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Sample records for br-2 reactor

  1. Reactor BR2

    International Nuclear Information System (INIS)

    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

  2. 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.

  3. 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.

  4. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    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

  5. BR2 Reactor: Introduction

    International Nuclear Information System (INIS)

    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 3He, 6Li 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

  6. BR2 reactor neutron beams

    International Nuclear Information System (INIS)

    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)

  7. BR2 Reactor: Irradiation of Fusion Materials

    International Nuclear Information System (INIS)

    In collaboration with the EFDA (European Fusion Development Agreement), SCK-CEN irradiates several materials in the BR2 reactor at different temperatures and up to different doses to study their mechanical and physical properties during and after irradiation. Those materials are candidates for the construction of different parts of the ITER fusion reactor and of the long-term DEMO (DEMOnstration) reactor. The objectives of research performed at SCK-CEN are to irradiate up to 2 dpa RAFM (Reduced Activity Ferritic Martensitic) steels joints and RAFM ODS (Oxide Dispersion Strengthening) at 300 degrees Celsius; to build and test an experimental rig to perform in-situ creep-fatigue tests under neutron irradiation and its out-pile equipment and to design a new irradiation basket to irradiate in BR2 copper/stainless steel joints and RAFM specimens with implanted helium at low dose

  8. BR2 reactor: medical and industrial applications

    International Nuclear Information System (INIS)

    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 1015 n/cm2.s), the BR2 reactor is considered as a major facility through its contribution for a continuous supply of products such 99Mo (99mTc), 131I, 133Xe, 192Ir, 186Re, 153Sm, 90Y, 32P, 188W (188Re), 203Hg, 82Br, 41Ar, 125I, 177Lu,89Sr, 60Co, 169Yb, 147Nd, 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

  9. BR2 reactor: medical and industrial applications

    International Nuclear Information System (INIS)

    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 1015 n/cm2.s), the BR2 reactor is considered as a major facility through its contribution for a continuous supply of products such as 99Mo (99mTc), 131I, 133Xe, 192Ir, 186Re, 90Sm, 131Y, 32P, 188W (188Re), 203Hg, 82Br, 41Ar, 125I, 177Lu, 89Sr, 60Co, 169Yb, 147Nd, ... 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

  10. Material test reactor fuel research at the BR2 reactor

    International Nuclear Information System (INIS)

    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

  11. Optimized Control Rods of the BR2 Reactor

    International Nuclear Information System (INIS)

    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.

  12. 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.

  13. Ongoing refurbishment activities and strategy for the future operation of the BR2 reactor

    International Nuclear Information System (INIS)

    The operation of the BR2 reactor with its second Be-matrix is foreseen up to mid-1995 or mid-1996. A life extension for another 15 years is envisaged considering programmatic, financial and technical aspects. At present, the second phase of the refurbishment programme is being executed. The major activities of this programme can be grouped under two headings: safety reassessment and ageing issues. The expected outcome end '93 is an assessment report defining extent, choosen options, prioritized activities, budget and a tentative planning for the preparation and execution of the refurbishment. These aspects together with the prospects of possible cooperation with other parties for the refurbishment programme and the future operation of BR2 will be evaluated by the CEN/SCK Board who has to take a decision early in 1994. Various scenarios are now being considered and evaluated for the refurbishment and the future BR2 operation regime. (author)

  14. Mixed core management: Use of 93% and 72% enriched uranium in the BR2 reactor

    International Nuclear Information System (INIS)

    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% 235U 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% 235U 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% 235U 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)

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

    International Nuclear Information System (INIS)

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

  16. Production of Radioisotopes and NTD-Silicon in the BR2 Reactor

    International Nuclear Information System (INIS)

    The BR2 reactor is a multipurpose 100 MWth high flux 'Materials Testing Reactor' operated by the Belgian Nuclear Research Centre (SCK·CEN) in which various research and commercial programmes are performed. The commercial activities such as radioisotope production and silicon doping have been actively developed since the early 1990s to generate additional revenues. Currently, they represent a significant contribution to the reactor operating costs and are carried out in accordance with a 'Quality System' that has been certified to the requirements of the ''EN ISO 9001:2000'' in December 2006. Due to its operating flexibility, its reliability and its production capacity, the BR2 reactor is considered as a major facility for these commercial activities worldwide. The availability of thermal neutron fluxes up to 1015 cm-2s-1 allows the production of a wide range of radioisotopes for various applications in nuclear medicine, industry and research such as 99Mo (99mTc), 131I, 133Xe, 192Ir, 75Se 186Re, 153Sm, 169Er, 90Y, 32P, 188W (188Re), 203Hg, 82Br, 79Kr, 41Ar, 125I, 177Lu, 117mSn,89Sr, 169Yb, 147Nd, etc. Some irradiation devices allow the loading and unloading of irradiated targets during the operation of the reactor. Hot-cells and storage facilities are available to prepare and organize the shipment of the irradiated targets to dedicated processing facilities. In the frame of the current 99Mo/99mTc global shortage, new dedicated irradiation devices have been installed in April 2010 to increase the 99Mo production capacity by 50%. Special efforts have also been made to develop the production of therapeutic radioisotopes as 177Lu which is supplied by both direct and indirect routes. Neutron Transmutation Doping (NTD) Silicon activities for the semiconductor industry started at SCK·CEN in 1992 with the commissioning of SIDONIE, a single channel light water device that is located in a 200 mm diameter beryllium channel within the reactor pressure vessel. Its design

  17. LTA'S manufacturing for JHR fuel qualification program in BR2 reactor

    International Nuclear Information System (INIS)

    In the frame of the JHR fuel qualification program, CEA awarded to AREVA-CERCA an order for the manufacture and delivery of twelve Lead Test Assemblies. These assemblies are made in MEU U3Si2 type fuel plates using AlFeNi cladding. They are dedicated to experimental tests performed in the BR2 reactor, with the purpose of demonstrating the good behaviour of the fuel elements under operating conditions representative of the JHR normal operation. The main challenges were to manufacture new designed fuel assemblies with specific constraints on the geometry and on the manufacturing margins, due to the high performances required for the JHR reactor; as well as to deal with the insertion of burnable poisons in the U3Si2 type fuel plates. This paper is focused on manufacturing developments performed along the project so as to finally successfully produce the LTA's fuel elements. (author)

  18. Transients and safety testing of LMFBR fuel pins in the reactor BR2

    International Nuclear Information System (INIS)

    Testing of the behaviour of LMFBR fuel pins under operational transients has been performed in the reactor BR2 at S.C.K./C.E.N.-Mol (Belgium) since 1981 in the framework of the DEBENE programme ''SNR-Betriebstransienten-experimente''. A special purpose sodium loop, called ''VIC'', has therefore been developed to allow off-nominal and transient experiments on single fuel pins under realistic fast reactor operating conditions. Two basic types of tests can be run, either separately or simultaneously: fission power alteration, e.g. steady overpower runs, power cycling and fast transient overpower (TOP); mismatch of the sodium cooling, e.g. operation with reduced sodium flow and transient loss of flow (LOF). The loop allows the loading and testing of pre-irradiated fuel pins. In the field of safety oriented tests, the programme ''MOL 7 C'' investigates the LMFBR fuel element behaviour under locally blocked cooling conditions and the possible failure propagation. The work is jointly carried out by the Karlsruhe center KfK (FRG) and S.C.K./C.E.N.-Mol (Belgium). The related in-pile tests in the reactor BR2 have started in 1977 and are performed in a fully integrated sodium loop. The test section contains a 30-rod bundle with fresh or pre-irradiated fuel pins. A local porous blockage within the fuel bundle initiates severe local damage to the central rods. Important informations are obtained with respect to the problems of pin to pin propagation and the long term behaviour of a fuel subassembly with defect pins. The MOL 7 C loop system can also be used to run operational transients on a fuel bundle with representative fuel pins. The paper describes the irradiation devices VIC and MOL 7 C from their technological point of view and depicts their field of testing applications. Also the major experiments already performed and relevant irradiation data are reviewed

  19. The BR2-material testing reactor and its major contribution to the reactor material, fuel and safety research

    International Nuclear Information System (INIS)

    The BR2 was shutdown at the end of June 1995 for a programme of extensive refurbishment after more than 30 years utilization. The beryllium matrix was replaced and the aluminum vessel inspected and requalified for the envisaged 15 years life extension. Other aspects of the refurbishment programme were aimed at reliability and availability of the installations, safety of operation and compliance with modem safety standards. The reactor was restarted in April 1997. This paper deals with aspects of this refurbishment in general as well as the ongoing experimental projects in the areas of reactor material, fuel behaviour and safety research. (author)

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

    International Nuclear Information System (INIS)

    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 235U 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 1021 n/cm2) 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 242Pu 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

  1. Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. R. [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); Van den Branden, G. [SCK CEN (Belgium); Kalcheva, S. [SCK CEN (Belgium); Sikik, E. [SCK CEN (Belgium); Koonen, E. [SCK CEN (Belgium)

    2015-12-01

    BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water (Figure 1). The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident.

  2. SoLid: Search for Oscillation with a 6Li Detector at the BR2 research reactor

    CERN Document Server

    Michiels, Ianthe

    2016-01-01

    In the past decades, various nuclear reactor neutrino experiments have measured a deficit in the flux of antineutrinos coming from the reactor at short reactor-detector distances, when compared to theoretical calculations. One of the experiments designed to investigate this reactor antineutrino anomaly is the SoLid experiment. It uses the compact BR2 research reactor from the SCK-CEN in Mol, Belgium, to perform reactor antineutrino flux measurements at very short baseline. These proceedings discuss the general detection concepts of the SoLid experiment and its novel detector technology. The performance of the SoLid design is demonstrated with some results of the analysis of the data gathered with the experiment's first large scale test module, SM1.

  3. Checking of the Leakage Rate from the Leakproof Building of the BR-2 Reactor

    International Nuclear Information System (INIS)

    After a brief description of the containment building of the BR-2 reactor, the paper defines the procedure for checking its leakage rate, quoting and analysing the observations made between January 1962 and December 1966. The leakage rate F is measured from the total-pressure variation in the building with a correction for temperature change; the correction for changes in relative humidity is negligible in comparison with that for temperature. The change in temperature is obtained from the readings of 24 alcohol thermometers suitably distributed throughout the containment building. With a nominal relative pressure of 230 mmHg we have: F = (1.7 ± 0.3) x 10-3/day which is 2.4 or 1.4 times the permissible leakage rate, depending on whether the leakage flow is laminar or turbulent. T o make this measurement, the building was taken out of service for two periods of 4 days and 3 days in February and April 1962. Maintenance of the leakage rate is checked by leak-tightness tests at the nominal relative pressure of 230 mmHg (building out of use for 12-16 h). Between August 1962 and July 1965, seven tests were carried out. Two of these tests indicated a high leakage rate (168 x 10-3/day and 25 x 10-3/day) owing to a fault in the hermetic joint of a ventilation shut-off valve. In the remaining five cases, the leakage rate had an average value of (2.5 ± 0.6 ) x 10-3/day with extreme values of 0.3 x 10-3/day and 3.5 x 10-3/day. Checking of the fixed and mobile leak-prevention devices can be carried out without hampering operation by regular visual inspections and by tests of operation at each cycle shutdown (∼ 1/month); in this way any faults liable to lead to a high leakage rate can be detected in the shortest possible time. (author)

  4. Qualification of the on-line power determination of fuel elements in irradiation devices in the BR2 reactor

    International Nuclear Information System (INIS)

    Fuel irradiation tests require an on-line monitoring of the fuel power. In the BR2 reactor, this is performed by continuously measuring the enthalpy change in the coolant of the irradiation device and complementing this information with data on power losses, heating of structure parts and spatial power profiles from mock-up test experiments and from calculations. Since a few years Monte Carlo codes (MCNP) are used, describing the BR2 core in great detail for every reactor cycle with its specific core load, yielding not only reliable relative values, but also calculated absolute local power values in agreement with data from PIE analyses. Several methods were conceived to combine the experimental and calculated data for the on-line calculation of the local linear power in the fuel elements; their internal consistency and the consistency with gamma spectroscopy data and data from radiochemical fission product analysis was checked. The data show that fuel irradiations in BR2 can be performed in a well-controlled way, with an accurate and reliable on-line follow-up of the fuel power. (author)

  5. Fuel characteristics needed for optimal operation of the BR2 reactor

    International Nuclear Information System (INIS)

    The standard BR2 fuel element contains 400 g 235U under the form of UAlx with burnable absorbers homogeneously mixed into the fuel meat. The uranium is highly enriched with a density of ∼1.30 g U/cm3. This fuel element was developed in the early seventies to satisfy the irradiation conditions required by many experimental programmes: large reactivity available, cycle length, hard neutron spectrum, limited motion of the control rods during the cycle thereby stabilizing the irradiation conditions. Another benefit is the reduction of the fuel consumption by increasing the burnup at discharge. BR2 has recently been restarted after the completion of an important refurbishment programme. Future utilization will again be concentrated on engineering R and D in the field of nuclear fuels, materials and safety, and on radioisotope production. Therefore the required irradiation conditions and the corresponding fuel characteristics remain essentially the same as in the past. (author)

  6. Fuel characteristics needed for optimal operation of the BR2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koonen, E.; Beeckmans, A.; Gubel, P. [SCK-CEN, B-2400 Mol (Belgium)

    1998-07-01

    The standard BR2 fuel element contains 400 g {sup 235}U under the form of UAl{sub x} with burnable absorbers homogeneously mixed into the fuel meat. The uranium is highly enriched with a density of {approx}1.30 g U/cm{sup 3}. This fuel element was developed in the early seventies to satisfy the irradiation conditions required by many experimental programmes: large reactivity available, cycle length, hard neutron spectrum, limited motion of the control rods during the cycle thereby stabilizing the irradiation conditions. Another benefit is the reduction of the fuel consumption by increasing the burnup at discharge. BR2 has recently been restarted after the completion of an important refurbishment programme. Future utilization will again be concentrated on engineering R and D in the field of nuclear fuels, materials and safety, and on radioisotope production. Therefore the required irradiation conditions and the corresponding fuel characteristics remain essentially the same as in the past. (author)

  7. First results of the deployment of a SoLid detector module at the SCK-CEN BR2 reactor

    CERN Document Server

    Ryder, Nick

    2015-01-01

    The SoLid experiment aims to resolve the reactor neutrino anomaly by searching for electron-to-sterile anti-neutrino oscillations. The search will be performed between 5.5 and 10 m from the highly enriched uranium core of the BR2 reactor at SCK-CEN. The experiment utilises a novel approach to anti-neutrino detection based on a highly segmented, composite scintillator detector design. High experimental sensitivity can be achieved using a combination of high neutron-gamma discrimination using 6 LiF:ZnS(Ag) and precise localisation of the inverse beta decay products. This compact detector system requires limited passive shielding as it relies on spacial topology to determine the different classes of backgrounds. The first full scale, 288 kg, detector module was deployed at the BR2 reactor in November 2014. A phased three tonne experimental deployment will begin in the second half of 2016, allowing a precise search for oscillations that will resolve the reactor anomaly using a three tonne detector running for thr...

  8. Validation of SCALE4.4a for Calculation of Xe-Sm Transients After a Scram of the BR2 Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-15

    The aim of this report is to validate the computational modules system SCALE4.4a for evaluation of reactivity changes, macroscopic absorption cross sections and calculations of the positions of the Control Rods during their motion in Xe-Sm transient after a scram of the BR-2 reactor. The rapid shutting down of the reactor by inserting of negative reactivity by the Control Rods is known as a reactor scram. Following reactor scram, a large xenon and samarium buildup occur in the reactor, which may appreciably affect the multiplication factor of the core due to enormous neutron absorption. The validation of the calculations of Xe-Sm transients by SCALE4.4a has been performed on the measurements of the positions of the Control Rods during their motion in Xe-Sm transients of the BR-2 reactor and on comparison with the calculations by the standard procedure XESM, developed at the BR-2 reactor. A final conclusion is made that the SCALE4.4a modules system can be used for evaluation of Xe-Sm transients of the BR-2 reactor. The utilization of the code is simple, the computational time takes from few seconds.

  9. Optimization strategies for sustainable fuel cycle of the BR2 Reactor

    International Nuclear Information System (INIS)

    The objective of the present study is to achieve a sustainable fuel cycle in a long term of reactor operation applying advanced in-core loading strategies. The optimization criteria concern mainly enhancement of nuclear safety by means of reactivity margins and minimization of the operational fuel cycle cost at a given (constant) power level and same or longer cycle length. An important goal is also to maintain the same or to improve the experimental performances. Current developments are focused on optimization of control rods localization; optimization of fresh and burnt fuel assemblies in-core distribution; optimization of azimuth and axial fuel burn up strategies, including fuel assembly rotating and flipping upside down. (authors)

  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)

    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. Practical limitations for the release of fission products during the operation of a research reactor: a case study of BR2

    Energy Technology Data Exchange (ETDEWEB)

    Joppen, F. [Health Physics and Safety Department, SCK-CEN, B-2400 Mol (Belgium)

    1998-07-01

    Failure of the cladding of a fuel element is an event occurring from time to time while operating a research reactor. As a consequence, fission products are released in the primary circuit of the reactor. This contamination means no direct hazard for the workers or for the environment in case the reactor has a closed primary circuit. The operator can decide to continue the irradiation to finish a scientific experiment or a commercial isotope production program. However, the operator cannot prolong the cycle regardless the concentration fission products in the primary loop. Beside the limitations imposed by the regulatory authorities, ALARA considerations should be taken into account. An untimely stop of the reactor can have serious financial consequences and prolonged operation causes higher radiation doses. This paper gives an overview of decision process applied in case of detection of fission products in the primary circuit of BR2. (author)

  12. Practical limitations for the release of fission products during the operation of a research reactor: a case study of BR2

    International Nuclear Information System (INIS)

    Failure of the cladding of a fuel element is an event occurring from time to time while operating a research reactor. As a consequence, fission products are released in the primary circuit of the reactor. This contamination means no direct hazard for the workers or for the environment in case the reactor has a closed primary circuit. The operator can decide to continue the irradiation to finish a scientific experiment or a commercial isotope production program. However, the operator cannot prolong the cycle regardless the concentration fission products in the primary loop. Beside the limitations imposed by the regulatory authorities, ALARA considerations should be taken into account. An untimely stop of the reactor can have serious financial consequences and prolonged operation causes higher radiation doses. This paper gives an overview of decision process applied in case of detection of fission products in the primary circuit of BR2. (author)

  13. Phase diagram study of CaBr2-SrBr2 binary system

    International Nuclear Information System (INIS)

    Hydride ion conducting electrolytes are used in the electrochemical sensors for detecting hydrogen in liquid sodium coolant of fast reactors. In this connection several combinations of alkali/alkaline earth metal halides mixed with alkali/alkaline earth metal hydrides were investigated for using them as hydride ion conducting electrolytes. Currently, a mixture of CaBr2-CaH2 is used as the solid electrolyte in electrochemical sensors. With a view to develop and understand electrolyte characteristics of CaBr2-SrBr2-CaH2, system, the binary phase diagram of CaBr2-SrBr2 is being investigated in this work. Samples were prepared in dry argon atmosphere glove box using purified CaBr2 and SrBr2 salts covering the composition range of 0 to 100 mol% of CaBr2 in SrBr2 in steps of ∼10 mol%. Approximately 30 mg of samples were loaded inside hermetically closed iron capsules and were analysed by DTA under argon+4% hydrogen gas at controlled heating and cooling rates. Results have shown that the system shows appreciable terminal solid solution at both CaBr2 and SrBr2 rich sides. The system exhibits a eutectic reaction involving the terminal solid solutions at 553℃ and at ∼40 mol% CaBr2 in SrBr2. Further investigations are in progress. (author)

  14. 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)

    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

  15. Refurbishment of BR2 (Phases 4 and 5)

    International Nuclear Information System (INIS)

    The BR2 is a materials testing reactor and is 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. In phase 4 of the refurbishment programme, various activities were performed to allow reactor start-up. In phase 5, remaining refurbishment works were carried out as well as the extra studies and upgradings required by the licensing authorities. Major achievements in 1997 are described and discussed

  16. MTR fuel testing in BR2

    International Nuclear Information System (INIS)

    New fuel design for MTR 's requires to be qualified under representative conditions, that is geometry, neutron spectrum, heat flux and thermo hydraulic conditions. An irradiation device for fuel plates has been designed to derive the maximum benefit from the BR2 irradiation capacities. The fuel plates can be easily extracted from their support during a shutdown to undergo additional tests. One of these tests is the measurement of the thickness changes along the fuel plate. To that purpose, a facility in the reactor water pool has been designed to measure the fuel swelling with an accuracy of 5 μm using inductive probes. At SCK-CEN, the full range of destructive and non-destructive PIE can be performed, including γ-scanning, wet sipping, surface examination and other methods. (author)

  17. Interaction in ternary HgBr2-BaBr2-CsBr system

    International Nuclear Information System (INIS)

    Series of polythermal sections in ternary HgBr2-BaBr2-CsBr system are investigated by the methods of physicochemical analysis. In double systems restricting ternary HgBr2-BaBr2-CsBr system formation of CsHg2Br5, CsHgBr3, Cs2HgBr4, CsBa2Br4 and CsBa2Br5 compounds is detected. Projection of liquidus surface of ternary HgBr2-BaBr2-CsBr system on triangle of compositions is plotted. This projection consists of eight fields of the primary crystallization of phases: HgBr2, BaBr2, CsBr, CsHgBr5, CsHgBr3, Cs2HgBr4, Cs2BaBr4 and CsBa2Br5. Coordinates of nonvariant points are determined

  18. Refurbishment of BR2 (Phases 4 and 5)

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

    The BR2 is a materials testing reactor and is 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. In phase 4 of the refurbishment programme, various activities were performed to allow reactor start-up. In phase 5, remaining refurbishment works were carried out as well as the extra studies and upgradings required by the licensing authorities. Major achievements in 1997 are described and discussed.

  19. BR2: Some aspects of structural mechanics

    International Nuclear Information System (INIS)

    This article discusses some of the important aspects of structural mechanics of BR2, namely: the follow-up of the beryllium matrix and of the reactor vessel and the seismic qualification. According the licence, a follow up program for the beryllium matrix is mandatory. This inspection is necessary because of the swelling of beryllium during irradiation. Due to this swelling, the individual beryllium blocks make contact between each other. This results in mechanical stresses and, because beryllium is a brittle material, cracks. At regular intervals inspection are made to evaluate the evolution of the swelling and the cracks. The maximum allowed neutron fluence is 6.4 1022 fast neutrons (energy more than 1 MeV) per cm2 . After this time the matrix has to be replaced. This has been done already twice. During the replacement an inspection of the reactor pressure vessel must be made. Last inspection was performed in 1996, using ultrasonic and eddy current inspections. On this occasion a fracture mechanics calculation was made and the minimum allowed fracture toughness of material was determined. Since very little information on irradiated aluminium 5052-O is available, a number of samples were cut out of a second wall around the vessel. This aluminium had received nearly the fluence. Out of the samples test pieces (tensile and charpy) were made. A number of them were tested immediately, while the other was loaded in the reactor for accelerated irradiation. In this way a material follow up program was started. This program still continues. During the period safety reassessment the authorities requested a seismic qualification. It was decided to make a full dynamic calculation, with input a 0.1g zero period peak ground acceleration and a regulatory guide 1.60 spectrum. The installation can withstand this earthquake, considered as a safe shutdown earthquake. A few structural reinforcements were necessary. The main ones were the primary piping outside the containment

  20. The possibilities of application of experimental Kfk results from BR2 on SNR designs

    International Nuclear Information System (INIS)

    A review is given of the relevant results of the technological application for the SNR300 reactor, since the BR2 reactor has been used as a test facility for the material development. Special emphasis has been laid on the fuel pin behavior under the aspect of chemical and mechanical fuel-clad interaction and on the specification of the cladding in terms of high temperature mechanical behavior in the SNR 300 reactor. A systematic analysis of urgent research topics in BR2 test facility reactor is presented. (A.F.)

  1. BR2: The first year of operation after refurbishment

    International Nuclear Information System (INIS)

    The BR2 reactor has resumed operation in April 97 after an extensive refurbishment shutdown, which lasted for nearly two years. The yearly operation is presently limited to major cycles of 21 efpd, plus some short cycles for special programmes. The reactor is mainly used for irradiations in the framework of the following programs: qualification of MOX fuel at high burn-up, the PWR vessel surveillance program and associated modelling activities, the IASCC program focused on PVVR vessel intervals. The major irradiation device is the CALLISTO loop, simulating PWR conditions and comprising three in-pile sections. Additionally production activities are carried out: radio-isotopes and silicon doping. Irradiations for the surveillance programmes of beryllium and aluminum are underway; they concerns unirradiated and preirradiated samples, with various lead factors. Several refurbishment actions are still continuing, mainly: - continuation of the renewal of the process instrumentation, - extension of the BR2 DAS, - follow-up of the seismic qualification study, - follow-up of the PSA study: some detailed studies on supporting systems. A formalised training programme for the reactor operators has been launched. Special attention is given to the new reactor control desk and the emergency control panel outside of the containment building. A solution for the evacuation of the spent fuel has been adopted and is being implemented: reprocessing in La Hague

  2. Optical properties of CdBr2:Eu and CdBr2:Eu, Mn crystals

    International Nuclear Information System (INIS)

    Optical and luminescent properties of the CdBr:Eu and CdBr2:Eu, Mn crystals, grown through the Stockbarger-Bridgman method in evacuated quartz ampoules, are studied within the temperature range of 85-295 K. The results obtained are compared with spectral characteristics of the CdBr2 and CdBr2:Mn crystals. The band with the maximum about 254 nm, observed in the absorption spectra of mono- and polyactivated crystals of cadmium bromide, is attributed to the 4f7 -> 4f65d electron transitions in the Eu2+ ions. The manganese sensitized luminescence is identified by excitation of the CdBr2:Eu, Mn crystals by the light from the area of this band. The nature of the capture centers, responsible for thermostimulated fluorescence, and excitation mechanisms of recombination luminescence in the studied crystals are considered

  3. Scientific activities in support of the BR2 operation and irradiation programmes

    International Nuclear Information System (INIS)

    One of the major characteristics of the BR2 reactor is the fact that the core configuration is essentially variable. This allows to optimize the irradiation conditions of various experiments and to minimize the fuel consumption. In order to do that, BR2 has its own autonomous reactor physics cell. In order to allow for on-line measurements of the major irradiation parameters, BR2 has extended its own proven data acquisition system to serve this purpose. This system, called BIDASSE (for BR2 Integrated Data Acquisition System for Survey and Experiments), originally designed for the follow-up of all BR2 operational parameters, is since several years extensively used for experiments. The object rives of research at the BR2 are to evaluate and adjust provisional irradiation conditions by adjustments of the environment, axial and azimuthal positioning of the samples, global power level, ... ; to deliver reliable, well defined irradiation condition and fluence data during and after irradiation; to assist the designer of new irradiation devices by simulations and neutronic optimisations of design options and o provide the experimenters with accurate on-line information on the evolution of their ongoing irradiation projects

  4. Raman spectra of ZnBr2-based glasses

    International Nuclear Information System (INIS)

    Raman spectra of ZnBr2-KBr and ZnBr2-KBr-CaBr2 glasses contain strong bands at 60 cm-1 and 155 or 174 cm-1 and some weak bands between 200-300 cm-1. From the compositional dependence of the spectra and comparison with vibrational modes of molten mixtures and crystals, the 155 and 174 cm-1 bands are assigned to symmetric stretching modes of tetrahedra consisting of four bridging and four non-bridging bromines, respectively. It is revealed that tetrahedra of bridging bromines exist in the glasses even at the composition of so large amount of bromine that the theoretical number of non-bridging bromine per zinc is beyond 4. (author) 6 refs., 4 figs

  5. The BR2 refurbishment programme: achievements and two years operation feedback

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P.; Dekeyser, J.; Koonen, E.; Van der Auwera, J. [SCK/CEN, Belgian Nuclear Research Center, Mol (Belgium)

    1999-08-01

    The BR2 reactor was shutdown end of June 1995 for an extensive refurbishment after more than 30 years utilization. The beryllium matrix needed to be replaced and the aluminium vessel inspected for an envisaged 15 year life extension. Other aspects of the refurbishment programme aimed at the reliability and availability of the installations, safety of operation and compliance with modern safety standards. The reactor was started again in' April '97 and operated only for three cycles in 1997. These first irradiation cycles were intended as a demonstration of the safety and reliability of all components and systems after refurbishment. Also during the extended shutdowns non-critical refurbishment tasks were allowed to be continued and finalized. At the request of the Safety Authorities, some modifications and studies are still in progress without perturbation of the reactor operation. (author)

  6. 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.

  7. Qualification program for JHR fuel elements: Irradiation of the first JHR test assembly in the BR2-Evita loop

    International Nuclear Information System (INIS)

    An experimental program has been designed by CEA to qualify the behaviour of the JHR fuel under conditions representative of the reactor operating ones. This program uses the SCK.CEN facilities, irradiating JHR lead test elements in the BR2 reactor, inside its central channel which has been particularly arranged for this objective (Evita loop). As a first step in the program, a two cycle irradiation (4 weeks by cycle) started mid-July 2009 and ended mid-November (EVITA-1). After a cooling phase, this first JHR lead test element will be submitted to post-irradiation examination. The second JHR test element began its irradiation in the first quarter of 2010; its unloading is planned before the end of 2010, after 5 cycles in the BR2 reactor. The results of these two experiments are expected as input information for the Safety Authority Report. This paper presents the qualification program with the objectives assigned to each phase (irradiation, examination). A first interpretation of the irradiation data for the first element is presented, so as the information available on the progress of the following phases of the programme. (author)

  8. KARAKTERISASI DAN AKTIVITAS KATALITIK BERBAGAI VARIASI KOMPOSISI KATALIS Ni DAN ZnBr2 DALAM Γ-Al2O3 UNTUK ISOMERISASI DAN HIDROGENASI (R-(+-SITRONELAL

    Directory of Open Access Journals (Sweden)

    ED Iftitah

    2014-06-01

    -EELS. The catalysts spesific surface area and porosity determined by adsorption-desorption of dinitrogen at 77 K. Pore distribution and volume were determined by the desorption isotherm at P/Po ≥ 0.3. The result showed that there was correlation between the catalyst characteristics and catalytic activity to (R-(+-Citronellal isomerisation and hydrogenation product. The activity test were performed in a mini fixed bed reactor with 0.5 g of catalyst and 3 mL of (R-(+-Citronellal using N2 and/or H2 gas atmosphere in 5 and 24 hours at each temperature 90 and 120 oC. The catalyst composition of the choice of gas atmosphere and temperature greatly influenced the activity as well as the selectivity of isomerisation and hydrogenation product formation. The highest conversion was achieved for A3=Ni/ZnBr2/γ-Al2O3 (2:3 with complete conversion of (R-(+-Citronellal were obtained when it was running in 5 hours (4 hours N2 + 1 hour H2 at 90 oC and 24 hours (4 hours N2 + 20 hour H2 at 120 oC.

  9. Spin-orbit relaxation of Br ((2)P(sub 1/2))

    Science.gov (United States)

    Johnson, R. O.; Katapski, S. M.; Perram, G. P.; Roh, W. B.; Tate, R. F.

    Pulsed and steady-state photolysis experiments have been conducted to determine the rate coefficients for collisional deactivation of the spin-orbit excited state atomic bromine, Br ((2)P(sub 1/2)). Pulsed lifetime studies for quenching by Br2 and CO2 established absolute rate coefficients at room temperature of k(sub Br2) = 1.2 +/- 10(exp -12) and k(sub CO2) = 1.5 +/- 0.3 x 10(exp -11)/cc/molecule-s. Steady-state photolysis methods were used to determine the quenching rates for the rare gases, N2, O2, H2, D2, NO, NO2, N2O, SF6, CF4, CH4, CO, CO2, COS, SO2, H2S, HBr, HCl, and HI relative to that for Br2.

  10. Temperature dependence of rate constants of the reactions Br(Br*)+IBr → Br2+I

    International Nuclear Information System (INIS)

    Rate constant and their temperature dependence in the range from -25 to +50 deg C for reactions Br(2P3/2)+IBr → Br2+I(2P2/3) and Br*(2P1/2)+IBr → Br2+I(2P3/2) have been measured by the method of laser atomic-resonance spectroscopy using radiation of iodine and bromine lasers. It has been detected that at 300 K the values of k1 and k2 agree with the known ones, meanwhile with the temperature growth both constants increase, moreover, for k2 the temperature dependence is much stronger. It is shown that the value of deceleration of the rate of reaction between Br atom and IBr in case of its spin-orbital excitation is the function of the temperature, decreasing with the temperature increase. 14 refs.; 3 figs.; 2 tabs

  11. CaBr2 hydrolysis for HBr production using a direct sparging contactor

    International Nuclear Information System (INIS)

    We investigated a novel, continuous hybrid cycle for hydrogen production employing both heat and electricity. Calcium bromide (CaBr2) hydrolysis, which is endothermic, generates hydrogen bromide (HBr), and this is electrolysed to produce hydrogen. CaBr2 hydrolysis at 1050 K is endothermic with a 181.5 KJ/mol heat of reaction and the free energy change is positive at 99.6 kJ/mol. What makes this hydrolysis reaction attractive is both its rate and the fact that well over half the thermodynamic requirements for water-splitting free energy of ΔGT = 285.8 KJ/mol are supplied at this stage using heat rather than electricity. These experiments provide support for a second order hydrolysis reaction in CaBr2 forming a complex involving CaBr2 and CaO and the system appears to be: 3CaBr2 + H2O → (CaBr2)2.CaO + 2HBr. This reaction is highly endothermic and the complex also includes some water of hydration. COMSOLTM multi-physics modelling of sparging steam into a calcium bromide melt guided the design of an experiment using a mullite tube (ID 70 mm) capable of holding 0.3-0.5 kg (1.5-2.5 10-3 kmol) CaBr2 forming a melt with a maximum 0.08 m depth. Half of the experiments employed packings. Sparging steam at a steam rate of 0.02-0.04 mol/mol of CaBr2 per minute into this molten bath promptly yielded HBr in a stable operation that converted up to 19 mol% of the calcium bromide. The kinetic constant derived from the experimental data was kinetic constant was 2.17 10-12 kmol s-1 m-2 MPa-1 for the hydrolysis reaction. (authors)

  12. Thermodynamic and structural properties of high temperature solid and liquid EuBr2

    DEFF Research Database (Denmark)

    Rycerz, L.; Gadzuric, S.; Ingier-Stocka, E.; Berg, Rolf W.; Gaune-Escard, M.

    Heat capacity of solid and liq. EuBr2 was measured by differential scanning calorimetry in the temp. range 300-1100 K. The temp. and enthalpy of fusion were also detd. exptl. By combination of these results with the literature data on the entropy at 298.15 K, S(o,m) (EuBr2, s, 298.15 K) , and the...

  13. Proton and helium stopping cross sections in Cl2 and Br2

    International Nuclear Information System (INIS)

    Proton and 4He stopping powers for gaseous Cl2 and Br2 were measured in the energy ranges 50-750 keV and 100-1000 keV, respectively, and fitted with the semi-empirical Andersen-Ziegler formula. The peak energies are located near the minima of the oscillating structure as a function of the target atomic number Z2 recently observed by Gowda et al. The stopping-power ratios Ssub(He)/Ssub(p) for Cl2 and Br2 as a function of ion velocity show a similar behavior as for the adjacent inert gases Ar and Kr. The low-energy helium stopping cross sections were described by the power functions S=kEsup(p), whereby p values of 0.5 for Cl2 and 0.67 for Br2 were found. (orig.)

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

    International Nuclear Information System (INIS)

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

  15. Reactor Physics

    International Nuclear Information System (INIS)

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

  16. Reactor Physics

    International Nuclear Information System (INIS)

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

  17. 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.

  18. Estimation of steady-state and transcient power distributions for the RELAP analyses of the 1963 loss-of-flow and loss-of-pressure tests at BR2.

    Energy Technology Data Exchange (ETDEWEB)

    Dionne, B.; Tzanos, C. P. (Nuclear Engineering Division)

    2011-05-23

    To support the safety analyses required for the conversion of the Belgian Reactor 2 (BR2) from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, the simulation of a number of loss-of-flow tests, with or without loss of pressure, has been undertaken. These tests were performed at BR2 in 1963 and used instrumented fuel assemblies (FAs) with thermocouples (TC) imbedded in the cladding as well as probes to measure the FAs power on the basis of their coolant temperature rise. The availability of experimental data for these tests offers an opportunity to better establish the credibility of the RELAP5-3D model and methodology used in the conversion analysis. In order to support the HEU to LEU conversion safety analyses of the BR2 reactor, RELAP simulations of a number of loss-of-flow/loss-of-pressure tests have been undertaken. Preliminary analyses showed that the conservative power distributions used historically in the BR2 RELAP model resulted in a significant overestimation of the peak cladding temperature during the transient. Therefore, it was concluded that better estimates of the steady-state and decay power distributions were needed to accurately predict the cladding temperatures measured during the tests and establish the credibility of the RELAP model and methodology. The new approach ('best estimate' methodology) uses the MCNP5, ORIGEN-2 and BERYL codes to obtain steady-state and decay power distributions for the BR2 core during the tests A/400/1, C/600/3 and F/400/1. This methodology can be easily extended to simulate any BR2 core configuration. Comparisons with measured peak cladding temperatures showed a much better agreement when power distributions obtained with the new methodology are used.

  19. Direct reduction of some benzoic acids to alcohols via NaBH4-Br2

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    Direct reduction of seven benzoic acids to alcohols via sodium borohydride-bromine (NaBH4-Br2) reagent was developed. The isolated yields for the seven acids to reduce reached 60.6-90.1 %. This new synthesis route has the advantages of simple of application, low cost, mild nature, and satisfactory yields.

  20. Caveats for poly(methimazolyl)borate chemistry: the novel inorganic heterocycles [H2C(mt)2BR2]Cl (mt = methimazolyl; BR2 = BH2, BH(mt), 9-BBN).

