary water as coolant is not a suitable coolant due to high neutrons absorbing substance. It is concluded that although Iron as a cladding material and methanol as a coolant meet the modelling out-of-pile criteria but are not satisfying neutronic properties. Therefore, use of them as a model clad and coolant are not suggested for research reactors. (orig.)

  12. A Conceptual Study on a Supercritical CO_2-cooled Micro Modular Reactor

    International Nuclear Information System (INIS)

    Yu, Hwanyeal; Hartanto, Donny; Kim, Yonghee

    2014-01-01

    A Micro Modular Reactor (MMR) using Supercritical-CO_2 (S-CO_2) as coolant has been investigated from the neutronics perspective. The MMR is designed to be transportable so it can reach the remote areas. The thermal power of the reactor is 36.2 M Wth. The size of the active core is limited to 1.2 m length and 93.16 cm width. The size of whole core is 2.8 m length and 166.9 cm width. The reactor lifetime design target is 20 years. To maximize the fuel volume fraction in the core, high density uranium nitride UN"1"5 was used. The PbO/MgO reflector was also utilized to improve the neutron economy. The S-CO_2 is chosen as the coolant because it offers a higher thermal efficiency. In this study, neutronics calculations and depletion using McCARD Monte Carlo code has been done to determine the lifetime and behavior of the core. Several important safety parameters such as Control Rod worth, Doppler reactivity coefficients and coolant void reactivity coefficient have also been analyzed. (author)

  13. Performance Estimation of Supercritical Co2 Micro Modular Reactor (MMR) for Varying Cooling Air Temperature

    International Nuclear Information System (INIS)

    Ahn, Yoonhan; Kim, Seong Gu; Cho, Seong Kuk; Lee, Jeong Ik

    2015-01-01

    A Small Modular Reactor (SMR) receives interests for the various application such as electricity co-generation, small-scale power generation, seawater desalination, district heating and propulsion. As a part of SMR development, supercritical CO2 Micro Modular Reactor (MMR) of 36.2MWth in power is under development by the KAIST research team. To enhance the mobility, the entire system including the power conversion system is designed for the full modularization. Based on the preliminary design, the thermal efficiency is 31.5% when CO2 is sufficiently cooled to the design temperature. A supercritical CO2 MMR is designed to supply electricity to the remote regions. The ambient temperature of the area can influence the compressor inlet temperature as the reactor is cooled with the atmospheric air. To estimate the S-CO2 cycle performance for various environmental conditions, A quasi-static analysis code is developed. For the off design performance of S-CO2 turbomachineries, the experimental result of Sandia National Lab (SNL) is utilized

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-07-01

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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  16. A gas-phase reactor powered by solar energy and ethanol for H2 production

    International Nuclear Information System (INIS)

    Ampelli, Claudio; Genovese, Chiara; Passalacqua, Rosalba; Perathoner, Siglinda; Centi, Gabriele

    2014-01-01

    In the view of H 2 as the future energy vector, we presented here the development of a homemade photo-reactor working in gas phase and easily interfacing with fuel cell devices, for H 2 production by ethanol dehydrogenation. The process generates acetaldehyde as the main co-product, which is more economically advantageous with respect to the low valuable CO 2 produced in the alternative pathway of ethanol photoreforming. The materials adopted as photocatalysts are based on TiO 2 substrates but properly modified with noble (Au) and not-noble (Cu) metals to enhance light harvesting in the visible region. The samples were characterized by BET surface area analysis, Transmission Electron Microscopy (TEM) and UV–visible Diffusive Reflectance Spectroscopy, and finally tested in our homemade photo-reactor by simulated solar irradiation. We discussed about the benefits of operating in gas phase with respect to a conventional slurry photo-reactor (minimization of scattering phenomena, no metal leaching, easy product recovery, etc.). Results showed that high H 2 productivity can be obtained in gas phase conditions, also irradiating titania photocatalysts doped with not-noble metals. - Highlights: • A gas-phase photoreactor for H 2 production by ethanol dehydrogenation was developed. • The photocatalytic behaviours of Au and Cu metal-doped TiO 2 thin layers are compared. • Benefits of operating in gas phase with respect to a slurry reactor are presented. • Gas phase conditions and use of not-noble metals are the best economic solution

  17. An improved slow neutron spectrometer at nuclear research reactor et-r r-1. Vol. 2

    Energy Technology Data Exchange (ETDEWEB)

    Abu El-Ela, M A [Reactor and Neutron Physics, Nuclear Research Center, AEA, Cairo (Egypt)

    1996-03-01

    An improved slow neutron selector has been aligned at channel number 6 of the nuclear research reactor ET-R R-1 Inshas. The flight path is 4 meter. The collimator-rotor-collimator system has the dimensions 0.3 x 2.5 x 70 cm with the rotor diameter 16 cm and 3 slits of 0.3 x 2.5 cm cross section. The rotor rotation rate varies between 600 r.p.m. the counting system has one of the best modern high electronic advanced technology time analyzer with minimum dwell time 2 sec, 8192 channels and a double detector inputs of TTL and NEG NIM standard pulses. The analyzer external triggering signals are of TTL standard type. A special design {sup 3} He detector for time of flight spectrometry has been used in the SNS. The reactor bare thermal neutron spectrum has been successfully measured, to show good agreement with the previous data. 6 figs.

  18. Benchmark testing of Canadol-2.1 for heavy water reactor

    International Nuclear Information System (INIS)

    Liu Ping

    1999-01-01

    The new version evaluated nuclear data library of ENDF-B 6.5 has been released recently. In order to compare the quality of evaluated nuclear data CENDL-2.1 with ENDF-B 6.5, it is necessary to do benchmarks testing for them. In this work, CENDL-2.1 and ENDF-B 6.5 were used to generated the WIMS-69 group library respectively, and benchmarks testing was done for the heavy water reactor, using WIMS5A code. It is obvious that data files of CENDL-2.1 is better than that of old WIMS library for the heavy water reactors calculations, and is in good agreement with those of ENDF-B 6.5

  19. Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Bissani, M; O' Kelly, D S

    2006-05-08

    A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to

  20. Experimental studies of U-Pu-Zr fast reactor fuel pins in the Experimental Breeder Reactor 2

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

    1990-01-01

    Argonne National Laboratory's Integral Fast Reactor (IFR) concept has been under demonstration in the Experimental Breeder Reactor II (EBR-II) since February 1985. Irradiation tests of U-Zr and U-Pu-Zr fuel pins to >15 at. pct burnup have demonstrated their viability as driver fuel prototypes in innovative design liquid metal reactors. A number of technically challenging irradiation effects have been observed and are now under study. Microstructural changes in the fuel are dominated early in exposure by grain boundary cavitation and fission gas bubble growth, producing large amounts of swelling. Irradiation creep and swelling of the austenitic (D9) and martensitic (HT-9) candidate cladding alloys have been measured and correlate well with property modeling efforts. Chemical interaction between the fuel and cladding alloys has been characterized to assess the magnitude of cladding wastage during steady-state irradiation. Significant interdiffusion of the uranium and zirconium occurs producing metallurgically distinct zones in the fuel

  1. Proceedings of the international topical meeting on advanced reactors safety: Volume 2

    International Nuclear Information System (INIS)

    1997-01-01

    In this volume, 89 papers are grouped under the following headings: advances in research/test reactor safety; advanced reactor accident management and emergency actions; advanced reactors instrumentation/controls/human factors; probabilistic risk/safety and reliability assessments; steam explosion research and issues; advanced reactor severe accident issues and research (analysis and assessments); advanced reactor thermal hydraulics; accelerator-driven source safety; liquid-metal reactor safety; structural assessments and issues; late papers

  2. Final report on in-reactor creep-fatigue deformation behaviour of a CuCrZr alloy: COFAT 2

    DEFF Research Database (Denmark)

    Singh, Bachu Narain; Johansen, Bjørn Sejr; Tähtinen, S.

    facilities for this purpose, in-reactor creep-fatigue tests have been performed at strain amplitudes of 0.25 and 0.35 % with a holdtime of 10s in the BR-2 reactor at Mol (Belgium). These tests were performed at the ambient temperatures of 326K and 323K. For comparison purposes corresponding out...

  3. Summary of IEA-R1 research a reactor licensing related to its power increase from 2 to 10 MW

    International Nuclear Information System (INIS)

    1989-04-01

    This work is a summary of IEA-R1 research reactor licensing related to its power increase from 2 to 10 MW. It reports also safety requirements, fuel elements, and reactor control modifications inherent to power increase. (A.C.A.S.)

  4. Micro-scale H2-CO2 dynamics in a hydrogenotrophic methanogenic membrane reactor

    DEFF Research Database (Denmark)

    Garcia-Robledo, Emilio; Ottosen, Lars Ditlev Mørck; Voigt, Niels Vinther

    2016-01-01

    Biogas production is a key factor in a sustainable energy supply. It is possible to get biogas with very high methane content if the biogas reactors are supplied with exogenous hydrogen, and one of the technologies for supplying hydrogen is through gas permeable membranes. In this study the activ......Biogas production is a key factor in a sustainable energy supply. It is possible to get biogas with very high methane content if the biogas reactors are supplied with exogenous hydrogen, and one of the technologies for supplying hydrogen is through gas permeable membranes. In this study...

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

    Science.gov (United States)

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

    2011-03-01

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

  6. Treatment patterns in moderate-to-severe plaque psoriasis: results from a Belgian cross-sectional study (DISCOVER).

    Science.gov (United States)

    Lambert, Julien; Ghislain, Pierre-Dominique; Lambert, Jo; Cauwe, Bénédicte; Van den Enden, Maria

    2017-08-01

    The present study aimed to evaluate current treatment patterns and achievement of treatment goals in Belgian patients with moderate-to-severe plaque psoriasis. This cross-sectional observational study (DISCOVER) was conducted in 2011 - 2012 in Belgian dermatology centers. Patient data were collected during a single visit and included information on psoriasis management and severity (PASI and DLQI). Treatment success was defined according to the current European consensus treatment goal algorithm. Of the 556 patients included in the study, 38.1% reported no current treatment or only topicals, 34.2% were being treated with traditional systemics and/or phototherapy, and 29.5% with biologics. Methotrexate (11.7%) was the most commonly prescribed traditional systemic and adalimumab (14.2%) was the most commonly prescribed biologic agent at the time of the study. The percentage of patients achieving treatment goals was significantly higher in biologic-treated patients (73.1%) compared to those using traditional systemics (50.6%), phototherapy (41.1%), or no treatment/only topicals (20.9%; p psoriasis in the DISCOVER study were undertreated despite the severity of their disease. Undertreatment of psoriasis remains a problem in Belgium and more effective educational strategies are needed to ensure the best treatment outcome for these patients. [Formula: see text].

  7. Performance of small reactors at universities for teaching, research, training and service (TRTS): thirty five years' experience with the Dalhousie University SLOWPOKE-2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chatt, A., E-mail: a.chatt@dal.ca [Dalhousie Univ., Trace Analysis Research Centre, Dept. of Chemistry, Halifax, Nova Scotia (Canada)

    2013-07-01

    The Dalhousie University SLOWPOKE-2 Reactor (DUSR) facility, operated during 1976-2011, was the only research reactor in Atlantic Canada as well as the only one associated with a chemistry department in a Canadian university. The most outstanding features of the facility included: a rapid (100 ms) cyclic pneumatic sample transfer system, a permanently installed Cd-site, and a Compton-suppression gamma-ray spectrometer. The usage encompassed fundamental as well as applied studies in various fields using neutron activation analysis (NAA). The facility was used for training undergraduate/graduate students, postdoctoral fellows, technicians, and visiting scientists, and for cooperative projects with other universities, research organizations and industries. (author)

  8. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'JOYO'. 2. Replacement of upper core structure

    International Nuclear Information System (INIS)

    Ushiki, Hiroshi; Ito, Hiromichi; Okuda, Eiji; Suzuki, Nobuhiro; Sasaki, Jun; Oota, Katsu; Kawahara, Hirotaka; Takamatsu, Misao; Nagai, Akinori; Okawa, Toshikatsu

    2015-01-01

    In the experimental fast reactor Joyo, it was confirmed that the top of the irradiation test sub-assembly of MARICO-2 (material testing rig with temperature control) had bent onto the in-vessel storage rack as an obstacle and had damaged the upper core structure (UCS) in 2007. As a part of the restoration work, UCS replacement was begun at March 24, 2014 and was completed at December 17. In-vessel repair (including observation) for sodium-cooled fast reactors (SFRs) is distinct from that for light water reactors and necessitates independent development. Application of developed in-vessel repair techniques to operation and maintenance of SFRs enhanced their safety and integrity. There is little UCS replacement experience in the world and this experience and insights, which were accumulated in the replacement work of in-vessel large structure (UCS) used for more than 30 years, are expected to improve the in-vessel repair techniques in SFRs. (author)

  9. A regression approach for Zircaloy-2 in-reactor creep constitutive equations

    International Nuclear Information System (INIS)

    Yung Liu, Y.; Bement, A.L.

    1977-01-01

    In this paper the methodology of multiple regressions as applied to Zircaloy-2 in-reactor creep data analysis and construction of constitutive equation are illustrated. While the resulting constitutive equation can be used in creep analysis of in-reactor Zircaloy structural components, the methodology itself is entirely general and can be applied to any creep data analysis. The promising aspects of multiple regression creep data analysis are briefly outlined as follows: (1) When there are more than one variable involved, there is no need to make the assumption that each variable affects the response independently. No separate normalizations are required either and the estimation of parameters is obtained by solving many simultaneous equations. The number of simultaneous equations is equal to the number of data sets. (2) Regression statistics such as R 2 - and F-statistics provide measures of the significance of regression creep equation in correlating the overall data. The relative weights of each variable on the response can also be obtained. (3) Special regression techniques such as step-wise, ridge, and robust regressions and residual plots, etc., provide diagnostic tools for model selections. Multiple regression analysis performed on a set of carefully selected Zircaloy-2 in-reactor creep data leads to a model which provides excellent correlations for the data. (Auth.)

  10. Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

    International Nuclear Information System (INIS)

    Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Pal, S.; Mishra, G.P.

    2012-01-01

    CERMET fuel with either PuO 2 or enriched UO 2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR’s). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R and D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO 2 dispersed in uranium metal matrix pellets for three different compositions i.e. U–20 wt%UO 2 , U–25 wt%UO 2 and U–30 wt%UO 2 . It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U–UO 2 compositions.

  11. Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

    Science.gov (United States)

    Sinha, V. P.; Hegde, P. V.; Prasad, G. J.; Pal, S.; Mishra, G. P.