    Science.gov (United States)

    Crossley, Ian R; Hill, Anthony F; Humphrey, Elizabeth R; Smith, Matthew K; Tshabang, Never; Willis, Anthony C

    2004-08-21

    Whilst frequently used for reactions of poly(methimazolyl)borates, dichloromethane is not an innocent solvent, but rather slowly forms heterocyclic salts [H(2)C(mt)(2)BR(2)]Cl, three examples of which (BR(2) = BH(2), BH(mt), 9-borabicyclononyl) have been structurally characterised to confirm the unprecedented B(NCS)(2)C connectivity. PMID:15306929

  1. Manganese dioxide causes spurious gold values in flame atomic-absorption readings from HBr-Br2 digestions

    Science.gov (United States)

    Campbell, W.L.

    1981-01-01

    False readings, apparently caused by the presence of high concentrations of manganese dioxide, have been observed in our current flame atomic-absorption procedure for the determination of gold. After a hydrobromic acid (HBr)-bromine (Br2) leach, simply heating the sample to boiling to remove excess Br2 prior to extraction with methyl-isobutyl-ketone (MIBK) eliminates these false readings. ?? 1981.

  2. Proton-stopping cross sections and mean excitation energies for gaseous Cl2 and Br2

    International Nuclear Information System (INIS)

    First measurements of proton stopping powers in gaseous Cl2 and Br2 are reported for the energy range between 50 and 750 keV with statistical and systematical errors of about 1% each. In the stopping-power maximum the experimental values are higher than those predicted by the Andersen-Ziegler tables; moreover, lower peak energies are found. Experimental shell corrections were deduced from the proton stopping cross sections and adjusted to the theoretical predictions of Bonderup, whereby higher-order Z1 correction terms were included. Within this procedure semiempirical values for the mean ionization potentials of 174 eV for Cl2 and 363 eV for Br2 were obtained

  3. Spin-wave resonance in the FeBr2 magnetic thin film

    International Nuclear Information System (INIS)

    The spin-wave resonance in the thin FeBr2 field-induced metamagnet in the paramagnetic phase with the (001) surfaces and at low temperatures is examined theoretically. It is found that the absorption spectrum is strongly affected by modifications of the surface exchange parameters, Also, the conditions for the appearance of various surface and bulk spin-wave features are discussed. (author)

  4. Electron-phonon scattering in HgBr2 and IBr graphite intercalation compounds

    International Nuclear Information System (INIS)

    The Hall coefficient and the dc resistivity in the basal plane and along the c axis of stage-2 IBr and stage-3 HgBr2 graphite intercalation compounds (GICs) were measured between 4.2 and 370 K. The resistivity of the IBr GIC indicates a metallic-like behaviour and a structural phase transition at 175 K. The Hall coefficient above 300 K of the HgBr2 GIC indicates deintercalation from an irreversible reduction of the carrier concentration. A study of the temperature dependence of the resistivity and Hall coefficient of the stage-2 IBr GIC above room temperature shows that the electron-out-of-plane phonon scattering plays a major role in this compound. The Hall coefficients are temperature independent below room temperature. The one-carrier model gives a value of 8.1x1026 m-3 for the IBr and a value of 7.9x1026 m-3 for the HgBr2 GIC for the carrier concentrations. (author)

  5. Photoluminescence properties of Sm2+-doped BaBr2 under hydrostatic pressure

    International Nuclear Information System (INIS)

    Photoluminescence spectra of Sm2+-doped BaBr2 have been measured under hydrostatic pressures up to 17 GPa at room temperature. In the low pressure range a red-shift of the broad 5d-4f transition of -145 cm-1/GPa is observed. From 5 to 8 GPa a phase mixture of the initial orthorhombic phase and the high-pressure monoclinic phase gives rise to two 5d-4f bands, which are strongly overlapping. Above 8 GPa the crystal is completely transformed to its high-pressure phase where two different Sm2+ sites exist, but only one broad 5d-4f transition is detected. It exhibits a red-shift of -36 cm-1/GPa. In addition, the line shifts of the 5D0→7FJ (J=0, 1, 2) transitions are investigated. Linear shifts of -19 cm-1/GPa for J=0, 2 and of -13 cm-1/GPa for J=1 are observed in the pressure range from 0 to 5 GPa. - Highlights: → Pressure-induced phase transition of BaBr2:Sm from orthorhombic to monoclinic. → BaBr2 phase transition is observed between 5 and 8 GPa. → Phase transition is monitored by photoluminescence spectroscopy. → Pressure-induced shift of the Sm2+ 5d-4f transition is analyzed.

  6. TRIP-Br2 promotes oncogenesis in nude mice and is frequently overexpressed in multiple human tumors

    Directory of Open Access Journals (Sweden)

    Peh Bee

    2009-01-01

    Full Text Available Abstract Background Members of the TRIP-Br/SERTAD family of mammalian transcriptional coregulators have recently been implicated in E2F-mediated cell cycle progression and tumorigenesis. We, herein, focus on the detailed functional characterization of the least understood member of the TRIP-Br/SERTAD protein family, TRIP-Br2 (SERTAD2. Methods Oncogenic potential of TRIP-Br2 was demonstrated by (1 inoculation of NIH3T3 fibroblasts, which were engineered to stably overexpress ectopic TRIP-Br2, into athymic nude mice for tumor induction and (2 comprehensive immunohistochemical high-throughput screening of TRIP-Br2 protein expression in multiple human tumor cell lines and human tumor tissue microarrays (TMAs. Clinicopathologic analysis was conducted to assess the potential of TRIP-Br2 as a novel prognostic marker of human cancer. RNA interference of TRIP-Br2 expression in HCT-116 colorectal carcinoma cells was performed to determine the potential of TRIP-Br2 as a novel chemotherapeutic drug target. Results Overexpression of TRIP-Br2 is sufficient to transform murine fibroblasts and promotes tumorigenesis in nude mice. The transformed phenotype is characterized by deregulation of the E2F/DP-transcriptional pathway through upregulation of the key E2F-responsive genes CYCLIN E, CYCLIN A2, CDC6 and DHFR. TRIP-Br2 is frequently overexpressed in both cancer cell lines and multiple human tumors. Clinicopathologic correlation indicates that overexpression of TRIP-Br2 in hepatocellular carcinoma is associated with a worse clinical outcome by Kaplan-Meier survival analysis. Small interfering RNA-mediated (siRNA knockdown of TRIP-Br2 was sufficient to inhibit cell-autonomous growth of HCT-116 cells in vitro. Conclusion This study identifies TRIP-Br2 as a bona-fide protooncogene and supports the potential for TRIP-Br2 as a novel prognostic marker and a chemotherapeutic drug target in human cancer.

  7. Synthesis, structure and Eu2+-doped luminescence properties of bromosilicate compound Ca3SiO4Br2

    International Nuclear Information System (INIS)

    The bromosilicate Ca3SiO4Br2 crystal has been grown, and this compound crystallizes in triclinic symmetry, space group P-1 (No. 2), with unit cell parameters: a=8.0051(18) Å, b=8.720(3) Å, c=11.749(3)Å, α=69.07(0)°, β=89.98(0)°, γ=75.46(0)°, and cell volume V=737.88(1 9 6)Å3, Z=3. The unit cell of the Ca3SiO4Br2 crystal is composed of the alternating layers of CaBr2 and Ca2SiO4, therefore, the luminescence of Ca3SiO4Br2:Eu2+ gives a broad emission band centered at 469 nm with some asymmetry on the long wavelength side with different coordination environment. Their detailed photoluminescence (PL) properties, PL decay curves and the temperature dependent PL behavior were also discussed. - Highlights: ► The Ca3SiO4Br2 crystal has been grown and the structure has been analyzed. ► Ca3SiO4Br2:Eu2+ gives a blue emission band centered at 469 nm. ► PL decay curves and the temperature dependent PL behavior of Ca3SiO4Br2:Eu2+ have been discussed.

  8. Three new chalcohalides, Ba4Ge2PbS8Br2, Ba4Ge2PbSe8Br2 and Ba4Ge2SnS8Br2: Syntheses, crystal structures, band gaps, and electronic structures

    International Nuclear Information System (INIS)

    Highlights: • Three new chalcohalides: Ba4Ge2PbS8Br2, Ba4Ge2PbSe8Br2 and Ba4Ge2SnS8Br2 have been synthesized. • The MQ5Br octahedra and GeQ4 tetrahedra form a three-dimensional framework with Ba2+ in the channels. • Band Gaps and electronic structures of the three compounds were studied. - Abstract: Single crystals of three new chalcohalides: Ba4Ge2PbS8Br2, Ba4Ge2PbSe8Br2 and Ba4Ge2SnS8Br2 have been synthesized for the first time. These isostructural compounds crystallize in the orthorhombic space group Pnma. In the structure, the tetra-valent Ge atom is tetrahedrally coordinated with four Q (Q = S, Se) atoms, while the bi-valent M atom (M = Pb, Sn) is coordinated with an obviously distorted octahedron of five Q (Q = S, Se) atoms and one Br atom, showing the stereochemical activity of the ns2 lone pair electron. The MQ5Br (M = Sn, Pb; Q = S, Se) distorted octahedra and the GeQ4 (Q = S, Se) tetrahedra are connected to each other to form a three-dimensional framework with channels occupied by Ba2+ cations. Based on UV–vis–NIR spectroscopy measurements and the electronic structure calculations, Ba4Ge2PbS8Br2, Ba4Ge2PbSe8Br2 and Ba4Ge2SnS8Br2 have indirect band gaps of 2.054, 1.952, and 2.066 eV respectively, which are mainly determined by the orbitals from the Ge, M and Q atoms (M = Pb, Sn; Q = S, Se)

  9. Ionothermal synthesis and structural characterization of [Cu(C4H6N2)4]Br2 and [Ni(C4H6N2)4]Br2

    Indian Academy of Sciences (India)

    Hong Wang; Bin Lu; Jingxiang Zhao; Qinghai Cai

    2015-07-01

    The ionothermal synthesis and spectroscopic, thermal and structural characterization of two new compounds [Cu(C4H6N2)4]Br2 (1) and [Ni(C4H6N2)4]Br2 (2) [(C4H6N2) = N-methylimidazole] are reported. In both 1 and 2, the central metal Cu (or Ni) ion adopts a square planar geometry and is bonded to the N-atoms of four terminal N-methylimidazole ligands.

  10. Phase diagram of the CsBr-CaBr2 system

    Institute of Scientific and Technical Information of China (English)

    BAO Xinhua; CHEN Nianyi; LU Wencong; CHENG Zhixuan; LUO Yunyun; LU Weiying; XIA Yiben

    2006-01-01

    The phase diagram of the CsBr-CaBr2 system was re-determined by using differential thermal analysis and high temperature and room temperature X-ray diffraction analysis. It is concluded that there are three intermediate compounds in this system: a congruently melting compound, CsCaBr3, with a melting point of 823℃ and two incongruently melting compounds, Cs2CaBr4 and Cs3Ca2Br7, whose peritectic points being 597℃ and 635℃, respectively. X-ray diffraction analysis indicated that compound CsCaBr3 is of slightly distorted perovskite structure.

  11. Synthesis and crystal structure determination of Br2SeIBr polyhalogen–chalcogen

    Indian Academy of Sciences (India)

    A A Alemi; E Solaimani

    2004-06-01

    In this paper polyhalogen–chalcogen Br2SeIBr was synthesized and the crystal structure was determined by single crystal X-ray diffraction method. This compound was prepared in the temperature range 150–50°C which was brownish-red in colour and crystallized in monoclinic crystal system and space group 21/c with four molecules per unit cell. Lattice parameters were: = 6.3711(1), = 6.7522(2), = 16.8850(5) Å, = = 90°, = 95·96°, = 722·45 Å3.

  12. Reactor Division semestrial progress report July - December 1987

    International Nuclear Information System (INIS)

    This report covers the activities of the reactor division at the SCK-CEN during the second semester of 1987. It deals with the BR-2 materials testing reactor, the BR-3 power plant, reactor physics, water cooled reactors, fast neutron reactors, fusion, non nuclear programmes, testing and commissioning, high and medium activities, and informatics. (MCB)

  13. Reactor Division semestrial progress report January - June 1987

    International Nuclear Information System (INIS)

    This report covers the activities of the reactor division at the SCK-CEN during the first semester of 1987. It deals with the BR-2 materials testing reactor, the BR-3 power plant, reactor physics, water cooled reactors, fast neutron reactors, fusion, non nuclear programmes, testing and commissioning, high and medium activities, and informatics. (MCB)

  14. Reactor Physics Programme

    International Nuclear Information System (INIS)

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

  15. Magnetic torque measurements of TPP[Fe(Pc)Br2]2

    International Nuclear Information System (INIS)

    Magnetic torque measurements of TPP[Fe(Pc)Br2]2 are reported. The torque curves for the magnetic field rotated within the ac plane exhibit a two-fold symmetry. The temperature dependences of the torque amplitudes divided by the square of the field strength (τφ=45/B2) exhibit a drastic field dependence below 12 K. This field dependence is attributable to the fluctuation in the spontaneous magnetization that appears below 4.5 K. The torque curves for the field rotated within the ab plane exhibit four-fold symmetry. The curves are sinusoidal above 18 K and exhibit saw-toothed shapes below ∼12 K and complicated shapes below 8 K. The complicated shapes suggest that both d-electrons and π-electrons form a magnetic order below 8 K.

  16. Photodissociation of Br2 molecules in an intense femtosecond laser field

    Science.gov (United States)

    Zhang, Jian; Zhang, Shian; Yang, Yan; Sun, Shengzhi; Wu, Hua; Li, Jing; Chen, Yuting; Jia, Tianqing; Wang, Zugeng; Kong, Fanao; Sun, Zhenrong

    2014-11-01

    We experimentally demonstrate the photodissociation process of Br2 molecules in the intense femtosecond laser field by a dc-sliced ion velocity map imaging technique. We show that four fragment ions B rn +(n =1 -4 ) are observed, and their kinetic energy increases while their angular distribution decreases with the increase of the charge number. We prove that the low (or high) charged fragment ions result from the photodissociation of the low (or high) charged parent ions. We explain the changes of the kinetic energy and angular distribution in these fragment ions by considering the potential energy curves of these parent ions that involve both the interaction of the Coulomb repulsive energy and chemical bonding energy. We also explain the experimental observation that the measured kinetic energy release in the experiment is much smaller than the theoretical calculation by enhanced ionization at a critical distance.

  17. Spin Excitations and Phonon Anomaly in Quasi-1D Spiral Magneti CuBr2

    Science.gov (United States)

    Li, Yuan; Wang, Chong; Yu, Daiwei; Wang, Lichen; Wang, Fa; Iida, Kazuki; Kamazawa, Kazuya; Wakimoto, Shuichi

    CuBr2 can be considered as a model quasi-one-dimensional (quasi-1D) spin-1/2 magnet, in which the frustrating ferromagnetic nearest-neighbor and antiferromagnetic next-nearest-neighbor exchange interactions give rise to a cycloidal magnetic order below TN = 73 K. The removal of inversion symmetry by the magnetic order also makes the material a type-II multiferroic system with a remarkably simple crystal structure. Using time-of-flight inelastic neutron scattering spectroscopy, we have determined the spin-wave as well as phonon spectra throughout the entire Brillouin zone. The spin-wave spectrum exhibits pronounced anisotropy and magnon damping, consistent with the material's quasi-1D nature and the non-colinear spin structure. The phonon spectrum exhibits dramatic discontinuities in the dispersion across the quasi-1D magnetic wave vector, indicative of strong magnetoelastic coupling and possibly of a spin-orbital texture that comes along with the spin correlations.

  18. Microstructure and Thermal Properties Analysis of (1-x)As2S3-xCdBr2 Glass System

    Institute of Scientific and Technical Information of China (English)

    TAO Hai-zheng; ZHAO Xiu-jian; JING Cheng-bin; WANG Xiao-hu

    2004-01-01

    The homogeneous glass sample for the (1-x)As2S3-xCdBr2,where x=0.015,0.035,0.05, was prepared by the conventional melt-quenched method.Amorphous (1-x)As2S3-xCdBr2 alloys were determined by X-ray diffraction, thermal comprehensive analysis and Raman scattering. The glass transition temperature (Tg) decreases a bit with the addition of CdBr2. Based on the experimental data, the microstructure is considered to be the discrete molecule species of AsBr3 and Cd-S atomic bonds or clusters are homogeneously dispersed in a disordered polymer network formed by AsS3 pyramids interlinked by sulfur bridges.

  19. Theoretical Investigations on the Interaction Nature of Br2 with HF, H2O and NH3

    Institute of Scientific and Technical Information of China (English)

    WU Wen-Sheng; WANG Zhao-Xu; ZHENG Bai-Shu; SHU Hua; LI Xiang; WU Jun-Yong

    2007-01-01

    The interactions of HF, H2O and NH3 with Br2 are investigated at the MP2(full)/ aug-cc-pVDZ level. It is found that two kinds of stable complexes, halogen-bonded and hydrogen- bonded complexes, exist between Br2 and HF and between Br2 and H2O. The interaction energy analysis and natural population analysis (NPA) are conducted to these two kinds of complexes, indicating the halogen-bonded complexes are more stable than the corresponding hydrogen-bonded ones, and the binding energies of the former increase in the order HF<H2O<NH3, different from HF>H2O for the latter.

  20. The role of MgBr2 to enhance the ionic conductivity of PVA/PEDOT:PSS polymer composite

    Directory of Open Access Journals (Sweden)

    Eslam M. Sheha

    2015-07-01

    Full Text Available A solid polymer electrolyte system based on poly(vinyl alcohol (PVA and poly(3,4-Etylenedioxythiophene:poly(styrenesulfonate (PEDOT:PSS complexed with magnesium bromide (MgBr2 salt was prepared using solution cast technique. The ionic conductivity is observed to increase with increasing MgBr2 concentration. The maximum conductivity was found to be 9.89 × 10−6 S/cm for optimum polymer composite film (30 wt.% MgBr2 at room temperature. The increase in the conductivity is attributed to the increase in the number of ions as the salt concentration is increased. This has been proven by dielectric studies. The increase in conductivity is also attributable to the increase in the fraction of amorphous region in the electrolyte films as confirmed by their structural, thermal, electrical and optical properties.

  1. Study of bromide salts solubility in the (m1KBr + m2CaBr2)(aq) system at T = 323.15 K. Thermodynamic model of solution behaviour and (solid + liquid) equilibria in the ternaries (m1KBr + m2CaBr2)(aq), and (m1MgBr2 + m2CaBr2)(aq), and in the quinary (Na + K + Mg + Ca + Br + H2O) systems to high concentration and temperature

    International Nuclear Information System (INIS)

    Highlights: ► We study the solubility in the mixed system (m1KBr + m2CaBr2)(aq) at T = 323.15 K. ► We construct T-variable model for (m1KBr + m2CaBr2)(aq) and (m1MgBr2 + m2CaBr2)(aq). ► The solubility modelling approach based on Pitzer equations is employed. ► We validate the model for all subsystems within quinary {Na + K + Mg + Ca + Br + H2O} system. ► Model calculations are in excellent agreement with the reference data and recommendations. - Abstract: The bromide minerals solubility in the mixed system (m1KBr + m2CaBr2)(aq) have been investigated at T = 323.15 K by the physico-chemical analysis method. The equilibrium crystallization of KBr(cr), and CaBr2·4H2O(cr) has been established. The results from solubility measurements obtained have been combined with experimental equilibrium solubility data available in the literature at T = 298.15 K to construct a chemical model that calculates (solid + liquid) equilibria in the ternary (m1KBr + m2CaBr2)(aq) system. The solubility modelling approach based on fundamental Pitzer specific interaction equations is employed. Temperature extrapolation of the mixed system model provides reasonable mineral solubilities at low (273.15 K) and high temperature (up to 373.15 K). The reference solubility data for (m1MgBr2 + m2CaBr2)(aq) system, which are available in the literature at T = (273.15, 298.15, and 323.15) K are used to evaluate mixing ion interaction parameters and to develop a model that calculates (solid + liquid) equilibria in this ternary system. The models for both ternary systems give a very good agreement with bromide salts equilibrium solubility data. Limitations of the mixed solution models due to data insufficiencies at high temperature are discussed. The mixed system models presented in this study expand the previously published temperature dependent sodium–potassium–magnesium–bromide model by evaluating potassium–calcium–bromide and magnesium–calcium–bromide mixing solution

  2. A Kinetic Study of the Gas-Phase Reaction of OH with Br2

    Science.gov (United States)

    Bryukov, Mikhail G.; Dellinger, Barry; Knyazev, Vadim D.

    2011-01-01

    An experimental, temperature-dependent kinetic study of the gas-phase reaction of the hydroxyl radical with molecular bromine (reaction 1) has been performed using a pulsed laser photolysis/pulsed-laser-induced fluorescence technique over a wide temperature range of 297 – 766 K, and at pressures between 6.68 and 40.29 kPa of helium. The experimental rate coefficients for reaction 1 demonstrate no correlation with pressure and exhibit a negative temperature dependence with a slight negative curvature in the Arrhenius plot. A non-linear least-squares fit with two floating parameters of the temperature dependent k1(T) data set using an equation of the form k1(T) = ATn yields the recommended expression k1(T) = 1.85×10−9T − 0.66 cm3 molecule−1 s−1 for the temperature dependence of the reaction 1 rate coefficient. The potential energy surface (PES) of reaction 1 was investigated using quantum chemistry methods. The reaction proceeds through formation of a weakly bound OH···Br2 complex and a PES saddle point with an energy below that of the reactants. Temperature dependence of the reaction rate coefficient was modeled using the RRKM method on the basis of the calculated PES. PMID:16854030

  3. The obesity-induced transcriptional regulator TRIP-Br2 mediates visceral fat endoplasmic reticulum stress-induced inflammation.

    Science.gov (United States)

    Qiang, Guifen; Kong, Hyerim Whang; Fang, Difeng; McCann, Maximilian; Yang, Xiuying; Du, Guanhua; Blüher, Matthias; Zhu, Jinfang; Liew, Chong Wee

    2016-01-01

    The intimate link between location of fat accumulation and metabolic disease risk and depot-specific differences is well established, but how these differences between depots are regulated at the molecular level remains largely unclear. Here we show that TRIP-Br2 mediates endoplasmic reticulum (ER) stress-induced inflammatory responses in visceral fat. Using in vitro, ex vivo and in vivo approaches, we demonstrate that obesity-induced circulating factors upregulate TRIP-Br2 specifically in visceral fat via the ER stress pathway. We find that ablation of TRIP-Br2 ameliorates both chemical and physiological ER stress-induced inflammatory and acute phase response in adipocytes, leading to lower circulating levels of inflammatory cytokines. Using promoter assays, as well as molecular and pharmacological experiments, we show that the transcription factor GATA3 is responsible for the ER stress-induced TRIP-Br2 expression in visceral fat. Taken together, our study identifies molecular regulators of inflammatory response in visceral fat that-given that these pathways are conserved in humans-might serve as potential therapeutic targets in obesity. PMID:27109496

  4. Finding the missing stratospheric Bry: a global modeling study of CHBr3 and CH2Br2

    Directory of Open Access Journals (Sweden)

    L. Ott

    2009-11-01

    Full Text Available Recent in situ and satellite measurements suggest a contribution of ~5 pptv to stratospheric inorganic bromine from short-lived bromocarbons. We conduct a modeling study of the two most important short-lived bromocarbons, bromoform (CHBr3 and dibromomethane (CH2Br2, with the Goddard Earth Observing System Chemistry Climate Model (GEOS CCM to account for this missing stratospheric bromine. We derive a "top-down" emission estimate of CHBr3 and CH2Br2 using airborne measurements in the Pacific and North American troposphere and lower stratosphere (LS obtained during previous NASA aircraft campaigns. Our emission estimate suggests that to reproduce the observed concentrations in the free troposphere, a global oceanic emission of 425 Gg Br yr−1 for CHBr3 and 57 Gg Br yr−1 for CH2Br2 is needed, with 60% of emissions from open ocean and 40% from coastal regions. Although our simple emission scheme assumes no seasonal variations, the model reproduces the observed seasonal variations of the short-lived bromocarbons with high concentrations in winter and low concentrations in summer. This indicates that the seasonality of short-lived bromocarbons is largely due to seasonality in their chemical loss and transport. The inclusion of CHBr3 and CH2Br2 contributes ~5 pptv bromine throughout the stratosphere. Both the source gases and inorganic bromine produced from the source gas degradation (BryVSLS in the troposphere are transported into the stratosphere, and are equally important. Inorganic bromine accounts for half (2.5 pptv of the bromine from the inclusion of CHBr3 and CH2Br2 near the tropical tropopause and its contribution rapidly increases to ~100% as altitude increases. More than 85% of the wet scavenging of BryVSLS occurs in large-scale precipitation below 500 hPa and BryVSLS in the stratosphere is not sensitive to convection.

  5. The utility of different reactor types for the research

    International Nuclear Information System (INIS)

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

  6. Summary of the studies carried out for the definition of the maximum allowed heat flux in the BR2 reactor

    International Nuclear Information System (INIS)

    The maximum allowed heat flux is defined following measurements of the cladding temperature performed in the year 1963 in steady hydraulic conditions and during the flow inversion occurring after a core cooling perturbation. Two-phase flow calculations have implemented the criterion of flow instability and the safety during an incident of power excursion with automatic power reduction by control rods. The influence of the thickness of the water gap on the maximum allowed heat flux is also given

  7. Ionothermal Syntheses and Crystal Structures of Two Cobalt(Ⅱ) Compounds:Co(2,2'-bpy)2Br2 and Co(1,10-phen)2Br2

    Institute of Scientific and Technical Information of China (English)

    YANG Er-Lei; ZHANG Na; SHAN Zeng-Mei; WANG Yu-Ling; HU Hai-Chun; LIU Qing-Yan

    2011-01-01

    Two cobalt(Ⅱ) compounds, Co(2,2'-bpy)2Br2 (1, 2,2'-bpy = 2,2'-bipyridyl) and Co(1,10- phen)2Br2 (2, 1,10-phen = 1,10-phenanthroline), have been prepared under ionothermal reactions using the 1-propyl-3-methylimidazolium bromide ionic liquid solvent.Single-crystal X-ray analyses reveal that 1 crystallizes in a monoclinic space group P21/c, with a = 8.5509(13), b =14.804(2), c = 15.650(2)(A),β = 97.119(2)°, V= 1965.8(5)(A)3, Z = 4, C20H16Br2N4Co, Me= 531.12,Dc= 1.795 g/cm3,μ = 4.950 mm-1, F(000) = 1044, the final R = 0.0467 and wR = 0.0736 for 2291observed reflections with I > 2σ(I).Complex 2 crystallizes in a monoclinic space group P21/n, with a = 10.4237(8), b = 16.8657(12), c = 12.4945(9) (A),β = 102.110(1)°, V = 2147.7(3) (A)3, Z = 4,C24H16Br2N4Co, Mr= 579.16, Dc= 1.791 g/cm3,/μ = 4.540 mm-1, F(000) = 1140, the final R =0.0431 and wR = 0.1042 for 3470 observed reflections with I > 2σ(I).The mononuclear molecules of 1 are linked by the C-H…Br hydrogen bonds and π-π interactions to form a three-dimensional supramolecular framework structure.The C-H…Br hydrogen bonds and π-π interactions link the mononuclear molecules of 2 to give a two-dimensional layer structure.

  8. Peculiarities of kinetics of destruction of excited Br*(2P2/2) atoms forming at IBr photolysis

    International Nuclear Information System (INIS)

    Kinetics of destruction of bromine atoms in excited Br*(2P1/2) and basic Br(2P3/2) states, depending on degree of IBr photodissociation α, has been studied under conditions of pulsed lamp UV-photolysis of IBr molecules at 293+-2 K by the method of laser atomic resonance spectroscopy. It has been ascertained that with α increase the rate of destruction of nonexcited Br atoms practically does not change, whereas for excited Br* atoms it increases by nearly an order with α increase from 3x10-3 to 8x10-2. The detected effect of strong dependence of Br* destruction rate on α is explained, making allowance for fast reaction of Br* atoms with highly vibration-excited molecules formed in the course of IBr photolysis. Ways of suppressing the effect mentioned are suggested. 21 refs.; 4 figs

  9. Synchrotron x-ray powder diffraction study of the structural phase transition in CaBr2

    International Nuclear Information System (INIS)

    The structures of CaBr2 have been investigated using high resolution synchrotron powder x-ray diffraction methods between room temperature and 800 deg. C. At room temperature CaBr2 has an orthorhombic CaCl2-type structure (Pnnm Z=2) with a=6.8847(1), b=6.5806(1), and c=4.3477(1) A. Heating above 560 deg. C results in a continuous transition to a tetragonal rutile type structure (P42/mnm Z=2) with a=6.810 95(7) A and c=4.417 70(5) A at 650 deg. C. Investigation, through either spontaneous strain or octahedral tilt angle, suggests that the transition is close to second order in nature, although the contribution from the sixth order term in the Landau potential cannot be neglected

  10. 3+ and [Sb13Se16Br2] 5+ - Double and quadruple spiro cubanes from ionic liquids

    KAUST Repository

    Ahmed, Ejaz

    2014-01-08

    The reaction of antimony and selenium in the bromine-rich Lewis acidic ionic liquid [BMIm]Br·4.7AlBr3 (BMIm: 1-butyl-3- methylimidazolium) in the presence of a small amount of NbCl5 at 160 °C yielded dark-red crystals of [Sb7Se8Br 2][AlX4]3. For X = Cl0.15(1)Br 0.85(1), the compound is isostructural to [Sb7S 8Br2][AlCl4]3 [P212 121, a = 12.5132(5) Å, b = 17.7394(6) Å, c = 18.3013(6) Å]. For a higher chlorine content, X = Cl 0.58(1)Br0.42(1), a slightly disordered variant with a bisected unit cell is found [P21212, a = 12.3757(3) Å, b = 17.4116(5) Å, c = 9.0420(2) Å]. The [Sb 7Se8Br2]3+ heteropolycation (C 2 symmetry) is a spiro double-cubane with an antimony atom on the shared corner. From this distorted octahedrally coordinated central atom, tricoordinate selenium and antimony atoms alternate in the bonding sequence. The terminal antimony atoms each bind to a bromine atom. Quantum chemical calculations confirm polar covalent Sb-Se bonding within the cubes and indicate three-center, four-electron bonds for the six-coordinate spiro atoms. The calculated charge distribution reflects the electron-donor role of the antimony atoms. The use of a chlorine-rich ionic liquid resulted in the formation of triclinic [Sb13Se16Br2][AlX4] 5 with X = Cl0.80(1)Br0.20(1) [P$\\\\bar {1}$, a = 9.0842(5) Å, b = 19.607(1) Å, c = 21.511(1) Å, α = 64.116(6), β = 79.768(7), γ = 88.499(7)]. The cationic cluster [Sb13Se16Br2]5+ is a bromine-terminated spiro quadruple-cubane. This 31 atom concatenation of four cubes is assumed to be the largest known discrete main group polycation. A similar reaction in a chloride-free system yielded [Sb7Se 8Br2][Sb13Se16Br2] [AlBr4]8. In its monoclinic structure [P2/c, a = 27.214(5) Å, b = 9.383(2) Å, c = 22.917(4) Å, β = 101.68(1)], the two types of polycations alternate in layers along the a axis. In the series [Sb4+3nSe4+4nBr2](2+n)+, these cations are the members with n = 1 and 3. Copyright © 2014 WILEY-VCH Verlag GmbH & Co. KGa

  11. The investigation of influence of iodine ions on a luminescence of Cd Br2 - Mn Cl2

    International Nuclear Information System (INIS)

    Optical and luminescence properties of the crystal Cd Br2 - Mn Cl2, Cd I2 are investigated in the 85 - 295 K temperature interval. It is found that additional doping Cd Br2 - Mn Cl2 with iodine results in increasing photo- and thermoluminescence efficiencies due to 4 Tlg (4 G) → 6 Alg(6S) electron-phonon transitions in Mn2+ -centers under LGI-21 nitric laser excitation. The additional band at 495 nm attributed to radiative recombination of excitons localized on I- ions is observed in X-ray luminescence spectra of the crystal at 85 K. The nature of activation bands in excitation spectra and the mechanism of the thermally stimulated luminescence are discussed

  12. The effects of CaCl2 and CaBr2 on the reproduction of Daphnia magna Straus.

    Science.gov (United States)

    Mažuran, Neda; Hršak, Vladimir; Kovačević, Goran

    2015-06-01

    Concentrated CaCl2 and CaBr2 salt solutions of densities up to 2.3 kg L-1 are regularly used to control hydrostatic pressure in oil wells during special operations in the exploration and production of natural gas and crude oil. Various concentrations of high density salts are frequently left in mud pits near the drilling site as waste, polluting fresh and ground waters by spillage and drainage. The toxic effects of these salts have already been observed. This study investigated the effects of CaCl2 and CaBr2 on water flea Daphnia magna Straus in a 21-day reproduction test. The three tested concentrations of CaCl2 (240, 481, and 1925 mg L-1) caused a significant dose-response decrease of reproduction (p<0.001). With CaBr2 (533 and 1066 mg L-1), only aborted eggs were produced, demonstrating the embryotoxicity of the substance. The results suggest that high concentrations of the tested chemicals are harmful to Daphnia's reproduction and could reduce its abundance. PMID:26110475

  13. The (p, ρ, T) properties and apparent molar volumes Vφ of (ZnBr2 + C2H5OH)

    International Nuclear Information System (INIS)

    The (p, ρ, T) properties and apparent molar volumes Vφ of ZnBr2 in ethanol at temperatures (293.15 to 393.15) K and pressures up to p = 40 MPa are reported. The measurements were made with a recently developed vibration-tube densimeter. The system was calibrated using double-distilled water, methanol, ethanol, and aqueous NaCl solutions. The experiments were carried out at molalities of m = (0.05681, 0.16958, 0.30426, 0.43835, 0.93055, 1.49016, and 1.88723) mol . kg-1 using zinc bromide. An empirical correlation for the density of (ZnBr2 + C2H5OH) with pressure, temperature, and molality has been derived. This equation of state was used to calculate other volumetric properties such as isothermal compressibility, isobaric thermal expansibility, the differences in specific heat capacities at constant pressures and volumes, apparent molar volumes of ZnBr2 in ethanol, and partial molar volumes of both components.

  14. Finding the missing stratospheric Bry: a global modeling study of CHBr3 and CH2Br2

    Directory of Open Access Journals (Sweden)

    D. R. Blake

    2010-03-01

    Full Text Available Recent in situ and satellite measurements suggest a contribution of ~5 pptv to stratospheric inorganic bromine from short-lived bromocarbons. We conduct a modeling study of the two most important short-lived bromocarbons, bromoform (CHBr3 and dibromomethane (CH2Br2, with the Goddard Earth Observing System Chemistry Climate Model (GEOS CCM to account for this missing stratospheric bromine. We derive a "top-down" emission estimate of CHBr3 and CH2Br2 using airborne measurements in the Pacific and North American troposphere and lower stratosphere obtained during previous NASA aircraft campaigns. Our emission estimate suggests that to reproduce the observed concentrations in the free troposphere, a global oceanic emission of 425 Gg Br yr−1 for CHBr3 and 57 Gg Br yr−1 for CH2Br2 is needed, with 60% of emissions from open ocean and 40% from coastal regions. Although our simple emission scheme assumes no seasonal variations, the model reproduces the observed seasonal variations of the short-lived bromocarbons with high concentrations in winter and low concentrations in summer. This indicates that the seasonality of short-lived bromocarbons is largely due to seasonality in their chemical loss and transport. The inclusion of CHBr3 and CH2Br2 contributes ~5 pptv bromine throughout the stratosphere. Both the source gases and inorganic bromine produced from source gas degradation (BryVSLS in the troposphere are transported into the stratosphere, and are equally important. Inorganic bromine accounts for half (2.5 pptv of the bromine from the inclusion of CHBr3 and CH2Br2 near the tropical tropopause and its contribution rapidly increases to ~100% as altitude increases. More than 85% of the wet scavenging of BryVSLS occurs in large-scale precipitation below 500 hPa. Our sensitivity study with wet scavenging in convective updrafts switched off suggests that BryVSLS in the stratosphere is not sensitive to convection. Convective scavenging only accounts for

  15. Reactors

    International Nuclear Information System (INIS)

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

  16. Laser induced fluorescence (LIF) of Hg2 and Hg3 via dissociation of HgBr2 at 157 nm

    Science.gov (United States)

    Skordoulis, C.; Sarantopoulou, E.; Spyrou, S. M.; Kosmidis, C.; Cefalas, A. C.

    1991-06-01

    Laser induced fluorescence of the mercury clusters Hg2 and Hg3 in the spectral range between 300 nm to 510 nm has been obtained from the dissociation of HgBr2 at 7.88 eV (157.5 nm) with an F2 molecular laser, together with fluorescence from mercury atomic transitions from highly excited states. The excitation process involves two photon absorption which dissociates the molecule at 15.76 eV total photon energy with the subsequent formation of the metallic clusters.

  17. Laser induced fluorescence (LIF) of Hg2 and Hg3 via dissociation of HgBr2 at 157 nm

    International Nuclear Information System (INIS)

    Laser induced fluorescence of the mercury clusters Hg2 and Hg3 in the spectral range between 300 nm to 510 nm has been obtained from the dissociation of HgBr2 at 7.88 eV (157.5 nm) with an F2 molecular laser, together with fluorescence from mercury atomic transitions from highly excited states. The excitation process involves two photon absorption which dissociates the molecule at 15.76 eV total photon energy with the subsequent formation of the metallic clusters. (orig.)

  18. Theoretical Study on the Reaction Mechanism of F2+2HBr=2HF+Br2

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    The gas phase reaction mechanism of F2 + 2HBr = 2HF + Br2 has been investigated by (U)MP2 at 6-311G** level, and a series of four-center and three-center transition states have been obtained. The reaction mechanism was achieved by comparing the activation energy of seven reaction paths, i.e. the dissociation energy of F2 is less than the activation energy of the bimolecular elementary reaction F2 + HBr → HF + BrF. Thus it is theoretically proved that the title reaction occurs more easily inthe free radical reaction with three medium steps.

  19. Ablation of TRIP-Br2, a novel regulator of fat lipolysis, thermogenesis and oxidative metabolism, prevents diet-induced obesity and insulin resistance

    OpenAIRE

    Liew, Chong Wee; Boucher, Jeremie; Cheong, Jit Kong; Vernochet, Cecile; Koh, Ho-Jin; Mallol, Cristina; Townsend, Kristy; Langin, Dominique; Kawamori, Dan; Hu, Jiang; Tseng, Yu-Hua; Hellerstein, Marc K.; Farmer, Stephen R.; Goodyear, Laurie; Doria, Alessandro

    2012-01-01

    SUMMARY Obesity develops due to altered energy homeostasis favoring fat storage. Here we describe a novel transcription co-regulator for adiposity and energy metabolism, TRIP-Br2 (also called SERTAD2). TRIP-Br2 null mice are resistant to obesity and obesity-related insulin resistance. Adipocytes of the knockout (KO) mice exhibited greater stimulated lipolysis secondary to enhanced expression of hormone sensitive lipase (HSL) and β3-adrenergic (Adrb3) receptors. The KOs also exhibit higher ene...