    2012-08-01

    CERMET fuel with either PuO2 or enriched UO2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR's). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R & D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO2 dispersed in uranium metal matrix pellets for three different compositions i.e. U-20 wt%UO2, U-25 wt%UO2 and U-30 wt%UO2. It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U-UO2 compositions.

  12. Integral Inherently Safe Light Water Reactor (I2S-LWR)

    International Nuclear Information System (INIS)

    Petrovic, Bojan; Memmott, Matthew; Boy, Guy; Charit, Indrajit; Manera, Annalisa; Downar, Thomas; Lee, John; Muldrow, Lycurgus; Upadhyaya, Belle; Hines, Wesley; Haghighat, Alierza

    2017-01-01

    This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project ''Integral Inherently Safe Light Water Reactors (I 2 S-LWR)''. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

  13. Direct In Situ Quantification of HO2 from a Flow Reactor.

    Science.gov (United States)

    Brumfield, Brian; Sun, Wenting; Ju, Yiguang; Wysocki, Gerard

    2013-03-21

    The first direct in situ measurements of hydroperoxyl radical (HO2) at atmospheric pressure from the exit of a laminar flow reactor have been carried out using mid-infrared Faraday rotation spectroscopy. HO2 was generated by oxidation of dimethyl ether, a potential renewable biofuel with a simple molecular structure but rich low-temperature oxidation chemistry. On the basis of the results of nonlinear fitting of the experimental data to a theoretical spectroscopic model, the technique offers an estimated sensitivity of reactor exit temperature range of 398-673 K. Accurate in situ measurement of this species will aid in quantitative modeling of low-temperature and high-pressure combustion kinetics.

  14. Integral Inherently Safe Light Water Reactor (I2S-LWR)

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Memmott, Matthew [Brigham Young Univ., Provo, UT (United States); Boy, Guy [Florida Inst. of Technology, Melbourne, FL (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States); Manera, Annalisa [Univ. of Michigan, Ann Arbor, MI (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Lee, John [Univ. of Michigan, Ann Arbor, MI (United States); Muldrow, Lycurgus [Morehouse College, Atlanta, GA (United States); Upadhyaya, Belle [Univ. of Tennessee, Knoxville, TN (United States); Hines, Wesley [Univ. of Tennessee, Knoxville, TN (United States); Haghighat, Alierza [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States)

    2017-10-02

    This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project “Integral Inherently Safe Light Water Reactors (I2S-LWR)”. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

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

    International Nuclear Information System (INIS)

    Takano, Hideki; Kaneko, Kunio.

    1989-10-01

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

  16. Techno-economic assessment of membrane assisted fluidized bed reactors for pure H_2 production with CO_2 capture

    International Nuclear Information System (INIS)

    Spallina, V.; Pandolfo, D.; Battistella, A.; Romano, M.C.; Van Sint Annaland, M.; Gallucci, F.

    2016-01-01

    Highlights: • Membrane reactors improve the overall efficiency of H_2 production up to 20%. • Respect to conventional reforming, the H_2 yield increases from 12% to 20%. • The COH is reduced of at least 220% using membrane reactors. • FBMR capture 72% of CO_2 with a specific cost of 8 eur/tonn_C_O_2_. • MA-CLR can reach 90% of CO_2 avoided with same cost of FTR. - Abstract: This paper addresses the techno-economic assessment of two membrane-based technologies for H_2 production from natural gas, fully integrated with CO_2 capture. In the first configuration, a fluidized bed membrane reactor (FBMR) is integrated in the H_2 plant: the natural gas reacts with steam in the catalytic bed and H_2 is simultaneously separated using Pd-based membranes, and the heat of reaction is provided to the system by feeding air as reactive sweep gas in part of the membranes and by burning part of the permeated H_2 (in order to avoid CO_2 emissions for heat supply). In the second system, named membrane assisted chemical looping reforming (MA-CLR), natural gas is converted in the fuel rector by reaction with steam and an oxygen carrier (chemical looping reforming), and the produced H_2 permeates through the membranes. The oxygen carrier is re-oxidized in a separate air reactor with air, which also provides the heat required for the endothermic reactions in the fuel reactor. The plants are optimized by varying the operating conditions of the reactors such as temperature, pressures (both at feed and permeate side), steam-to-carbon ratio and the heat recovery configuration. The plant design is carried out using Aspen Simulation, while the novel reactor concepts have been designed and their performance have been studied with a dedicated phenomenological model in Matlab. Both configurations have been designed and compared with reference technologies for H_2 production based on conventional fired tubular reforming (FTR) with and without CO_2 capture. The results of the analysis show

  17. Reactor pressure vessel steels ASTM A533B and A508 Cl.2

    International Nuclear Information System (INIS)

    Pelli, R.; Kemppainen, M.; Toerroenen, K.

    1979-11-01

    This report presents the tensile test results of steels ASTM A533B and A508 Cl.2 obtained in connection with a programme initiated to gather and create information needed for the assessment of the structural integrity of the reactor pressure vessels. The tensile properties were studied between -196 and 300 degC varying austenitizing and tempering temperatures and having two different carbon contents for the heats of A533B. (author)

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  19. Technical evolution and operation of French CO2 cooled reactors (UNGG)

    International Nuclear Information System (INIS)

    Berthion, Y.

    1986-10-01

    The technical evolution of the five French CO 2 cooled reactors (UNGG) from 1981 to 1986 needs to be outlined. These technical evolutions concerned the fuel element of Bugey 1 which is now slightly enriched, as well as the load reduction operation required by the grid. In addition work in underway to increase the safety at the two St Laurent units, or to repair the hot steel upper-structures of Chinon-3 unit

  20. Tests to control the power distribution in the IRT-2,000 reactor

    International Nuclear Information System (INIS)

    Filipcuk, E.V.; Potapenko, P.T.; Krjukov, A.P.; Trofimov, A.P.; Kosilov, A.N.; Nebojan, V.T.; Timochin, E.S.

    1976-01-01

    Results of the investigations of a few structures of such control systems carried out with the help of the IRT 2,000 MIFI reactor in the years 1973/74 are presented in the present work. Within the framework of this study, the successful test of using the transmitter of the direct loading in equipment to control the neutron field was carried out. (orig./TK) [de

  1. Development of a new control software package for Pakistan Research Reactor-2

    International Nuclear Information System (INIS)

    Qazi, M.K.

    1993-05-01

    The development of a new control software package for Pakistan Research Reactor-2 is presented. The software operates in different modes which comprises of surveillance, pre-operational self tests, operator, supervisor and robotic control. The control logic critically damp the system minimizing power overshoots. The software, handles multiple abnormal conditions, provides an elaborate access control and maintains startup/shutdown record. The report describes the functional details and covers the operational aspects of the new control software. (author)

  2. Y2K issues for real time computer systems for fast breeder test reactor

    International Nuclear Information System (INIS)

    Swaminathan, P.

    1999-01-01

    Presentation shows the classification of real time systems related to operation, control and monitoring of the fast breeder test reactor. Software life cycle includes software requirement specification, software design description, coding, commissioning, operation and management. A software scheme in supervisory computer of fast breeder test rector is described with the twenty years of experience in design, development, installation, commissioning, operation and maintenance of computer based supervision control system for nuclear installation with a particular emphasis on solving the Y2K problem

  3. Power optimization in the STAR-LM modular natural convection reactor system. Topic 2.1 advanced reactor power plants

    International Nuclear Information System (INIS)

    Spencer, B.W.; Sienicki, J.J.; Farmer, M.T.