  20. Ablation of TRIP-Br2, a novel regulator of fat lipolysis, thermogenesis and oxidative metabolism, prevents diet-induced obesity and insulin resistance

    OpenAIRE

    Liew, Chong Wee; Boucher, Jeremie; Cheong, Jit Kong; Vernochet, Cecile; Koh, Ho-Jin; Mallol, Cristina; Townsend, Kristy; Langin, Dominique; Kawamori, Dan; Hu, Jiang; Tseng, Yu-Hua; Hellerstein, Marc K.; Farmer, Stephen R.; Goodyear, Laurie; Doria, Alessandro

    2013-01-01

    SUMMARY Obesity develops due to altered energy homeostasis favoring fat storage. Here we describe a novel transcription co-regulator for adiposity and energy metabolism, TRIP-Br2 (also called SERTAD2). TRIP-Br2 null mice are resistant to obesity and obesity-related insulin resistance. Adipocytes of the knockout (KO) mice exhibited greater stimulated lipolysis secondary to enhanced expression of hormone sensitive lipase (HSL) and β3-adrenergic (Adrb3) receptors. The KOs also exhibit higher ene...

  1. Bi6(SeO3)3O5Br2: A new bismuth oxo-selenite bromide

    International Nuclear Information System (INIS)

    A new bismuth oxo-selenite bromide Bi6(SeO3)3O5Br2 was synthesized and structurally characterized. The crystal structure belongs to the triclinic system (space group P1-bar , Z=2, a=7.1253(7) Å, b=10.972(1) Å, c=12.117(1) Å, α=67.765(7)°, β=82.188(8)°, γ=78.445(7)°) and is unrelated to those of other known oxo-selenite halides. It can be considered as an open framework composed of BiOx or BiOyBrz polyhedrons forming channels running along [1 0 0] direction which contain the selenium atoms in pyramidal shape oxygen coordination (SeO3E). The spectroscopic properties and thermal stability were studied. The new compound is stable up to 400 °C. - graphical abstract: New bismuth oxo-selenite bromide with new open framework structure. Highlights: ► New bismuth oxo-selenite bromide was found and structurally characterized. ► Bi6(SeO3)3O5Br2 exhibit a new open framework structure type. ► BiOx or BiOyBrz polyhedrons form channels in the structure which are decorated by [SeO3E] groups.

  2. Impact of membrane characteristics on the performance and cycling of the Br-2-H-2 redox flow cell

    Energy Technology Data Exchange (ETDEWEB)

    Tucker, MC; Cho, KT; Spingler, FB; Weber, AZ; Lin, GY

    2015-06-15

    The Br-2/H-2 redox flow cell shows promise as a high-power, low-cost energy storage device. In this paper, the effect of various aspects of material selection and processing of proton exchange membranes on the operation of the Br-2/H-2 redox flow cell is determined. Membrane properties have a significant impact on the performance and efficiency of the system. In particular, there is a tradeoff between conductivity and crossover, where conductivity limits system efficiency at high current density and crossover limits efficiency at low current density. The impact of thickness, pretreatment procedure, swelling state during cell assembly, equivalent weight, membrane reinforcement, and addition of a microporous separator layer on this tradeoff is assessed. NR212 (50 mu m) pretreated by soaking in 70 degrees C water is found to be optimal for the studied operating conditions. For this case, an energy efficiency of greater than 75% is achieved for current density up to 400 mA cm(-2), with a maximum obtainable energy efficiency of 88%. A cell with this membrane was cycled continuously for 3164 h. Membrane transport properties, including conductivity and bromine and water crossover, were found to decrease moderately upon cycling but remained higher than those for the as-received membrane. (C) 2015 Elsevier B.V. All rights reserved.

  3. Impact of membrane characteristics on the performance and cycling of the Br2-H2 redox flow cell

    Science.gov (United States)

    Tucker, Michael C.; Cho, Kyu Taek; Spingler, Franz B.; Weber, Adam Z.; Lin, Guangyu

    2015-06-01

    The Br2/H2 redox flow cell shows promise as a high-power, low-cost energy storage device. In this paper, the effect of various aspects of material selection and processing of proton exchange membranes on the operation of the Br2/H2 redox flow cell is determined. Membrane properties have a significant impact on the performance and efficiency of the system. In particular, there is a tradeoff between conductivity and crossover, where conductivity limits system efficiency at high current density and crossover limits efficiency at low current density. The impact of thickness, pretreatment procedure, swelling state during cell assembly, equivalent weight, membrane reinforcement, and addition of a microporous separator layer on this tradeoff is assessed. NR212 (50 μm) pretreated by soaking in 70 °C water is found to be optimal for the studied operating conditions. For this case, an energy efficiency of greater than 75% is achieved for current density up to 400 mA cm-2, with a maximum obtainable energy efficiency of 88%. A cell with this membrane was cycled continuously for 3164 h. Membrane transport properties, including conductivity and bromine and water crossover, were found to decrease moderately upon cycling but remained higher than those for the as-received membrane.

  4. Molecular elimination of Br2 in photodissociation of CH2BrC(O)Br at 248 nm using cavity ring-down absorption spectroscopy

    International Nuclear Information System (INIS)

    The primary elimination channel of bromine molecule in one-photon dissociation of CH2BrC(O)Br at 248 nm is investigated using cavity ring-down absorption spectroscopy. By means of spectral simulation, the ratio of nascent vibrational population in v = 0, 1, and 2 levels is evaluated to be 1:(0.5 ± 0.1):(0.2 ± 0.1), corresponding to a Boltzmann vibrational temperature of 581 ± 45 K. The quantum yield of the ground state Br2 elimination reaction is determined to be 0.24 ± 0.08. With the aid of ab initio potential energy calculations, the obtained Br2 fragments are anticipated to dissociate on the electronic ground state, yielding vibrationally hot Br2 products. The temperature-dependence measurements support the proposed pathway via internal conversion. For comparison, the Br2 yields are obtained analogously from CH3CHBrC(O)Br and (CH3)2CBrC(O)Br to be 0.03 and 0.06, respectively. The trend of Br2 yields among the three compounds is consistent with the branching ratio evaluation by Rice-Ramsperger-Kassel-Marcus method. However, the latter result for each molecule is smaller by an order of magnitude than the yield findings. A non-statistical pathway so-called roaming process might be an alternative to the Br2 production, and its contribution might account for the underestimate of the branching ratio calculations.

  5. Br2 molecular elimination in photolysis of (COBr)2 at 248 nm by using cavity ring-down absorption spectroscopy: A photodissociation channel being ignored

    International Nuclear Information System (INIS)

    A primary dissociation channel of Br2 elimination is detected following a single-photon absorption of (COBr)2 at 248 nm by using cavity ring-down absorption spectroscopy. The technique contains two laser beams propagating in a perpendicular configuration. The tunable laser beam along the axis of the ring-down cell probes the Br2 fragment in the B3Πou+-X1Σg+ transition. The measurements of laser energy- and pressure-dependence and addition of a Br scavenger are further carried out to rule out the probability of Br2 contribution from a secondary reaction. By means of spectral simulation, the ratio of nascent vibrational population for v = 0, 1, and 2 levels is evaluated to be 1:(0.65 ± 0.09):(0.34 ± 0.07), corresponding to a Boltzmann vibrational temperature of 893 ± 31 K. The quantum yield of the ground state Br2 elimination reaction is determined to be 0.11 ± 0.06. With the aid of ab initio potential energy calculations, the pathway of molecular elimination is proposed on the energetic ground state (COBr)2 via internal conversion. A four-center dissociation mechanism is followed synchronously or sequentially yielding three fragments of Br2+ 2CO. The resulting Br2 is anticipated to be vibrationally hot. The measurement of a positive temperature effect supports the proposed mechanism.

  6. Fuel swelling and interaction layer formation in the SELENIUM Si and ZrN coated U(Mo) dispersion fuel plates irradiated at high power in BR2

    Energy Technology Data Exchange (ETDEWEB)

    Leenaers, A., E-mail: aleenaer@sckcen.be [Nuclear Materials Science Institute, SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Van den Berghe, S.; Koonen, E.; Kuzminov, V. [Nuclear Materials Science Institute, SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Detavernier, C. [Department of Solid State Sciences, Ghent University, Krijgslaan 281/S1, 9000 Ghent (Belgium)

    2015-03-15

    In the framework of the SELENIUM project two full size flat fuel plates were produced with respectively Si and ZrN coated U(Mo) particles and irradiated in the BR2 reactor at SCK• CEN. Non-destructive analysis of the plates showed that the fuel swelling profiles of both SELENIUM plates were very similar to each other and none of the plates showed signs of pillowing or excessive swelling at the end of irradiation at the highest power position (local maximum 70% {sup 235}U). The microstructural analysis showed that the Si coated fuel has less interaction phase formation at low burn-up but at the highest burn-ups, defects start to develop on the IL–matrix interface. The ZrN coated fuel, shows a virtual absence of reaction between the U(Mo) and the Al, up to high fission densities after which the interaction layer formation starts and defects develop in the matrix near the U(Mo) particles. It was found and is confirmed by the SELENIUM (Surface Engineering of Low ENrIched Uranium–Molybdenum) experiment that there are two phenomena at play that need to be controlled: the formation of an interaction layer and swelling of the fuel. As the interaction layer formation occurs at the U(Mo)–matrix interface, applying a diffusion barrier (coating) at that interface should prevent the interaction between U(Mo) and the matrix. The U(Mo) swelling, observed to proceed at an accelerating rate with respect to fission density accumulation, is governed by linear solid state swelling and fission gas bubble swelling due to recrystallization of the fuel. The examination of the SELENIUM fuel plates clearly show that for the U(Mo) dispersion fuel to be qualified, the swelling rate at high burn-up needs to be reduced.

  7. Fuel swelling and interaction layer formation in the SELENIUM Si and ZrN coated U(Mo) dispersion fuel plates irradiated at high power in BR2

    International Nuclear Information System (INIS)

    In the framework of the SELENIUM project two full size flat fuel plates were produced with respectively Si and ZrN coated U(Mo) particles and irradiated in the BR2 reactor at SCK• CEN. Non-destructive analysis of the plates showed that the fuel swelling profiles of both SELENIUM plates were very similar to each other and none of the plates showed signs of pillowing or excessive swelling at the end of irradiation at the highest power position (local maximum 70% 235U). The microstructural analysis showed that the Si coated fuel has less interaction phase formation at low burn-up but at the highest burn-ups, defects start to develop on the IL–matrix interface. The ZrN coated fuel, shows a virtual absence of reaction between the U(Mo) and the Al, up to high fission densities after which the interaction layer formation starts and defects develop in the matrix near the U(Mo) particles. It was found and is confirmed by the SELENIUM (Surface Engineering of Low ENrIched Uranium–Molybdenum) experiment that there are two phenomena at play that need to be controlled: the formation of an interaction layer and swelling of the fuel. As the interaction layer formation occurs at the U(Mo)–matrix interface, applying a diffusion barrier (coating) at that interface should prevent the interaction between U(Mo) and the matrix. The U(Mo) swelling, observed to proceed at an accelerating rate with respect to fission density accumulation, is governed by linear solid state swelling and fission gas bubble swelling due to recrystallization of the fuel. The examination of the SELENIUM fuel plates clearly show that for the U(Mo) dispersion fuel to be qualified, the swelling rate at high burn-up needs to be reduced

  8. Pathways for the OH + Br2 → HOBr + Br and HOBr + Br → HBr + BrO Reactions.

    Science.gov (United States)

    Wang, Hongyan; Qiu, Yudong; Schaefer, Henry F

    2016-02-11

    The OH radical reaction with Br2 and the subsequent reaction HOBr + Br are of exceptional importance to atmospheric chemistry and environmental chemistry. The entrance complex, transition state, and exit complex for both reactions have been determined using the coupled-cluster method with single, double, and perturbative triple excitations CCSD(T) with correlation consistent basis sets up to size cc-pV5Z and cc-pV5Z-PP. Coupled cluster effects with full triples (CCSDT) and full quadruples (CCSDTQ) are explicitly investigated. Scalar relativistic effects, spin-orbit coupling, and zero-point vibrational energy corrections are evaluated. The results from the all-electron basis sets are compared with those from the effective core potential (ECP) pseudopotential (PP) basis sets. The results are consistent. The OH + Br2 reaction is predicted to be exothermic 4.1 ± 0.5 kcal/mol, compared to experiment, 3.9 ± 0.2 kcal/mol. The entrance complex HO···BrBr is bound by 2.2 ± 0.2 kcal/mol. The transition state lies similarly well below the reactants OH + Br2. The exit complex HOBr···Br is bound by 2.7 ± 0.6 kcal/mol relative to separated HOBr + Br. The endothermicity of the reaction HOBr + Br → HBr + BrO is 9.6 ± 0.7 kcal/mol, compared with experiment 8.7 ± 0.3 kcal/mol. For the more important reverse (exothermic) HBr + BrO reaction, the entrance complex BrO···HBr is bound by 1.8 ± 0.6 kcal/mol. The barrier for the HBr + BrO reaction is 6.8 ± 0.9 kcal/mol. The exit complex (Br···HOBr) for the HBr + BrO reaction is bound by 1.9 ± 0.2 kcal/mol with respect to the products HOBr + Br. PMID:26766412

  9. MnBr2/18-crown-6 coordination complexes showing high room temperature luminescence and quantum yield.

    Science.gov (United States)

    Hausmann, David; Kuzmanoski, Ana; Feldmann, Claus

    2016-04-12

    The reaction of manganese(ii) bromide and the crown ether 18-crown-6 in the ionic liquid [(n-Bu)3MeN][N(Tf)2] under mild conditions (80-130 °C) resulted in the formation of three different coordination compounds: MnBr2(18-crown-6) (), Mn3Br6(18-crown-6)2 () and Mn3Br6(18-crown-6) (). In general, the local coordination and the crystal structure of all compounds are driven by the mismatch between the small radius of the Mn(2+) cation (83 pm) and the ring opening of 18-crown-6 as a chelating ligand (about 300 pm). This improper situation leads to different types of coordination and bonding. MnBr2(18-crown-6) represents a molecular compound with Mn(2+) coordinated by two bromine atoms and only five oxygen atoms of 18-crown-6. Mn3Br6(18-crown-6)2 falls into a [MnBr(18-crown-6)](+) cation - with Mn(2+) coordinated by six oxygen atoms and Br - and a [MnBr(18-crown-6)MnBr4](-) anion. In this anion, Mn(2+) is coordinated by five oxygen atoms of the crown ether as well as by two bromine atoms, one of them bridging to an isolated (MnBr4) tetrahedron. Mn3Br6(18-crown-6), finally, forms an infinite, non-charged [Mn2(18-crown-6)(MnBr6)] chain. Herein, 18-crown-6 is exocyclically coordinated by two Mn(2+) cations. All compounds show intense luminescence in the yellow to red spectral range and exhibit remarkable quantum yields of 70% (Mn3Br6(18-crown-6)) and 98% (Mn3Br6(18-crown-6)2). The excellent quantum yield of Mn3Br6(18-crown-6)2 and its differentiation from MnBr2(18-crown-6) and Mn3Br6(18-crown-6) can be directly correlated to the local coordination. PMID:26956783

  10. Band gap opening in silicene on MgBr2(0001) induced by Li and Na

    KAUST Repository

    Zhu, Jiajie

    2014-11-12

    Silicene consists of a monolayer of Si atoms in a buckled honeycomb structure and is expected to be well compatible with the current Si-based technology. However, the band gap is strongly influenced by the substrate. In this context, the structural and electronic properties of silicene on MgBr2(0001) modified by Li and Na are investigated by first-principles calculations. Charge transfer from silicene (substrate) to substrate (silicene) is found for substitutional doping (intercalation). As compared to a band gap of 0.01 eV on the pristine substrate, strongly enhanced band gaps of 0.65 eV (substitutional doping) and 0.24 eV (intercalation) are achieved. The band gap increases with the dopant concentration.

  11. Nonadiabatic quantum dynamics of Br2 in solid Ar: A four-dimensional study of the B to C state predissociation

    International Nuclear Information System (INIS)

    Quantum dynamics simulations are performed for a diatomics-in-molecules based model of Br2 in solid Ar which incorporates four nuclear degrees of freedom and four electronic states. The nuclear motions comprise two large amplitude coordinates describing the Br2 bond distance and an effective symmetry-preserving matrix mode. Two symmetry-lowering harmonic modes are added in the spirit of linear vibronic coupling theory. Initiating the dynamics on the B state by means of an ultrafast laser pulse, nonadiabatic transitions to the two degenerate C states are monitored and the effect of vibrational preexcitation in the electronic ground state is investigated

  12. Nonadiabatic quantum dynamics of Br 2 in solid Ar: A four-dimensional study of the B to C state predissociation

    Science.gov (United States)

    Borowski, A.; Kühn, O.

    2008-05-01

    Quantum dynamics simulations are performed for a diatomics-in-molecules based model of Br 2 in solid Ar which incorporates four nuclear degrees of freedom and four electronic states. The nuclear motions comprise two large amplitude coordinates describing the Br 2 bond distance and an effective symmetry-preserving matrix mode. Two symmetry-lowering harmonic modes are added in the spirit of linear vibronic coupling theory. Initiating the dynamics on the B state by means of an ultrafast laser pulse, nonadiabatic transitions to the two degenerate C states are monitored and the effect of vibrational preexcitation in the electronic ground state is investigated.

  13. Assessing the occurrence of the dibromide radical (Br2−·) in natural waters: Measures of triplet-sensitised formation, reactivity, and modelling

    International Nuclear Information System (INIS)

    The triplet state of anthraquinone-2-sulphonate (AQ2S) is able to oxidise bromide to Br·/Br2−·, with rate constant (2–4) ⋅ 109 M−1 s−1 that depends on the pH. Similar processes are expected to take place between bromide and the triplet states of naturally occurring chromophoric dissolved organic matter (3CDOM*). The brominating agent Br2−· could thus be formed in natural waters upon oxidation of bromide by both ·OH and 3CDOM*. Br2−· would be consumed by disproportionation into bromide and bromine, as well as upon reaction with nitrite and most notably with dissolved organic matter (DOM). By using the laser flash photolysis technique, and phenol as model organic molecule, a second-order reaction rate constant of ∼ 3 ⋅ 102 L (mg C)−1 s−1 was measured between Br2−· and DOM. It was thus possible to model the formation and reactivity of Br2−· in natural waters, assessing the steady-state [Br2−·] ≈ 10−13–10−12 M. It is concluded that bromide oxidation by 3CDOM* would be significant compared to oxidation by ·OH. The 3CDOM*-mediated process would prevail in DOM-rich and bromide-rich environments, the latter because elevated bromide would completely scavenge ·OH. Under such conditions, ·OH-assisted formation of Br2−· would be limited by the formation rate of the hydroxyl radical. In contrast, the formation rate of 3CDOM* is much higher compared to that of ·OH in most surface waters and would provide a large 3CDOM* reservoir for bromide to react with. A further issue is that nitrite oxidation by Br2−· could be an important source of the nitrating agent ·NO2 in bromide-rich, nitrite-rich and DOM-poor environments. Such a process could possibly account for significant aromatic photonitration observed in irradiated seawater and in sunlit brackish lagoons. Highlights: ► The triplet state of anthraquinone-2-sulphonate oxidises bromide to Br2−·. ► Dissolved organic matter is the main Br2−· scavenger in surface waters

  14. Excitation of Rydberg states of HgCl2 and HgBr2 by electron impact

    International Nuclear Information System (INIS)

    We have examined the electronic structure of HgCl2 and HgBr2 from 5 eV to 14 eV by high-resolution electron energy-loss spectroscopy. Our measurements include the near-UV region, which has not been examined by any previous technique and which is found to be rich in Rydbery states. In particular, in each molecule we identify members of two optically allowed Rydberg series and one forbidden series with electronic structures sigma/sub g/2 sigma/sub u/2 π/sub u/4 π/sub g/3 [npπ/sub u/, npsigma/sub u/ and ndπ/sub g/, or (n+1)ssigma/sub g/] that converge to the 2PI/sub g/ ground ionic state. In addition, other structures that form the dominant energy-loss mechanisms in our spectra are identified as arising from optically allowed Rydberg states associated with the excited ionic states 2PI/sub u/, 2Σ/sub u/, and 2Σ/sub g/. Measurements at energy losses below 9 eV confirm previous valence states observed in photoabsorption and suggest the existence of two new valence states in each molecule. Angular measurements facilitate identification of many of the newly observed structures in our electron energy-loss spectra

  15. Ablation of TRIP-Br2, a regulator of fat lipolysis, thermogenesis and oxidative metabolism, prevents diet-induced obesity and insulin resistance.

    Science.gov (United States)

    Liew, Chong Wee; Boucher, Jeremie; Cheong, Jit Kong; Vernochet, Cecile; Koh, Ho-Jin; Mallol, Cristina; Townsend, Kristy; Langin, Dominique; Kawamori, Dan; Hu, Jiang; Tseng, Yu-Hua; Hellerstein, Marc K; Farmer, Stephen R; Goodyear, Laurie; Doria, Alessandro; Blüher, Matthias; Hsu, Stephen I-Hong; Kulkarni, Rohit N

    2013-02-01

    Obesity develops as a result of altered energy homeostasis favoring fat storage. Here we describe a new transcription co-regulator for adiposity and energy metabolism, SERTA domain containing 2 (TRIP-Br2, also called SERTAD2). TRIP-Br2-null mice are resistant to obesity and obesity-related insulin resistance. Adipocytes of these knockout mice showed greater stimulated lipolysis secondary to enhanced expression of hormone sensitive lipase (HSL) and β3-adrenergic (Adrb3) receptors. The knockout mice also have higher energy expenditure because of increased adipocyte thermogenesis and oxidative metabolism caused by upregulating key enzymes in their respective processes. Our data show that a cell-cycle transcriptional co-regulator, TRIP-Br2, modulates fat storage through simultaneous regulation of lipolysis, thermogenesis and oxidative metabolism. These data, together with the observation that TRIP-Br2 expression is selectively elevated in visceral fat in obese humans, suggests that this transcriptional co-regulator is a new therapeutic target for counteracting the development of obesity, insulin resistance and hyperlipidemia. PMID:23291629

  16. Structure Studies of PbBr2-PbF2-P2O5 Glasses%PbBr2-PbF2-P2O5玻璃的结构研究

    Institute of Scientific and Technical Information of China (English)

    赵宏生; 李艳青; 周万城; 罗发

    2003-01-01

    采用红外光谱(IR)和X射线光电子能谱(XPS)等方法研究了PbBr2-PbF2-P2O5系铅卤磷酸盐玻璃的结构.结果表明,Pb2+离子在玻璃中起着网络修饰阳离子和网络形成体的双重作用.当P2O5含量为60mol%时,Pb2+离子主要是作为网络修饰体;当P2O5含量降低到50 mol%时,一部分pb2+离子能够进入玻璃网络形成PbO4]四面体或P-O-Pb键.Br-和F-离子达到一定浓度时就会进入玻璃网络,形成[PO4-nXn](X=Br或F,n=0~4)四面体使磷酸盐链长变短.玻璃中P2O5的含量不变时P-O-P键的比例也基本保持不变;当P2O5的含量降低时,P-O-P键和P-O-键的含量都减少,P-O-Pb键的含量则明显增加.

  17. Composition Analysis of Glasses in the PbBr2-PbCl2-PbF2-PbO-P2O5 System%PbBr2-PbCl2-PbF2-PbO-P2O5系统玻璃成分分析*

    Institute of Scientific and Technical Information of China (English)

    朱冬梅; 周万城; 赵宏生

    2001-01-01

    研究了PbBr2-PbCl2-PbF2-PbO-P2O5系统的玻璃在熔融过程中的变化.发现在玻璃的熔制过程中,元素P、Pb和Br的损失量最大.化学分析和理论计算表明,这些元素是以PbC12、PbBr2和P2O5的形式挥发的.而且,随着配料成分中含量的增加,PbCl2、PbBr2和P2O5的挥发量也随之增加.

  18. Temperature-induced structural changes in CaCl2,CaBr2, and CrCl2: A synchrotron x-ray powder diffraction study

    International Nuclear Information System (INIS)

    The halides CaCl2,CaBr2, and CrCl2 all adopt, at room temperature, the same distorted rutile structure, in orthorhombic space group Pnnm, known as the calcium chloride structure. Upon heating, CaCl2 and CaBr2 each undergoes a continuous transformation to the true tetragonal rutile structure, in space group P42/mnm, the transition temperatures being 235 and 553 deg. C, respectively. By contrast, the structural change in CrCl2 upon heating is just further elongation of octahedra already lengthened by Jahn-Teller effects, and no phase transition occurs. The orthorhombic structure is maintained by a strong and temperature-dependent geometrical coupling of the orthorhombic strain to the order parameter, represented by the tilt angle of the CrCl6 octahedron

  19. Synthesis, Structure and Photoluminescent Properties of the 2D Coordination Polymers Based on Cu2Br2 Unit with Flexible Thioether%基于柔性硫醚与Cu2Br2单元二维配位聚合物的合成、结构和荧光性质

    Institute of Scientific and Technical Information of China (English)

    李冬青; 时文娟

    2009-01-01

    A complex [Cu2Br2(L)2]2 (1) (L=bis (2-pyrimidinylthio)methane) has been synthesized and structurally characterized. Complex 1 contains dinuclear Cu2Br2 units, which are linked by ditopic L to form a 2D layer structure with a 36-membered macrometallocycle. The adjacent layers are further connected through interpyrimidyl rings C-H strong green solid-state photoluminescence, due to metal-to-ligand charge-transfer (MLCT) at room temperature. CCDC: 711434.

  20. Solubility measurement and solid-liquid equilibrium model for the ternary system MgBr2 + MgSO4 + H2O at 288.15 K

    Directory of Open Access Journals (Sweden)

    Li Dan

    2014-06-01

    Full Text Available The solubility of magnesium minerals and the refractive index of the ternary system MgBr2 + MgSO4 + H2O at 288.15 K were investigated using an isothermal dissolution method. It was found that there are two invariant points in the phase diagram and the solubility isotherm of this ternary system consists of three branches, corresponding to equilibrium crystallization of Epsomite (MgSO4·7H2O, Eps, hexahydrite (MgSO4·6H2O, Hex and magnesium bromide hexahydrate (MgBr2·6H2O, Mb. Neither solid solutions nor double salts were found. The refractive indices calculated from empirical equation are in good agreement with the experimental data. Combining the results from solubility measurements with the single-salt parameters for MgBr2 and MgSO4, and the mixed ion-interaction parameter θBr,S0(4, the parameter ψMg,Br,S0(4 at 288.15 K was fitted using the Pitzer theory and Harvie-Weare (HW approach. In addition, the average equilibrium constants of the stable equilibrium solids at 288.15 K were obtained by a method using the activity product constant. A chemical model, which combined the Pitzer parameters and the average equilibrium constants, was constructed to calculate the solid + liquid equilibria in the ternary system MgBr2 + MgSO4 + H2O at 288.15 K. The model agreed well with the equilibrium solubility data for the magnesium salts.

  1. Utilization of reactors overseas for study of radiation effects in materials

    International Nuclear Information System (INIS)

    This paper presents the current status of utilization of overseas reactors for study of radiation effects in materials as joint-use research at International Research Center for Nuclear Materials Science, Institute for Materials Research (IMR), Tohoku University (IMR-Oarai Center). Japan Materials Testing Reactor (JMTR) which had been mainly utilized for material irradiation at IMR-Oarai Center since 1970 was shut down in August 2006. Since then an overseas reactor, BR2 at SCK/CEN in Belgium, has been used in order to continue our irradiation programs and perform advanced material irradiation in collaboration with SCK/CEN for development of new irradiation techniques in BR2. In this paper, the features of irradiation facilities/devices in BR2 such as CALLISTO, MISTRAL, ROBIN and BAMI and the irradiation programs in 2006-2008 with use of each facility/device are described. (author)

  2. PROPERTY OF GLASSES IN THE PbBr2-PbCl2-PbF2-PbO-P2O5 SYSTEM%PbBr2-PbCl2-PbF2-PbO-P2O5系统玻璃性能研究

    Institute of Scientific and Technical Information of China (English)

    朱冬梅; 周万城; 赵宏生

    1999-01-01

    研究了PbBr2-PbCl2-PbF2-PbO-P2O5系统的玻璃转变温度、密度、耐水性和透光率.PbBr2-PbCl2-P2O5系统的玻璃转变温度低达146 ℃,密度高达4.75 g/cm3.加入PbF2和/或PbO可显著提高玻璃的转变温度和密度,其中PbO的影响更为显著,可使玻璃的密度增加到5.48 g/cm3.多数PbBr2-PbCl2-PbF2-PbO-P2O5系统玻璃的耐水性都很好,在水中的溶解速率为10-5 mm/day.玻璃的透光性较好,加入PbBr2使玻璃的紫外截断波长明显向长波方向移动.

  3. Effect of the calcium halides, CaCl2 and CaBr2, on hydrogen desorption in the Li–Mg–N–H system

    International Nuclear Information System (INIS)

    Highlights: • H2 desorption from 2LiNH2–MgH2–xCaX2 (x = 0, 0.1, 0.15; X = Cl, Br) samples studied. • Addition of calcium halides reduced the desorption temperature in all samples. • Peak H2 release was around 150 °C lower in ball-milled than in hand-ground samples. • The 2LiNH2–MgH2–0.15CaBr2 sample showed the lowest peak desorption temperature. • CaBr2 reduced the activation energy to 78.8 kJ mol−1, 24% less than the undoped sample. - Abstract: Calcium-halide-doped lithium amide–magnesium hydride samples were prepared both by hand-grinding and ball-milling 2LiNH2–MgH2–xCaX2 (x = 0, 0.1, and 0.15; X = Cl or Br). The addition of calcium halides reduced the hydrogen desorption temperature in all samples. The ball-milled undoped sample (2LiNH2–MgH2) began to desorb hydrogen at around 125 °C and peaked at 170 °C. Hydrogen desorption from the 0.15 mol CaCl2-containing sample began ca 30 °C lower than that of the undoped sample and peaked at 150 °C. Both the onset and peak temperatures of the CaBr2 sample (x = 0.15) were reduced by 15 °C compared to the chloride. Kissinger’s method was used to calculate the effective activation energy (Ea) for the systems: Ea for the 0.15 mol CaCl2-containing sample was found to be 91.8 kJ mol−1 and the value for the 0.15 mol CaBr2-containing sample was 78.8 kJ mol−1.

  4. Et2NH2C6H3(CO23SnBr2.4H2O: SYNTHESIS AND INFRARED STUDY

    Directory of Open Access Journals (Sweden)

    DAOUDA NDOYE

    2014-01-01

    Full Text Available The title compound has been obtained on allowing [C6H3(CO23(Et2NH23] to react with SnBr4. The molecular structure of Et2NH2C6H3(CO23SnBr2.4H2O has been determined on the basis of the infrared data. The suggested structure is a dimer in which each tin atom is hexacoordinated by two chelating C6H3(CO233- anions and two Br atoms. Cy2NH2+cations are involved through hydrogen bonds with non-coordinating CO2 groups. The suggested structure is a cage.

  5. PbBr2-PbCl2-PbF2-PbO-P2O5系统玻璃的热性质和化学稳定性%Thermal Properties and Water Durability of Glasses in the PbBr2-PbCl2-PbF2-PbO-P2O5 System

    Institute of Scientific and Technical Information of China (English)

    朱冬梅; 周万城; 赵宏生; 吴静波

    2000-01-01

    重点研究了PbBr2-PbCl2-PbF2-PbO-P2O5系统玻璃的热性质、耐水性和抗潮解性. 结果表明:该系统玻璃的热膨胀系数较大,一般在25×10-6°C-1左右. PbBr2-PbCl2-P2O5系统的玻璃具有较低的玻璃转变温度,可低达146°C. 加入PbF2和/或PbO可显著提高玻璃的转变温度和密度,其中PbO对试样的影响更为显著. PbBr2-PbCl2-PbF2-PbO-P2O5系统玻璃的抗潮解性一般较好. 多数玻璃在水中的溶解速率可达10-5mm/day, 具有较好的耐水性.%A series of new glasses in the PbBr2-PbCl2-PbF2-PbO-P2O5 system were prepared and some properties of the glasses in the PbBr2-PbCl2-PbF2-PbO-P2O5 system were studied. Their composition dependence of glass transition temperature(T g), thermal expansion coefficient and water durability was reported. Glasses in the PbBr2-PbCl2-P2O5 ternary system have relative low glass transition temperature(>146C). When PbO and/or PbF2 are introduced into the ternary system, the glass transition temperature increases considerably. Glasses containing PbO have higher transition temperature than those containing PbF2. The glasses in this system have relative high thermal expansion coefficient(25×10-6/C). Most of the glasses exhibit good water durability, and the dissolution rate in deionised water at 25C is in the order of 10-4~10-5mm/day. The introduction of PbO and/or PbF2 into the glasses does not change their water durability much. With the increase of P2O5, the water durability of the glass becomes bad obviously.

  6. A chopper system for shortening the duration of pulsed supersonic beams seeded with NO or Br2 down to 13 microseconds

    CERN Document Server

    Lam, Jessica; Softley, Tim

    2015-01-01

    A chopper wheel construct is used to shorten the duration of a molecular beam to 13 microseconds. Molecular beams seeded with NO or with Br2 and an initial pulse width of greater or equal to 200 microseconds were passed through a spinning chopper wheel, which was driven by a brushless DC in vacuo motor at a range of speeds, from 3,000 rpm to 80,000 rpm. The resulting duration of the molecular-beam pulses measured at the laser detection volume ranged from 80 microseconds to 13 microseconds, and was the same for both NO and Br2. The duration is consistent with a simple analytical model, and the minimum pulse width measured is limited by the spreading of the beam between the chopper and the detection point as a consequence of the longitudinal velocity distribution of the beam. The setup adopted here effectively eliminates buildup of background gas without the use of a differential pumping stage, and a clean narrow pulse is obtained with low rotational temperature.

  7. A chopper system for shortening the duration of pulsed supersonic beams seeded with NO or Br2 down to 13 μs

    International Nuclear Information System (INIS)

    A chopper wheel construct is used to shorten the duration of a molecular beam to 13 μs. Molecular beams seeded with NO or with Br2 and an initial pulse width of ≥200 μs were passed through a spinning chopper wheel, which was driven by a brushless DC in vacuo motor at a range of speeds, from 3000 rpm to 80 000 rpm. The resulting duration of the molecular-beam pulses measured at the laser detection volume ranged from 80 μs to 13 μs and was the same for both NO and Br2. The duration is consistent with a simple analytical model, and the minimum pulse width measured is limited by the spreading of the beam between the chopper and the detection point as a consequence of the longitudinal velocity distribution of the beam. The setup adopted here effectively eliminates buildup of background gas without the use of a differential pumping stage, and a clean narrow pulse is obtained with low rotational temperature

  8. Modeling and simulation of output power of a high-power He-SrBr2 laser by using multivariate adaptive regression splines

    Science.gov (United States)

    Iliev, I. P.; Gocheva-Ilieva, S. G.

    2013-02-01

    Due to advancement in computing technology researchers are focusing onto novel predictive models and techniques for improving the design and performance of the devices. This study examines a recently developed high-powered SrBr2 laser excited in a nanosecond pulse longitudinal He-SrBr2 discharge. Based on the accumulated experiment data, a new approach is proposed for determining the relationship between laser output power and basic laser characteristics: geometric design, supplied electric power, helium pressure, etc. Piece-wise linear and nonlinear statistical models have been built with the help of the flexible predictive MARS technique. It is shown that the best nonlinear MARS model containing second degree terms provides the best description of the examined data, demonstrating a coefficient of determination of over 99%. The resulting theoretical models are used to estimate and predict the experiment as well as to analyze the local behavior of the relationships between laser output power and input laser characteristics. The second order model is applied to the design and optimization of the laser in order to enhance laser generation.

  9. A chopper system for shortening the duration of pulsed supersonic beams seeded with NO or Br2 down to 13 μs

    Science.gov (United States)

    Lam, Jessica; Rennick, Christopher J.; Softley, Timothy P.

    2015-05-01

    A chopper wheel construct is used to shorten the duration of a molecular beam to 13 μs. Molecular beams seeded with NO or with Br2 and an initial pulse width of ≥200 μs were passed through a spinning chopper wheel, which was driven by a brushless DC in vacuo motor at a range of speeds, from 3000 rpm to 80 000 rpm. The resulting duration of the molecular-beam pulses measured at the laser detection volume ranged from 80 μs to 13 μs and was the same for both NO and Br2. The duration is consistent with a simple analytical model, and the minimum pulse width measured is limited by the spreading of the beam between the chopper and the detection point as a consequence of the longitudinal velocity distribution of the beam. The setup adopted here effectively eliminates buildup of background gas without the use of a differential pumping stage, and a clean narrow pulse is obtained with low rotational temperature.

  10. The glass-forming ability of the P2O5-PbBr2-PbF2 system%P2O5-PbBr2-PbF2系玻璃抗析晶能力的研究

    Institute of Scientific and Technical Information of China (English)

    赵宏生; 周万城; 李艳青; 罗发

    2003-01-01

    为了提高P2O5-PbBr2-PbF2系铅卤磷酸盐玻璃的热稳定性,采用非等温差热分析实验研究了P2O5-PbBr2-PbF2玻璃的抗析晶能力和非等温析晶特性与成分之间的关系.研究结果表明:在不同的加热速率下,采用ΔT、H′和Hr等玻璃热稳定性判据判断玻璃的抗析晶能力均难以得到一致的结果;kD(T)和kb(T)判据结合了析晶动力学和热力学因素,不受玻璃成分和升温速率影响,是比较好的玻璃抗析晶能力的判据;采用kD(T)和kb(T)判据发现玻璃的成分为PbF2/(PbBr2+PbF2)≈0.3时,样品具有最强的玻璃形成能力;P2O5-PbBr2-PbF2玻璃析晶的Avrami指数n约为2,表明玻璃在加热过程中同时存在表面析晶和整体析晶.