    2001-01-01

    The secure, transportable, autonomous reactor (STAR) project addresses the needs of developing countries and independent power producers for a small (300 MWt), multi-purpose energy system. The STAR-LM variant described here is a liquid metal cooled, fast spectrum reactor system. Previous development of a reference STAR-LM design resulted in a 300 MWt modular, pool- type reactor based on criteria for factory fabrication of modules, full transportability of modules (barge, rail, overland), fast construction and startup, and semi-autonomous operation. Earlier work on the reference 300 MWt concept focused first on addressing whether 100% natural circulation heat transport was achievable under the module size constraints for full transportability and under the coolant and cladding peak temperature limitations imposed by the existing Russian database for ferritic-martensitic core material with oxide-layer corrosion protection. Secondly, owing to uncertainties and limitations in the available Russian materials compatibility database, the objective of the reference design was to address how low the coolant and cladding peak temperatures could be commensurate with achieving 300 MWt power level with 100% natural circulation in a fully transportable module size. In the present work we have refocused the approach to attempt to maximize the power achievable in the reactor module based on preserving the criteria for full module transportability and remaining within the materials compatibility database limits. (author)

  4. Methodology for reactor core physics analysis - part 2; Metodologia de analise fisica do nucleo - etapa 2

    Energy Technology Data Exchange (ETDEWEB)

    Ponzoni Filho, P; Fernandes, V B; Lima Bezerra, J de; Santos, T I.C.

    1992-12-01

    The computer codes used for reactor core physics analysis are described. The modifications introduced in the public codes and the technical basis for the codes developed by the FURNAS utility are justified. An evaluation of the impact of these modifications on the parameter involved in qualifying the methodology is included. (F.E.). 5 ref, 7 figs, 5 tabs.

  5. Geomechanical Analysis of Underground Coal Gasification Reactor Cool Down for Subsequent CO2 Storage

    Science.gov (United States)

    Sarhosis, Vasilis; Yang, Dongmin; Kempka, Thomas; Sheng, Yong

    2013-04-01

    Underground coal gasification (UCG) is an efficient method for the conversion of conventionally unmineable coal resources into energy and feedstock. If the UCG process is combined with the subsequent storage of process CO2 in the former UCG reactors, a near-zero carbon emission energy source can be realised. This study aims to present the development of a computational model to simulate the cooling process of UCG reactors in abandonment to decrease the initial high temperature of more than 400 °C to a level where extensive CO2 volume expansion due to temperature changes can be significantly reduced during the time of CO2 injection. Furthermore, we predict the cool down temperature conditions with and without water flushing. A state of the art coupled thermal-mechanical model was developed using the finite element software ABAQUS to predict the cavity growth and the resulting surface subsidence. In addition, the multi-physics computational software COMSOL was employed to simulate the cavity cool down process which is of uttermost relevance for CO2 storage in the former UCG reactors. For that purpose, we simulated fluid flow, thermal conduction as well as thermal convection processes between fluid (water and CO2) and solid represented by coal and surrounding rocks. Material properties for rocks and coal were obtained from extant literature sources and geomechanical testings which were carried out on samples derived from a prospective demonstration site in Bulgaria. The analysis of results showed that the numerical models developed allowed for the determination of the UCG reactor growth, roof spalling, surface subsidence and heat propagation during the UCG process and the subsequent CO2 storage. It is anticipated that the results of this study can support optimisation of the preparation procedure for CO2 storage in former UCG reactors. The proposed scheme was discussed so far, but not validated by a coupled numerical analysis and if proved to be applicable it could

  6. Mass transfer of ammonia escape and CO2 absorption in CO2 capture using ammonia solution in bubbling reactor

    International Nuclear Information System (INIS)

    Ma, Shuangchen; Chen, Gongda; Zhu, Sijie; Han, Tingting; Yu, Weijing

    2016-01-01

    Highlights: • Mass transfer coefficient models of ammonia escape were built. • Influences of temperature, inlet CO 2 and ammonia concentration were studied. • Mass transfer coefficients of ammonia escape and CO 2 absorption were obtained. • Studies can provide the basic data as a reference guideline for process application. - Abstract: The mass transfer of CO 2 capture using ammonia solution in the bubbling reactor was studied; according to double film theory, the mass transfer coefficient models and interface area model were built. Through our experiments, the overall volumetric mass transfer coefficients were obtained, while the interface areas in unit volume were estimated. The volumetric mass transfer coefficients of ammonia escaping during the experiment were 1.39 × 10 −5 –4.34 × 10 −5 mol/(m 3 s Pa), and the volumetric mass transfer coefficients of CO 2 absorption were 2.86 × 10 −5 –17.9 × 10 −5 mol/(m 3 s Pa). The estimated interface area of unit volume in the bubbling reactor ranged from 75.19 to 256.41 m 2 /m 3 , making the bubbling reactor a viable choice to obtain higher mass transfer performance than the packed tower or spraying tower.

  7. Mixed core management: Use of 93% and 72% enriched uranium in the BR2 reactor

    International Nuclear Information System (INIS)

    Ponsard, B.

    2000-01-01

    The BR2 reactor, put into operation in 1963 and refurbished from July 1995 till April 1997, is a 100 MW high-flux Materials Testing Reactor, using 93% 235 U enriched uranium as standard fuel, light water as coolant and beryllium as moderator. The present operating regime consists of five irradiation cycles per year at an operating power between 50 and 70 MW; each cycle is characterized by 21 days operation. In the framework of a 'qualification programme', six 72% 235 U fuel elements fabricated with uranium recovered from the reprocessing of BR2 spent fuel at UKAEA-Dounreay have been successfully irradiated in the period 1994-1995 reaching a maximum mean burnup of 48% without the release of fission products. Since 1998, this type of fuel element is irradiated routinely together with standard 93% 235 U fuel elements in order to optimize the utilization of the available HEU inventory. The purpose of this paper is to present the strategy developed in order to optimize the mixed core management of the BR2 reactor. (author)

  8. MASTER-2.0: Multi-purpose analyzer for static and transient effects of reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Byung Oh; Song, Jae Seung; Joo, Han Gyu [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    MASTER-2.0 (Multi-purpose Analyzer for Static and Transient Effects of Reactors) is a nuclear design code based on the two group diffusion theory to calculate the steady-state and transient pressurized water reactor core in a 3-dimensional Cartesian or hexagonal geometry. Its neutronics model solves the space-time dependent neutron diffusion equations with NIM(Nodal Integration Method), NEM (Nodal Expansion Method), AFEN (Analytic Function Expansion Nodal Method)/NEM Hybrid Method, NNEM (Non-linear Nodal Expansion Method) or NANM (Non-linear Analytic Nodal Method) for a Cartesian geometry and with AFEN/NEM Hybrid Method or NLFM (Non-linear Local Fine-Mesh Method) for a hexagonal one. Coarse mesh rebalancing, Krylov Subspace method and asymptotic extrapolation method are implemented to accelerate the convergence of iteration process. Master-2.0 performs microscopic depletion calculations using microscopic cross sections provided by CASMO-3 or HELIOS and also has the reconstruction capability of pin information by use of MSS-IAS (Method of Successive Smoothing with Improved Analytic Solution). For the thermal-hydraulic calculation, fuel temperature table or COBRA3-C/P model can be used selectively. In addition, MASTER-2.0 is designed to cover various PWRs including SMART as well as WH-and CE-type reactors, providing all data required in their design procedures. (author). 39 refs., 12 figs., 4 tabs.

  9. Testing of HTR UO{sub 2} TRISO fuels in AVR and in material test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kania, Michael J., E-mail: MichaelJKania@googlemail.com [Retired from Lockheed Martin Corp, 20 Beach Road, Averill Park, NY 12018 (United States); Nabielek, Heinz, E-mail: heinznabielek@me.com [Retired from Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Verfondern, Karl [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Allelein, Hans-Josef [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, 52072 Aachen (Germany)

    2013-10-15

    The German High Temperature Reactor Fuel Development Program successfully developed, licensed and manufactured many thousands of spherical fuel elements that were used to power the experimental AVR reactor and the commercial THTR reactor. In the 1970s, this program extended the performance envelope of HTR fuels by developing and qualifying the TRISO-coated particle system. Irradiation testing in real-time AVR tests and accelerated MTR tests demonstrated the superior manufacturing process of this fuel and its irradiation performance. In the 1980s, another program direction change was made to a low enriched UO{sub 2} TRISO-coated particle system coupled with high-quality manufacturing specifications designed to meet new HTR plant design needs. These needs included requirements for inherent safety under normal operation and accident conditions. Again, the German fuel development program met and exceeded these challenges by manufacturing and qualifying the low-enriched UO{sub 2} TRISO-fuel system for HTR systems with steam generation, gas-turbine systems and very high temperature process heat applications. Fuel elements were manufactured in production scale facilities that contained near defect free UO{sub 2} TRISO coated particles, homogeneously distributed within a graphite matrix with very low levels of uranium contamination. Good irradiation performance for these elements was demonstrated under normal operating conditions to 12% FIMA and under accident conditions not exceeding 1600 °C.

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  11. A study of UO2 wafer fuel for very high-power research reactors

    International Nuclear Information System (INIS)

    Hsieh, T.C.; Jankus, V.Z.; Rest, J.; Billone, M.C.

    1983-01-01

    The Reduced Enrichment Research and Test Reactor Program is aimed at reducing fuel enrichment to 2 caramel fuel is one of the most promising new types of reduced-enrichment fuel for use in research reactors with very high power density. Parametric studies have been carried out to determine the maximum specific power attainable without significant fission-gas release for UO 2 wafers ranging from 0.75 to 1.50 mm in thickness. The results indicate that (1) all the fuel designs considered in this study are predicted not to fail under full power operation up to a burnup, of 1.9x10 21 fis/cm 3 ; (2) for all fuel designs, failure is predicted at approximately the same fuel centerline temperature for a given burnup; (3) the thinner the wafer, the wider the margin for fuel specific power between normal operation and increased-power operation leading to fuel failure; (4) increasing the coolant pressure in the reactor core could improve fuel performance by maintaining the fuel at a higher power level without failure for a given burnup; and (5) for a given power level, fuel failure will occur earlier at a higher cladding surface temperature and/or under power-cycling conditions. (author)

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

    Science.gov (United States)

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

    2015-06-01

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

  13. A Conceptual Study of a Supercritical CO2-Cooled Micro Modular Reactor

    Directory of Open Access Journals (Sweden)

    Hwanyeal Yu

    2015-12-01

    Full Text Available A neutronics conceptual study of a supercritical CO2-cooled micro modular reactor (MMR has been performed in this work. The suggested MMR is an extremely compact and truck-transportable nuclear reactor. The thermal power of the MMR is 36.2 MWth and it is designed to have a 20-year lifetime without refueling. A salient feature of the MMR is that all the components including the generator are integrated in a small reactor vessel. For a minimal volume and long lifetime of the MMR core, a fast neutron spectrum is utilized in this work. To enhance neutron economy and maximize the fuel volume fraction in the core, a high-density uranium mono-nitride U15N fuel is used in the fast-spectrum MMR. Unlike the conventional supercritical CO2-cooled fast reactors, a replaceable fixed absorber (RFA is introduced in a unique way to minimize the excess reactivity and the power peaking factor of the core. For a compact core design, the drum-type control absorber is adopted as the primary reactivity control mechanism. In this study, the neutronics analyses and depletions have been performed by using the continuous energy Monte Carlo Serpent code with the evaluated nuclear data file ENDF/B-VII.1 Library. The MMR core is characterized in view of several important safety parameters such as control system worth, fuel temperature coefficient (FTC and coolant void reactivity (CVR, etc. In addition, a preliminary thermal-hydraulic analysis has also been performed for the hottest channel of the Korea Advanced Institute of Science and Technology (KAIST MMR.

  14. Use of unmanned aerial vehicle (UAV) for the detection and surveillance of marine oil spills in the Belgian part of the North Sea

    International Nuclear Information System (INIS)

    Donnay, E.

    2009-01-01

    This paper discussed the use of unmanned aerial vehicles (UAV) deployed by the Belgian Army in order to detect oil spills as well as for intelligence, reconnaissance and surveillance missions. The UAV are fitted with a dual sensor gyro-stabilized turret which combines a daylight camera and a thermal infrared camera. Live images of the sensors are transmitted in real time to control stations. All Belgian marine pollution surveillance platforms are coordinated by the Maritime Security Center of the Belgian Coast Guard. Satellite surveillance services provide real time information related to potential oil spills and other anomalies on the sea surface. Stand-by helicopters are also used for the rapid assessment of reported spills. The B-Hunter system is undetectable due to its small size and low noise signature, and can be used for the continuous monitoring of specific areas or for the tracking of suspect vessels. The system will also be used to monitor the progress of oil spill response operations as well as to provide information and guidance to response vessels. 6 refs., 2 figs

  15. Use of unmanned aerial vehicle (UAV) for the detection and surveillance of marine oil spills in the Belgian part of the North Sea

    Energy Technology Data Exchange (ETDEWEB)

    Donnay, E. [Federal Public Service Health, Food Chain Safety and Environment, Brussels (Belgium)

    2009-07-01

    This paper discussed the use of unmanned aerial vehicles (UAV) deployed by the Belgian Army in order to detect oil spills as well as for intelligence, reconnaissance and surveillance missions. The UAV are fitted with a dual sensor gyro-stabilized turret which combines a daylight camera and a thermal infrared camera. Live images of the sensors are transmitted in real time to control stations. All Belgian marine pollution surveillance platforms are coordinated by the Maritime Security Center of the Belgian Coast Guard. Satellite surveillance services provide real time information related to potential oil spills and other anomalies on the sea surface. Stand-by helicopters are also used for the rapid assessment of reported spills. The B-Hunter system is undetectable due to its small size and low noise signature, and can be used for the continuous monitoring of specific areas or for the tracking of suspect vessels. The system will also be used to monitor the progress of oil spill response operations as well as to provide information and guidance to response vessels. 6 refs., 2 figs.

  16. Use of highly enriched uranium in the material testing reactor BR2

    International Nuclear Information System (INIS)

    Beeckmans de West-Meerbeeck, A.

    1979-05-01

    In the material testing reactor BR2, the use of highly enriched uranium is determined by the consideration of the fast, epithermal and thermal neutron flux effectively available for the experimental devices. The choice of the core configuration is defined by combining the localisation of the experimental devices and of fuel elements of various burnup, such as to satisfy the irradiation conditions of the experimental load, compatible with an economic use of the fuel elements and safe operation of the reactor. Taking into account the present manufacturing technology for MTR fuels (37 Wt % uranium density in the fuel meat) the highly enriched uranium cannot be avoided; if higher concentration of uranium could be realised by some new manufacturing technology, the 235 U density of fuel elements at elimination should be kept at the required level and the enrichment could be reduced accordingly

  17. Use of highly enriched uranium in the material testing reactor BR2

    International Nuclear Information System (INIS)

    Beeckmans de West-Meerbeeck, A.