  11. New High-Density and Low-Melting Glassesin the PbO-PbBr2-PbF2-P2O5 Quaternary System%PbO-PbBr2-PbF2-P2O5系统高密度低熔融温度玻璃的研究

    Institute of Scientific and Technical Information of China (English)

    赵宏生; 周万城

    1999-01-01

    研究了PbO-PbBr2-PbF2-P2O5系统的玻璃形成能力,并探索出了玻璃形成区.测试了一些玻璃的特征温度、密度和紫外截止波长.实验结果表明PbBr2-PbF2-P2O5系统的玻璃形成能力很强,加入10mol%的PbO可以提高玻璃的密度并降低玻璃转变温度.当P2O5的含量小于40mol%时,玻璃的密度在5.10~5.81g/cm3之间;Tg可降至210℃.PbF2-P2O5二元玻璃的紫外截止波长约为273nm,加入PbBr2使玻璃的紫外截止波长向长波方向移动.

  12. Synthesis and structure of diorganotin dibromides, R2SnBr2 (R = 2,4,6-trimethylphenyl or 2,4,6-trimethylbenzyl): Hydrolysis of (2,4,6-Me3C6H2)2SnBr2

    Indian Academy of Sciences (India)

    Vadapalli Chandrasekhar; Ramalingam Thirumoorthi

    2010-09-01

    The reaction of SnBr4 with in situ generated 2,4,6-trimethylphenylmagnesium bromide afforded a mixture of (2,4,6-Me3C6H2)2SnBr2 (1) and (2,4,6-Me3C6H2)3SnBr (2) which could be separated from each other by their solubility differences in diethyl ether. On the other hand, the reaction of tin metal with 2,4,6-Me3C6H2CH2Br afforded (2,4,6-Me3C6H2CH2)2SnBr2 (3). Hydrolysis of the latter using triethylamine as the base afforded [{(2,4,6-Me3C6H2CH2)2Sn}2(-O)(Br)(-OH)]2.2CH2Cl2 (4) while the use of NaOH as the base afforded [{(2,4,6-Me3C6H2CH2)2Sn}2(-O)(OH)(-OH)]2.2CH2Cl2 (5). Compounds 4 and 5 are dimeric tetraorganodistannoxanes consisting of a central distannoxane (Sn2O2) motif.

  13. Low-dimensional charge transport of the ferroic NH 2(C 2H 5) 4CoCl 2Br 2 nanocrystals

    Science.gov (United States)

    Tkaczyk, S. W.; Kityk, I. V.; Rudyk, V.; Kapustianyk, V.

    2010-06-01

    DC-conductivity of the nanocomposites consisting of ferroic nanocrystals NH 2(C 2H 5) 4CoCl 2Br 2 (TEA-CCB) incorporated into the polymer PMMA matrices is investigated. The investigations are performed for different number of nanocrystallite chromophores and thickness of the samples. The thickness of the nanocomposite films was varied within 0.44-0.48 μm. The measurements were done at different temperatures and different applied voltages. As a dominant mechanism thermoemission was considered, which is determined by temperature of the samples.. The gold electrodes were used as principle electrodes. The observed phenomena were explained within a framework of hopping between the trapping levels and possible contribution of self-trapped excitons.

  14. O K and Cu LIII edge study of itinerant holes in I2-, HgI2- and HgBr2- intercalated BSCCO(2212) single crystals

    International Nuclear Information System (INIS)

    Intercalation effect on BSCCO (2212) system, although it increases the interlayer distance and the c-axis remarkably, produces only a small change in the transition temperature. Thus, amongst other things, intercalation provides an effective method to investigate the influence of the interblock coupling. Electrons are transferred from the host Cu-O2 layers to the guest molecules I2, HgBr2, HgI2 leading to evolution of the Tc. For this we have made high resolution XANES study on the O K and Cu L3 edges to estimate the density of the doping holes. We attempt on basis of our and earlier results the evolution of Tc in these as also the much larger decrease produced in Tc for I2-intercalation for which the increase in basal spacing is the smallest of the three halides. (au)

  15. Vapor pressures and isopiestic molalities of concentrated CaCl2(aq), CaBr2(aq), and NaCl(aq) to T = 523 K

    International Nuclear Information System (INIS)

    The Oak Ridge National Laboratory high-temperature isopiestic apparatus was outfitted with precise pressure gauges to allow for direct vapor pressure measurements. Vapor pressures over concentrated solutions of CaCl2(aq), and CaBr2(aq) were measured at temperatures between (380.15 and 523.15) K in the range of water activities between 0.2 and 0.85. Isopiestic molalities were used to determine osmotic coefficients at the conditions where NaCl reference standard solutions remained undersaturated. The main goal of this work was to improve the accuracy of isopiestic comparisons based on the calcium chloride reference standard. Osmotic coefficients for CaCl2(aq) and CaBr2(aq) calculated from both isopiestic and direct vapor pressure results were combined with the literature data and used to build general thermodynamic models based on a variant of extended Pitzer ion-interaction equations and valid at the saturation pressure of water. While these empirical models approach the accuracy of the experimental data in a wider range of concentrations and temperatures than any previously published equations, considerable amounts of accurate data and a substantial effort are required in order to obtain a satisfactory representation using power series-based virial equations. The effect of experimental uncertainties on the accuracy of the direct vapor pressure results is discussed, including in particular the error caused by the presence in the apparatus of a small amount of CO2. The substantial decrease of the solubility product of CaCO3 in concentrated chloride solutions at temperatures above 423 K is a serious defect of calcium chloride as a water activity reference standard

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

    International Nuclear Information System (INIS)

    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. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

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

  18. Composition Change During Melting of PbO-PbBr2-PbF2-P2O5 System Glass%熔化过程中 PbO-PbBr2-PbF2-P2O5系统玻璃成分的变化

    Institute of Scientific and Technical Information of China (English)

    赵宏生; 周万城; 朱冬梅

    2000-01-01

    为了研究PbO-PbBr2-PbF2-P2O5系统玻璃在熔化过程中成分的变化,本文应用定量化学分析法测出了玻璃在熔化后的成分. 结果表明,PbO-PbBr2-PbF2-P2O5系统玻璃熔化时氟和磷的损失较大,铅的损失相对较小; 溴的损失随配料中各元素含量的多少而变化. 氧含量的变化受P2O5的挥发以及氧、水分与玻璃熔体中卤化物的反应程度两个因素影响. 相关化学反应的热力学计算与分析结果相吻合.%To study the composition change during melting of PbO-PbBr2-PbF2-P2O5 system glass, the quantitative chemical analysis was complied to measure the composition of the glass after melting. The results show that the loss of lead during melting is mainly ascribed to the volatilization of PbBr2. When a sample contains a high content of phosphorus and a low content of fluorine, the loss of bromine and phosphorus is remarkable, mainly in the form of PBr3. On the other hand, if a sample contains a high content of fluorine and a low content of bromine or a low content of phosphorus, bromine will not lose much while a large amount of phosphorus and fluorine are lost during melting, mainly in the form of PF5. The reaction between phosphorus and fluorine is much easier than that between phosphorus and bromine. The results of thermodynamic calculation are consistent with the above results and discussions. The content of oxygen is influenced by two factors. Volatilization of P2O5 decreases the content of oxygen, while the reaction between oxygen or water in air and halides in melts increases the content of oxygen.

  19. An investigation on the structure, spectroscopy and thermodynamic aspects of Br2((-))(H2O)n clusters using a conjunction of stochastic and quantum chemical methods.

    Science.gov (United States)

    Naskar, Pulak; Chaudhury, Pinaki

    2016-06-28

    In this work we obtained global as well as local structures of Br2((-))(H2O)n clusters for n = 2 to 6 followed by the study of IR-spectral features and thermochemistry for the structures. The way adopted by us to obtain structures is not the conventional one used in most cases. Here we at first generated excellent quality pre-optimized structures by exploring the suitable empirical potential energy surface using stochastic optimizer simulated annealing. These structures are then further refined using quantum chemical calculations to obtain the final structures, and spectral and thermodynamic features. We clearly showed that our approach results in very quick and better convergence which reduces the computational cost and obviously using the strategy we are able to get one [i.e. global] or more than one [i.e. global and local(s)] energetically lower structures than those which are already reported for a given cluster size. Moreover, IR-spectral results and the evolutionary trends in interaction energy, solvation energy and vertical detachment energy for global structures of each size have also been presented to establish the utility of the procedure employed. PMID:27251059

  20. Synthesis and structure of [(NH2)2CSSC(NH2)2]2[OsBr6]Br2 . 3H2O

    International Nuclear Information System (INIS)

    The complex [(NH2)2CSSC(NH2)2]2[OsBr6]Br2 . 3H2O is synthesized by the reaction of K2OsBr6 with thiocarbamide in concentrated HBr and characterized using electronic absorption and IR absorption spectroscopy. Its crystal structure is determined by X-ray diffraction. The crystals are orthorhombic, a = 11.730(2) A, b = 14.052(3) A, c = 16.994(3) A, space group Cmcm, and Z = 4. The [OsBr6]2- anionic complex has an octahedral structure. The Os-Br distances fall in the range 2.483-2.490 A. The α,α'-dithiobisformamidinium cation is a product of the oxidation of thiocarbamide. The S-S and C-S distances are 2.016 and 1.784 A, respectively. The H2O molecules, Br-ions, and NH2 groups of the cation are linked by hydrogen bonds.

  1. Computational fluid dynamics analyses of lateral heat conduction, coolant azimuthal mixing and heat transfer predictions in a BR2 fuel assembly geometry.

    Energy Technology Data Exchange (ETDEWEB)

    Tzanos, C. P.; Dionne, B. (Nuclear Engineering Division)

    2011-05-23

    To support the analyses related to the conversion of the BR2 core from highly-enriched (HEU) to low-enriched (LEU) fuel, the thermal-hydraulics codes PLTEMP and RELAP-3D are used to evaluate the safety margins during steady-state operation (PLTEMP), as well as after a loss-of-flow, loss-of-pressure, or a loss of coolant event (RELAP). In the 1-D PLTEMP and RELAP simulations, conduction in the azimuthal and axial directions is not accounted. The very good thermal conductivity of the cladding and the fuel meat and significant temperature gradients in the lateral directions (axial and azimuthal directions) could lead to a heat flux distribution that is significantly different than the power distribution. To evaluate the significance of the lateral heat conduction, 3-D computational fluid dynamics (CFD) simulations, using the CFD code STAR-CD, were performed. Safety margin calculations are typically performed for a hot stripe, i.e., an azimuthal region of the fuel plates/coolant channel containing the power peak. In a RELAP model, for example, a channel between two plates could be divided into a number of RELAP channels (stripes) in the azimuthal direction. In a PLTEMP model, the effect of azimuthal power peaking could be taken into account by using engineering factors. However, if the thermal mixing in the azimuthal direction of a coolant channel is significant, a stripping approach could be overly conservative by not taking into account this mixing. STAR-CD simulations were also performed to study the thermal mixing in the coolant. Section II of this document presents the results of the analyses of the lateral heat conduction and azimuthal thermal mixing in a coolant channel. Finally, PLTEMP and RELAP simulations rely on the use of correlations to determine heat transfer coefficients. Previous analyses showed that the Dittus-Boelter correlation gives significantly more conservative (lower) predictions than the correlations of Sieder-Tate and Petukhov. STAR-CD 3-D

  2. Overlapping $B^{3}_{0u}← X^{1}^{+}_{g}$ and $^{1}_{1u} ← X^{1}^{+}_{g}$ non-radiative characteristic of Br2 vapour in the wavelength region 505–541 nm

    Indian Academy of Sciences (India)

    Ramesh C Sharma; S N Thakur

    2001-01-01

    The vibronic vapour phase photoacoustic spectrum of Br2 in the wavelength region 505–541 nm (19796–18480 cm-1) has been recorded using microphone as well as pump-probe method. Discrete vibronic bands superimposed on a monotonically increasing continuum background towards the dissociation limit results from the overlapping $B^{3}^{+}_{0u}← X^{1}^{+}_{g}$ and $^{1}_{1u}← X^{1}^{+}_{g}$ electronic transitions. Vibronic bands originating from '' = 0 have been used to estimate the relative rate of non-radiative relaxation as a function of the excited state $^{3}_{0u}$ vibrational quantum number '. A comparison with the optical absorption spectroscopy of Br2 leads to the identification of three broad spectral regions between 505 and 541 nm (19796 and 18480 cm-1) on the basis of different non-radiative relaxation processes.

  3. Test programme for the nuclear fuel of the Jules Horowitz reactor starts

    International Nuclear Information System (INIS)

    In Cadarache in the south of France, an advanced material test reactor is currently being built. The Jules Horowitz reactor is an initiative of the French Atomic Energy Commission ((CEA), supported by the European Commission and research centres in the Czech Republic, Spain, Finland and Belgium, plus a number of large energy companies. On the request of CEA, SCK-CEN started characterising the nuclear fuel for this reactor, in its own BR2. The article describes SCK-CEN's programme on the characterisation of new nuclear fuel elements.

  4. N Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The last of Hanfordqaodmasdkwaspemas7ajkqlsmdqpakldnzsdflss nine plutonium production reactors to be built was the N Reactor.This reactor was called a dual purpose...

  5. (p, ρ, T) properties, and apparent molar volumes V φ of ZnBr2 in methanol at T = (298.15 to 398.15) K and pressures up to p = 40 MPa

    International Nuclear Information System (INIS)

    The (p, ρ, T) properties of pure methanol, the (p, ρ, T) properties and apparent molar volumes V φ of ZnBr2 in methanol at T (298.15 to 398.15) K and pressures up to p = 40 MPa are reported, and apparent molar volumes have been evaluated. The experimental (p, ρ, T, m) values were described by an equation of state. For the solutions the experiments were carried out at molalities m = (0.05772, 0.37852, 0.71585 and 1.95061) mol . kg-1 of zinc bromide

  6. 2D numerical modelling of the gas temperature in a high-temperature high-power strontium atom laser excited by nanosecond pulsed longitudinal discharge in a He-SrBr2 mixture

    Science.gov (United States)

    Chernogorova, T. P.; Temelkov, K. A.; Koleva, N. K.; Vuchkov, N. K.

    2014-05-01

    Assuming axial symmetry and a uniform power input, a 2D model (r, z) is developed numerically for determination of the gas temperature in the case of a nanosecond pulsed longitudinal discharge in He-SrBr2 formed in a newly-designed large-volume high-temperature discharge tube with additional incompact ZrO2 insulation in the discharge-free zone, in order to find the optimal thermal mode for achievement of maximal output laser parameters. The model determines the gas temperature of a nanosecond pulsed longitudinal discharge in helium with small additives of strontium and bromine.

  7. The radial distribution of the neutron field in the core of Dalat reactor

    International Nuclear Information System (INIS)

    Determining the radial distribution of the thermal neutron field in the core of the Dalat reactor was done by the Cu foil activation method. The measured data were fitted by the least square method to determine some physical parameters of the reactor, as follows: 1. Laplacian: Br2 = (84.6 +- 5.5)10-4/,cm2. 2. The effective radius: Reff = (27.6 +- 1.0)cm. 3. The extrapolation distance: λr = (8.7 +- 1.0)cm. 4. The unequal coefficient of the effective multiplication: kr = 1.77 +- 0.11. (author). 3 refs., 4 figs., 1 tab

  8. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

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

  9. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    2013-01-01

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

  10. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-09-01

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

  11. Research reactors

    International Nuclear Information System (INIS)

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

  12. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

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

  13. Methane hydrate phase equilibrium in the presence of NaBr, KBr, CaBr2, K2CO3, and MgCl2 aqueous solutions: Experimental measurements and predictions of dissociation conditions

    International Nuclear Information System (INIS)

    In this communication, experimental data for dissociation conditions of methane hydrates in the presence of 0.05 and 0.1 mass fractions NaBr, KBr, K2CO3, and MgCl2 aqueous solutions and in the presence of 0.05 and 0.15 mass fractions CaBr2 aqueous solutions are reported. The experimental data were generated using an isochoric pressure-search method. The new experimental dissociation data for methane hydrates in the presence of 0.1 mass fraction MgCl2 aqueous solution are compared with some selected experimental data from the literature and the agreements are generally found acceptable. Some of new data are finally compared with the predictions of a correlation, which is generally used in the absence of experimental data, and acceptable agreements between the experimental and predicted data are observed.

  14. Self-diffusion of Co2+ ions in CoBr2, CoI2 and electrolyte-diffusion of ZnSO4 in agar gel medium

    International Nuclear Information System (INIS)

    Self-diffusion of Co2+ ions in CoBr2 and CoI2 is reported in the concentration range of 10-5 to 0.25 M in 1% agar gel at 25 deg C. The deviations observed between the experimental and theoretical values of diffusion coefficients are explained by considering different types of interactions occurring in the ion-gel water system. The applicability of the transition state theory to the diffusion of ZnSO4 in agar gel medium is tested by varying the temperature as well as the gel concentration at high concentration of the electrolyte. The activation energy E and D0 value decrease with increasing gel concentration in agreement with the theory. (author)

  15. 2D numerical modelling of gas temperature in a nanosecond pulsed longitudinal He-SrBr2 discharge excited in a high temperature gas-discharge tube for the high-power strontium laser

    Science.gov (United States)

    Chernogorova, T. P.; Temelkov, K. A.; Koleva, N. K.; Vuchkov, N. K.

    2016-05-01

    An active volume scaling in bore and length of a Sr atom laser excited in a nanosecond pulse longitudinal He-SrBr2 discharge is carried out. Considering axial symmetry and uniform power input, a 2D model (r, z) is developed by numerical methods for determination of gas temperature in a new large-volume high-temperature discharge tube with additional incompact ZrO2 insulation in the discharge free zone, in order to find out the optimal thermal mode for achievement of maximal output laser parameters. A 2D model (r, z) of gas temperature is developed by numerical methods for axial symmetry and uniform power input. The model determines gas temperature of nanosecond pulsed longitudinal discharge in helium with small additives of strontium and bromine.

  16. Research reactors

    International Nuclear Information System (INIS)

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world's research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted

  17. Reactor container

    International Nuclear Information System (INIS)

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

  18. BR-10 Reactor Sodium and Sodium-Potassium Removal and Passivation. Appendix I

    International Nuclear Information System (INIS)

    Research reactor BR-10 is located in the State Scientific Centre of the Russian Federation ‘A.I. Leipunski Institute of Physics and Power Engineering’ (IPPE), Russian Federation, Obninsk. BR-10 operated from 1959 to 2002, during which time a significant amount of radioactive waste in the form of liquid-metal alkali coolants was accumulated by the research institute. A schematic illustration of the BR-10 systems is shown. The sodium coolant of the primary circuit was changed three times, and the cold traps were regenerated with the resulting spent sodium drained into tanks for long-term storage. The tanks also received sodium potassium (NaK) eutectic alloy coolant from the secondary circuit, contaminated with mercury. This was the result of de-pressurization of a double-wall steam generator tube with a mercury layer, which was part of the BR-2 reactor. The BR-2 reactor was located in the building where the BR-10 reactor was built. The volumes and activities of the drained materia are indicated

  19. MCNPX 2.6.C versus MCNPX and ORIGEN-S: state of the art for reactor core management

    International Nuclear Information System (INIS)

    This paper discusses the application of the Monte Carlo burnup code MCNPX-2.6.C for the criticality and depletion reactor core analysis of the Material Testing Research Reactor BR2 in Mol, Belgium. A comparison with the combined MCNP and ORIGEN-S fuel depletion method is presented. The accuracy of both methods, the consumption of calculation time, the depletion capabilities, the advantages and disadvantages of use of both methods are discussed. The accuracy of the criticality calculations by MCNPX-2.6.C is still lower in comparison with the validated MCNP and ORIGEN-S method

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

    International Nuclear Information System (INIS)

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

  1. Reactor building

    International Nuclear Information System (INIS)

    The whole reactor building is accommodated in a shaft and is sealed level with the earth's surface by a building ceiling, which provides protection against penetration due to external effects. The building ceiling is supported on walls of the reactor building, which line the shaft and transfer the vertical components of forces to the foundations. The thickness of the walls is designed to withstand horizontal pressure waves in the floor. The building ceiling has an opening above the reactor, which must be closed by cover plates. Operating equipment for the reactor can be situated above the building ceiling. (orig./HP)

  2. Heterogeneous reactors

    International Nuclear Information System (INIS)

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

  3. Synthesis of Tributyl Citrate Catalyzed by MgBr2·6H2O%六水合溴化镁催化合成柠檬酸三丁酯

    Institute of Scientific and Technical Information of China (English)

    骈继鑫; 邓功达; 王湘波; 王卫; 何林

    2014-01-01

    Green plasticizer is a hot topic in currently academic research and tributyl citrate (abbreviated as TBC) as a new“green”plastic plasticizer, gets the favour of people of all ages.The esterification reaction of citric acid and n-butyl alcohol catalyzed by MgBr2·6H2O was studied in this article.Different factors such as catalyst loading,temperature,the ratio of citric acid and n-butyl alcohol,reaction time were investigated through controlled experiments and the optimal reaction conditions were obtained.When the reaction was conducted for 4 h at 140 ℃ in the presence of 10 mol% MgBr2·6H2O and a 5.5-fold excess of n-butyl alcohol, the desired tributyl citrate could be obtained above 98% yield.The method has the advantages of short reaction time,fewer side effects,high efficiency of esterification,cheapness,etc.%绿色增塑剂是当前学术界研究的一个热点,而柠檬酸三丁酯作为一种新型绿色的塑料增塑剂受到了人们的青睐,为此,本文研究了六水合溴化镁催化一水合柠檬酸与正丁醇酯化合成绿色增塑剂柠檬酸三丁酯的反应条件,通过实验考察了催化剂用量、反应温度、酸醇物质的量比、反应时间等因素对反应的影响,确定出最佳反应条件为:催化剂用量10 mol%,酸醇物质的量比1∶5.5,反应温度140℃,反应时间为4 h,酯化率可达到98%以上。上述方法具有反应时间短,副反应少,酯化率高,价廉易得等优点。

  4. Plasma reactor

    OpenAIRE

    Molina Mansilla, Ricardo; Erra Serrabasa, Pilar; Bertrán Serra, Enric

    2008-01-01

    [EN] A plasma reactor that can operate in a wide pressure range, from vacuum and low pressures to atmospheric pressure and higher pressures. The plasma reactor is also able to regulate other important settings and can be used for processing a wide range of different samples, such as relatively large samples or samples with rough surfaces.

  5. Reactor physics

    International Nuclear Information System (INIS)

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

  6. A novel 3D Cu(I) coordination polymer based on Cu6Br2 and Cu2(CN)2 SBUs: in situ ligand formation and use as a naked-eye colorimetric sensor for NB and 2-NT.

    Science.gov (United States)

    Song, Jiang-Feng; Li, Yang; Zhou, Rui-Sha; Hu, Tuo-Ping; Wen, Yan-Liang; Shao, Jia; Cui, Xiao-Bing

    2016-01-14

    A novel coordination polymer with the chemical formula [Cu4Br(CN)(mtz)2]n (mtz = 5-methyl tetrazole) (), has been synthesized under solvothermal conditions and characterized by elemental analysis, infrared (IR) spectroscopy, thermal gravimetric analysis, powder X-ray diffraction and single-crystal X-ray diffraction. Interestingly, the Cu(i), CN(-) and mtz(-) in compound are all generated from an in situ translation of the original precursors: Cu(2+), acetonitrile and 1-methyl-5-mercapto-1,2,3,4-tetrazole (Hmnt). The in situ ring-to-ring conversion of Hmnt into mtz(-) was found for the first time. Structural analysis reveals that compound is a novel 3D tetrazole-based Cu(i) coordination polymer, containing both metal halide cluster Cu6Br2 and metal pseudohalide cluster Cu2(CN)2 secondary building units (SBUs), which shows an unprecedented (3,6,10)-connected topology. Notably, a pseudo-porphyrin structure with 16-membered rings constructed by four mtz(-) anions and four copper(i) ions was observed in compound . The fluorescence properties of compound were investigated in the solid state and in various solvent emulsions, the results show that compound is a highly sensitive naked-eye colorimetric sensor for NB and 2-NT (NB = nitrobenzene and 2-NT = 2-nitrotoluene). PMID:26600452

  7. Production of CH (A2 Δ) by multi-photon dissociation of (CH3)2CO, CH3NO2, CH2Br2, and CHBr3 at 213 nm

    International Nuclear Information System (INIS)

    In order to realize electrostatic Stark deceleration of CH radicals and study cold chemistry, the fifth harmonic of a YAG laser is used to prepare CH (A2Δ) molecules through using the multi-photon dissociation of (CH3)2CO, CH3NO2, CH2Br2, and CHBr3 at ∼ 213 nm. The CH product intensity is measured by using the emission spectrum of CH (A2 Δ → X2 Π). The dependence of fluorescence intensity on laser power is studied, and the probable dissociation channels are analyzed. The relationship between the fluorescence intensity and some parameters, such as the temperature of the beam source, stagnation pressure, and the time delay between the opening of pulse valve and the photolysis laser, are also studied. The influence of three different carrier gases on CH signal intensity is investigated. The vibrational and rotational temperatures of the CH (A2 Δ) product are obtained by comparing experimental data with the simulated ones from the LIFBASE program. (atomic and molecular physics)

  8. Compact Reactor

    International Nuclear Information System (INIS)

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date

  9. NEUTRONIC REACTOR

    Science.gov (United States)

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  10. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  11. Nuclear reactors

    International Nuclear Information System (INIS)

    This draft chart contains graphical symbols from which the type of (nuclear) reactor can be seen. They will serve as illustrations for graphical sketches. Important features of the individual reactor types are marked out graphically. The user can combine these symbols to characterize a specific reactor type. The basic graphical symbol is a square with a point in the centre. Functional groups can be depicted for closer specification. If two functional groups are not clearly separated, this is symbolized by a dotted line or a channel. Supply and discharge lines for coolant, moderator and fuel are specified in accordance with DIN 2481 and can be further specified by additional symbols if necessary. The examples in the paper show several different reactor types. (orig./AK)

  12. Multifunctional reactors

    OpenAIRE

    Westerterp, K.R.

    1992-01-01

    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much emphasis in research in the last decade. A survey is given of modern developments and the first successful applications on a large scale. It is explained why their application in many instances is ...

  13. NUCLEAR REACTOR

    Science.gov (United States)

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    In order to reduce neutron embrittlement of the pressue vessel of an LWR, blanked off elements are fitted at the edge of the reactor core, with the same dimensions as the fuel elements. They are parallel to each other, and to the edge of the reactor taking the place of fuel rods, and are plates of neutron-absorbing material (stainless steel, boron steel, borated Al). (HP)

  15. Breeder reactors

    International Nuclear Information System (INIS)

    The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components

  16. Research reactors - an overview

    Energy Technology Data Exchange (ETDEWEB)

    West, C.D.

    1997-03-01

    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.

  17. Reactor utilization

    International Nuclear Information System (INIS)

    In 1962, the RA reactor was operated almost three times more than in 1961, producing total of 25 555 MWh. Diagram containing comparative data about reactor operation for 1960, 1961, and 1962, percent of fuel used and U-235 burnup shows increase in reactor operation. Number of samples irradiated was 659, number of experiments done was 16. mean powered level was 5.93 MW. Fuel was added into the core twice during the reporting year. In fact the core was increased from 56 to 68 fuel channels and later to 84 fuel channels. Fuel was added to the core when the reactivity worth decreased to the minimum operation level due to burnup. In addition to this 5 central fuel channels were exchanged with fresh fuel in february for the purpose of irradiation in the VISA-2 channel

  18. Reactor Neutrinos

    CERN Document Server

    Lasserre, T; Lasserre, Thierry; Sobel, Henry W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrino oscillation physics in the last years. It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the last undetermined neutrino mixing angle theta13. We conclude by opening on possible use of neutrinos for Society: NonProliferation of Nuclear materials and Geophysics.

  19. Nuclear reactors

    International Nuclear Information System (INIS)

    A nuclear reactor has a large prompt negative temperature coefficient of reactivity. A reactor core assembly of a plurality of fluid-tight fuel elements is located within a water-filled tank. Each fuel element contains a solid homogeneous mixture of 50-79 w/o zirconium hydride, 20-50 w/o uranium and 0.5-1.5 W erbium. The uranium is not more than 20 percent enriched, and the ratio of hydrogen atoms to zirconium atoms is between 1.5:1 and 7:1. The core has a long lifetime, E.G., at least about 1200 days

  20. Nuclear reactors

    International Nuclear Information System (INIS)

    In a liquid cooled nuclear reactor, the combination is described for a single-walled vessel containing liquid coolant in which the reactor core is submerged, and a containment structure, primarily of material for shielding against radioactivity, surrounding at least the liquid-containing part of the vessel with clearance therebetween and having that surface thereof which faces the vessel make compatible with the liquid, thereby providing a leak jacket for the vessel. The structure is preferably a metal-lined concrete vault, and cooling means are provided for protecting the concrete against reaching a temperature at which damage would occur. (U.S.)

  1. Nuclear reactor

    International Nuclear Information System (INIS)

    In an improved reactor core for a high conversion BWR reactor, Pu-breeding type BWR type reactor, Pu-breeding type BWR type rector, FEBR type reactor, etc., two types of fuel assemblies are loaded such that fuel assemblies using a channel box of a smaller irradiation deformation ratio are loaded in a high conversion region, while other fuel assemblies are loaded in a burner region. This enables to suppress the irradiation deformation within an allowable limit in the high conversion region where the fast neutron flux is high and the load weight from the inside of the channel box due to the pressure loss is large. At the same time, the irradiation deformation can be restricted within an allowable limit without deteriorating the neutron economy in the burner region in which fast neutron flux is low and the load weight from the inside of the channel box is small since a channel box with smaller neutron absorption cross section or reduced wall thickness is charged. As a result, it is possible to prevent structural deformations such as swelling of the channel box, bending of the entire assemblies, bending of fuel rods, etc. (K.M.)

  2. The Properties of Glasses in the PoO-PbB2-PbF2-P2O5System%PbO-PbBr2-PbF2-P2O5系统玻璃的性质

    Institute of Scientific and Technical Information of China (English)

    赵宏生; 周万城; 朱冬梅

    2000-01-01

    对新研制的PbO-PbBr2-PbF2-P2O5系铅卤磷酸盐玻璃的化学成分以及玻璃转变温度、密度、抗潮解性和紫外光谱等性质进行了分析.结果表明,PbO-PbBr2-PbF2-P2O5系铅卤磷酸盐玻璃具有较低的玻璃转变温度,较好的抗潮解性,密度在5.0g.Cm-3以上,并且在可见光区域具有较好的透光率.

  3. Reactor container

    International Nuclear Information System (INIS)

    A reactor container has a suppression chamber partitioned by concrete side walls, a reactor pedestal and a diaphragm floor. A plurality of partitioning walls are disposed in circumferential direction each at an interval inside the suppression chamber, so that independent chambers in a state being divided into plurality are formed inside the suppression chamber. The partition walls are formed from the bottom portion of the suppression chamber up to the diaphragm floor to isolate pool water in a divided state. Operation platforms are formed above the suppression chamber and connected to an access port. Upon conducting maintenance, inspection or repairing, a pump is disposed in the independent chamber to transfer pool water therein to one or a plurality of other independent chambers to make it vacant. (I.N.)

  4. Reactor building

    International Nuclear Information System (INIS)

    The present invention concerns a structure of ABWR-type reactor buildings, which can increase the capacity of a spent fuel storage area at a low cost and improved earthquake proofness. In the reactor building, the floor of a spent fuel pool is made flat, and a depth of the pool water satisfying requirement for shielding is ensured. In addition, a depth of pool water is also maintained for a equipment provisionally storing pool for storing spent fuels, and a capacity for a spent fuel storage area is increased by utilizing surplus space of the equipment provisionally storing pool. Since the flattened floor of the spent fuel pool is flushed with the floor of the equipment provisionally storing pool, transfer of horizontal loads applied to the building upon occurrence of earthquakes is made smooth, to improve earthquake proofness of the building. (T.M.)

  5. Nuclear reactors

    International Nuclear Information System (INIS)

    Disclosed is a nuclear reactor cooled by a freezable liquid has a vessel for containing said liquid and comprising a structure shaped as a container, and cooling means in the region of the surface of said structure for effecting freezing of said liquid coolant at and for a finite distance from said surface for providing a layer of frozen coolant on and supported by said surface for containing said liquid coolant. In a specific example, where the reactor is sodium-cooled, the said structure is a metal-lined concrete vault, cooling is effected by closed cooling loops containing NaK, the loops extending over the lined surface of the concrete vault with outward and reverse pipe runs of each loop separated by thermal insulation, and air is flowed through cooling pipes embedded in the concrete behind the metal lining. 7 claims, 3 figures

  6. NEUTRONIC REACTORS

    Science.gov (United States)

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  7. Nuclear reactor

    International Nuclear Information System (INIS)

    The liquid metal (sodium) cooled fast breeder reactor has got fuel subassemblies which are bundled and enclosed by a common can. In order to reduce bending of the sides of the can because of the load caused by the coolant pressure the can has got a dodecagon-shaped crosssection. The surfaces of the can may be of equal width. One out of two surfaces may also be convex towards the center. (RW)

  8. Nuclear reactor

    International Nuclear Information System (INIS)

    A detector having high sensitivity to fast neutrons and having low sensitivity to thermal neutrons is disposed for reducing influences of neutron detector signals on detection values of neutron fluxes when the upper end of control rod pass in the vicinity of the neutron flux detector. Namely, the change of the neutron fluxes is greater in the thermal neutron energy region while it is smaller in the fast neutron energy region. This is because the neutron absorbing cross section of B-10 used as neutron absorbers of control rods is greater in the thermal neutron region and it is smaller in the fast neutron region. As a result, increase of the neutron detection signals along with the local neutron flux change can be reduced, and detection signals corresponding to the reactor power can be obtained. Even when gang withdrawal of operating a plurality of control rods at the same time is performed, the reactor operation cycle can be measured accurately, thereby enabling to shorten the reactor startup time. (N.H.)

  9. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    In a BWR type nuclear reactor, the number of first fuel assemblies (uranium) loaded in a reactor core is smaller than that of second fuel assemblies (mixed oxide), the average burnup degree upon take-out of the first fuel assemblies is reduced to less than that of the second fuel assemblies, and the number of the kinds of the fuel rods constituting the first fuel assemblies is made smaller than that of the fuel rods constituting the second fuel assemblies. As a result, the variety of the plutonium enrichment degree is reduced to make the distribution of the axial enrichment degree uniform, thereby enabling to simplify the distribution of the enrichment degree. Then the number of molding fabrication steps for MOX fuel assemblies can be reduced, thereby enabling to reduce the cost for molding and fabrication. (N.H.)

  10. Achievements and projects in Belgium in the field of reactor shielding

    International Nuclear Information System (INIS)

    Four reactors are in operation in Belgium: the BR-1 and BR-2 which are research and materials testing reactors, the BR-3 which has a power output of 11 MW(e) and the BR-02 which is the nuclear mock-up of the BR-2. In 1965 the BR-3 will be operating on the spectral-shift principle. The VENUS reactor, the critical mock-up of the future BR-3 core, should commence operation in April 1964. By the end of 1964, the University of Ghent will have at its disposal a swimming-pool type reactor. The fast reactors MASURCA and HARMONIE were studied on behalf of the French Atomic Energy Commission and EURATOM. Most of these reactors raise no shielding problems other than the conventional ones. The VULCAIN prototype shield will be designed in accordance with the specifications characteristic of naval reactors, in which an optimization of volume, weight and cost is of primary interest. Study of these problems has begun. The radiation damage of the BR-3 pressure vessel is a problem when equipped with the future spectral-shift core. The level and spectrum of neutron irradiation will be experimentally determined in the VENUS facility. All shielding studies were carried out by conventional methods and in no case was any research project, either experimental or theoretical, undertaken along these lines. Works were accepted on the basis of their practical utility and no specific experiment for the verification of the theoretical studies was ever carried out. Among the main theoretical problems encountered in shielding design, the following should be cited: Gamma and neutron attenuation along straight and stepped ducts; Back-scattering of gammas and neutrons from air (skyshine), solid and liquid materials, and shielding of the reflected radiations; Gamma and neutron propagation at long distance in air; Capture gamma spectra as a function of the energy of the incident neutron; and Gamma radiations from neutron inelastic scattering. In conjunction with a Ferranti-Mercury computer, the

  11. Types of Nuclear Reactors

    International Nuclear Information System (INIS)

    The presentation is based on the following areas: Types of Nuclear Reactors, coolant, moderator, neutron spectrum, fuel type, pressurized water reactor (PWR), boiling water reactor (BWR) reactor pressurized heavy water (PHWR), gas-cooled reactor, RBMK , Nuclear Electricity Generation,Challenges in Nuclear Technology Deployment,EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR.

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    A nuclear reactor is described in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assemblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters in the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters in the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance

  13. Nuclear research reactors

    International Nuclear Information System (INIS)

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.)

  14. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Cover gas spaces for primary coolant vessel, such as a reactor container, a pump vessel and an intermediate heat exchanger vessel are in communication with each other by an inverted U-shaped pressure conduit. A transmitter and a receiver are disposed to the pressure conduit at appropriate positions. If vibration frequencies (pressure vibration) from low frequency to high frequency are generated continuously from the transmitter to the inside of the communication pipe, a resonance phenomenon (air-column resonance oscillation) is caused by the inherent frequency or the like of the communication pipe. The frequency of the air-column resonance oscillation is changed by the inner diameter and the clogged state of the pipelines. Accordingly, by detecting the change of the air-column oscillation characteristics by the receiver, the clogged state of the flow channels in the pipelines can be detected even during the reactor operation. With such procedures, steams of coolants flowing entrained by the cover gases can be prevented from condensation and coagulation at a low temperature portion of the pipelines, otherwise it would lead clogging in the pipelines. (I.N.)