    1979-05-01

    In the material testing reactor BR2, the use of highly enriched uranium is determined by the consideration of the fast, epithermal and thermal neutron flux effectively available for the experimental devices. The choice of the core configuration is defined by combining the localisation of the experimental devices and of fuel elements of various burnup, such as to satisfy the irradiation conditions of the experimental load, compatible with an economic use of the fuel elements and safe operation of the reactor. Taking into account the present manufacturing technology for MTR fuels (37 Wt % uranium density in the fuel meat) the highly enriched uranium cannot be avoided: if higher concentration of uranium could be realised by some new manufacturing technology, the 235 U density of fuel elements at elimination should be kept at the required level and the enrichment could be reduced accordingly. (author)

  18. Re criticality assessment following reactor core damage in Fukushima unit 2

    International Nuclear Information System (INIS)

    Jeong, Hae Sun; Song, Jin Ho; Park, Chang Je; Ha, Kwang Soon; Song, Yong Mann; Ryu, Eun Hyun

    2012-01-01

    Following the severe core damage accident at the Fukushima nuclear power plants (NPPs), many researchers have studied a possible Re criticality caused by core melting or corium. However, no one can accurately examine the internal conditions of the reactor vessel, and thus there have been different opinions from some organizations depending on their assumption and analysis methods. If there is a potential Re criticality in the reactor vessel, some counter plans for the accident management should be established to prevent and mitigate re criticality, and to return the plant to a safe and stable state. In this study, the criticality level following a severe core damage accident was first analyzed using the MCNPX 2.6.0 code. Based on this result, practical strategies in terms of accident management were obtained by charging soluble boron (H 3B O 3) into re flooded water

  19. Application of CO2 laser beam weld for repair of fuel element of nuclear reactor 'YAYOI'

    International Nuclear Information System (INIS)

    Hashimoto, Mitsuo; Yanagi, Hideharu; Sukegawa, Toshio; Saito, Isao; Sasuga, Norihiko; Aizawa, Nagaaki; Miya, Kenzo

    1986-01-01

    The present studies are to develop CO 2 laser beam welding techniques in order to apply for repoint of nuclear reactor fuel of Fast Neutron Source Reactor YAYOI. For that purpos, many experiments were conduted to obtain various effects of laser welding variables with use of SUS 304 plates, pipes and simulated dumy fuels. These experiments provided us an optimal welding condition through metallurgical observations, non-destructive and mechanical tests. It was found that the laser welds exhibited properties equivalent to those of the base metal, in addition they provided us a favorable system than that of electron beam welds against a cladding of radioactive nuclear fuel in a hot cell. The present paper reports on the characteristics of laser welds, structural analysis of fuel element and a system design of remotely operated devices setting in a hot cell. (author)

  20. Dose assessment around TR-2 reactor due to maximum credible accident

    International Nuclear Information System (INIS)

    Turgut, M. H.; Adalioglu, U.; Aytekin, A.

    2001-01-01

    The revision of safety analysis report of TR-2 research reactor had been initiated in 1995. The whole accident analysis and accepted scenario for maximum credible accident has been revised according to the new safety concepts and the impact to be given to the environment due to this scenario has been assessed. This paper comprises all results of these calculations. The accepted maximum credible accident scenario is the partial blockage of the whole reactor core which resulted in the release of 25% of the core inventory. The DOSER code which uses very conservative modelling of atmospheric distributions were modified for the assessment calculations. Pasquill conditions based on the local weather observations, topography, and building affects were considered. The thyroid and whole body doses for 16 sectors and up to 10 km of distance around CNAEM were obtained. Release models were puff and a prolonged one of two hours of duration. Release fractions for the active isotopes were chosen from literature which were realistic

  1. Experience with the RE fuel transition at the Studsvik R2 reactor

    International Nuclear Information System (INIS)

    Pazsit, I.; Saltvedt, K.

    1991-01-01

    Irradiation of 7 LEU fuel elements is underway in the Studsvik R2 reactor. Four of these have 490 g U-235, and three 320 g U-235 loading, and the enrichment is 19.7% for all of them. The irradiation of LEU fuel started in 1987. The heavier elements have burnup figures 67% (CERCA), 50% (B and W), 47% (NUKEM) and 19% (B and W). One of the lighter elements has reached a burnup of 65%. To support the whole-core conversion process, reactor physical calculations were performed to see if a one-step conversion is possible with a suitable fuel management strategy such that all HEU fuel is burned up. The calculations show that it is possible to perform such a conversion with fuel elements containing 400 g U-235. (orig.)

  2. Stage 2: dismantling of reactor case of the experimental F.B.R. Rapsodie

    International Nuclear Information System (INIS)

    Roger, J.

    1994-01-01

    This document defines the main objectives of stage 2 dismantling of the Rapsodie experimental fast neutron reactor and specifies its time schedule. The work already in progress consists in containing the reactor vessel and its internal equipment, as well as the neutron protection concrete, inside the two leak-tight barriers, and in dismantling all the systems and equipment systems contaminated by sodium. This work, which includes the destruction of 37 metric tons of contaminated sodium from the primary system, was begun in 1987 and will be completed in 1994. The duration of the waiting period for complete dismantling (stage 3) has not been defined. However, the containment and monitoring means implemented should allow a safe waiting period of several decades. (author). 4 figs

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

    CERN Document Server

    2013-01-01

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

  4. Shipment of Taiwanese research reactor spent nuclear fuel (Phase 2): Environmental assessment

    International Nuclear Information System (INIS)

    1988-06-01

    The proposed action is to transport approximately 1100 spent fuel rods from a foreign research reactor in Taiwan by sea to Hampton Roads, Virginia, and then overland by truck to the receiving basin for offsite fuels at the Savannah River Plant (SRP) for reprocessing to recover uranium and plutonium. The analysis of the impacts of the proposed action have been evaluated and shown to have negligible impact on the local environments. The calculations have been completed using the RADTRAN III code. PWR spent fuel was analyzed as a benchmark to link the calculations in this analysis to those in earlier environmental documentation. Cumulative total, maximum annual, and per shipment risks were calculated. The results indicate that the PWR spent fuel shipment risks are somewhat lower than those previously estimated. The cumulative and maximum annual normal, or incident-free, risks associated with the shipment of Taiwanese research reactor spent fuel is a factor of 10 lower than that for PWR fuel, and the cumulative and maximum annual accident radiological risks are a factor of about 2.2 lower than that for PWR spent fuel. As a result, the port risks are about a factor of 10 larger than the risk of overland transport. All of the risks calculated are small. The PWR risk values are similar to those judged by the NRC to be small enough not to warrant increased stringency in regulations. The Taiwanese research reactor spent fuel shipment risk values are smaller yet. 51 refs., 22 tabs

  5. Evaluation of the Three Mile Island Unit 2 reactor building decontamination process

    International Nuclear Information System (INIS)

    Dougherty, D.; Adams, J.W.

    1983-08-01

    Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams. Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings. The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61. It appears that much of the Epicor-II ion-exchange resin being disposed of in commerical land burial will be Class B and require stabilization if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B. Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation

  6. Evaluation of the Three Mile Island Unit 2 reactor building decontamination process

    Energy Technology Data Exchange (ETDEWEB)

    Dougherty, D.; Adams, J. W.

    1983-08-01

    Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams. Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings. The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61. It appears that much of the Epicor-II ion-exchange resin being disposed of in commerical land burial will be Class B and require stabilization if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B. Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation.

  7. Accidents of loss of flow for the ETTR-2 reactor; deterministic analysis

    International Nuclear Information System (INIS)

    El-Messiry, A.M.

    2000-01-01

    The main objective for reactor safety is to keep the fuel in a thermally safe condition with adequate safety margins during all operational modes (normal-abnormal and accidental states). To achieve this purpose an accident analysis of different design base accident (DBA) as loss of flow accident (LOFA), is required assessing reactor safety. The present work concerns this transients applied to Egypt Test and Research Reactor ETRR-3 (new reactor). An accident analysis code FLOWTR is developed to investigate the thermal behaviour of the core during such flow transients. The active core is simulated by two channels: 1 - hot channel (HC), and 2 - average channel (AC) representing the remainder of the core. Each channel is divided into four axial sections. The external loop, core plenums, and core chimney are simulated by different dynamic loops. The code includes modules for pump cast down, flow regimes, decay heat, temperature distributions, and feedback coefficients. FLOWTR is verified against results from RETRAN code, THERMIC code and commissioning tests for null transient case. The comparison shows a good agreement. The study indicates that for LOFA transients, provided the scram system is available, the core is shutdown safely by low flow signal (496.6 kg/s) at 1.4 s were the HC temperature reaches the maximum value, 45.64 o C after shutdown. On the other hand, if the scram system is unavailable, and at t = 47.33 s, the core flow decreases to 67.41 kg/s, the HC temperature increases to 164.02 o C, and the HC clad surface heat flux exceeds its critical value of 400.00 W/cm 2 resulting of fuel burnout. (author)

  8. Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor

    International Nuclear Information System (INIS)

    Čerba, Štefan; Vrban, Branislav; Lüley, Jakub; Dařílek, Petr; Zajac, Radoslav; Nečas, Vladimír; Haščik, Ján

    2014-01-01

    Highlights: • Verification of the MCNPX, HELIOS and SCALE codes. • MOX and ceramic fuel assembly. • Gas-cooled fast reactor. • Burnup calculation. - Abstract: The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice

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

    Science.gov (United States)

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

    2013-03-10

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

  10. Improved Emergency Preparedness For Management Of The Food chain Via Stakeholder Involvement: Belgian and European Experience

    Energy Technology Data Exchange (ETDEWEB)

    Hardeman, Frank; Carle, Benny [SCK.CEN, the Belgian Nuclear Research Centre, Boeretang 200, 2400 Mol (Belgium); Turcanu, Catrinel [Universite Libre de Bruxelles, Av. F. Roosevelt 50, 1050 Brussels (Belgium); Vandecasteele, Christian [FANC, Federal Agency for Nuclear Control, Ravensteinstraat 36, 1000 Brussels (Belgium)

    2006-07-01

    Initiatives involving stakeholder engagement have gained increasing importance in sustainable decision making for many risk-related issues. This paper describes a Belgian experience within a European context related to food management options in the event of a radioactive contamination of the food chain. Under the auspices of the European Commission's 5. Framework Programme, the F.A.R.M.I.N.G. (F.A.R.M.I.N.G. 2000) project (co-ordinated by H.P.A.) a stakeholder network was established in a number of European countries, following a successful approach originally adopted in the UK. In a comparable approach, national working groups were thus established in Belgium, Finland, France and Greece in order to organise stakeholder panels and to discuss the outcomes of scientific and technical research related to management options for the food chain. The results of these panels were exchanged between participating Member States and on a wider international basis at the W.I.S.D.O.M.2. workshop in 2003. The F.A.R.M.I.N.G. project had many achievements and there were also several important lessons learned for Belgium (Vandecasteele et al., 2005): Firstly, many stakeholders showed a real interest in tackling problems relating to food chain contamination; Secondly, the Belgian agricultural system is very intensive and technically and economically optimised, making many of the options envisaged difficult to implement; thirdly, the applicability of management options is also limited by political and legal issues (e.g. competencies, environmental legislation), operational constraints (e.g. waste treatment, supplies of materials), societal and ethical aspects (e.g. milk disposal to sea, animal welfare), and economics (e.g. who pays the intervention cost?); fourthly, there is a now a greater awareness of these problems in both the food production sector and among the experts involved in emergency management; Fifthly, increased attention is now given in Belgium to the medium and

  11. Improved Emergency Preparedness For Management Of The Food chain Via Stakeholder Involvement: Belgian and European Experience

    International Nuclear Information System (INIS)

    Hardeman, Frank; Carle, Benny; Turcanu, Catrinel; Vandecasteele, Christian

    2006-01-01

    Initiatives involving stakeholder engagement have gained increasing importance in sustainable decision making for many risk-related issues. This paper describes a Belgian experience within a European context related to food management options in the event of a radioactive contamination of the food chain. Under the auspices of the European Commission's 5. Framework Programme, the F.A.R.M.I.N.G. (F.A.R.M.I.N.G. 2000) project (co-ordinated by H.P.A.) a stakeholder network was established in a number of European countries, following a successful approach originally adopted in the UK. In a comparable approach, national working groups were thus established in Belgium, Finland, France and Greece in order to organise stakeholder panels and to discuss the outcomes of scientific and technical research related to management options for the food chain. The results of these panels were exchanged between participating Member States and on a wider international basis at the W.I.S.D.O.M.2. workshop in 2003. The F.A.R.M.I.N.G. project had many achievements and there were also several important lessons learned for Belgium (Vandecasteele et al., 2005): Firstly, many stakeholders showed a real interest in tackling problems relating to food chain contamination; Secondly, the Belgian agricultural system is very intensive and technically and economically optimised, making many of the options envisaged difficult to implement; thirdly, the applicability of management options is also limited by political and legal issues (e.g. competencies, environmental legislation), operational constraints (e.g. waste treatment, supplies of materials), societal and ethical aspects (e.g. milk disposal to sea, animal welfare), and economics (e.g. who pays the intervention cost?); fourthly, there is a now a greater awareness of these problems in both the food production sector and among the experts involved in emergency management; Fifthly, increased attention is now given in Belgium to the medium and long

  12. PR-EDB: Power Reactor Embrittlement Data Base, Version 2. Revision 2, Program description

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-01-01

    Investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes Standard Review Plans (SRP`s) and Guides for license renewal can be greatly expedited by the use of a well-designed computerized data base. Also, such a data base is essential for the validation of embrittlement prediction models by researchers. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The current version of the PR-EDB contains the Charpy test data that were irradiated in 252 capsules of 96 reactors and consists of 207 data points for heat-affected-zone (HAZ) materials (98 different HAZ), 227 data points for weld materials (105 different welds), 524 data points for base materials (136 different base materials), including 297 plate data points (85 different plates), 119 forging data points (31) different forging), and 108 correlation monitor materials data points (3 different plates). The data files are given in dBASE format and can be accessed with any computer using the DOS operating system. ``User-friendly`` utility programs are used to retrieve and select specific data, manipulate data, display data to the screen or printer, and to fit and plot Charpy impact data. The results of several studies investigated are presented in Appendix D.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hellstrand, E; Andersson, T L; Brunfelter, B; Kockum, J; Londen, S O; Tiren, L I

    1965-12-15

    In a first part of this report, published as AE-195, an account was given of critical mass determinations and measurements of flux distribution and reaction ratios in the first assemblies of the fast zero power reactor FR0. This second part of the report deals with various investigations involving the measurement of reactivity. Control rod calibrations have been made using the positive period, the inverse multiplication, the rod drop and the pulsed source techniques, and show satisfactory agreement between the various methods. The reactivity worths of samples of different materials and different sizes have been measured at the core centre. Comparisons with perturbation calculations show that the regular and adjoint fluxes are well predicted in the central region of the core. The variation in the prompt neutron life-time with reactivity has been studied by means of the pulsed source and the Rossi-{alpha} techniques. Comparison with one region calculations reveals large discrepancies, indicating that this simple model is inadequate. Some investigations of streaming effects in an empty channel in the reactor and of interaction effects between channels have been made and are compared with theoretical estimates. Measurements of the reactivity worth of an air gap between the reactor halves and of the temperature coefficient are also described in the report. The work has been performed as a joint effort by AB Atomenergi and the Research Institute of National Defence.