  16. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  17. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  18. In-reactor performance of prototype SBR MOX fuel

    International Nuclear Information System (INIS)

    As part of the international Callisto experiment, BNFL have undertaken the base irradiation, ramp testing and post-irradiation examination of two fuel rods, manufactured by BNFL's Short Binderless Route (SBR) for MOX fuel. Although of only short (lm) length, the rods were in other respects of standard PWR geometry. Base irradiation was performed in the Callisto loop of the BR2 reactor and achieved burn-ups of 17 GWd/tM, 28 GWd/tM peak pellet, at moderate to high power levels. No problems were encountered during base irradiation, confirmed by intermediate examination. In a separate BR2 capsule, a ramp test was then performed on one of the two rods. Again, the rod showed no problems and was discharged intact. The ramp test conditions employed were somewhat more onerous than the level corresponding to the best-estimate failure level for standard UO2 fuel. The survival of the rod confirms the general observation that the PCI failure resistance of MOX fuel is superior to that of standard UO2. The irradiation has been followed by a detailed programme of non-destructive and destructive hot-cell examinations on both of the rods. Profilometry, ECT and gamma-scanning confirmed the overall satisfactory performance of the rods and the absence of incipient damage. Puncture testing and gas analysis showed the fission gas release levels in both the ramped and unramped rods to be in line with what would be expected for a UO2 rod. The satisfactory microstructure of the fuel was confirmed via optical ceramography and autoradiography. This paper will describe the fuel manufacture, base irradiation and ramp test conditions, and will provide a summary of the results of the PIE programme. (author)

  19. Reactor container

    International Nuclear Information System (INIS)

    Purpose: To prevent shocks exerted on a vent head due to pool-swell caused within a pressure suppression chamber (disposed in a torus configuration around the dry well) upon loss of coolant accident in BWR type reactors. Constitution: The following relationship is established between the volume V (m3) of a dry well and the ruptured opening area A (m2) at the boundary expected upon loss of coolant accident: V >= 30340 (m) x A Then, the volume of the dry well is made larger than the ruptured open area, that is, the steam flow rate of leaking coolants upon loss of coolant accident to decrease the pressure rise in the dry well at the initial state where loss of coolant accident is resulted. Accordingly, the pressure of non-compressive gases jetted out from the lower end of the downcomer to the pool water is decreased to suppress the pool-swell. (Ikeda, J.)

  20. Hybrid reliability model for nuclear reactor safety system

    International Nuclear Information System (INIS)

    The dependability of critical safety systems needs to be quantitatively determined in order to verify their effectiveness, e.g. with regard to regulatory requirements. Since modular redundant safety systems are not required for normal operation, their reliability is strongly dependent on periodic inspection. Several modeling methods for the quantitative assessment of dependability are described in the literature, with a broad variation in complexity and modeling power. Static modeling techniques such as fault tree analysis (FTA) or reliability block diagrams (RBD) are not capable of capturing redundancy and repair or test activities. Dynamic state space based models such as continuous time Markov chains (CTMC) are more powerful but often result in very large, intractable models. Moreover, exponentially distributed state residence times are not a correct representation of actual residence times associated with repair activities or periodic inspection. In this study, a hybrid model combines a system level RBD with a CTMC to describe the dynamics. The effects of periodic testing are modeled by redistributing state probabilities at deterministic test times. Applying the method to the primary safety shutdown system of the BR2(Belgian Reactor 2)—nuclear research reactor, resulted in a quantitative as well as a qualitative assessment of its reliability.

  1. Survey of research reactors

    International Nuclear Information System (INIS)

    A survey of reasearch reactors based on the IAEA Nuclear Research Reactor Data Base (RRDB) was done. This database includes information on 273 operating research reactors ranging in power from zero to several hundred MW. From these 273 operating research reactors 205 reactors have a power level below 5 MW, the remaining 68 reactors range from 5 MW up to several 100 MW thermal power. The major reactor types with common design are: Siemens Unterrichtsreaktors, 1.2 Argonaut reactors, Slowpoke reactors, the miniature neutron source reactors, TRIGA reactors, material testing reactors and high flux reactors. Technical data such as: power, fuel material, fuel type, enrichment, maximum neutron flux density and experimental facilities for each reactor type as well as a description of their utilization in physics and chemistry, medicine and biology, academic research and teaching, training purposes (students and physicists, operating personnel), industrial application (neutron radiography, silicon neutron transmutation doping facilities) are provided. The geographically distribution of these reactors is also shown. As conclusions the author discussed the advantages (low capital cost, low operating cost, low burn up, simple to operate, safe, less restrictive containment and sitting requirements, versatility) and disadvantages (lower sensitivity for NAA, limited radioisotope production, limited use of neutron beams, limited access to the core, licensing) of low power research reactors. 24 figs., refs. 15, Tab. 1 (nevyjel)

  2. Department of reactor technology

    International Nuclear Information System (INIS)

    The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  3. RB reactor noise analysis

    International Nuclear Information System (INIS)

    Statistical fluctuations of reactivity represent reactor noise. Analysis of reactor noise enables determining a series of reactor kinetic parameters. Fluctuations of power was measured by ionization chamber placed next to the tank of the RB reactor. The signal was digitized by an analog-digital converter. After calculation of the mean power, 3000 data obtained by sampling were analysed

  4. Final report on in-reactor uniaxial tensile deformation of pure iron and Fe-Cr alloy

    DEFF Research Database (Denmark)

    Singh, Bachu Narain; Huang, X.; Tähtinen, S.;

    performed a series of uniaxial tensile tests on Fe-Cr and pure iron specimens in the BR-2 reactor at Mol (Belgium). The present report first provides a brief description of the test facilities and the procedure used for performing the in-reactor tests. The results on the mechanical response of materials......Traditionally, the effect of irradiation on mechanical properties of metals and alloys is determined using post-irradiation tests carried out on pre-irradiated specimens and in the absence of irradiation environment. The results of these tests may not be representative of deformation behaviour of...... during these tests are presented in the form of stress-displacement dose and the conventional stress-strain curves. For comparison, the results of post-irradiation tests and tests carried out on unirradiated specimens are also presented. Results of microstructural investigations on the unirradiated and...

  5. Research Nuclear Reactors

    International Nuclear Information System (INIS)

    Published in English and in French, this large report first proposes an overview of the use and history of research nuclear reactors. It discusses their definition, and presents the various types of research reactors which can be either related to nuclear power (critical mock-ups, material test reactors, safety test reactors, training reactors, prototypes), or to research (basic research, industry, health), or to specific particle physics phenomena (neutron diffraction, isotope production, neutron activation, neutron radiography, semiconductor doping). It reports the history of the French research reactors by distinguishing the first atomic pile (ZOE), and the activities and achievements during the fifties, the sixties and the seventies. It also addresses the development of instrumentation for research reactors (neutron, thermal, mechanical and fission gas release measurements). The other parts of the report concern the validation of neutronics calculations for different reactors (the EOLE water critical mock-up, the MASURCA air critical mock-up dedicated to fast neutron reactor study, the MINERVE water critical mock-up, the CALIBAN pulsed research reactor), the testing of materials under irradiation (OSIRIS reactor, laboratories associated with research reactors, the Jules Horowitz reactor and its experimental programs and related devices, irradiation of materials with ion beams), the investigation of accident situations (on the CABRI, Phebus, Silene and Jules Horowitz reactors). The last part proposes a worldwide overview of research reactors

  6. Reactor Physics Training

    International Nuclear Information System (INIS)

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  7. Introduction of Nuclear Reactor Engineering

    International Nuclear Information System (INIS)

    This book introduces development, status, supply and demand and resource of nuclear reactor. It deals with basic knowledge of nuclear reactor, which are reactor system, heat recovery in reactor core, structural feature in reactor, materials of structure in reactor, shielding of gamma ray, shielding of reactor, safety and environmental problem of nuclear power plant, nuclear fuel and economical efficiency of nuclear energy.

  8. Final report on in-reactor creep-fatigue deformation behaviour of a CuCrZr alloy: COFAT 2

    International Nuclear Information System (INIS)

    The main objective of the present work was to determine experimentally the mechanical response and resulting microstructural changes in CuCrZr (HT1) alloy exposed concurrently to flux of neutrons and creep-fatigue cyclic loading directly in a fission reactor. Using specially designed test 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-of-reactor creep-fatigue tests were also carried out. In the following we first describe the details of the creep-fatigue experiments. We then present the main results on the mechanical response of the material in the form of hysteresis loops and the maximum stress amplitude as a function of the number of creep-fatigue cycles during the out-of-reactor and the in-reactor tests carried out at different strain amplitudes. Finally, the dependence of the number of cycles to failure (i.e. creep-fatigue lifetime) on the strain amplitudes is shown. The details of microstructure of the specimens tested out-of-reactor as well as in the reactor were investigated using transmission electron microscopy. The main results on the mechanical response as well as changes in the microstructure are briefly discussed. The main conclusion emerging from the present work is that the lifetime of the in-reactor tested specimens is by a factor of about two longer than in the case of corresponding out-of-reactor tests. (au)

  9. Safeguarding research reactors

    International Nuclear Information System (INIS)

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

  10. Research nuclear reactors

    International Nuclear Information System (INIS)

    Since the divergence of the first nuclear reactor in 1942, about 600 research or test reactors have been built throughout the world. Today 255 research reactors are operating in 57 countries and about 70% are over 25 years old. Whereas there are very few reactor types for power plants because of rationalization and standardisation, there is a great diversity of research reactors. We can divide them into 2 groups: heavy water cooled reactors and light water moderated reactors. Heavy water cooled reactors are dedicated to the production of high flux of thermal neutrons which are extracted from the core by means of neutronic channels. Light water moderated reactors involved pool reactors and slightly pressurized closed reactors, they are polyvalent but their main purposes are material testing, technological irradiations, radionuclide production and neutron radiography. At the moment 8 research reactors are being built in Canada, Germany, Iran, Japan, Kazakhstan, Morocco, Russia and Slovakia and 8 others are planned in 7 countries (France, Indonesia, Nigeria, Russia, Slovakia, Thailand and Tunisia. Different research reactors are described: Phebus, Masurca, Phenix and Petten HFR. The general principles of nuclear safety applied to test reactors are presented. (A.C.)

  11. Nuclear reactor building

    International Nuclear Information System (INIS)

    Purpose: To prevent seismic vibrations of external buildings from transmitting to the side walls of a reactor container in a tank type FBR reactor building. Constitution: The reactor building is structured such that the base mat for a reactor container chamber and a reactor container is separated from the base mat for the walls of building, and gas-tight material such as silicon rubber is filled in the gap therebetween. With such a constitution, even if the crane-supporting wall vibrates violently upon occurrence of earthqualkes, the seismic vibrations do not transmit toward the reactor container chamber. (Horiuchi, T.)

  12. Ship propulsion reactors technology

    International Nuclear Information System (INIS)

    This paper takes the state of the art on ship propulsion reactors technology. The french research programs with the corresponding technological stakes, the reactors specifications and advantages are detailed. (A.L.B.)

  13. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

  14. Process heat reactors

    International Nuclear Information System (INIS)

    The consumption of heat, for industrial and domestic needs, takes up half of the national energy supply; direct utilization of the heat produced by nuclear reactors could therefore contribute to reduce the deficit in the energetic results. The restraints proper to heat consumption (dispersal and variety of consumers, irregular demand) involve the development of the heat transport system structures and adequate nuclear reactors. With this in view, the Commissariat a l'Energie Atomique and Technicatome are developing the CAS reactor series, pressurized water reactors (PWR), (CAS 3G reactor with a power of 420 MW.th.), and the Thermos reactor (100 MW.th.), directly conceived to produce heat at 1200C and whose technology derives from the experimental pool reactors type. In order to prove the value of the Thermos design, an experimental reactor should soon be constructed in the Saclay nuclear research centre

  15. Reactor System Design

    International Nuclear Information System (INIS)

    SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

  16. Nuclear Reactor RA Safety Report, Vol. 11, Reactor operation

    International Nuclear Information System (INIS)

    This volume includes the following chapters describing: Organisation of reactor operation (including operational safety, fuel management, and regulatory rules for RA reactor operation); Control and maintenance of reactor components (reactor core, nuclear fuel, heavy water and cover gas systems, mechanical structures, electric power supply system, reactor instrumentation); Quality assurance and Training of the reactor personnel

  17. The Chernobylsk reactor accident

    International Nuclear Information System (INIS)

    The construction, the safety philosophy, the major reactor physical parameters of RBMK-1000 type reactor units and the detailed description of the Chernobylsk-4 reactor accident, its causes and conclusions, the efforts to reduce the consequences on the reactor site and in the surroundings are discussed based on different types of Soviet documents including the report presented to the IAEA by the Soviet Atomic Energy Agency in August 1986. (V.N.)

  18. Zero energy reactor 'RB'

    International Nuclear Information System (INIS)

    In 1958 the zero energy reactor RB was built with the purpose of enabling critical experiments with various reactor systems to be carried out. The first core assembly built in this reactor consists of heavy water as moderator and natural uranium metal as fuel. In order to be able to obtain very accurate results when measuring the main characteristics of the assembly the reactor was built as a completely bare system. (author)

  19. High solids fermentation reactor

    Science.gov (United States)

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-01-01

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  20. Fossil nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Maurette, M.

    1976-01-01

    The discussion of fossil nuclear reactors (the Oklo phenomenon) covers the earth science background, neutron-induced isotopes and reactor operating conditions, radiation-damage studies, and reactor modeling. In conclusion possible future studies are suggested and the significance of the data obtained in past studies is summarized. (JSR)

  1. Fusion reactor studies

    International Nuclear Information System (INIS)

    A review is given of fusion reactor systems studies, the objectives of these studies are outlined and some recent conceptual reactor designs are described. The need for further studies in greater depth is indicated so that progress towards a commercial fusion reactor may be consolidated. (U.K.)

  2. Reactor power measuring device

    International Nuclear Information System (INIS)

    The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

  3. Crystal structures of hydrates of simple inorganic salts. II. Water-rich calcium bromide and iodide hydrates: CaBr2 · 9H2O, CaI2 · 8H2O, CaI2 · 7H2O and CaI2 · 6.5H2O.

    Science.gov (United States)

    Hennings, Erik; Schmidt, Horst; Voigt, Wolfgang

    2014-09-01

    Single crystals of calcium bromide enneahydrate, CaBr(2) · 9H2O, calcium iodide octahydrate, CaI(2) · 8H2O, calcium iodide heptahydrate, CaI(2) · 7H2O, and calcium iodide 6.5-hydrate, CaI(2) · 6.5H2O, were grown from their aqueous solutions at and below room temperature according to the solid-liquid phase diagram. The crystal structure of CaI(2) · 6.5H2O was redetermined. All four structures are built up from distorted Ca(H2O)8 antiprisms. The antiprisms of the iodide hydrate structures are connected either via trigonal-plane-sharing or edge-sharing, forming dimeric units. The antiprisms in calcium bromide enneahydrate are monomeric. PMID:25186361

  4. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  5. Nuclear reactor repairing device

    International Nuclear Information System (INIS)

    Purpose: To enable free repairing of an arbitrary position in an LMFBR reactor. Constitution: A laser light emitted from a laser oscillator installed out of a nuclear reactor is guided into a portion to be repaired in the reactor by using a reflecting mirror, thereby welding or cutting it. The guidance of the laser out of the reactor into the reactor is performed by an extension tube depending into a through hole of a rotary plug, and the guidance of the laser light into a portion to be repaired is performed by the transmitting and condensing action of the reflecting mirror. (Kamimura, M.)

  6. Fundamentals of reactor chemistry

    International Nuclear Information System (INIS)

    In the Nuclear Engineering School of JAERI, many courses are presented for the people working in and around the nuclear reactors. The curricula of the courses contain also the subject material of chemistry. With reference to the foreign curricula, a plan of educational subject material of chemistry in the Nuclear Engineering School of JAERI was considered, and the fundamental part of reactor chemistry was reviewed in this report. Since the students of the Nuclear Engineering School are not chemists, the knowledge necessary in and around the nuclear reactors was emphasized in order to familiarize the students with the reactor chemistry. The teaching experience of the fundamentals of reactor chemistry is also given. (author)

  7. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  8. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPRTM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENATM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENATM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

  9. The Maple reactor project

    International Nuclear Information System (INIS)

    MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)

  10. High temperature reactors

    International Nuclear Information System (INIS)

    With the advent of high temperature reactors, nuclear energy, in addition to producing electricity, has shown enormous potential for the production of alternate transport energy carrier such as hydrogen. High efficiency hydrogen production processes need process heat at temperatures around 1173-1223 K. Bhabha Atomic Research Centre (BARC), is currently developing concepts of high temperature reactors capable of supplying process heat around 1273 K. These reactors would provide energy to facilitate combined production of hydrogen, electricity, and drinking water. Compact high temperature reactor is being developed as a technology demonstrator for associated technologies. Design has been also initiated for a 600 MWth innovative high temperature reactor. High temperature reactor development programme has opened new avenues for research in areas like advanced nuclear fuels, high temperature and corrosion resistant materials and protective coatings, heavy liquid metal coolant technologies, etc. The paper highlights design of these reactors and their material related requirements

  11. Spinning fluids reactor

    Science.gov (United States)

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  12. Reactor Safety: Introduction

    International Nuclear Information System (INIS)

    The programme of the Reactor Safety Division focuses on the development of expertise on materials behaviour under irradiation for fission and fusion oriented applications. Furthermore, as nuclear energy needs international public acceptance with respect to safety and efficient management of natural resources and wants to reduce the burden of nuclear waste, the Reactor Safety Division enhanced its efforts to develop the MYRRHA project. MYRRHA, an accelerator driven sub-critical system, might have the potential to cope in Europe with the above mentioned constraints on acceptability and might serve as a technological platform for GEN IV reactor development, in particular the Liquid Metal Fast Reactor.The Reactor Safety Division gathers three research entities that are internationally recognised: the Reactor Materials Research department, the Reactor Physics and MYRRHA department and the Instrumentation department.The objectives of Reactor Materials Research are: to evaluate the integrity and behaviour of structural materials and nuclear fuels used in present and future nuclear power industry; to perform research to unravel and understand the parameters that determine the material and fuel behaviour under or after irradiation; to contribute to the interpretation and modelling of the materials and fuels behaviour in order to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the Reactor Materials Research department concentrate on four distinct disciplines: Reactor Pressure Vessel Steel embrittlement Stress corrosion cracking in reactor coolant environment, including Irradiation Assisted Stress Corrosion Cracking; Nuclear Fuel characterisation and development of new fuel types for commercial and test reactors. Development of materials for Fusion and advanced nuclear fission reactors. The safe operation of present nuclear power plants relies primarily on the integrity of the reactor pressure vessel

  13. Development of chemical sensors for Fast Breeder Reactor Technology

    International Nuclear Information System (INIS)

    Fast breeder reactors use liquid sodium as heat transfer medium and generate high pressure steam at the steam generator to run the turbine. This high pressure steam is separated from sodium coolant by ferritic steel tubes of 4 to 5 mm wall thickness. Development of any material defect in these heat exchanger tubes during their service would result in the ingress of high pressure steam into the sodium circuit leading to sodium-water reactions. A high temperature electrochemical hydrogen sensor based on CaBr2-CaHBr solid electrolyte and capable of measuring ppb levels of dissolved hydrogen in sodium has been developed at the laboratory. A very sensitive system, using thermal conductivity detector and semiconducting oxide based sensor has also been developed for continuous monitoring of hydrogen levels in argon cover gas. An electrochemical carbon sensor using a molten carbonate electrolyte and an oxygen sensor based on yttria doped thoria oxide electrolyte are also under advanced stage of development for measuring carbon and oxygen levels in sodium. Materials chemistry issues involved in developing these sensors and their operational experience in sodium system are highlighted in this presentation

  14. Research reactors in Argentina

    International Nuclear Information System (INIS)

    Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

  15. The experimental testing of the long-term behaviour of cemented radioactive waste from nuclear research reactors in the geological disposal conditions of the boom clay

    International Nuclear Information System (INIS)

    Liquid wastes, resulting from the reprocessing of spent nuclear fuel from the BR-2 Materials Testing Reactor, will be conditioned in a cement matrix at the dedicated cementation facility of UKAEA at Dounreay. In Belgium, the Boom clay formation is studied as a potential host rock for the final geological disposal of cemented research reactor waste. In view of evaluating the safety of disposal, laboratory leach experiments and in situ tests have been performed. Leach experiments in synthetic clay water indicate that the leach rates of calcium and silicium are relatively low compared to those of sodium and potassium. In situ experiments on inactive samples are performed in order to obtain information on the microchemical and mineralogical changes of the cemented waste in contact with the Boom clay. Finally, results from a preliminary performance assessment calculation suggest a non-negligible maximum dose rate of 5 10-9 Sv/a for 129I. (author)

  16. Thai research reactor

    International Nuclear Information System (INIS)

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

  17. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  18. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.)

  19. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  20. Evaluation of research reactors

    International Nuclear Information System (INIS)

    The present status of research reactors with highly enriched (93%) uranium fuel at JAERI, JRR-2 and JMTR is described. JRR-2 is a heterogeneous type of reactor, using heavy water as moderator and coolant. It uses both MTR type and cylindrical type of fuel elements. The maximum thermal power and the thermal neutron flux are 10 MW and 2x1014 n/cm2 see respectively. The reactor has been used for various experiments such as solid state physics, material irradiation, reactor fuel irradiation and radioisotope production. The JMTR is a multi-purpose tank type material testing reactor, and light water moderator and coolant, operated at 50 MW. The evaluation of lower enriched fuel and its consequences for both reactors is considered more especially

  1. Multipurpose research reactors

    International Nuclear Information System (INIS)

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

  2. The nuclear soliton reactor

    International Nuclear Information System (INIS)

    The basic reactor physics of a completely novel nuclear fission reactor design - the soliton-reactor - is presented on the basis of a simple model. In such a reactor, the neutrons in the critical region convert either fertile material in the adjacent layers into fissile material or reduce the poisoning of fissile material in such a manner that successively new critical regions emerge. The result is an autocatalytically driven burn-up wave which propagates throughout the reactor. Thereby, the relevant characteristic spatial distributions (neutron flux, specific power density and the associated particle densities) are solitons - wave phenomena resulting from non-linear partial differential equations which do not change their shape during propagation. A qualitativley new kind of harnessing nuclear fission energy may become possible with fuel residence times comparable with the useful lifetime of the reactor system. In the long run, fast breeder systems which exploit the natural uranium and thorium resources, without any reprocessing capacity are imaginable. (orig.)

  3. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  4. Fusion reactor research

    International Nuclear Information System (INIS)

    This work covers four separate areas: (1) development of technology for processing liquid lithium from blankets, (2) investigation of hydrogen isotope permeation in candidate structural metals and alloys for near-term fusion reactors, (3) analytical studies encompassing fusion reactor thermal hydraulics, tritium facility design, and fusion reactor safety, and (4) studies involving dosimetry and damage analysis. Recent accomplishments in each of these areas are summarized

  5. The Integral Fast Reactor

    International Nuclear Information System (INIS)

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

  6. The replacement research reactor

    International Nuclear Information System (INIS)

    As a consequences of the government decision in September 1997. ANSTO established a replacement research reactor project to manage the procurement of the replacement reactor through the necessary approval, tendering and contract management stages This paper provides an update of the status of the project including the completion of the Environmental Impact Statement. Prequalification and Public Works Committee processes. The aims of the project, management organisation, reactor type and expected capabilities are also described

  7. PFBR reactor protection

    International Nuclear Information System (INIS)

    Design philosophy adopted for Prototype Fast breeder Reactor (PFBR) is a classical one and has the following features: triplicated sensors for measuring important safety parameters; two independent reactor protection Logic Systems based on solid state devices; reactivity control achieved by control rods; gas equipped modules at the core blanket interface providing negative reactivity. Design verification of these features showed that safety of the reactor can be achieved by a traditional approach since the inherent features of LMFBR make this easy

  8. TRIGA reactor characteristics

    International Nuclear Information System (INIS)

    This module describes the general design, characteristics and parameters of TRIGA reactors and fuels. It is recommended that most of this information should be incorporated into any reactor operator training program and, in many cases, the facility Safety Analysis Report. It is oriented to teach the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and the differences between TRIGA fuels and others more traditional reactor fuels. (nevyjel)

  9. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported

  10. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

  11. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

  12. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)

  13. New reactor concepts

    International Nuclear Information System (INIS)

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

  14. Reactor construction steels

    International Nuclear Information System (INIS)

    The basic functions of light water reactor components are shown on the example of a pressurized water reactor and the requirements resulting therefrom for steel, the basic structural material, are derived. A detailed analysis of three main groups of reactor steels is presented and the applications are indicated of low-alloyed steels, high-alloyed austenitic steels, and steels with a high content of Ni and of alloying additions for steam generator pipes. An outline is given of prospective fast breeder reactor steels. (J.K.)

  15. Commercialization of fast reactors

    International Nuclear Information System (INIS)

    Comparative analysis has been performed of capital and fuel cycle costs for fast BN-type and pressurized light water VVER-type reactors. As a result of materials demand and components costs comparison of NPPs with VVER-1000 and BN-600 reactors, respectively, conclusion was made, that under equal conditions of the comparison, NPP with fast reactor had surpassed the specific capital cost of NPP with VVER by about 30 - 40 %. Ways were determined for further decrease of this difference, as well as for the fuel cycle cost reduction, because at present it is higher than that of VVER-type reactors. (author)

  16. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant

  17. Natural convection type reactor

    International Nuclear Information System (INIS)

    In a natural convection type nuclear reactor, a reactor core is disposed such that the top of the reactor core is always situated in a flooded position even if pipelines connected to the pressure vessel are ruptured and the level at the inside of the reactor vessel is reduced due to flashing. Further, a lower dry well situated below the pressure vessel is disposed such that it is in communication with a through hole to a pressure suppression chamber situated therearound and the reactor core is situated at the level lower than that of the through hole. If pipelines connected to the pressure vessel are ruptured to cause loss of water, although the water level is lowered after the end of the flashing, the reactor core is always flooded till the operation of a pressure accummulation water injection system to prevent the top of the reactor core even from temporary exposure. Further, injected water is discharged to the outside of the pressure vessel, transferred to the lower dry well, and flows through the through hole to the pressure control chamber and cools the surface of the reactor pressure vessel from the outside. Accordingly, the reactor core is cooled to surely and efficiently remove the after-heat. (N.H.)

  18. INVAP's Research Reactor Designs

    International Nuclear Information System (INIS)

    INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper summarizes the general features and utilization of several INVAP research reactor designs, from subcritical and critical assemblies to high-power reactors IAEA safety

  19. Reactor power control device

    International Nuclear Information System (INIS)

    The present invention provides a control device which can conduct scram and avoid lowering of the power of a nuclear power plant upon occurrence of earthquakes. Namely, the device of the present invention comprises, in addition to an existent power control device, (1) an earthquake detector for detecting occurrence and annihilation of earthquakes and (2) a reactor control device for outputting control rod operation signals and reactor core flow rate control signals depending on the earthquake detection signals from the detector, and reactor and plant information. With such a constitution, although the reactor is vibrated by earthquakes, the detector detects slight oscillations of the reactor by initial fine vibration waves as premonitory symptoms of serious earthquakes. The earthquake occurrence signals are outputted to the reactor control device. The reactor control device, receiving the signals, changes the position of control rods by way of control rod driving mechanisms to make the axial power distribution in the reactor core to a top peak type. As a result, even if the void amount in the reactor core is reduced by the subsequent actual earthquakes, since the void amount is moved, effects on the increase of neutron fluxes by the actual earthquakes is small. (I.S.)

  20. Nuclear reactor internals arrangement

    International Nuclear Information System (INIS)

    A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

  1. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

  2. The research reactors their contribution to the reactors physics

    International Nuclear Information System (INIS)

    The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

  3. One piece reactor removal

    International Nuclear Information System (INIS)

    Japan Research Reactor No.3 (JRR-3) was the first reactor consisting of 'Japanese-made' components alone except for fuel and heavy water. After reaching its initial critical state in September 1962, JRR-3 had been in operation for 21 years until March 1983. It was decided that the reactor be removed en-bloc in view of the work schedule, cost and management of the reactor following the removal. In the special method developed jointly by the Japanese Atomic Energy Research Institute and Shimizu Construction Co., Ltd., the reactor main unit was cut off from the building by continuous core boring, with its major components bound in the block with biological shield material (heavy concrete), and then conveyed and stored in a large waste store building constructed near the reactor building. Major work processes described in this report include the cutting off, lifting, horizontal conveyance and lowering of the reactor main unit. The removal of the JRR-3 reactor main unit was successfully carried out safely and quickly by the en-block removal method with radiation exposure dose of the workers being kept at a minimum. Thus the high performance of the en-bloc removal method was demonstrated and, in addition, valuable knowhow and other data were obtained from the work. (Nogami, K.)

  4. Reactor Materials Research

    International Nuclear Information System (INIS)

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

  5. The fusion reactor

    International Nuclear Information System (INIS)

    Basic principles of the fusion reactor are outlined. Plasma heating and confinement schemes are described. These confinement systems include the linear Z pinch, magnetic mirrors and Tokamaks. A fusion reactor is described and a discussion is given of its environmental impact and its fuel situation. (R.L.)

  6. Polymerization Reactor Engineering.

    Science.gov (United States)

    Skaates, J. Michael

    1987-01-01

    Describes a polymerization reactor engineering course offered at Michigan Technological University which focuses on the design and operation of industrial polymerization reactors to achieve a desired degree of polymerization and molecular weight distribution. Provides a list of the course topics and assigned readings. (TW)

  7. 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)

  8. Gas-cooled reactors

    International Nuclear Information System (INIS)

    The present study is the second part of a general survey of Gas Cooled Reactors (GCRs). In this part, the course of development, overall performance and present development status of High Temperature Gas Cooled Reactors (HTCRs) and advances of HTGR systems are reviewed. (author)

  9. Light water reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  10. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of two main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents. Main achievements in 1999 are reported

  11. Light water type reactor

    International Nuclear Information System (INIS)

    The nuclear reactor of the present invention prevents disruption of a reactor core even in a case of occurrence of entire AC power loss event, and even if a reactor core disruption should occur, it prevents a rupture of the reactor container due to excess heating. That is, a high pressure water injection system and a low pressure water injection system operated by a diesel engine are disposed in the reactor building in addition to an emergency core cooling system. With such a constitution, even if an entire AC power loss event should occur, water can surely be injected to the reactor thereby enabling to prevent the rupture of the reactor core. Even if it should be ruptured, water can be sprayed to the reactor container by the low pressure water injection system. Further, if each of water injection pumps of the high pressure water injection system and the low pressure water injection system can be driven also by motors in addition to the diesel engine, the pump operation can be conducted more certainly and integrally. (I.S.)

  12. Naval propulsion reactors

    International Nuclear Information System (INIS)

    This article deals with the design and exploitation of naval propulsion reactors, mainly of PWR-type. The other existing or conceivable types of reactors are also presented: 1 - specificities of nuclear propulsion (integration in the ship, marine environment, maneuverability, instantaneous availability, conditions of exploitation-isolation, nuclear safety, safety authority); 2 - PWR-type reactor (stable operation, mastered technology, general design, radiation protection); 3 - other reactor types; 4 - compact or integrated loops architecture; 5 - radiation protection; 6 - reactor core; 7 - reactivity control (core lifetime, control means and mechanisms); 8 - core cooling (natural circulation, forced circulation, primary flow-rate program); 9 - primary loop; 10 - pressurizer; 11 - steam generators and water-steam secondary loop; 12 - auxiliary and safety loops; 13 - control instrumentation; 14 - operation; 15 - nuclear wastes and dismantling. (J.S.)

  13. Iris reactor conceptual design

    International Nuclear Information System (INIS)

    IRIS (International Reactor Innovative and Secure) is a modular, integral, light water cooled, low-to-medium power (100-350 MWe) reactor which addresses the requirements defined by the US DOE for Generation IV reactors, i.e., proliferation resistance, enhanced safety, improved economics and fuel cycle sustainability. It relies on the proven technology of light water reactors and features innovative engineering, but it does not require new technology development. This paper discusses the current reference IRIS design, which features a 1000 MWt thermal core with proven 5%-enriched uranium oxide fuel and five-year long straight burn fuel cycle, integral reactor vessel housing helical tube steam generators and immersed spool pumps. Other major contributors to the high level of safety and economic attractiveness are the safety by design and optimized maintenance approaches, which allow elimination of some classes of accidents, lower capital cost, long operating cycle, and high capacity factors. (author)

  14. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 1014 n/cm2/sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

  15. Reactor core monitoring method

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Michitsugu [Tokyo Electric Power Co., Inc. (Japan); Kanemoto, Shigeru; Enomoto, Mitsuhiro; Ebata, Shigeo

    1998-05-06

    The present invention provides a method of monitoring the state of coolant flow in a reactor of a BWR power plant. Namely, a plurality of local power region monitors (LPRM) are disposed to the inside of the reactor core for monitoring a power distribution. Signals of at least two optional LPRM detectors situated at positions different in axial or radial positions of the reactor core are obtained. General fluctuation components which nuclear hydrothermally fluctuate in overall reactor core are removed from the components of the signals. Then, correlational functions between these signals are determined. The state of coolant flow in the reactor is monitored based on the correlational function. When the axial flowing rate and radial flow interference are monitored, the accuracy upon monitoring axial and radial local behaviors of coolants can be improved by thus previously removing the general fluctuation components from signals of LPRM detectors and extracting local void information near to LPRM detectors at high accuracy. (I.S.)

  16. Physics of nuclear reactors

    International Nuclear Information System (INIS)

    This manual covers all the aspects of the science of neutron transport in nuclear reactors and can be used with great advantage by students, engineers or even reactor experts. It is composed of 18 chapters: 1) basis of nuclear physics, 2) the interactions of neutrons with matter, 3) the interactions of electromagnetic radiations and charged-particles with matter, 4) neutron slowing-down, 5) resonant absorption, 6) Doppler effect, 7) neutron thermalization, 8) Boltzmann equation, 9) calculation methods in neutron transport theory, 10) neutron scattering, 11) reactor reactivity, 12) theory of the critical homogenous pile, 13) the neutron reflector, 14) the heterogeneous reactor, 15) the equations of the fuel cycle, 16) neutron counter-reactions, 17) reactor kinetics, and 18) calculation methods in neutron scattering

  17. Mirror reactor surface study

    International Nuclear Information System (INIS)

    A general survey is presented of surface-related phenomena associated with the following mirror reactor elements: plasma first wall, ion sources, neutral beams, director converters, vacuum systems, and plasma diagnostics. A discussion of surface phenomena in possible abnormal reactor operation is included. Several studies which appear to merit immediate attention and which are essential to the development of mirror reactors are abstracted from the list of recommended areas for surface work. The appendix contains a discussion of the fundamentals of particle/surface interactions. The interactions surveyed are backscattering, thermal desorption, sputtering, diffusion, particle ranges in solids, and surface spectroscopic methods. A bibliography lists references in a number of categories pertinent to mirror reactors. Several complete published and unpublished reports on surface aspects of current mirror plasma experiments and reactor developments are also included

  18. FBR type reactor

    International Nuclear Information System (INIS)

    A circular neutron reflector is disposed vertically movably so as to surround the outer circumference of a reactor core barrel. A reflector driving device comprises a driving device main body attracted to the outer wall surface of the reactor barrel by electromagnetic attraction force and an inertia body disposed above the driving device main body vertically movably. A reflector is connected below the reactor driving device. At the initial stage, a spontaneous large current is supplied to upper electromagnetic repulsion coils of the reflector driving device, impact electromagnetic repulsion force is caused between the inertia body and the reflector driving device, so that the driving device main body moves downwardly by a predetermined distance and stopped. The reflector driving device can be lowered in a step-like manner to an appropriate position suitable to restart the reactor during stoppage of the reactor core by conducting spontaneous supply of current repeatedly to the upper electromagnetic repulsion coils. (I.N.)

  19. TRIGA research reactors

    International Nuclear Information System (INIS)

    TRIGA (Training, Research, Isotope production, General-Atomic) has become the most used research reactor in the world with 65 units operating in 24 countries. The original patent for TRIGA reactors was registered in 1958. The success of this reactor is due to its inherent level of safety that results from a prompt negative temperature coefficient. Most of the neutron moderation occurs in the nuclear fuel (UZrH) because of the presence of hydrogen atoms, so in case of an increase of fuel temperature, the neutron spectrum becomes harder and neutrons are less likely to fission uranium nuclei and as a consequence the power released decreases. This inherent level of safety has made this reactor fit for training tool in university laboratories. Some recent versions of TRIGA reactors have been designed for medicine and industrial isotope production, for neutron therapy of cancers and for providing a neutron source. (A.C.)

  20. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

    1997-08-01

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

  1. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  2. Compact torsatron reactors

    International Nuclear Information System (INIS)

    Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

  3. Multi-purpose reactor

    International Nuclear Information System (INIS)

    The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MWth, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co60) production capacity is 50000 Ci/yr, 200Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

  4. The CAREM reactor and present currents in reactor design

    International Nuclear Information System (INIS)

    INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author)

  5. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in the Division of Reactor Engineering in fiscal 1977 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committee on Reactor Physics. (Author)

  6. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1983 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

  7. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities conducted in Reactor Engineering Division in fiscal 1975 are summarized in this report. Works in the division are closely related to the development of multi-purpose High-temperature Gas Cooled Reactor, the development of Liquid Metal Fast Breeder Reactor by Power Reactor and Nuclear Fuel Development Corporation, and engineering research of thermonuclear fusion reactor. Many achievements are described concerning nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of the Committee on Reactor Physics. (auth.)

  8. Reactor performance calculations for water reactors

    International Nuclear Information System (INIS)

    The principles of nuclear, thermal and hydraulic performance calculations for water cooled reactors are discussed. The principles are illustrated by describing their implementation in the UKAEA PATRIARCH scheme of computer codes. This material was originally delivered as a course of lectures at the Technical University of Helsinki in Summer of 1969.

  9. Nanoparticles of [Fe(NH2-trz)3]Br2.3H2O (NH2-trz=2-amino-1,2,4-triazole) prepared by the reverse micelle technique: influence of particle and coherent domain sizes on spin-crossover properties.

    Science.gov (United States)

    Forestier, Thibaut; Kaiba, Abdellah; Pechev, Stanislav; Denux, Dominique; Guionneau, Philippe; Etrillard, Céline; Daro, Nathalie; Freysz, Eric; Létard, Jean-François

    2009-06-15

    This paper describes the synthesis of iron(II) spin-crossover nanoparticles prepared by the reverse micelle technique by using the non-ionic surfactant Lauropal (Ifralan D0205) from the polyoxyethylenic family. By changing the surfactant/water ratio, the size of the particles of [Fe(NH2-trz)3]Br2.3H2O (with NH2trz=4-amino-1,2,4-triazole) can be controlled. On the macroscopic scale this complex exhibits cooperative thermal spin crossovers at 305 and 320 K. We find that when the size is reduced down to 50 nm, the spin transition becomes gradual and no hysteresis can be detected. For our data it seems that the critical size, for which the existence of a thermal hysteresis can be detected, is around 50 nm. Interestingly, the change of the particle size induces almost no change in the temperature of the thermal spin transition. A systematic determination of coherent domain size carried out on the nanoparticles by powder X-ray diffraction indicates that at approximately 30 nm individual particles consist of one coherent domain. PMID:19504472

  10. Fourth Generation Reactor Concepts

    International Nuclear Information System (INIS)

    Concerns over energy resources availability, climate changes and energy supply security suggest an important role for nuclear energy in future energy supplies. So far nuclear energy evolved through three generations and is still evolving into new generation that is now being extensively studied. Nuclear Power Plants are producing 16% of the world's electricity. Today the world is moving towards hydrogen economy. Nuclear technologies can provide energy to dissociate water into oxygen and hydrogen and to production of synthetic fuel from coal gasification. The introduction of breeder reactors would turn nuclear energy from depletable energy supply into an unlimited supply. From the early beginnings of nuclear energy in the 1940s to the present, three generations of nuclear power reactors have been developed: First generation reactors: introduced during the period 1950-1970. Second generation: includes commercial power reactors built during 1970-1990 (PWR, BWR, Candu, Russian RBMK and VVER). Third generation: started being deployed in the 1990s and is composed of Advanced LWR (ALWR), Advanced BWR (ABWR) and Passive AP600 to be deployed in 2010-2030. Future advances of the nuclear technology designs can broaden opportunities for use of nuclear energy. The fourth generation reactors are expected to be deployed by 2030 in time to replace ageing reactors built in the 1970s and 1980s. The new reactors are to be designed with a view of the following objectives: economic competitiveness, enhanced safety, minimal radioactive waste production, proliferation resistance. The Generation IV International Forum (GIF) was established in January 2000 to investigate innovative nuclear energy system concepts. GIF members include Argentina, Brazil, Canada, Euratom, France Japan, South Africa, South Korea, Switzerland, United Kingdom and United States with the IAEA and OECD's NEA as permanent observers. China and Russia are expected to join the GIF initiative. The following six systems

  11. Safety of research reactors

    International Nuclear Information System (INIS)

    The number of research reactors that have been constructed worldwide for civilian applications is about 651. Of the reactors constructed, 284 are currently in operation, 258 are shut down and 109 have been decommissioned. More than half of all operating research reactors worldwide are over thirty years old. During this long period of time national priorities have changed. Facility ageing, if not properly managed, has a natural degrading effect. Many research reactors face concerns with the obsolescence of equipment, lack of experimental programmes, lack of funding for operation and maintenance and loss of expertise through ageing and retirement of the staff. Other reactors of the same vintage maintain effective ageing management programmes, conduct active research programmes, develop and retain high calibre personnel and make important contributions to society. Many countries that operate research reactors neither operate nor plan to operate power reactors. In most of these countries there is a tendency not to create a formal regulatory body. A safety committee, not always independent of the operating organization, may be responsible for regulatory oversight. Even in countries with nuclear power plants, a regulatory regime differing from the one used for the power plants may exist. Concern is therefore focused on one tail of a continuous spectrum of operational performance. The IAEA has been sending missions to review the safety of research reactors in Member States since 1972. Some of the reviews have been conducted pursuant to the IAEA' functions and responsibilities regarding research reactors that are operated within the framework of Project and Supply Agreements between Member States and the IAEA. Other reviews have been conducted upon request. All these reviews are conducted following procedures for Integrated Safety Assessment of Research Reactors (INSARR) missions. The prime objective of these missions has been to conduct a comprehensive operational safety

  12. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1992 (April 1, 1992-March 31, 1993). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  13. Reactor engineering department annual report

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1989 (April 1, 1989 - March 31, 1990). One of major Department's programs is the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Development of a high energy proton linear accelerator for the nuclear engineering including is also TRU incineration promoted. Other major tasks of the Department are various basic researches on nuclear data and group constants, theoretical methods and code development, on reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  14. Slurry reactor design studies

    Energy Technology Data Exchange (ETDEWEB)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  15. FBR type reactor

    International Nuclear Information System (INIS)

    The present invention provides an FBR type reactor in which the combustion of reactor core fuels is controlled by reflectors, and the position of a reflector driving device can be controlled even during shut down of the reactor. Namely, the reflector driving device is attracted to the outer wall surface of a reactor core barrel by electromagnetic attraction force. An inertia body is disposed vertically movably to the upper portion of the reflector driving device. Magnetic repulsive coils generate instantaneous magnetic repulsive force between the inertia body and the reflector driving device. With such a constitution, the reflector driving device can be driven by using magnetic repulsion of the electromagnetic repulsive coils and inertia of the inertia body. As a result, not only the reflectors can be elevated at an ultraslow speed during normal reactor operation, but also fine position adjustment for the reflector driving device, as well as fine position adjustment of the reflectors required upon restart of the reactor can be conducted by lowering the reflector driving device during shut down of the reactor. (I.S.)

  16. Reactor water sampling device

    International Nuclear Information System (INIS)

    The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

  17. Test reactor technology

    International Nuclear Information System (INIS)

    The Reactor Development Program created a need for engineering testing of fuels and materials. The Engineering Test Reactors were developed around the world in response to this demand. The design of the test reactors proved to be different from that of power reactors, carrying the fuel elements closer to the threshold of failure, requiring more responsive instrumentation, more rapid control element action, and inherent self-limiting behavior under accident conditions. The design of the experimental facilities to exploit these reactors evolved a new, specialized, branch of engineering, requiring a very high-lvel scientific and engineering team, established a meticulous concern with reliability, the provision for recovery from their own failures, and detailed attention to possible interactions with the test reactors. This paper presents this technology commencing with the Materials Testing Reactor (MTR) through the Fast Flux Test Facility, some of the unique experimental facilities developed to exploit them, but discusses only cursorily the experiments performed, since sample preparation and sample analyses were, and to some extent still are, either classified or proprietary. The Nuclear Engineering literature is filled with this information

  18. Advanced reactor licensing issues

    International Nuclear Information System (INIS)

    In July 1986 the US Nuclear Regulatory Commission issued a Policy Statement on the Regulation of Advanced Nuclear Power Plants. As part of this policy advanced reactor designers were encouraged to interact with NRC early in the design process to obtain feedback regarding licensing requirements for advanced reactors. Accordingly, the staff has been interacting with the Department of Energy (DOE) and its contractors on the review of three advanced reactor conceptual designs: one modular High Temperature Gas-Cooled Reactor (MHTGR) and two Liquid Metal Reactors (LMRs). As a result of these interactions certain safety issues associated with these advanced reactor designs have been identified as key to the licensability of the designs as proposed by DOE. The major issues in this regard are: (1) selection and treatment of accident scenarios; (2) selection of siting source term; (3) performance and reliability of reactor shutdown and decay heat removal systems; (4) need for conventional containment; (5) need for conventional emergency evacuation; (6) role of the operator; (7) treatment of balance of plant; and (8) modular approach. This paper provides a status of the NRC review effort, describes the above issues in more detail and provides the current status and approach to the development of licensing guidance on each

  19. Nuclear reactor power monitor

    International Nuclear Information System (INIS)

    The device of the present invention monitors phenomena occurred in a nuclear reactor more accurately than usual case. that is, the device monitors a reactor power by signals sent from a great number of neutron monitors disposed in the reactor. The device has a means for estimating a phenomenon occurred in the reactor based on the relationship of a difference of signals between each of the great number of neutron monitors to the positions of the neutron monitors disposed in the reactor. The estimation of the phenomena is conducted by, for example, conversion of signals sent from the neutron monitors to a code train. Then, a phenomenon is estimated rapidly by matching the code train described above with a code train contained in a data base. Further. signals sent from the neutron monitors are processed statistically to estimate long term and periodical phenomena. As a result, phenomena occurred in the reactor are monitored more accurately than usual case, thereby enabling to improve reactor safety and operationability. (I.S.)

  20. Reactor Sharing Program

    International Nuclear Information System (INIS)

    Support utilization of the RINSC reactor for student and faculty instructions and research. The Department of Energy award has provided financial assistance during the period 9/29/1995 to 5/31/2001 to support the utilization of the Rhode Island Nuclear Science Center (RINSC) reactor for student and faculty instruction and research by non-reactor owning educational institutions within approximately 300 miles of Narragansett, Rhode Island. Through the reactor sharing program, the RINSC (including the reactor and analytical laboratories) provided reactor services and laboratory space that were not available to the other universities and colleges in the region. As an example of services provided to the users: Counting equipment, laboratory space, pneumatic and in-pool irradiations, demonstrations of sample counting and analysis, reactor tours and lectures. Funding from the Reactor Sharing Program has provided the RINSC to expand student tours and demonstration programs that emphasized our long history of providing these types of services to the universities and colleges in the area. The funding have also helped defray the cost of the technical assistance that the staff has routinely provided to schools, individuals and researchers who have called on the RINSC for resolution of problems relating to nuclear science. The reactor has been featured in a Public Broadcasting System documentary on Pollution in the Arctic and how a University of Rhode Island Professor used Neutron Activation Analysis conducted at the RINSC to discover the sources of the ''Arctic Haze''. The RINSC was also featured by local television on Earth Day for its role in environmental monitoring

  1. Determination of research reactor safety parameters by reactor calculations

    International Nuclear Information System (INIS)

    Main research reactor safety parameters such as power density peaking factors, shutdown margin and temperature reactivity coefficients are treated. Reactor physics explanation of the parameters is given together with their application in safety evaluation performed as part of research reactor operation. Reactor calculations are presented as a method for their determination assuming use of widely available computer codes. (author)

  2. Reactor de plasma

    OpenAIRE

    Erra Serrabasa, Pilar; Molina Mansilla, Ricardo; Beltrán Serra, Eric

    2008-01-01

    Reactor de plasma. Se trata de un reactor de plasma que puede trabajar en un amplio rango de presión, desde el vacío y presiones reducidas hasta la presión atmosférica y presiones superiores. Adicionalmente el reactor de plasma tiene la capacidad de regular otros parámetros importantes y permite su uso para el tratamiento de muestras de tipología muy diversa, como por ejemplo las de tamaño relativamente grande o de superficie rugosa.

  3. Integral nuclear reactor

    International Nuclear Information System (INIS)

    The invention deals with an inprovement of the design of an integral pressurized water nuclear reactor. A typical embodyment of the invention includes a generally cylindrical pressure vessel that is assembled from three segments which are bolted together at transverse joints to form a pressure tight unit that encloses the steam generator and the reactor. The new construction permits primary to secondary coolant heat exchange and improved control rod drive mecanisms which can be exposed for full service access during reactor core refueling, maintenance and inspection

  4. Microfluidic electrochemical reactors

    Science.gov (United States)

    Nuzzo, Ralph G.; Mitrovski, Svetlana M.

    2011-03-22

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  5. Licensed operating reactors

    International Nuclear Information System (INIS)

    The Operating Units Status Report --- Licensed Operating Reactors provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management from the Headquarters staff on NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non- power reactors in the US

  6. First Algerian research reactor

    International Nuclear Information System (INIS)

    In 1985, both the Algerian Commissariat of New Energies and the Argentine National Atomic Energy Commission plus the firm INVAP S.E., started a series of mutual visits aimed at defining the mechanisms for cooperation in the nuclear field. Within this framework, a commercial contract was undersigned covering the supply of a low-power reactor (RUN), designed for basic and applied research in the fields of reactor physics and nuclear engineering. The reactor may also be used for performing experiences with neutron beams, for the irradiation of several materials and for the training of technicians, scientists and operators

  7. Course on reactor physics

    International Nuclear Information System (INIS)

    In Germany only few students graduate in nuclear technology, therefore the NPP operating companies are forced to develop their own education and training concepts. AREVA NP has started together with the Technical University of Dresden a one-week course ''reactor physics'' that includes the know-how of the nuclear power plant construction company. The Technical University of Dresden has the training reactor AKR-2 that is retrofitted by modern digital instrumentation and control technology that allows the practical training of reactor control.

  8. Fast Breeder Reactor studies

    International Nuclear Information System (INIS)

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts

  9. Nuclear reactor theory

    International Nuclear Information System (INIS)

    This textbook is composed of two parts. Part 1 'Elements of Nuclear Reactor Theory' is composed of only elements but the main resource for the lecture of nuclear reactor theory, and should be studied as common knowledge. Much space is therefore devoted to the history of nuclear energy production and to nuclear physics, and the material focuses on the principles of energy production in nuclear reactors. However, considering the heavy workload of students, these subjects are presented concisely, allowing students to read quickly through this textbook. (J.P.N.)

  10. PWR type reactor

    International Nuclear Information System (INIS)

    From a PWR with a primary circuit, consisting of a reactor pressure vessel, a steam generator and a reactor coolant pump, hot coolant is removed by means of an auxiliary system containing h.p. pumps for feeding water into the primary circuit and being connected with a pipe, originating at the upper part, which has got at least one isolating value. This is done by opening an outlet in a part of the auxiliary system that has got a lower pressure than the reactor vessel. Preferably a water jet pump is used for mixing with the water of the auxiliary system. (orig.)

  11. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the behaviour of fusion reactor materials and components during and after irradiation. Ongoing projects include: the study of the mechanical behaviour of structural materials under neutron irradiation; the investigation of the characteristics of irradiated first wall material such as beryllium; the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; and the study of dismantling and waste disposal strategy for fusion reactors. Progress and achievements in these areas in 2000 are discussed

  12. International tokamak reactor

    International Nuclear Information System (INIS)

    Since 1978, the US, the European Communities, Japan, and the Soviet Union have collaborated on the definition, conceptual design, data base assessment, and analysis of critical technical issues for a tokamak engineering test reactor, called the International Tokamak Reactor (INTOR). During 1985-1986, this activity has been expanded in scope to include evaluation of concept innovations that could significantly improve the tokamak as a commercial reactor. The purposes of this paper are to summarize the present INTOR design concept and to summarize the work on concept innovations

  13. Joyo experimental reactor tour

    International Nuclear Information System (INIS)

    JAEA cooperation in remote monitoring focuses on the Joyo Experimental Reactor at the O'arai Research and Development Center. Joyo performs irradiation of test fuels to support development of the fast reactor cycle in Japan, both in international cooperation and in support of the Monju fast reactor, which is now undergoing reconstruction. The tour included an introduction at the model, a visit to the control room, entry into the containment vessel, and viewing of remote monitoring equipment in the Fresh Fuel Storage and at one of the Spent Fuel Ponds. (author)

  14. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  15. nuclear reactor design calculations

    International Nuclear Information System (INIS)

    In this work , the sensitivity of different reactor calculation methods, and the effect of different assumptions and/or approximation are evaluated . A new concept named error map is developed to determine the relative importance of different factors affecting the accuracy of calculations. To achieve this goal a generalized, multigroup, multi dimension code UAR-DEPLETION is developed to calculate the spatial distribution of neutron flux, effective multiplication factor and the spatial composition of a reactor core for a period of time and for specified reactor operating conditions. The code also investigates the fuel management strategies and policies for the entire fuel cycle to meet the constraints of material and operating limitations

  16. Nuclear reactor internal structures

    International Nuclear Information System (INIS)

    The upper internal structures of the reactor are connected to the closing head so as to be readily removed with the latter and a skirt connected to the lower portion of said upper structures so as to surround the latter, extends under the control rods when they are removed from the reactor core. Through such an arrangement the skirt protects the control rods and supports the vessel closing-head and the core upper structures, whenever the head is severed from the vessel and put beside the latter in order to discharge the reactor

  17. Reactor monitoring system

    International Nuclear Information System (INIS)

    The present invention concerns a device for monitoring the inside of an FBR type reactor which can not be monitored by a usual optical camera. An ultrasonic camera having an excellent propagating property in a liquid metal sodium is scanned, and reflected waves of the ultrasonic waves are received as signals. The signals are processed by using a virtual realistic feeling (VR) technique such as a head mounting type image display (HMD) and a three dimensional pointing device. With such procedures, the inside of the FBR type reactor can be observed with such a realistic feeling that the inside of the FBR type reactor were seen directly. (I.S.)

  18. Research reactor support

    International Nuclear Information System (INIS)

    Research reactors (RRs) have been used in a wide range of applications including nuclear power development, basic physics research, education and training, medical isotope production, geology, industry and other fields. However, many research reactors are fuelled with High Enriched Uranium (HEU), are underutilized and aging, and have significant quantities of spent fuel. HEU inventories (fresh and spent) pose security risks Unavailability of a high-density-reprocessable fuel hinders conversion and limits back-end options and represents a survival dilemma for many RRs. Improvement of interim spent fuel storage is required at some RRs. Many RRs are under-utilized and/or inadequately funded and need to find users for their services, or permanently shut down and eventually decommission. Reluctance to decommission affect both cost and safety (loss of experienced staff ) and many shut down but not decommissioned RR with fresh and/or spent fuel at the sites invoke serious concern. The IAEA's research reactor support helps to ensure that research reactors can be operated efficiently with fuels and targets of lower proliferation and security concern and that operators have appropriate technology and options to manage RR fuel cycle issues, especially on long term interim storage of spent research reactor fuel. Availability of a high-density-reprocessable fuel would expand and improve back end options. The International Atomic Energy Agency provides assistance to Member States to convert research reactors from High Enriched Uranium fuel and targets (for medical isotope production) to qualified Low Enriched Uranium fuel and targets while maintaining reactor performance levels. The assistance includes provision of handbooks and training in the performance of core conversion studies, advice for the procurement of LEU fuel, and expert services for LEU fuel acceptance. The IAEA further provides technical and administrative support for countries considering repatriation of its

  19. Study of power reactor dynamics by stochastic reactor oscillator method

    International Nuclear Information System (INIS)

    Stochastic reactor oscillator and cross correlation method were used for determining reactor dynamics characteristics. Experimental equipment, fast reactor oscillator (BOR-1) was activated by random pulses from the GBS-16 generator. Tape recorder AMPEX-SF-300 and data acquisition tool registered reactor response to perturbations having different frequencies. Reactor response and activation signals were cross correlated by digital computer for different positions of stochastic oscillator and ionization chamber

  20. MYRRHA a fast reactor to be operated by the young generation

    International Nuclear Information System (INIS)

    MYRRHA, the Multi-purpose hybrid Research Reactor for High-tech Applications, is an innovative research facility that is able to carry out a wide variety of applications: 1) transmutation of minor actinides, 2) qualification of fuel for new reactors, 3) material testing for Gen IV and fusion reactors, 4) production of medical radio-isotopes and 5) silicon doping for renewable electrical power. By means of these applications MYRRHA is the successor of BR2, it will validate the ADS (Accelerator Driven System) full concept and the use of heavy metal coolants (particularly the lead-bismuth eutectic - LBE). The MYRRHA plant consists of 3 main parts being the primary systems, the accelerator and the balance of plant which groups all structures, systems and components that are not included in the first two items. The scientific and technological knowledge that will be acquired with MYRRHA will be the basis for new lead cooled reactors having a higher performance and for industrial transmutation facilities capable of significantly reducing the amount and the burden time of high-level waste produced in power plants. It is the responsibility of the young and next generations to operate these upcoming facilities but our duty to obtain the crucial data. To profit from nuclear applications in the coming decades, the safety of these systems obviously has to be guaranteed. The safety approach of MYRRHA is based on the defence-in-depth principle. Initiating events are grouped in probability classes in order to determine the number of lines of defence needed to mitigate their consequences. Initiating events are preferably prevented, certainly those with potentially severe consequences or difficult to mitigate. For the design of MYRRHA we consider events up to a probability of occurrence of 10-6/year. This article describes the present state of the MYRRHA design. The article is followed by the slides of the presentation

  1. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    An improved nuclear power reactor fuel element is described which consists of fuel rods, rod guide tubes and an end plate. The system allows direct access to an end of each fuel rod for inspection purposes. (U.K.)

  2. Reactor power control device

    International Nuclear Information System (INIS)

    The present invention concerns a method of controlling reactor power to shift it into a partial power operation upon occurrence of recycling pump tripping or loss of generator load. Operation state of a reactor is classified into a plurality of operation states based on values of the reactor core flow rate and the reactor power. Different insertion patterns for selected control rods are determined on every classified operation states. Then, an insertion pattern corresponding to the operation state upon occurrence of recycling pump tripping or loss of power generator load is carried out to shift into partial power operation. The operation is shifted to a load operation solely in the station while avoiding risks such as TPM scram. Then neutron fluxes are suppressed upon transient to increase margin of fuel integrity. Selected control rod pattern of the optimum reactivity is set to each of operation regions, thereby enabling to conduct flexible countermeasure so as to attain optimum operationability. (N.H.)

  3. Reactor pressure boundary materials

    International Nuclear Information System (INIS)

    With a long-term operation of nuclear power plants, the component materials are degraded under severe reactor conditions such as neutron irradiation, high temperature, high pressure and corrosive environment. It is necessary to establish the reliable and practical technologies for improving and developing the component materials and for evaluating the mechanical properties. Especially, it is very important to investigate the technologies for reactor pressure boundary materials such as reactor vessel and pipings in accordance with their critical roles. Therefore, this study was focused on developing and advancing the microstructural/micro-mechanical evaluation technologies, and on evaluating the neutron irradiation characteristics and radiation effects analysis technology of the reactor pressure boundary materials, and also on establishing a basis of nuclear material property database

  4. Reactor parameter simulation system

    International Nuclear Information System (INIS)

    A reactor parameter simulation system (RPSS) has been built with the capability of analyzing any reactor signals, decomposing those signals into their deterministic and stochastic components, then reconstructing new, simulated signals that possess the same statistical and correlation structure as the original plant variables. Important uses of the RPSS are for integration with reactor simulation software to provide tools for plant control strategy development, and for safety-study investigations of scenarios that can arise involving signal faults generated from degraded sensors. A third use of the RPSS is for frequency-domain filtering of reactor process variables contaminated with serially correlated noise, which is important for our ongoing development of expert systems for sensor-operability surveillance. 5 refs., 4 figs., 3 tabs

  5. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2002-04-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

  6. New reactor type proposed

    CERN Multimedia

    2003-01-01

    "Russian scientists at the Research Institute of Nuclear Power Engineering in Moscow are hoping to develop a new reactor that will use lead and bismuth as fuel instead of uranium and plutonium" (1/2 page).

  7. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    A fuel assembly construction for liquid metal cooled fast breeder reactors is described in which the sub-assemblies carry a smaller proportion of parasitic material than do conventional sub-assemblies. (U.K.)

  8. Ageing of research reactors

    International Nuclear Information System (INIS)

    Historically, many of the research institutions were centred on a research reactor facility as main technological asset and major source of neutrons for research. Important achievements were made in time in these research institutions for development of nuclear materials technology and nuclear safety for nuclear energy. At present, ageing of nuclear research facilities among these research reactors and ageing of staff are considerable factors of reduction of competence in research centres. The safe way of mitigation of this trend deals with ageing management by so called, for power reactors, Plant Life Management and new investments in staff as investments in research, or in future resources of competence. A programmatic approach of ageing of research reactors in correlation with their actual and future utilisation, will be used as a basis for safety evaluation and future spending. (author)

  9. Experience with Kamini reactor

    International Nuclear Information System (INIS)

    Kamini is a 233U fuelled, 30 kW(th) research reactor. It is one of the best neutron source facility with a core average flux of 1012 n/cm2/s in IGCAR used for neutron radiography of active and nonradioactive objects, activation analysis and radiation physics research. The core consists of nine plate type fuel elements with a total fuel inventory of 590 g of 233U. Two safety control plates made of cadmium are used for start up and shutdown of the reactor. Three beam tubes, two-thimble irradiation site outside reflector and one irradiation site nearer to the core constitute the testing facilities of Kamini. Kamini attained first criticality on 29th October 96 and nominal power of 30 kW in September 1997. This paper covers the design features of the reactor, irradiation facilities and their utilities and operating experience of the reactor. (author)

  10. Dossier: research reactors

    International Nuclear Information System (INIS)

    Research reactors are used at the CEA (the French atomic energy commission) since many years. Their number has been reduced but they remain unique tools that CEA valorize continuously. The results of the programs involving such reactors are of prime importance for the operation of Electricite de France (EdF) park of existing power plants but also for the design of future nuclear power plants and future research reactors. This dossier presents three examples of research reactors in use at the CEA: Osiris and Orphee (CEA-Saclay), devoted to nuclear energy and fundamental research, respectively, and the critical mockups Eole, Minerve and Masurca (CEA-Cadarache) devoted to nuclear data libraries and neutronic calculation. (J.S.)

  11. Reactor vessel sealing plug

    International Nuclear Information System (INIS)

    This invention relates to an apparatus and method for sealing the cold leg nozzles of a nuclear reactor pressure vessel from a remote location during maintenance and inspection of associated steam generators and pumps while the pressure vessel and refueling canal are filled with water. The apparatus includes a sealing plug for mechanically sealing the cold leg nozzle from the inside of a reactor pressure vessel. The sealing plugs include a primary and a secondary O-ring. An installation tool is suspended within the reactor vessel and carries the sealing plug. The tool telescopes to insert the sealing plug within the cold leg nozzle, and to subsequently remove the plug. Hydraulic means are used to activate the sealing plug, and support means serve to suspend the installation tool within the reactor vessel during installation and removal of the sealing plug

  12. Future Reactor Experiments

    CERN Document Server

    He, Miao

    2013-01-01

    The measurement of the neutrino mixing angle $\\theta_{13}$ opens a gateway for the next generation experiments to measure the neutrino mass hierarchy and the leptonic CP-violating phase. Future reactor experiments will focus on mass hierarchy determination and the precision measurement of mixing parameters. Mass hierarchy can be determined from the disappearance of reactor electron antineutrinos based on the interference effect of two separated oscillation modes. Relative and absolute measurement techniques have been explored. A proposed experiment JUNO, with a 20 kton liquid scintillator detector of $3%/$$\\sqrt{E(MeV)}$ energy resolution, $\\sim$ 53 km far from reactors of $\\sim$ 36 GW total thermal power, can reach to a sensitivity of $\\Delta\\chi^{2}>16$ considering the spread of reactor cores and uncertainties of the detector response. Three of mixing parameters are expected to be measured to better than 1% precision. There are multiple detector options for JUNO under investigation. The technical challenges...

  13. Reactor hot spot analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, R.B.

    1985-08-01

    The principle methods for performing reactor hot spot analysis are reviewed and examined for potential use in the Applied Physics Division. The semistatistical horizontal method is recommended for future work and is now available as an option in the SE2-ANL core thermal hydraulic code. The semistatistical horizontal method is applied to a small LMR to illustrate the calculation of cladding midwall and fuel centerline hot spot temperatures. The example includes a listing of uncertainties, estimates for their magnitudes, computation of hot spot subfactor values and calculation of two sigma temperatures. A review of the uncertainties that affect liquid metal fast reactors is also presented. It was found that hot spot subfactor magnitudes are strongly dependent on the reactor design and therefore reactor specific details must be carefully studied. 13 refs., 1 fig., 5 tabs.

  14. Research Reactor Benchmarks

    International Nuclear Information System (INIS)

    A criticality benchmark experiment performed at the Jozef Stefan Institute TRIGA Mark II research reactor is described. This experiment and its evaluation are given as examples of benchmark experiments at research reactors. For this reason the differences and possible problems compared to other benchmark experiments are particularly emphasized. General guidelines for performing criticality benchmarks in research reactors are given. The criticality benchmark experiment was performed in a normal operating reactor core using commercially available fresh 20% enriched fuel elements containing 12 wt% uranium in uranium-zirconium hydride fuel material. Experimental conditions to minimize experimental errors and to enhance computer modeling accuracy are described. Uncertainties in multiplication factor due to fuel composition and geometry data are analyzed by sensitivity analysis. The simplifications in the benchmark model compared to the actual geometry are evaluated. Sample benchmark calculations with the MCNP and KENO Monte Carlo codes are given

  15. Nuclear reactor (1960)

    International Nuclear Information System (INIS)

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

  16. Reactor Neutrino Spectra

    CERN Document Server

    Hayes, A C

    2016-01-01

    We present a review of the antineutrino spectra emitted from reactors. Knowledge of these and their associated uncertainties are crucial for neutrino oscillation studies. The spectra used to-date have been determined by either conversion of measured electron spectra to antineutrino spectra or by summing over all of the thousands of transitions that makeup the spectra using modern databases as input. The uncertainties in the subdominant corrections to beta-decay plague both methods, and we provide estimates of these uncertainties. Improving on current knowledge of the antineutrino spectra from reactors will require new experiments. Such experiments would also address the so-called reactor neutrino anomaly and the possible origin of the shoulder observed in the antineutrino spectra measured in recent high-statistics reactor neutrino experiments.

  17. Pulsed fusion reactors

    International Nuclear Information System (INIS)

    This summer school specialized in examining specific fusion center systems. Papers on scientific feasibility are first presented: confinement of high-beta plasma, liners, plasma focus, compression and heating and the use of high power electron beams for thermonuclear reactors. As for technological feasibility, lectures were on the theta-pinch toroidal reactors, toroidal diffuse pinch, electrical engineering problems in pulsed magnetically confined reactors, neutral gas layer for heat removal, the conceptual design of a series of laser fusion power plants with ''Saturn'', implosion experiments and the problem of the targets, the high brightness lasers for plasma generation, and topping and bottoming cycles. Some problems common to pulsed reactors were examined: energy storage and transfer, thermomechanical and erosion effects in the first wall and blanket, the problems of tritium production, radiation damage and neutron activation in blankets, and the magnetic and inertial confinement

  18. Reactor fueling of BWR type reactors

    International Nuclear Information System (INIS)

    Purpose: To enable the pattern exchange for control rods during burning in Control Cell Core type BWR reactors. Constitution: A plurality of control cells are divided into a plurality of groups such that the control cells is aparted from each other by way of at least two fuel assemblies other than the control cells with respect to the vertical and lateral directions of the reactor core cross section, as well as they are in adjacent with control cells of other groups with respect to the orthogonal direction. This enables to perform the pattern exchange for the control rods during burning in the control cell core with ease, and the control blade and the story effect harmful to the mechanical soundness of fuels can thus be suppressed. (Moriyama, K.)

  19. The risk of shortage of radioelements at medical use must not lead to overlook the reactors safety that produce them; Le risque de penurie de radioelements a usage medical ne doit pas conduire a faire l'impasse sur la surete des reacteurs qui les produisent

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-07-01

    As the reactors supplying the world production of radioelements for medical use have over 40 years of operation, the nuclear safety authority alerts the stake holders on the necessity to prevent the conflicts between public health and nuclear safety in the production of these elements; Asn estimates that the solution is not to extend the lifetime of the reactors but goes for a new international concerted approach. The most of the present production comes from five old reactors: N.R.U. at Chalk river (Canada, 40%), H.F.R. at Petten (Netherlands, 30%), Safari at Pelindaba (South Africa, 10%) B.R.2 at Mol (Belgium, 9%) and Osiris at Saclay (France, 5%). In this context, Asn organised in january 2009 a seminar on the safety-availability of facilities of radio-isotopes production with safety authorities of the concerned countries. Nea organised a seminar on the radiopharmaceuticals supply at the end of january 2009. (N.C.)

  20. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in fiscal 1974 in Reactor Engineering Division of eight laboratories and computing center are described. Works in the division are closely related with the development of a multi-purpose High-temperature Gas Cooled Reactor, the development of a Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation, and engineering of thermonuclear fusion reactors. They cover nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and aspects of the computing center. (auth.)

  1. Special lecture on nuclear reactor

    International Nuclear Information System (INIS)

    This book gives a special lecture on nuclear reactor, which is divided into two parts. The first part has explanation on nuclear design of nuclear reactor and analysis of core with theories of integral transports, diffusion Nodal, transports Nodal and Monte Carlo skill parallel computer and nuclear calculation and speciality of transmutation reactor. The second part deals with speciality of nuclear reactor and control with nonlinear stabilization of nuclear reactor, nonlinear control of nuclear reactor, neural network and control of nuclear reactor, control theory of observer and analysis method of Adomian.

  2. The replacement research reactor

    International Nuclear Information System (INIS)

    The contract for the design, construction and commissioning of the Replacement Research Reactor was signed in July 2000. This was followed by the completion of the detailed design and an application for a construction licence was made in May 2001. This paper will describe the main elements of the design and their relation to the proposed applications of the reactor. The future stages in the project leading to full operation are also described

  3. OECD Halden reactor project

    International Nuclear Information System (INIS)

    This report summarizes the activities of the OECD Halden Reactor Project for the year 1976. The main items reported on are: a) the process supervision and control which have focused on core monitoring and control, and operator-process communication; b) the fuel performance and safety behavior which have provided data and analytical descriptions of the thermal, mechanical and chemical behavior of fuel under various operating conditions; c) the reactor operations and d) the administration and finance

  4. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  5. Small reactor return

    International Nuclear Information System (INIS)

    Current state of the development of present-day small reactors in different countries is performed. Various designs of low and middle power reactors, among which are CAREM (25 MW, PWR), KLT-40 (40 MW, PWR), MRX (30 MW, PWR), IRIS (50 MW, PWR), SMART (1000 MW, PWR), Modular SBWR (50 MW, BWR), PBMR (120 MW, HTGR), GT-HMR (285 MW, HTGR), are discussed

  6. Reactor lattice transport calculations

    International Nuclear Information System (INIS)

    The present lecture is a continuation of the lecture on Introduction to the Neutron Transport Phenomena. It comprises three aspects of lattice calculations. First the idea of a reactor lattice is introduced. Then the main definitions used in reactor lattice analysis are given, and finally two basic methods applied for solution of the transport equations are defined. Several remarks on secondary results from lattice transport calculations are added. (author)

  7. Thermal or epithermal reactor

    International Nuclear Information System (INIS)

    In a thermal or epithermal heavy-water reactor of the pressure tube design the reactivity is to be increased by different means: replacement of the moderator by additional rods with heavy metal in the core or in the reflector; separation of the moderator (heavy water) from the coolant (light water) by means of shroud tubes. In light-water reactor types neutron losses are to be influenced by using the heavy elements in different configurations. (orig./PW)

  8. Future reactor experiments

    International Nuclear Information System (INIS)

    The non-zero neutrino mixing angle θ13 has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper

  9. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    The description is given of a water cooled nuclear reactor comprising a core, cooling water that rises through the core, vertical guide tubes located inside the core and control rods vertically mobile in the guide tubes. In this reactor the cooling water is divided into a first part introduced at the bottom end of the core and rising through it and a second part introduced at the top end of the guide tubes so as to drop in them

  10. Jet-Stirred Reactors

    OpenAIRE

    Herbinet, Olivier; Guillaume, Dayma

    2013-01-01

    The jet-stirred reactor is a type of ideal continuously stirred-tank reactor which is well suited for gas phase kinetic studies. It is mainly used to study the oxidation and the pyrolysis of hydrocarbon and oxygenated fuels. These studies consist in recording the evolution of the conversion of the reactants and of the mole fractions of reaction products as a function of different parameters such as reaction temperature, residence time, pressure and composition of the inlet gas. Gas chromatogr...

  11. Generation IV reactors: economics

    International Nuclear Information System (INIS)

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

  12. Future reactor experiments

    Science.gov (United States)

    Wen, Liangjian

    2015-07-01

    The non-zero neutrino mixing angle θ13 has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper.

  13. Department of Reactor Technology

    DEFF Research Database (Denmark)

    Risø National Laboratory, Roskilde

    The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included.......The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included....

  14. AVR reactor physics

    International Nuclear Information System (INIS)

    A process for reactivity control was developed and used for fuelling the AVR reactor core, which is largely based on experimentally determined values. By adding fuel elements with different quantities of heavy metals paired with various experimental requirements, great demands were made of reactivity control. Although only a small range of control was available, this was sufficient to operate the reactor and to shut it down safely in the required power and temperature range. (orig.)

  15. Moon base reactor system

    Science.gov (United States)

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

    1989-01-01

    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  16. BWR type nuclear reactor

    International Nuclear Information System (INIS)

    Purpose: To simplify the structure of an emergency core cooling system while suppressing the flow out of coolants upon rapture accidents in a coolant recycling device of BWR type reactors. Constitution: Recirculation pumps are located at a position higher than the reactor core in a pressure vessel, and the lower plenum is bisected vertically by a partition plate. Further, a gas-liquid separator is surrounded with a wall and the water level at the outer side of the wall is made higher than the water level in the inside of the wall. In this structure, coolants are introduced from the upper chamber in the lower plenum into the reactor core, and the steams generated in the reactor core are separated in the gas-liquid separator, whereby the separated liquid is introduced as coolants by way of the inner chamber into the lower chamber of the lower plenum and further sent by way of the outer chamber into the reactor core. Consequently, idle rotation of the recycling pumps due to the flow-in of saturated water is prevented and loss of coolants in the reactor core can also be prevented upon raptures in the pipeway and the driving section of the pump connected to the pressure vessel and in the bottom of the pressure vessel. (Horiuchi, T.)

  17. Emergency reactor scram system

    International Nuclear Information System (INIS)

    The present invention provides an emergency reactor scram system capable of shut down a reactor safely upon occurrence of pump trip by improving a passive scram performance for an FBR-type reactor. Namely, a driving motor and an electric generator are connected to a main pump of a primary system. An AC/DC convertor is connected to the electric generator. A shielding plug is disposed to the upper end opening of a reactor container, a control rod drive mechanism is erected on the shielding plug, and an extension pipe is attached to scram magnets of the control rod drive mechanism. The extension pipe is connected to a control rod. The rotation of the shaft of the pump is used as a direct rotator to provide an integrated-type electric generator. The electric generator is electrically connected with the power source of a scram magnet of the emergency scram system. Accordingly, the control rod of the emergency scram system is automatically and rapidly inserted to the reactor core using the power source of the electric generator upon trip of the main pump thereby enabling to scram the reactor safely. (I.S.)