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

    Science.gov (United States)

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

    2002-01-15

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

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

    Science.gov (United States)

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

    2018-01-01

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

  16. Study of steam, helium and supercritical CO2 turbine power generations in prototype fusion power reactor

    International Nuclear Information System (INIS)

    Ishiyama, Shintaro; Muto, Yasushi; Kato, Yasuyoshi; Nishio, Satoshi; Hayashi, Takumi; Nomoto, Yasunobu

    2008-01-01

    Power generation systems such as steam turbine cycle, helium turbine cycle and supercritical CO 2 (S-CO 2 ) turbine cycle are examined for the prototype nuclear fusion reactor. Their achievable cycle thermal efficiencies are revealed to be 40%, 34% and 42% levels for the heat source outlet coolant temperature of 480degC, respectively, if no other restriction is imposed. In the current technology, however, low temperature divertor heat source is included. In this actual case, the steam turbine system and the S-CO 2 turbine system were compared in the light of cycle efficiency and plant cost. The values of cycle efficiency were 37.7% and 36.4% for the steam cycle and S-CO 2 cycle, respectively. The construction cost was estimated by means of component volume. The volume became 16,590 m 3 and 7240 m 3 for the steam turbine system and S-CO 2 turbine system, respectively. In addition, separation of permeated tritium from the coolant is much easier in S-CO 2 than in H 2 O. Therefore, the S-CO 2 turbine system is recommended to the fusion reactor system than the steam turbine system. (author)

  17. Model study of an automatic controller of the IBR-2 pulsed reactor

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Popov, A.K.

    2007-01-01

    For calculation of power transients in the IBR-2 reactor a special mathematical model of dynamics taking into account the discontinuous jump of reactivity by an automatic controller with the step motor is created. In the model the nonlinear dependence of the energy of power pulse on the reactivity and the influence of warming up of the reactor on the reactivity by means of introduction of a nonlinear feedback 'power-pulse energy - reactivity' are taken into account. With the help of the model the transients of relative deviation of power-pulse energy are calculated at various (random, mixed and regular) reactivity disturbances at the reactor mean power 1.475 MW. It is shown that to improve the quality of processes the choice of such regular values of parameters of the automatic controller is expedient, at which the least effect of smoothing of a signal acting on an automatic controller and the least speed of an automatic controller are provided, and the reduction of efficiency of one step of the automatic controller and introduction of a five-percent dead space are also expedient

  18. In-reactor creep of zirconium-2.5 wt% niobium at 570 K

    International Nuclear Information System (INIS)

    Coleman, C.E.; Causey, A.R.; Fidleris, V.

    1976-01-01

    The effect of fast neutron flux at 570 K on the creep rate of specimens of zirconium-2.5 wt% niobium alloy taken from tubes in various metallurgical conditions has been measured using both constant load tensile creep machines and bent-beam stress relaxation. Creep rates calculated from stress relaxation fit on the trend line for the constant load creep data. Between 114 MPa and 450 MPa the creep rate is proportional to neutron flux. The creep rate of specimens from the longitudinal direction is about twice that of specimens from the circumferential direction of a tube. This anisotropy in creep strength is attributed partly to crystallographic texture and partly to deformation substructure. Cold-work is detrimental to in-reactor creep strength; as-extruded material has higher creep strength. In cold-worked material at stresses below 100 MPa the stress exponent, n, is about 1; n gradually increases with stress being about 10 at 525 MPa and about 100 at 660 MPa. In laboratory tests, rupture ductility correlates inversely with n; the lower n the higher the ductility. In-reactor tests support this correlation thus pressure tubes in CANDU reactors, operating at 117 MPa where n approximately 1, should have good ductility. (Auth.)

  19. A Study of the Temperature Distribution in UO{sub 2} Reactor Fuel Elements

    Energy Technology Data Exchange (ETDEWEB)

    Devold, I

    1968-05-15

    Thermal conductivity is one of the most important properties of nuclear reactor fuels. Accurate knowledge of this property is vital because, among other things, it determines the maximum power that can be taken out of the fuel element per unit length of the material without exceeding the safety limits of the fuel elements. This report consists of a study of the thermal behaviour of uranium dioxide in the form of reactor fuel. The experimental part of the report describes measurements performed at the OECD Halden Reactor Project, Halden, Norway. The experiment was originally set up in order to measure the temperature at the center of a UO{sub 2} fuel element as a function of element power, in order to determine the safe operation limit of the fuel assembly. However, in analysing the data obtained, very interesting thermal conductivity values were obtained and comparison with existing correlations could be performed. This comparison shows that a certain agreement is obtained between the measured data at Halden and a theory published by J.L. Bates in 1961, which predicts an increase in the thermal conductivity above 1500 deg C. The data obtained below 1300 deg C are also in good agreement with measurements performed by Vogt, Grandell and Runfors in 1964. The report contains a mathematical description of the heat transfer mechanisms in cylindrical fuel elements. The model is coded in FORTRAN IV-code and referred to as FUELTEMP.

  20. Study on Reactor Physics Characteristic of the PWR Core Using UO2

    International Nuclear Information System (INIS)

    Tukiran Surbakti

    2009-01-01

    Study on reactor physics characteristic of the PWR core using UO 2 fuel it is necessary to be done to know the characteristic of geometry, condition and configuration of pin cell in the fuel assembly Because the geometry, configuration and condition of the pin cell in fuel core determine the loading strategy of in-core fuel management Calculation of k e ff is a part of the neutronic core parameter calculation to know the reactor physics characteristic. Generally, core calculation is done using computer code starts from modelling one unit fuel lattice cell, fuel assembly, reflector, irradiation facility and until core reactor. In this research, the modelling of pin cell and fuel assembly of the PWR 17 ×17 is done homogeneously. Calculation of the k-eff is done with variation of the fuel volume fraction, fuel pin diameter, fuel enrichment. The calculation is using by NITAWL and CENTRM, and then the results will be compared to KENOVI code. The result showed that the value of k e ff for pin cell and fuel assembly PWR 17 ×17 is not different significantly with homogenous and heterogenous models. The results for fuel volume fraction of 0.5; rod pitch 1.26 cm and fuel pin diameter of 9.6 mm is critical with burn up of 35,0 GWd/t. The modeling and calculation method accurately is needed to calculation the core physic parameter, but sometimes, it is needed along time to calculate one model. (author)