  18. A modular reactor plant

    International Nuclear Information System (INIS)

    This paper describes a new concept in liquid metal reactors that is being developed by General Electric under contract to the Department of Energy. This concept is called the Modular Reactor Plant. While this effort is not expected to have a near-term impact, it is directed toward three principal issues currently affecting nuclear power in the United States. First, plant costs have escalated to the point where the startup of new plants require large electric rate increases. Second, the cost of new plants coming on-line today vary by as much as a factor of three. And, third, nuclear construction times often exceed the utilities prudent planning cycle. This paper describes how General Electric's Modular Reactor Plant addreses these issues through shop fabrication and assembly, rail shipment to the site for rapid installation of nuclear components and inherent reactor protection. In addition, it is expected the modular reactor plant will reduce the current cost of development and demonstration of liquid metal reactors to an affordable level

  19. New fission reactor designs

    International Nuclear Information System (INIS)

    A number of critical challenges to the expanded or continued use of nuclear power have developed. These can be categorized as: regulatory restrictions and complications; negative public attitudes; plant complexity; plant life, operations, and maintenance; uncertain load growth, financing; waste management. Solutions to these challenges through advanced reactor design centre around four key technical responses. Passive safety systems are being introduced which use the laws of physics to provide emergency reactor coding, control and shutdown thus eliminating the possibility of human error. Modular construction promises cuts in costs and construction time by shifting the major part of component manufacture from the site to the factory. Standardization also cuts capital costs and in addition operations and repair costs and expedites reactor licensing. Improvements to the fuel cycle include improved fuel types, designs and fabrication, and the reprocessing of and recycling spent fuel back into energy production, thus extending uranium resources and offering a partial solution to the problem of waste disposal. Examples of evolutionary and advanced water-cooled reactors, modular high temperature gas-cooled reactors, and advanced liquid metal cooled fast breeder reactors which are being developed round the world are presented. (author)

  20. OECD Halden reactor project

    International Nuclear Information System (INIS)

    This is the nineteenth annual Report on the OECD Halden Reactor Project, describing activities at the Project during 1978, the last year of the 1976-1978 Halden Agreement. Work continued in two main fields: test fuel irradiation and fuel research, and computer-based process supervision and control. Project research on water reactor fuel focusses on various aspects of fuel behavior under normal, and off-normal transient conditions. In 1978, participating organisations continued to submit test fuel for irradiation in the Halden boiling heavy-water reactor, in instrumented test assemblies designed and manufactured by the Project. Work included analysis of the impact of fuel design and reactor operating conditions on fuel cladding behavior. Fuel performance modelling included characterization of thermal and mechanical behavior at high burn-up, of fuel failure modes, and improvement of data qualification procedures to reduce and quantify error bands on in-reactor measurements. Instrument development yielded new or improved designs for measuring rod temperature, internal pressure, axial neutron flux shape determination, and for detecting cladding defects. Work on computer-based methods of reactor supervision and control included continued development of a system for predictive core surveillance, and of special mathematical methods for core power distribution control

  1. Reactor power measuring device

    International Nuclear Information System (INIS)

    The device of the present invention efficiently calibrates a fixed type gamma ray thermometer of a reactor power measuring device of a BWR type reactor. Namely, the device of the present invention calculates peripheral fuel rod power distribution by calibrating the reactor power distribution by heat generation amount, the reactor power distribution being obtained by a calculation based on a reactor model for converting the signals of a plurality of the gamma ray thermometers in the reactor core based on a conversion formula. In this case, the conversion formula is a relational formula between the power of a thermocouple of the gamma ray thermometer, gamma ray heat generation amount, thermocouple zero power sensitivity relative to a temperature coefficient. A conversion efficient calculation means makes a calibration heater to generate heat at a predetermined power, and the thermocouple zero power sensitivity and the temperature coefficient are obtained based on the output of the gamma ray thermometer in this case. The calibration means updates to conversion type thermocouple zero power sensitivity and temperature coefficient. A calibration execution means executes the operations described above successively, and when the thermocouple zero power sensitivity and the temperature coefficient are out of an allowable range, the means informs it and eliminates the corresponding gamma ray thermometer from the measuring meters. (I.S.)

  2. Reactor safety engineering

    International Nuclear Information System (INIS)

    The concept of the work is such that the basic safety philosophy for nuclear power plants as well as the safety features of both types of light water reactors, pressurized and boiling water reactors, and of the fast breeder reactor are dealt with. With the pressurized and boiling water reactors also variations, due to different supplies are mentioned. The state of development considered is characterized by the results of the American reactor safety study having very much influenced the way of presentation and the validity of the information contained. In the introduction the attentive reader is made familiar with the basic traits of safety engineering, the traditional deterministic way of proceeding being supplemented by a detailed illustration of probabilistic means used in the safety analysis. Added to this are comparative descriptions of the individual safety features, their design and mode of operation. There are, e.g., detailed discussion of the emergency core cooling systems, the power supply systems, the reactor protection system, and the containment. Special chapters are attributed to transients with and without the fast shutdown system working and to loss of coolant. The so-called external events are treated somewhat shortly whereas much space is given to core melting problems. The treatment of important events from the safety point of view, including the section on Harrisburg added for reasons of immediate interest, is limited to phenomenological description. (orig.)

  3. Regulations for RA reactor operation

    International Nuclear Information System (INIS)

    Regulations for RA reactor operation are written in accordance with the legal regulations defined by the Law about radiation protection and related legal acts, as well as technical standards according to the IAEA recommendations. The contents of this book include: fundamental data about the reactor; legal regulations for reactor operation; organizational scheme for reactor operation; general and detailed instructions for operation, behaviour in the reactor building, performing experiments; operating rules for operation under steady state and accidental conditions

  4. The reactor Cabri

    International Nuclear Information System (INIS)

    It has become necessary to construct in France a reactor which would permit the investigation of the conditions of functioning of future installations, the choice, the testing and the development of safety devices to be adopted. A water reactor of a type corresponding to the latest CEA constructions in the field of laboratory or university reactors was decided upon: it appeared important to be able to evaluate the risks entailed and to study the possibilities of increasing the power, always demanded by the users; on the other hand, it is particularly interesting to clarify the phenomena of power oscillation and the risks of burn out. The work programme for CABRI will be associated with the work carried out on the American Sperts of the same type, during its construction, very useful contacts were made with the American specialists who designed the se reactors. A brief description of the reactor is given in the communication as well as the work programme for the first years with respect to the objectives up to now envisaged. Rough description of the reactor. CABRI is an open core swimming-pool reactor without any lateral protection, housed in a reinforced building with controlled leakage, in the Centre d'Etudes Nucleaires de Cadarache. It lies alone in the middle of an area whose radius is 300 meters long. Control and measurements equipment stand out on the edge of that zone. It consumes MTR fuel elements. The control-safety rods are propelled by compressed air. The maximum flow rate of cooling circuit is 1500 m3/h. Transient measurements are recorded in a RW330 unit. Aims and work programme. CABRI is meant for: - studies on the safety of water reactors - for the definition of the safety margins under working conditions: research of maximum power at which a swimming-pool reactor may operate with respect to a cooling accident, of local boiling effect on the nuclear behaviour of the reactor, performances of the control and safety instruments under exceptional

  5. REACTOR GROUT THERMAL PROPERTIES

    Energy Technology Data Exchange (ETDEWEB)

    Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

    2011-01-28

    Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

  6. Materials for nuclear reactors

    International Nuclear Information System (INIS)

    The improved performance of present generation nuclear reactors and the realization of advanced reactor concepts, both, require development of better materials. Physical metallurgy/materials science principles which have been exploited in meeting the exacting requirements of nuclear reactor materials (fuels and structural materials), are outlined citing a few specific examples. While the incentive for improvement of traditional fuels (e.g., UO2 fuel) is primarily for increasing the average core burn up, the development of advanced fuels (e.g., MOX, mixed carbide, nitride, silicide and dispersion fuels) are directed towards better utilization of fissile and fertile inventories through adaptation of innovative fuel cycles. As the burn up of UO2 fuel reaches higher levels, a more detailed and quantitative understanding of the phenomena such as fission gas release, fuel restructuring induced by radiation and thermal gradients and pellet-clad interaction is being achieved. Development of zirconium based alloys for both cladding and pressure tube applications is discussed with reference to their physical metallurgy, fabrication techniques and in-reactor degradation mechanisms. The issue of radiation embrittlement of reactor pressure vessels (RPVs) is covered drawing a comparison between the western and eastern specifications of RPV steels. The search for new materials which can stand higher rates of atomic displacement due to radiation has led to the development of swelling resistant austenitic and ferritic stainless steels for fast reactor applications as exemplified by the development of the D-9 steel for Indian fast breeder reactor. The presentation will conclude by listing various materials related phenomena, which have a strong bearing on the successful development of future nuclear energy systems. (author)

  7. Final report on in-reactor uniaxial tensile deformation of pure iron and Fe-Cr alloy

    International Nuclear Information System (INIS)

    Traditionally, the effect of irradiation on mechanical properties of metals and alloys is determined using post-irradiation tests carried out on pre-irradiated specimens and in the absence of irradiation environment. The results of these tests may not be representative of deformation behaviour of materials used in the structural components of a fission or fusion reactor where the materials will be exposed concurrently to displacement damage and external and/or internal stresses. In an effort to evaluate and understand the dynamic response of materials under these conditions, we have recently performed a series of uniaxial tensile tests on Fe-Cr and pure iron specimens in the BR-2 reactor at Mol (Belgium). The present report first provides a brief description of the test facilities and the procedure used for performing the in-reactor tests. The results on the mechanical response of materials during these tests are presented in the form of stress-displacement dose and the conventional stress-strain curves. For comparison, the results of post-irradiation tests and tests carried out on unirradiated specimens are also presented. Results of microstructural investigations on the unirradiated and deformed, irradiated and undeformed, post-irradiation deformed and the in-reactor deformed specimens are also described. During the in-reactor tests the specimens of both Fe-Cr alloy and pure iron deform in a homogeneous manner and do not exhibit the phenomenon of yield drop. An increase in the pre-yield dose increases the yield stress but not the level of maximum flow stress during the in-reactor deformation of Fe-Cr alloy. Neither the in-reactor nor the post-irradiation deformed specimens of Fe-Cr alloy and pure iron showed any evidence of cleared channel formation. Both in Fe-Cr and pure iron, the in-reactor deformation leads to accumulation of dislocations in a homogeneous fashion and only to a modest density. No dislocation cells are formed during the in-reactor or post

  8. Reactor physics and economic aspects of the CANDU reactor system

    International Nuclear Information System (INIS)

    A history of the development of the CANDU system is given along with a fairly detailed description of the 600 MW(e) CANDU reactor. Reactor physics calculation methods are described, as well as comparisons between calculated reactor physics parameters and those measured in research and power reactors. An examination of the economics of CANDU in the Ontario Hydro system and a comparison between fossil fuelled and light water reactors is presented. Some physics, economics and resources aspects are given for both low enriched uranium and thorium-fuelled CANDU reactors. Finally the RβD program in Advanced Fuel Cycles is briefly described

  9. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    Energy Technology Data Exchange (ETDEWEB)

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  10. Fast breeder reactor research

    International Nuclear Information System (INIS)

    Full text: The meeting was attended by 15 participants from seven countries and two international organizations. The Eighth Annual Meeting of the International Working Group on Fast Reactors (IWGFR) was attended by representatives from France, Fed. Rep. Germany, Italy, Japan, United Kingdom, Union of Soviet Socialist Republics and the United States of America - countries that have made significant progress in developing the technology and physics of sodium cooled fast reactors and have extensive national programmes in this field - as well as by representatives of the Commission of the European Communities and the IAEA. The design of fast-reactor power plants is a more difficult task than developing facilities with thermal reactors. Different reactor kinetics and dynamics, a hard neutron spectrum, larger integral doses of fuel and structural material irradiation, higher core temperatures, the use of an essentially novel coolant, and, as a result of all these factors, the additional reliability and safety requirements that are imposed on the planning and operation of sodium cooled fast reactors - all these factors pose problems that can be solved comprehensively only by countries with a high level of scientific and technical development. The exchange of experience between these countries and their combined efforts in solving the fundamental problems that arise in planning, constructing and operating fast reactors are promoting technical progress and reducing the relative expenditure required for various studies on developing and introducing commercial fast reactors. For this reason, the meeting concentrated on reviewing and discussing national fast reactor programmes. The situation with regard to planning, constructing and operating fast experimental and demonstration reactors in the countries concerned, the experience accumulated in operating them, the difficulties arising during operation and ways of over-coming them, the search for optimal designs for the power

  11. Irradiation effects on the ductility of fusion reactor structural materials

    International Nuclear Information System (INIS)

    Austenitic and ferritic-martensitic stainless steels have been proposed as first wall structural materials for the next generation of fusion devices. In order to study the effect of high temperature irradiation on their tensile properties, specimens of the steel AISI 316 L (CEC reference), of the martensitic steel W. Nr 1.4914 and of the duplex ferritic-martensitic steel EM12 have been irradiated in the BR2 reactor in Mol. The austenitic steel was irradiated at 4700C to about 1.1 1022n/cm2 ( E>0.1 MeV) while the ferritic-martensitic steels were irradiated at 5900C to about 7.7 1022n/cm2 (E>0.1 MeV). The tensile tests of the 316 L steel have been performed between 250 and 7500C. Below around 5500C, the yield stress after irradiation increased from about 160 to 270 MPa and the total elongation decreased from 42 to about 26%. At 7500C, the yield stress increase was small but the total elongation decreased from 60 to only 10%. At this temperature, the rupture of the irradiated specimen was intergranular while all the other specimens presented a transgranular rupture. At 6500C the variations were intermediate. The change of the ultimate tensile strength was small at all test temperatures. The EM12 and W. Nr 1.4914 steels tested only at 5500C, showed a decrease of the yield and tensile strength as well as an increase of the total elongation. The same tests performed on specimens which have been heat treated in parallel showed that the observed changes were due, in a large part, if not completely, to the maintenance of steels at high temperature

  12. Reactor coolant cleanup facility

    International Nuclear Information System (INIS)

    A depressurization device is disposed in pipelines upstream of recycling pumps of a reactor coolant cleanup facility to reduce a pressure between the pressurization device and the recycling pump at the downstream, thereby enabling high pressure coolant injection from other systems by way of the recycling pumps. Upon emergency, the recycling pumps of the coolant cleanup facility can be used in common to an emergency reactor core cooling facility and a reactor shutdown facility. Since existent pumps of the emergency reactor core cooling facility and the reactor shutdown facility which are usually in a stand-by state can be removed, operation confirmation test and maintenance for equipments in both of facilities can be saved, so that maintenance and reliability of the plant are improved and burdens on operators can also be mitigated. Moreover, low pressure design can be adopted for a non-regenerative heat exchanger and recycling coolant pumps, which enables to improve the reliability and economical property due to reduction of possibility of leakage. (N.H.)

  13. HTGR type reactor

    International Nuclear Information System (INIS)

    A reactor core is disposed at the center of a reactor container, a reflector is disposed on the outer side thereof, a steam generator is disposed further outer side thereof coaxially, and they are constituted as an integrated one container. A gas circulator and control rod drives are protruded at the outer side of the lower portion of the integrated container. Heat insulators are disposed on the inner side of the container wall in the upper portion of the reactor container. Helium gas risen in the reactor core and heated to a high temperature descends in a circular steam generator and undergoes heat exchange with water, and is then pressurized in the gas circulator after the lowering of the temperature, and returned to the inlet of the reactor core from the lower central portion of the container. With such procedures, the helium gas as primary coolants circulates only in the container to improve confinement. The device can be reduced in the size and the cost. (I.N.)

  14. Reactor container spray device

    International Nuclear Information System (INIS)

    Purpose: To enable decrease in the heat and the concentration of radioactive iodine released from the reactor vessel into the reactor container in the spray device of BWR type reactors. Constitution: A plurality of water receiving trays are disposed below the spray nozzle in the dry well and communicated to a pressure suppression chamber by way of drain pipeways passing through a diaphragm floor. When the recycling system is ruptured and coolants in the reactor vessel and radioactive iodine in the reactor core are released into the dry well, spray water is discharged from the spray nozzle to eliminate the heat and the radioactive iodine in the dry well. In this case, the receiving trays collect the portions of spray water whose absorption power for the heat and radioactive iodine is nearly saturated and falls them into the pool water of the pressure suppression chamber. Consequently, other portions of the spray water that still possess absorption power can be jetted with no hindrance, to increase the efficiency for the removal of the heat and iodine of the spray droplets. (Horiuchi, T.)

  15. PROTEUS research reactor

    International Nuclear Information System (INIS)

    The PROTEUS zero power reactor at the Paul Scherrer Institute (PSI) in Switzerland achieved first criticality in 1968 and since then has been operated as an experimental tool for reactor physics research on test lattices representative of a wide range of reactor concepts. Reactor design codes and their associated data libraries are validated on the basis of the experimental results obtained. PROTEUS is normally configured as a driven system, in which a subcritical test zone is made critical by the surrounding driver zones. The advantages of driven systems can be summarized as follows: - Smaller amount of test fuel is required; - Large range of test zone conditions (including k∞ < 1 states) can be investigated by changes in the driver loading alone, thus avoiding undesirable perturbations to the test zone which would influence the measurement conditions and thus affect the interpretability of the results; - Necessary reactor control and instrumentation equipment (usually perturbing from the experimental viewpoint) can be located in the outer driver regions, thereby avoiding disturbance of the test lattice

  16. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

    From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

  17. China experimental fast reactor

    International Nuclear Information System (INIS)

    The Chinese experimental fast reactor (CEFR) is a pool-type sodium-cooled fast reactor whose short term purposes are: -) the validation of computer codes, -) the check of the relevance of standards, and -) the gathering of experimental data on fast reactors. On the long term the expectations will focus on: -) gaining experience in fast reactor operations, -) the testing of nuclear fuels and materials, and -) the study of sodium compounds. The main technical features of CEFR are: -) thermal power output: 65 MW (electrical power output: 20 MW), -) size of the core: height: 45 cm, diameter: 60 cm, -) maximal linear output: 430 W/cm, -) neutron flux: 3.7*1015 n/cm2/s, -) input/output sodium temperature: 360 / 530 Celsius degrees, -) 2 loops for the primary system and 2 loops for the secondary system. The temperature coefficient and the power coefficient are settled to stay negative for any change in the values of the core parameters. The installation of the reactor vessel will be completed by mid 2007. The first criticality of CEFR is expected during the first semester of 2010. (A.C.)

  18. EBT reactor analysis

    International Nuclear Information System (INIS)

    This report summarizes the results of a recent ELMO Bumpy Torus (EBT) reactor study that includes ring and core plasma properties with consistent treatment of coupled ring-core stability criteria and power balance requirements. The principal finding is that constraints imposed by these coupling and other physics and technology considerations permit a broad operating window for reactor design optimization. Within this operating window, physics and engineering systems analysis and cost sensitivity studies indicate that reactors with approx. 6 to 10%, P approx. 1200 to 1700 MW(e), wall loading approx. 1.0 to 2.5 MW/m2, and recirculating power fraction (including ring-sustaining power and all other reactors auxiliaries) approx. 10 to 15% are possible. A number of concept improvements are also proposed that are found to offer the potential for further improvement of the reactor size and parameters. These include, but are not limited to, the use of: (1) supplementary coils or noncircular mirror coils to improve magnetic geometry and reduce size, (2) energetic ion rings to improve ring power requirements, (3) positive potential to enhance confinement and reduce size, and (4) profile control to improve stability and overall fusion power density

  19. Modern research reactors in the world and RA research reactor

    International Nuclear Information System (INIS)

    This paper covers the following topics: fundamentals of research reactors, thermal neutron flux density, classification of research reactors in the world, properties of research reactors of higher power in the world according to IAEA data for 1995, their application, and trend of development, experimental feasibility and status of RA reactor. Trend of research reactors development in the world (after 1980) is directed towards increasing the neutron production quality factor, i.e. ratio between thermal neutron flux density and reactor power, which is achieved by designing compact reactor cores. With the aim of renewal of RA reactor (without analysis of reactor components and staff aging, possibility of restart and commercialization), according to the analysis in this paper, it can be concluded: there is very few reactors under construction in the world, all the important countries in Europe have research reactors; RA reactor is not very interesting for development of reactor physics; nowadays RA reactor is in the group of reactors which are 30-40 years old; its inventories of fuel and heavy water are enough for about 20 years of operation; it has achieved high quality factor of neutron production with low and highly enriched fuel; core transfer from low highly enriched to low enriched fuel should be carefully studies from operation, experimental and economical point of view; it is necessary to use the advantages of RA reactor (minimum investment): volume of the core and reflector which enables availability of neutron flux for the users (numerous experimental loops), fuel in shape of slugs enabling efficient fuel management and flexible neutron flux distribution in the core in the reflector, reactor operation should be directed towards commercial applications. Bibliography of more than 140 relevant papers used is included in this paper

  20. Sodium-cooled nuclear reactors

    International Nuclear Information System (INIS)

    This book first explains the choice of sodium-cooled reactors by outlining the reasons of the choice of fast neutron reactors (fast neutrons instead of thermal neutrons, recycling opportunity for plutonium, full use of natural uranium, nuclear waste optimization, flexibility of fast neutron reactors in nuclear material management, fast neutron reactors as complements of water-cooled reactors), and by outlining the reasons for the choice of sodium as heat-transfer material. Physical, chemical, and neutron properties of sodium are presented. The second part of the book first presents the main design principles for sodium-cooled fast neutron reactors and their core. The third part proposes an historical overview and an assessment of previously operated sodium-cooled fast neutron reactors (French reactors from Rapsodie to Superphenix, other reactors in the world), and an assessment of the main incidents which occurred in these reactors. It also reports the experience and lessons learned from the dismantling of various sodium-cooled fast breeder reactors in the world. The next chapter addresses safety issues (technical and safety aspects related to the use of sodium) and environmental issues (dosimetry, gaseous and liquid releases, solid wastes, and cooling water). Then, various technological aspects of these reactors are addressed: the energy conversion system, main components, sodium chemistry, sodium-related technology, advances in in-service inspection, materials used in reactors and their behaviour, and fuel system. The next chapter addresses the fuel cycle in these reactors: its integrated specific character, report of the French experience in fast neutron reactor fuel processing, description of the transmutation of minor actinides in these reactors. The last chapter proposes an overview of reactors currently projected or under construction in the world, presents the Astrid project, and gives an assessment of the economy of these reactors. A glossary and an index

  1. Scaleable, High Efficiency Microchannel Sabatier Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — A Microchannel Sabatier Reactor System (MSRS) consisting of cross connected arrays of isothermal or graded temperature reactors is proposed. The reactor array...

  2. Methanation assembly using multiple reactors

    Science.gov (United States)

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24

    A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

  3. Mimic of OSU research reactor

    International Nuclear Information System (INIS)

    The Ohio State University research reactor (OSURR) is undergoing improvements in its research and educational capabilities. A computer-based digital data acquisition system, including a reactor system mimic, will be installed as part of these improvements. The system will monitor the reactor system parameters available to the reactor operator either in digital parameters available to the reactor operator either in digital or analog form. The system includes two computers. All the signals are sent to computer 1, which processes the data and sends the data through a serial port to computer 2 with a video graphics array VGA monitor, which is utilized to display the mimic system of the reactor

  4. MINT research reactor safety program

    Energy Technology Data Exchange (ETDEWEB)

    Mohamad Idris bin Taib [Division of Special Project, Malaysian Institute for Nuclear Technology Research (MINT), Bangi (Malaysia)

    2000-11-01

    Malaysian Institute for Nuclear Technology Research (MINT) Research Reactor Safety Program has been done along with Reactor Power Upgrading Project, Reactor Safety Upgrading Project and Development of Expert System for On-Line Nuclear Process Control Project. From 1993 up to date, Neutronic and Thermal-hydraulics analysis, Probabilistic Safety Assessment as well as installation of New 2 MW Secondary Cooling System were done. Installations of New Reactor Building Ventilation System, Reactor Monitoring System, Updating of Safety Analysis Report and Upgrading Primary Cooling System are in progress. For future activities, Reactor Modeling will be included to add present activities. (author)

  5. RB research reactor Safety Report

    International Nuclear Information System (INIS)

    This RB reactor safety report is a revised and improved version of the Safety report written in 1962. It contains descriptions of: reactor building, reactor hall, control room, laboratories, reactor components, reactor control system, heavy water loop, neutron source, safety system, dosimetry system, alarm system, neutron converter, experimental channels. Safety aspects of the reactor operation include analyses of accident causes, errors during operation, measures for preventing uncontrolled activity changes, analysis of the maximum possible accident in case of different core configurations with natural uranium, slightly and highly enriched fuel; influence of possible seismic events

  6. Fusion reactor materials

    International Nuclear Information System (INIS)

    This is the fifteenth in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following progress reports: Alloy Development for Irradiation Performance; Damage Analysis and Fundamental Studies; Special purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials programs being conducted in support of the Magnetic Fusion Energy Program of the U.S. Department of Energy. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide

  7. Chernobyl reactor accident

    International Nuclear Information System (INIS)

    On April 26, 1986, an explosion occurred at the newest of four operating nuclear reactors at the Chernobyl site in the USSR. The accident initiated an international technical exchange of almost unprecedented magnitude; this exchange was climaxed with a meeting at the International Atomic Energy Agency in Vienna during the week of August 25, 1986. The meeting was attended by more than 540 official representatives from 51 countries and 20 international organizations. Information gleaned from that technical exchange is presented in this report. A description of the Chernobyl reactor, which differs significantly from commercial US reactors, is presented, the accident scenario advanced by the Russian delegation is discussed, and observations that have been made concerning fission product release are described

  8. International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    An international design team comprised of members from Canada, Europe, Japan, the Soviet Union, and the United States of America, are designing an experimental fusion test reactor. The engineering and testing objectives of this International Thermonuclear Experimental Reactor (ITER) are to validate the design and to demonstrate controlled ignition, extended burn of a deuterium and tritium plasma, and achieve steady state using technology expected to be available by 1990. The concept maximizes flexibility while allowing for a variety of plasma configurations and operating scenarios. During physics phase operation, the machine produces a 22 MA plasma current. In the technology phase, the machine can be reconfigured with a thicker shield and a breeding blanket to operate with an 18 MA plasma current at a major radius of 5.5 meters. Canada's involvement in the areas of safety, facility design, reactor configuration and maintenance builds on our internationally recognized design and operational expertise in developing tritium processes and CANDU related technologies

  9. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  10. Licensed operating reactors

    International Nuclear Information System (INIS)

    THE OPERATING UNITS STATUS REPORT - LICENSED OPERATING REACTORS provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US

  11. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  12. Colliding Beam Fusion Reactors

    Science.gov (United States)

    Rostoker, Norman; Qerushi, Artan; Binderbauer, Michl

    2003-06-01

    The recirculating power for virtually all types of fusion reactors has previously been calculated [1] with the Fokker-Planck equation. The reactors involve non-Maxwellian plasmas. The calculations are generic in that they do not relate to specific confinement devices. In all cases except for a Tokamak with D-T fuel the recirculating power was found to exceed the fusion power by a large factor. In this paper we criticize the generality claimed for this calculation. The ratio of circulating power to fusion power is calculated for the Colliding Beam Reactor with fuels D-T, D-He3 and p-B11. The results are respectively, 0.070, 0.141 and 0.493.

  13. The MNSR reactor

    International Nuclear Information System (INIS)

    This tank-in-pool reactor is based on the same design concept as the Canadian Slowpoke. The core is a right circular cylinder, 24 cm diameter by 25 cm long, containing 411 fuel pin positions. The pins are HEU-Aluminium alloy, 0.5 cm in diameter. Critical mass is about 900 g. The reactor has a single cadmium control rod. The back-up shutdown system is the insertion of a cadmium capsule in a core position. Excess reactivity is limited to 3.5mk. In both the MNSR and Slowpoke, the insertion of the maximum excess reactivity results in a power transient limited by the coolant/moderator temperature to safe values, independent of any operator action. This reactor is used primarily in training and neutron activation analysis. Up to 64 elements have been analyzed in a great variety of different disciplines. (author)

  14. Welding and reactor safety

    International Nuclear Information System (INIS)

    The high safety requirements which must be demanded of the quality of the welded joints in reactor technique have so far not been fulfilled in all cases. The errors occuring have caused considerable loss of availability and high material costs. They were not, however, so serious that one need have feared any immediate danger to the personnel or to the environment. The safety devices of reactor plants were only called upon in a few cases and to these they responded perfectly. The intensive efforts to complete and improve the specifications are to contribute to that in future, the reactor plants can be counted even more so as one of the safest technical plants ever. (orig./LH)

  15. Reactor operation experience

    International Nuclear Information System (INIS)

    Since the TRIGA Users Conference in Helsinki 1970 the TRIGA reactor Vienna was in operation without any larger undesired shutdown. The integrated thermal power production by August 15 1972 accumulated to 110 MWd. The TRIGA reactor is manly used for training of students, for scientific courses and research work. Cooperation with industry increased in the last two years either in form of research or in performing training courses. Close cooperation is also maintained with the IAEA, samples are irradiated and courses on various fields are arranged. Maintenance work was performed on the heat exchanger and to replace the shim rod magnet. With the view on the future power upgrading nine fuel elements type 110 have been ordered recently. Experiments, performed currently on the reactor are presented in details

  16. Nuclear Rocket Engine Reactor

    CERN Document Server

    Lanin, Anatoly

    2013-01-01

    The development of a nuclear rocket engine reactor (NRER ) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  17. Reactor accidents in perspective

    International Nuclear Information System (INIS)

    In each of the three major reactor accidents which have led to significant releases to the environment, and discussed in outline in this note, the reactor has been essentially destroyed - certainly Windscale and Chernobyl reactors will never operate and the cleanup operation for Three Mile Island is currently estimated to have cost in excess of US Pound 500 000 000. In each of the accidents there has not been any fatality off site in the short term and any long-term health detriment is unlikely to be seen in comparison with the natural cancer incidence rate. At Chernobyl, early fatalities did occur amongst those concerned with fighting the incident on site and late effects are to be expected. The assumption of a linear non-threshold risk, and hence no level of zero risk is the main problem in communication with the public, and the author calls for simplification of the presentation of the concepts of radiological protection. (U.K.)

  18. Reactor safety equipments

    International Nuclear Information System (INIS)

    Purpose: To positively recover radioactive substances discharged in a dry well at the time of failure of a reactor. Constitution: In addition to the emergency gas treating system fitted to a reactor building, a purification system connected through a pipeline to the dry well is arranged in the reactor building. This purification system is connected through pipes fitted to the dry well to forced circulation device, heat exchanger, and purification device. The atmosphere of high pressure steam gases in the dry well is derived to the heat exchanger for cooling, and then radioactive substances which are contained in the gases are removed by filter sets charged with the HEPA filters and the HECA filters. At last, there gases are returned to dry well by circulation pump, repeat this process. (Kamimura, M.)

  19. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  20. Reactor protection system

    International Nuclear Information System (INIS)

    The report describes the reactor protection system (RPS-II) designed for use on Babcock and Wilcox 145-, later 177-, and 205-fuel assembly pressurized water reactors. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low-pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, a description of the software programmed in the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W

  1. Backfitting swimming pool reactors

    International Nuclear Information System (INIS)

    Calculations based on measurements in a critical assembly, and experiments to disclose fuel element surface temperatures in case of accidents like stopping of primary coolant flow during full power operation, have shown that the power of the swimming pool type research reactor FRG-2 (15 MW, operating since 1967) might be raised to 21 MW within the present rules of science and technology, without major alterations of the pool buildings and the cooling systems. A backfitting program is carried through to adjust the reactor control systems of FRG-2 and FRG-1 (5 MW, housed in the same reactor hall) to the present safety rules and recommendations, to ensure FRG-2 operation at 21 MW for the next decade. (author)

  2. MERCHANT MARINE SHIP REACTOR

    Science.gov (United States)

    Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

    1961-05-01

    A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

  3. Safety systems of heavy water reactors and small power reactors

    International Nuclear Information System (INIS)

    After introductional descriptions of heavy water reactors and natural circulation boiling water reactors the safety philosophy and safety systems like ECCS, residual heat removal, protection systems etc., are described. (RW)

  4. Study of future reactors

    International Nuclear Information System (INIS)

    Today, more than 420 large reactors with a gross output of close to 350 GWe supply 20 percent of world electricity needs, accounting for less than 5 percent of primary energy consumption. These figures are not expected to change in the near future, due to suspended reactor construction in many countries. Nevertheless, world energy needs continue to grow: the planet's population already exceeds five billion and is forecast to reach ten billion by the middle of the next century. Most less developed countries have a very low rate of energy consumption and, even though some savings can be made in industrialized countries, it will become increasingly difficult to satisfy needs using fossil fuels only. Furthermore, there has been no recent breakthrough in the energy landscape. The physical feasibility of the other great hope of nuclear energy, fusion, has yet to be proved; once this has been done, it will be necessary to solve technological problems and to assess economic viability. Although it is more ever necessary to pursue fusion programs, there is little likelihood of industrial applications being achieved in the coming decades. Coal and fission are the only ways to produce massive amounts of energy for the next century. Coal must overcome the pollution problems inherent in its use; fission nuclear power has to gain better public acceptance, which is obviously colored by safety and waste concerns. Most existing reactors were commissioned in the 1970s; reactor lifetime is a parameter that has not been clearly established. It will certainly be possible to refurbish some to extend their operation beyond the initial target of 30 or 40 years. But normal advances in technology and safety requirements will make the operation of the oldest reactors increasingly difficult. It becomes necessary to develop new generations of nuclear reactors, both to replace older ones and to revive plant construction in their countries that are not yet equipped or that have halted their

  5. AREVA's nuclear reactors portfolio

    International Nuclear Information System (INIS)

    A reasonable assumption for the estimated new build market for the next 25 years is over 340 GWe net. The number of prospect countries is growing almost each day. To address this new build market, AREVA is developing a comprehensive portfolio of reactors intended to meet a wide range of power requirements and of technology choices. The EPR reactor is the flagship of the fleet. Intended for large power requirements, the four first EPRs are being built in Finland, France and China. Other countries and customers are in view, citing just two examples: the Usa where the U.S. EPR has been selected as the technology of choice by several U.S utilities; and the United Kingdom where the Generic Design Acceptance process of the EPR design submitted by AREVA and EDF is well under way, and where there is a strong will to have a plant on line in 2017. For medium power ranges, the AREVA portfolio includes a boiling water reactor and a pressurized water reactor which both offer all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation cost: -) KERENA (1250+ MWe), developed in collaboration with several European utilities, and in particular with Eon; -) ATMEA 1 (1100+ MWe), a 3-loop evolutionary PWR which is being developed by AREVA and Mitsubishi. AREVA is also preparing the future and is deeply involved into Gen IV concepts. It has developed the ANTARES modular HTR reactor (pre-conceptual design completed) and is building upon its vast Sodium Fast Reactor experience to take part into the development of the next prototype. (author)

  6. Operating US power reactors

    International Nuclear Information System (INIS)

    This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the US which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of Dec. 31, 1986, the end of the three-month period covered in this report. Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three months (October, November, and December 1986) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. They are defined as follows: In addition to the tabular data, this article discusses significant occurrences and developments that affected licensed US power reactors during this reporting period. It includes, but is not limited to, changes in operating status, regulatory actions and decisions, and legal actions involving the status of power reactors. We do not have space here for routine problems of operation and maintenance, but such information is available at the Nuclear Regulatory Commission (NRC) Public Document Room, 1717 H Street, NW, Washington, DC 20555. Some significant operating events are summarized elsewhere in this section in the article ''Selected Safety-Related Events,'' and a report on activities relating to facilities still in the construction process is given in the article ''Status of Power-Reactor Projects Undergoing Licensing Review'' in the last section of each issue of this journal. The reader's attention is also called to the regular feature ''General Administrative Activities,'' which deals with more general aspects of regulatory and legal matters that are not covered elsewhere in the journal

  7. Oscillatory flow chemical reactors

    Directory of Open Access Journals (Sweden)

    Slavnić Danijela S.

    2014-01-01

    Full Text Available Global market competition, increase in energy and other production costs, demands for high quality products and reduction of waste are forcing pharmaceutical, fine chemicals and biochemical industries, to search for radical solutions. One of the most effective ways to improve the overall production (cost reduction and better control of reactions is a transition from batch to continuous processes. However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat and mass transfer and narrow residence time distribution. The oscillatory flow reactors (OFR are newer type of tube reactors which can offer solution by providing continuous operation with approximately plug flow pattern, low shear stress rates and enhanced mass and heat transfer. These benefits are the result of very good mixing in OFR achieved by vortex generation. OFR consists of cylindrical tube containing equally spaced orifice baffles. Fluid oscillations are superimposed on a net (laminar flow. Eddies are generated when oscillating fluid collides with baffles and passes through orifices. Generation and propagation of vortices create uniform mixing in each reactor cavity (between baffles, providing an overall flow pattern which is close to plug flow. Oscillations can be created by direct action of a piston or a diaphragm on fluid (or alternatively on baffles. This article provides an overview of oscillatory flow reactor technology, its operating principles and basic design and scale - up characteristics. Further, the article reviews the key research findings in heat and mass transfer, shear stress, residence time distribution in OFR, presenting their advantages over the conventional reactors. Finally, relevant process intensification examples from pharmaceutical, polymer and biofuels industries are presented.

  8. Nuclear reactor simulator

    International Nuclear Information System (INIS)

    The Nuclear Reactor Simulator was projected to help the basic training in the formation of the Nuclear Power Plants operators. It gives the trainee the opportunity to see the nuclear reactor dynamics. It's specially indicated to be used as the support tool to NPPT (Nuclear Power Preparatory Training) from NUS Corporation. The software was developed to Intel platform (80 x 86, Pentium and compatible ones) working under the Windows operational system from Microsoft. The program language used in development was Object Pascal and the compiler used was Delphi from Borland. During the development, computer algorithms were used, based in numeric methods, to the resolution of the differential equations involved in the process. (author)

  9. Experimental reactor physics

    International Nuclear Information System (INIS)

    Neutronic experiments in moderators, subcritical assemblies, critical assemblies, and nuclear reactors are described, as well as the techniques of radiation measurements necessary to perform these experiments. Previously dispersed data from government reports, journal articles, and specialized monographs are codified. Original information drawn from the author's experience is included, especially on the pulsed source technique, spectrum measurements, research reactors, and exponential assemblies. The book provides the essential information for carrying out, analyzing, and understanding the experiments. Theory is kept to a minimum. Emphasis is placed on the physics of the situation, and the importance of estimating error as well as the mean value of a measured quantity

  10. Diagnostics for hybrid reactors

    International Nuclear Information System (INIS)

    The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

  11. Perspectives on reactor safety

    International Nuclear Information System (INIS)

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

  12. Clinch River Breeder Reactor

    International Nuclear Information System (INIS)

    Mr. Baron says the administration's effort to terminate the Clinch River Breeder Reactor (CRBR) project is symptomatic; they have also placed restrictions on fusion, coal, solar, and other areas of energy development in which technological advances are held back in order to force conservation. Because the breeder reactor, unlike solar and fusion energy, is both economically and technically feasible, a demonstration plant is needed. The contentions that the CRBR design is obsolete, that its proposed size is inappropriate, or that plutonium can be diverted for weapons proliferation are argued to be invalid. Failure to complete the CRBR will have both economic and national security repercussions

  13. Netherlands Interuniversity Reactor Institut

    International Nuclear Information System (INIS)

    This is the annual report of the Interuniversity Reactor Institute in the Netherlands for the Academic Year 1977-78. Activities of the general committee, the daily committee and the scientific advice board are presented. Detailed reports of the scientific studies performed are given under five subjects - radiation physics, reactor physics, radiation chemistry, radiochemistry and radiation hygiene and dosimetry. Summarised reports of the various industrial groups are also presented. Training and education, publications and reports, courses, visits and cooperation with other institutes in the area of scientific research are mentioned. (C.F.)

  14. Reactor Materials Research

    International Nuclear Information System (INIS)

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel

  15. Perspectives on reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

    1994-03-01

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

  16. Reactor neutron dosimetry

    International Nuclear Information System (INIS)

    An analysis of requirements and possibilities for experimental neutron spectrum determination during the reactor pressure vessel surveil lance programme is given. Fast neutron spectrum and neutron dose rate were measured in the Fast neutron irradiation facility of our TRIGA reactor. It was shown that the facility can be used for calibration of neutron dosimeters and for irradiation of samples sensitive to neutron radiation. The investigation of the unfolding algorithm ITER was continued. Based on this investigations are two specialized unfolding program packages ITERAD and ITERGS written this year. They are able to unfold data from activation detectors and NaI(T1) gamma spectrometer respectively

  17. Decay of reactor neutrinos

    OpenAIRE

    Vogel, P.

    1984-01-01

    We consider the decay of massive neutrinos which couple to electrons and are, therefore, produced in nuclear reactors. Lifetime limits for the γ and electron-positron decay modes of these neutrinos are deduced from the experimental limit on the singles count rate in the detector used to study neutrino oscillations at the Gösgen reactor. The dominantly coupled neutrinos are light, and their invariant-lifetime limit tc.m. / mν is 1-3 sec/eV. The subdominantly coupled heavy neutrinos with mass 1...

  18. Small mirror fusion reactors

    International Nuclear Information System (INIS)

    Basic requirements for the pilot plants are that they produce a net product and that they have a potential for commercial upgrade. We have investigated a small standard mirror fusion-fission hybrid, a two-component tandem mirror hybrid, and two versions of a field-reversed mirror fusion reactor--one a steady state, single cell reactor with a neutral beam-sustained plasma, the other a moving ring field-reversed mirror where the plasma passes through a reaction chamber with no energy addition

  19. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2002-04-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  20. Nuclear reactor constructions

    International Nuclear Information System (INIS)

    A nuclear reactor construction comprising a reactor core submerged in a pool of liquid metal coolant in a primary vessel which is suspended from the roof structure of a containment vault. Control rods supported from the roof structure are insertable in the core which is carried on a support structure from the wall of the primary vessel. To prevent excessive relaxation of the support structure whereby the control rods would be displaced relative to the core, the support structure incorporates a normally inactive secondary structure designed to become effective in bracing the primary structure against further relaxation beyond a predetermined limit. (author)

  1. Reactor PIK construction

    International Nuclear Information System (INIS)

    The construction work at the 100 MW researches reactor PIK in year 2002 was in progress. The main activity was concentrated on mechanical, ventilation and electrical equipment. Some systems and subsystems are under adjustment. Hydraulic driving gear for beam shutters are finished in installation, rinsing, and adjusting. Regulating rods test assembling was done. On the critical assembly the first reactor fueling was tested to evaluate the starting neutron source intensity and a sufficiency of existing control and instrument board. Mainline of the PIK facility design and neutron parameters are presented. (author)

  2. Reactor pressure vessel materials

    International Nuclear Information System (INIS)

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 3 offers a detailed treatment of the selection criteria and properties of reactor pressure vessel materials. The main attention is directed towards steel and ingot making and the subsequent material processing

  3. Reactor gamma spectrometry: status

    International Nuclear Information System (INIS)

    Current work is described for Compton Recoil Gamma-Ray Spectrometry including developments in experimental technique as well as recent reactor spectrometry measurements. The current status of the method is described concerning gamma spectromoetry probe design and response characteristics. Emphasis is given to gamma spectrometry work in US LWR and BR programs. Gamma spectrometry in BR environments are outlined by focussing on start-up plans for the Fast Test Reactor (FTR). Gamma spectrometry results are presented for a LWR pressure vessel mockup in the Poolside Critical Assembly (PCA) at Oak Ridge National Laboratory

  4. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised

  5. Space-time reactor kinetics for heterogeneous reactor structure

    International Nuclear Information System (INIS)

    An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods

  6. Risk prevention during reactor shutdown

    International Nuclear Information System (INIS)

    During reactor shutdown potential risks are issued of a number of maintenance operations. In this text we analyse these operations and give the modifications of technical specifications to ameliorate the reactor safety. 4 figs

  7. New fast-reactor approach

    International Nuclear Information System (INIS)

    The design parameters for a 1000 MW LMFBR type reactor are presented. The design requires the multiple primary coolant pumps and heat exchangers to be located around the core within the reactor vessel

  8. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1982 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Since fiscal 1982, Systematic research and development work on safeguards technology has been added to the activities of the Department. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

  9. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  10. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors

  11. High Flux Isotope Reactor (HFIR)

    Data.gov (United States)

    Federal Laboratory Consortium — The HFIR at Oak Ridge National Laboratory is a light-water cooled and moderated reactor that is the United States’ highest flux reactor-based neutron source. HFIR...

  12. The IR-8 reactor operation

    Energy Technology Data Exchange (ETDEWEB)

    Ryazantsev, E.P.; Egorenkov, P.M.; Yashin, A.F. [Reactor Technology and Materials Research Inst. of RRC ' KI' , Moscow (Russian Federation)

    1997-07-01

    At the Russian Research Center 'Kurchatov Institute' (RRC 'KI') the IR-8 reactor commissioning was carried out in 1981. The reactor was developed in return for earlier existing at RRC 'KI' of the IRT-M reactor (modernized IRT reactor, constructed in 1957). The IRT-M reactor was used for investigations in nuclear physics, solid state physics, radiation chemistry, biology as well as to produce isotopes. Under developing the IR-8 reactor the IRT biological shielding with beam tubes and its process systems were used. The IR-8 reactor creation was founded on application developed by then new fuel assemblies (FA) of IRT-3M type, having two times as great surface of heat transfer and 1.75 times higher U-235 load than the FA of the IRT-2M type, which were used in IRT-M reactor. (author)

  13. Power calibrations for TRIGA reactors

    International Nuclear Information System (INIS)

    The purpose of this paper is to establish a framework for the calorimetric power calibration of TRIGA reactors so that reliable results can be obtained with a precision better than ± 5%. Careful application of the same procedures has produced power calibration results that have been reproducible to ± 1.5%. The procedures are equally applicable to the Mark I, Mark II and Mark III reactors as well as to reactors having much larger reactor tanks and to TRIGA reactors capable of forced cooling up to 3 MW in some cases and 15 MW in another case. In the case of forced cooled TRIGA reactors, the calorimetric power calibration is applicable in the natural convection mode for these reactors using exactly the same procedures as are discussed below for the smaller TRIGA reactors (< 2 MW)

  14. Irradiation performance of (Th,Pu)O2 fuel under Pressurized Water Reactor conditions

    Science.gov (United States)

    Boer, B.; Lemehov, S.; Wéber, M.; Parthoens, Y.; Gysemans, M.; McGinley, J.; Somers, J.; Verwerft, M.

    2016-04-01

    This paper examines the in-pile safety performance of (Th,Pu)O2 fuel pins under simulated Pressurized Water Reactor (PWR) conditions. Both sol-gel and SOLMAS produced (Th,Pu)O2 fuels at enrichments of 7.9% and 12.8% in Pu/HM have been irradiated at SCK·CEN. The irradiation has been performed under PWR conditions (155 bar, 300 °C) in a dedicated loop of the BR-2 reactor. The loop is instrumented with flow and temperature monitors at inlet and outlet, which allow for an accurate measurement of the deposited enthalpy. Between 2004 and 2006, four fuel segments were irradiated, with on-line recording of centerline temperature and rod pressure of the two instrumented rods and intermittent non-destructive hot-cell investigations of the other two non-instrumented rods. At the end of 2006, the instrumented rods were unloaded for hot-cell investigations. The hot-cell investigations reduced uncertainties in the power history to build a reliable and consistent irradiation history which can be used to assess and validate fuel performance codes. The on-line recorded temperatures of the instrumented rods are presented in this paper and are compared to corresponding calculations on the basis of the power history. One of the non-instrumented rods was re-inserted in the reactor in 2012 and attained a peak burnup level of 37 GWd/tHM by the end of 2014. The combined data set of on-line measurements and post irradiation examinations enables further code validation. In this context, the results of the in-house MACROS code of SCK·CEN have been compared with the experimental results. The code contains dedicated (Th,Pu)O2 models for the calculation of the thermal conductivity as a function of the burnup and models that determine the radial power profile within the pellet.

  15. Reactor operation safety information document

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  16. Reactor safety in Eastern Europe

    International Nuclear Information System (INIS)

    The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. All papers are indexed separately in report GRS--117. (HP)

  17. VVER and RBMK reactors

    International Nuclear Information System (INIS)

    The safety of VVER and RBMK reactors has been discussed a lot after Chernobyl accident. Some improvements have been performed since that especially in RBMK-reactors and extensive programmes for backfitting have been planned and are partly underway. There are two different sizes of VVER reactors, 440 MW and 1000 MW. The design bases and designs itself vary inside the family of two size classes depending on the age of the plant. The oldest VVER-440 is called model 230 and the newest model 213. The oldest VVER-1000 units (two units) are prototypes that have some unique, nonfavorable features. The next stage of VVER-1000 developement (three units) is model V-302 and the remaining 15 plants in operation are model V-320, but even within this latest model there are some differences. The design bases and designs vary also inside the family of the RBMK reactors exactly the same way as in VVERs. The most important design bases of nuclear power plants designed in the former Soviet Union is presented in this paper. Also some safety advantages and disadvantages of these NPPs are discussed. (au). (5 figs.)

  18. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  19. Studies on reactor physics

    International Nuclear Information System (INIS)

    Most of the peaceful applications of atomic energy are inherently dependent on advances in the science and technology of nuclear reactors, and aspects of this development are part of a major programme of the International Atomic Energy Agency. The most useful role that the Agency can play is as a co-ordinating body or central forum where the trends can be reviewed and the results assessed. Some of the basic studies are carried out by members of the Agency's own scientific staff. The Agency also convenes groups of experts from different countries to examine a particular problem in detail and make any necessary recommendations. Some of the important subjects are discussed at international scientific meetings held by the Agency. One of the subjects covered by such studies is the physics of nuclear reactors and a specific topic recently discussed was Codes for Reactor Computations, on which a seminar was held in Vienna in April this year. Another The members of the Panel described the development of heavy water reactors, the equipment and methods of research currently used, and plans for further development in their respective countries meeting of Panel of Experts on Heavy Water Lattices was held in Vienna in August 1959

  20. Nuclear power reactor physics

    International Nuclear Information System (INIS)

    The purpose of this book is to explain the physical working conditions of nuclear reactors for the benefit of non-specialized engineers and engineering students. One of the leading ideas of this course is to distinguish between two fundamentally different concepts: - a science which could be called neutrodynamics (as distinct from neutron physics which covers the knowledge of the neutron considered as an elementary particle and the study of its interactions with nuclei); the aim of this science is to study the interaction of the neutron gas with real material media; the introduction will however be restricted to its simplified expression, the theory and equation of diffusion; - a special application: reactor physics, which is introduced when the diffusing and absorbing material medium is also multiplying. For this reason the chapter on fission is used to introduce this section. In practice the section on reactor physics is much longer than that devoted to neutrodynamics and it is developed in what seemed to be the most relevant direction: nuclear power reactors. Every effort was made to meet the following three requirements: to define the physical bases of neutron interaction with different materials, to give a correct mathematical treatment within the limit of necessary simplifying hypotheses clearly explained; to propose, whenever possible, numerical applications in order to fix orders of magnitude

  1. Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. [eds.

    1992-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  2. Thermal Reactor Safety

    International Nuclear Information System (INIS)

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods

  3. Nuclear reactor container

    International Nuclear Information System (INIS)

    In a container of a BWR type reactor, spray water is stored in a pedestal cavity. A perforated hole is formed on the side wall of the pedestal, and a stirrer is disposed in the pedestal cavity to stir the stored spray water. During reactor operation, the door on the side wall of the pedestal is closed to prevent discharge of fission products to the dry well when a severe accident should occur. During periodical inspection for the plant, the door is opened to enable an operator to access to the inside of the pedestal. When a molten reactor core should drop to the pedestal cavity, fission products generated from the failed reactor core left in a pressure vessel pass through the spray water in the pedestal cavity. Then, most of the fission products are held in the spray water by a scrubbing effect when they pass through the spray water. In addition, the stored spray water is stirred by the stirrer to enhance the scrubbing effect thereby enabling to further decrease the amount of the fission products discharged to the dry well. (N.H.)

  4. ICF tritium production reactor

    International Nuclear Information System (INIS)

    The conceptual design of an ICF tritium production reactor is described. The chamber design uses a beryllium multiplier and a liquid lithium breeder to achieve a tritium breeding ratio of 2.08. The annual net tritium production of this 532 MW/sub t/ plant is 16.9 kg, and the estimated cost of tritium is $8100/g

  5. Fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1989-01-01

    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.

  6. Nuclear reactor building

    Science.gov (United States)

    Gou, Perng-Fei; Townsend, Harold E.; Barbanti, Giancarlo

    1994-01-01

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

  7. Nuclear reactor constructions

    International Nuclear Information System (INIS)

    A method of constructing a radiation shielding plug for use in the roof of the coolant containment vault of liquid metal cooled fast breeder reactors is described. The construction allows relative movement of that part of service cables and pipes which are carried by the fixed roof and that part which is carried by the rotatable plug. (U.K.)

  8. Fast reactor programme

    International Nuclear Information System (INIS)

    This progress report summarizes the fast reactor research carried out by ECN during the period covering the year 1980. This research is mainly concerned with the cores of sodium-cooled breeders, in particular the SNR-300, and its related safety aspects. It comprises six items: A programme to determine relevant nuclear data of fission- and corrosion-products; A fuel performance programme comprising in-pile cladding failure experiments and a study of the consequences of loss-of-cooling and overpower; Basic research on fuel; Investigation of the changes in the mechanical properties of austenitic stainless steel DIN 1.4948 due to fast neutron doses, this material has been used in the manufacture of the reactor vessel and its internal components; Study of aerosols which could be formed at the time of a fast reactor accident and their progressive behaviour on leaking through cracks in the concrete containment; Studies on heat transfer in a sodium-cooled fast reactor core. As fast breeders operate at high power densities, an accurate knowledge of the heat transfer phenomena under single-phase and two-phase conditions is sought. (Auth.)

  9. Fusion reactor materials

    International Nuclear Information System (INIS)

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

  10. Reactors. Nuclear propulsion ships

    International Nuclear Information System (INIS)

    This article has for object the development of nuclear-powered ships and the conception of the nuclear-powered ship. The technology of the naval propulsion P.W.R. type reactor is described in the article B.N.3 141 'Nuclear Boilers ships'. (N.C.)

  11. Cermet fuel reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

  12. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs

  13. Integral Fast Reactor Program

    International Nuclear Information System (INIS)

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1992. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R ampersand D

  14. Stabilized Spheromak Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, T

    2007-04-03

    The U.S. fusion energy program is focused on research with the potential for studying plasmas at thermonuclear temperatures, currently epitomized by the tokamak-based International Thermonuclear Experimental Reactor (ITER) but also continuing exploratory work on other plasma confinement concepts. Among the latter is the spheromak pursued on the SSPX facility at LLNL. Experiments in SSPX using electrostatic current drive by coaxial guns have now demonstrated stable spheromaks with good heat confinement, if the plasma is maintained near a Taylor state, but the anticipated high current amplification by gun injection has not yet been achieved. In future experiments and reactors, creating and maintaining a stable spheromak configuration at high magnetic field strength may require auxiliary current drive using neutral beams or RF power. Here we show that neutral beam current drive soon to be explored on SSPX could yield a compact spheromak reactor with current drive efficiency comparable to that of steady state tokamaks. Thus, while more will be learned about electrostatic current drive in coming months, results already achieved in SSPX could point to a productive parallel development path pursuing auxiliary current drive, consistent with plans to install neutral beams on SSPX in the near future. Among possible outcomes, spheromak research could also yield pulsed fusion reactors at lower capital cost than any fusion concept yet proposed.

  15. SRP reactor safety evolution

    International Nuclear Information System (INIS)

    The Savannah River Plant reactors have operated for over 100 reactor years without an incident of significant consequence to on or off-site personnel. The reactor safety posture incorporates a conservative, failure-tolerant design; extensive administrative controls carried out through detailed operating and emergency written procedures; and multiple engineered safety systems backed by comprehensive safety analyses, adapting through the years as operating experience, changes in reactor operational modes, equipment modernization, and experience in the nuclear power industry suggested. Independent technical reviews and audits as well as a strong organizational structure also contribute to the defense-in-depth safety posture. A complete review of safety history would discuss all of the above contributors and the interplay of roles. This report, however, is limited to evolution of the engineered safety features and some of the supporting analyses. The discussion of safety history is divided into finite periods of operating history for preservation of historical perspective and ease of understanding by the reader. Programs in progress are also included. The accident at Three Mile Island was assessed for its safety implications to SRP operation. Resulting recommendations and their current status are discussed separately at the end of the report. 16 refs., 3 figs

  16. Fusion reactor materials

    International Nuclear Information System (INIS)

    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics

  17. Department of reactor technology

    International Nuclear Information System (INIS)

    The activities of the Department of Reactor Technology at Risoe during 1980 are described. The work is presented in three chapters: General Information on the Department, Summary of the Department's Development during 1980, and Activities of the Department. Lists of staff, publications, computer programs, and test facilities are included. (author)

  18. The AP1000 reactor

    International Nuclear Information System (INIS)

    The design of the AP1000 reactor began 20 years ago when Westinghouse launched the AP600 reactor project. In fact by re-assessing AP600's safety margins Westinghouse realized that the its power output could be raised without putting at risk its safety standard. The AP1000 was born, it yields 1100 MWe. The main AP1000's design features is its passive safety (particularly after the Fukushima accident) and its modularity. The passive safety of the AP1000 implies: -) no humane intervention needed for 72 hours at least after the incident; -) no necessity for redundant complex safety systems. The modularity means that the plant, the reactor and other buildings are constructed from a choice of 300 modular units. These units can be built off-site and fit together on site. The modularity allows more construction activities to be led simultaneously and more chances to cope with the construction schedule. The NRC has approved the operation license for 30 years of the first AP1000 being built in the Usa (Vogtle plant in Georgia). 4 AP1000 are being built in China (Sanmen and Haiyang sites) and 6 others are planned in the Usa. Westinghouse is convinced that the AP1000's passive safety makes it more attractive. Let us not forget that Westinghouse was at the origin of the concept of pressurized water reactors, an idea adopted for half the nuclear power stations in the world and for all the plants now active in France. (A.C.)

  19. Fast reactor programme

    International Nuclear Information System (INIS)

    This progress report summarises the fast reactor research carried out at the Netherlands Energy Research Centre during the year 1981. The neutron and fission product cross sections of various isotopes have been evaluated. In the fuel performance programme, some preliminary results are given and irradiation facilities described. Creep experiments on various stainless steel components are reported

  20. Pressure tube type reactor

    International Nuclear Information System (INIS)

    Heretofore, a pressure tube type reactor has a problem in that the evaluation for the reactor core performance is complicate and no sufficient consideration is made for the economical property, to increase the size of a calandria tank and make the cost expensive. Then, in the present invention, the inner diameter of a pressure tube is set to greater than 50% of the lattice gap in a square lattice like arrangement, and the difference between the inner and the outer diameters of the calandria tube is set smaller than 20% of the lattice gap. Further, the inner diameter of the pressure tube is set to greater than 40% and the difference between the inner and the outer diameters of the calandria tube is set smaller than 30% of the lattice gap in a triangle lattice arrangement. Then, heavy water-to-fuel volume ratio can be determined appropriately and the value for the coolant void coefficient is made more negative side, to improve the self controllability inherent to the reactor. In particular, when 72 to 90 fuel rods are arranged per one pressure tube, the power density per one fuel rod is can be increased by about twice. Accordingly, the number of the pressure tubes can be reduced about to one-half, thereby enabling to remarkably decrease the diameter of the reactor core and to reduce the size of the calandria, which is economical. (N.H.)

  1. TRIGA reactor characteristics

    International Nuclear Information System (INIS)

    The general design, characteristics and parameters of TRIGA reactors and fuel are described. This is a training module with the learning objectives: to understand the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and realize the differences between TRIGA fuels and other more traditional. 10 figs., 6 tabs. (nevyjel)

  2. SNAP Nuclear Space Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Corliss, William R

    1966-01-01

    This booklet describes the principles of nuclear-reactor space power plants and shows how they will contribute to the exploration and use of space. It compares them with chemical fuels, solar cells, and systems using energy from radioisotopes. The SNAP (Systems for Nuclear Auxiliary Power) Program, begun in 1955, is described.

  3. Fast breeder reactor

    International Nuclear Information System (INIS)

    This paper outlined the present status of FBR development in six countries and reviewed Japanese activities on FBR development. Joyo experimental FBR has accumulated a lot of technical data including irradiation tests of advanced fuels and was now long shut down due to the partial obstruction of rotating plug movement. Monju prototype FBR reactor experienced a sodium leakage in its secondary heat transfer system during performance tests in December 1995 and had been shut down until May 2010. Feasibility study on commercialized FBR cycle system ended in March 2006 and proposed the concept of commercialized FBR cycle technologies. In order to plan a demonstration reactor, research and development of innovative technologies are conducted as the FaCT (Fast Reactor Cycle Technology Development) Project. In connection with the results of this research and development, a 5-party council of Japan was established to discuss processes of demonstration and commercialization of FBR cycle systems in Japan. Joint efforts were made for a demonstration reactor to be committed in 2015, in addition to start operation around 2025 aiming at the commercialization of FBR before 2050. (T. Tanaka)

  4. The Chernobyl reactor accident

    International Nuclear Information System (INIS)

    The documentation abstracted contains a complete survey of the broadcasts transmitted by the Russian wire service of the Deutsche Welle radio station between April 28 and Mai 15, 1986 on the occasion of the Chernobyl reactor accident. Access is given to extracts of the remarkable eastern and western echoes on the broadcasts of the Deutsche Welle. (HP)

  5. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. The concept evolved in the 1960's with the objective of developing a reactor design which could be used for a wide range of mobile power generation systems including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests and in-reactor irradiation tests using cermet fuel were carried out by General Electric in the 1960's as part of the 710 Development Program and by Argonne National laboratory in a subsequent activity. Cermet fuel development programs are currently underway at Argonne National laboratory and Pacific Northwest Laboratory as part of the Multi-Megawatt Space Power Program. Key features of the cermet fueled reactor design are 1) the ability to achieve very high coolant exit temperatures, and 2) thermal shock resistance during rapid power changes, and 3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, there is a potential for achieving a long operating life because of 1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and 2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core

  6. Reactor power measuring device

    International Nuclear Information System (INIS)

    The device of the present invention comprises a γ-thermometer disposed in a BWR type reactor, a first amplifier for amplifying the output thereof, a fission ionization chamber disposed in the reactor separately from the γ-thermometer, a second amplifier for amplifying the output thereof, a differential circuit for differentiating the output signal of the second amplifier and a first adding circuit for adding an output signal of the differential circuit and an output signal of the first amplifier. Alternatively, a γ-ray self-powered neutron detector may be disposed instead of the fission ionization chamber. A second adding circuit is also disposed for adding the output signals of plurality of differentiation circuits and inputting the result to the first adding circuit. A sensitivity controller is disposed upstream of the first adder for controlling the sensitivity of the fission ionization chamber. Then, even if time delay should be caused in the γ-thermometer, output signals with good time response characteristic can be obtained by using signals of LPRM or SPND, and currently changing output of the reactor can be measured accurately to provide an effect on the improvement of the safety and operation controllability of the reactor. (N.H.)

  7. Fast reactors: potential for power

    International Nuclear Information System (INIS)

    The subject is discussed as follows: basic facts about conventional and fast reactors; uranium economy; plutonium and fast reactors; cooling systems; sodium coolant; safety engineering; handling and recycling plutonium; safeguards; development of fast reactors in Britain and abroad; future progress. (U.K.)

  8. Reactor physics problems on HCPWR

    International Nuclear Information System (INIS)

    Reactor physics problems on high conversion pressurized water reactors (HCPWRs) are discussed. Described in this report are outline of the HCPWR, expected accuracy for the various reactor physical qualities, and method for K-effective calculation in the resonance energy area. And requested further research problems are shown. The target value of the conversion ratio are also discussed. (author)

  9. Nuclear reactor with control rods

    International Nuclear Information System (INIS)

    The invention relates to liquid cooled nuclear reactors. In particular, it concerns reactors with mobile control rods in a straight line and guide tubes to guide these control rods through the internal upper components of the reactor vessel and in the aligned fuel assemblies of the core

  10. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

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

  11. Reactor regulating and protection system for a light water reactor

    International Nuclear Information System (INIS)

    Microprocessor based systems are developed for reactor regulation and protection of LWR. A triple modular redundancy approach is followed for the design of this system. This system is functionally partitioned into two sub-systems - Reactor Regulating System (RRS) and Reactor Trip Logic System (RTLS). RRS controls the reactor power as per demand and RTLS generates the reactor trip on abnormal process conditions. This paper describes the details of RRS and RTLS system architecture and fault tolerant and fail-safe features used in the system design. (author)

  12. The Fast-spectrum Transmutation Experimental Facility FASTEF: Main design achievements (part 2: Reactor building design and plant layout) within the FP7-CDT collaborative project of the European Commission

    International Nuclear Information System (INIS)

    MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) is the flexible experimental accelerator-driven system (ADS) in development at SCK-CEN in replacement of its material testing reactor BR2. SCK-CEN in association with 17 European partners from industry, research centres and academia, responded to the FP7 (Seventh Framework Programme) call from the European Commission to establish a Central Design Team (CDT) for the design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) able to demonstrate efficient transmutation and associated technology through a system working in subcritical and/or critical mode. The project has started on April 01, 2009 for a period of three years. In this paper, we present the latest concept of the reactor building and the plant layout. The FASTEF facility has evolved quite a lot since the intermediate reporting done at the ICAPP'10 and ICAPP'11 conferences 1,2. Many iterations have been performed to take into account the safety requirements. The present configuration enables an easy operation and maintenance of the facility, including the possibility to change large components of the reactor. In a companion paper 3, we present the latest configuration of the reactor core and primary system. (authors)

  13. The reactor Pegase

    International Nuclear Information System (INIS)

    The reactor Pegase is designed for testing fuel elements for gas-cooled power reactors. Experience has shown that the classical multi-purpose test reactors are not well adapted to these tests. On another side, the introduction of these test elements into the existing power reactors involves numerous problems, which limits their interest. Pegase, which is designed to satisfy these experimental needs, is composed of a parallelepipedal core of enriched Uranium, moderated and cooled by pressurized water. This core is used as a neutron source for eight autonomous loops, containing the elements to be tested, and situated around the core. The core and the eight loops are immersed in a irradiation pool. The loops are placed on the bottom of the pool so, it is possible to move a loop away from the core, or to remove it from the pool without interfering with the operation of the other loops. The irradiation conditions are adjusted, making the synthesis of the following development works. - Experimental studies on Peggy, a zero power critical facility, mock up of Pegase in operation since 1961: measurements of neutron flux level, radial and axial fly distributions on the experiments. Effect of burnable poisons and of movements of the control rods; adjustment of devices (reflectors, screens etc..) needed for optimum performances. - Experimental work on two prototype autonomous loops, heated electrically to the nominal operating power (in operation since 1961): development of the thermodynamic measurements, thermal balances parameters for control of the operating conditions, natural convection. - Studies on Pegase operating under power; thermodynamic measurements on the core circuits on the independent loop circuits; neutronic measurements, etc... The reactor Pegase went critical on the 4. of April 1963 and reached the nominal power of 30 MW on the 28. of May 1963. (authors)

  14. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    Science.gov (United States)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  15. Reactor physics activities in Japan

    International Nuclear Information System (INIS)

    This report reviews the research activity in reactor physics field in Japan during July, 1992 - July, 1993. The review was performed in the following fields : nuclear data evaluation, calculational method development, fast reactor physics, thermal reactor physics, advanced core design, fusion reactor neutronics, nuclear criticality safety, shielding, incineration of radioactive nuclear wastes, noise analysis and control and national programs. The main references were taken from journals and reports published during this period. The research committee of reactor physics is responsible for the review work. (author)

  16. Advanced reactor experimental facilities

    International Nuclear Information System (INIS)

    For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

  17. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    This report summarizes main research achievements in the 48th fiscal year which were made by Reactor Engineering Division consisted of eight laboratories and Computing Center. The major research and development projects, with which the research programmes in the Division are associated, are development of High Temperature Gas Cooled Reactor for multi-purpose use, development of Liquid Metal Fast Breeder Reactor conducted by Power Reactor and Nuclear Fuel Development Corporation, and Engineering Research Programme for Thermonuclear Fusion Reactor. Many achievements are reported in various research items such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of Computing Center. (auth.)

  18. Methanogenesis in Thermophilic Biogas Reactors

    DEFF Research Database (Denmark)

    Ahring, Birgitte Kiær

    1995-01-01

    Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process...... as indicated by a lower concentration of volatile fatty acids in the effluent from the reactors. The specific methanogenic activity in a thermophilic pilot-plant biogas reactor fed with a mixture of cow and pig manure reflected the stability of the reactor. The numbers of methanogens counted by the most...... against Methanothrix soehngenii or Methanothrix CALS-I in any of the thermophilic biogas reactors examined. Studies using 2-14C-labeled acetate showed that at high concentrations (more than approx. 1 mM) acetate was metabolized via the aceticlastic pathway, transforming the methyl-group of acetate...

  19. Reactor simulator development. Workshop material

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has established a programme in nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in reactor operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and training material, sponsors training courses and workshops, and distributes documentation and computer programs. This publication consists of course material for workshops on development of such reactor simulators. Participants in the workshops are provided with instruction and practice in the development of reactor simulation computer codes using a model development system that assembles integrated codes from a selection of pre-programmed and tested sub-components. This provides insight and understanding into the construction and assumptions of the codes that model the design and operational characteristics of various power reactor systems. The main objective is to demonstrate simple nuclear reactor dynamics with hands-on simulation experience. Using one of the modular development systems, CASSIMtm , a simple point kinetic reactor model is developed, followed by a model that simulates the Xenon/Iodine concentration on changes in reactor power. Lastly, an absorber and adjuster control rod, and a liquid zone model are developed to control reactivity. The built model is used to demonstrate reactor behavior in sub-critical, critical and supercritical states, and to observe the impact of malfunctions of various reactivity control mechanisms on reactor dynamics. Using a PHWR simulator, participants practice typical procedures for a reactor startup and approach to criticality. This workshop material consists of an introduction to systems used for developing reactor simulators, an overview of the dynamic simulation

  20. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  1. Feasible reactor power cutback logic development for an integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Han, Soon-Kyoo [KHNP Co., Ltd., Uljin-gun, Gyeong-buk (Korea, Republic of); Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok [Korea Atomic Energy Research Institute (KAERI), Daedeokdaero, Yuseong, Daejeon (Korea, Republic of)

    2013-07-15

    Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

  2. Feasible reactor power cutback logic development for an integral reactor

    International Nuclear Information System (INIS)

    Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

  3. Elk River Reactor dismantling

    International Nuclear Information System (INIS)

    The dismantling program was carried out in three overlapping phases: the planning phase which included the preliminary planning and selection of the dismantling approach, the dismantling phase which included all work performed to remove the reactor facility and restore the site to its pre-reactor condition, and the closeout phase which included the final site survey and efforts necessary to terminate the AEC license and contract. Of particular interest was the use of a remotely operated plasma cutting torch to section the pressure vessel internals, the pressure vessel and the outer thermal shield, the use of explosives in removal of the biological shield and the method of establishment of the criteria for material disposal

  4. Selecting reactor operator trainees

    International Nuclear Information System (INIS)

    Reactor operator trainee selection tends to be more effective if tailored to a utility's unique needs, and offers the organization a better chance for compliance with Federal regulations than if selection methods are adopted without benefit of local research. The costs of operator training range from $50,000 to $100,000. The test validity relative to a variety of training grades and performance measures is reviewed. Of interest is the degree to which tests differentiate reactor operators with respect to simulator training grades and performance in simulator operation; forms of evaluation which have become fairly standard throughout the power industry. The tests administered to each individual were selected because of their presumed relevance to training grades, and the aptitude measures are intended to assess an individual's potential to benefit from training. Tests, availability, form, the abilities they measure, and the time limit are described. (MCW)

  5. Embattled breeder reactor

    International Nuclear Information System (INIS)

    A commercial fuel-cloning machine, a nuclear breeder reactor, is yet to produce electricity in the United States. It is expensive in capital and fuel costs, its fuel that must be reprocessed can become a link to nuclear weapons manufacture, and its safety is no greater than conventional nuclear reactors. The breeder has had on-again/off-again administrative support from Washington. Opponents worry about escalating costs and failure to develop alternatives like solar energy. Proponents say fossil-fuel depletion will eventually force long-term renewable resources such as the breeder anyway. Some who share parts of both views oppose present policy regarding the Clinch River Breeder demonstration plant specifically. The correct choices on breeder concept development and commercialization will be known in 2050. 3 figures

  6. Robot for reactor dismantling

    International Nuclear Information System (INIS)

    Purpose: To enable to attain the operation on a cylindrical coordinate system thereby performing dismantling operation exactly and at a high reliability. Constitution: A reactor dismantling robot is suspended by ropes by an elevating device to the inside of reactor shielding walls. The robot has a fixed portion having a plurality of legs abutting against the inner surface of the shieling walls while extending and shrinking radially in the horizontal direction and an arm portion having an operation arm disposed with a shielding wall breaking operation device. The arm portion is disposed with a mechanism for vertically moving the operation arm and a mechanism for forwarding and backwarding the operation arm in the horizontal direction and the arm portion itself is constituted so as to be rotatable around a vertical axis. (Seki, T.)

  7. Compact fusion reactors

    CERN Document Server

    CERN. Geneva

    2015-01-01

    Fusion research is currently to a large extent focused on tokamak (ITER) and inertial confinement (NIF) research. In addition to these large international or national efforts there are private companies performing fusion research using much smaller devices than ITER or NIF. The attempt to achieve fusion energy production through relatively small and compact devices compared to tokamaks decreases the costs and building time of the reactors and this has allowed some private companies to enter the field, like EMC2, General Fusion, Helion Energy, Lawrenceville Plasma Physics and Lockheed Martin. Some of these companies are trying to demonstrate net energy production within the next few years. If they are successful their next step is to attempt to commercialize their technology. In this presentation an overview of compact fusion reactor concepts is given.

  8. The Pegase reactor loops

    International Nuclear Information System (INIS)

    After 4 years operation, experimentation and maintenance of the gas loops built especially for the nuclear fuel testing reactor Pegase, it appears desirable not only to gather together in a single document the essential characteristics and particularities of these devices and of their associated equipment, but also to give the reasons for the technical modifications and the way in which they were carried out; this is done here by the persons themselves who were responsible, day after day, for operating these loops. This essentially practically experience thus complements the careful research and preliminary testing carried out on these loops or on their prototypes. It should be of interest to those who deal with problems concerned with the design or operation of irradiation loops in experimental reactors or of similar equipment. (authors)

  9. Spherical torus fusion reactor

    Science.gov (United States)

    Peng, Yueng-Kay M.

    1989-01-01

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  10. Decay of reactor neutrinos

    International Nuclear Information System (INIS)

    We consider the decay of massive neutrinos which couple to electrons and are, therefore, produced in nuclear reactors. Lifetime limits for the γ and electron-positron decay modes of these neutrinos are deduced from the experimental limit on the singles count rate in the detector used to study neutrino oscillations at the Goesgen reactor. The dominantly coupled neutrinos are light, and their invariant-lifetime limit t/sup c.m.//m/sub ν/ is 1--3 sec/eV. The subdominantly coupled heavy neutrinos with mass 1--4 MeV could decay into electron-positron pairs. These pairs were not observed, and from the absence of such a signal we deduce restrictions on the corresponding mixing parameters

  11. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  12. Fast reactor database

    International Nuclear Information System (INIS)

    This publication contains detailed data on liquid metal cooled fast reactors (LMFRs), specifically plant parameters and design details. Each LMFR power plant is characterized by about 400 parameters, by design data and by relevant materials. The report provides general and detailed design characteristics including structural materials, data on experimental, demonstration, prototype and commercial size LMFRs. The focus is on practical issues that are useful to engineers, scientists, managers and university students and professors. The report includes updated information contained in IAEA previous publications on LMFR plant parameters: IWGRF/51 (1985) and IWGFR/80 (1991) and reflects experience gained from two consultants meetings held in Vienna (1993,1994). This compilation of data was produced by members of the IAEA International Working Group on Fast Reactors (IWGFR)

  13. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    In 1998, the NNSA organized to complete the nuclear safety review on the test loop in-reactor operation of the High-flux Engineering Experimental Reactor (HFEER) and the re-operation of the China Pulsed Reactor and the Uranium-water Criticality Facility. The NNSA conducted the nuclear safety review on the CP application of the China Experimental Fast Reactor (CEFR) and the siting of China Advanced Research Reactor (CARR), and carried out the construction supervision on HTR-10, and dealt with the event about the technological tube breakage of HWRR and other events

  14. Gaseous fuel reactor research

    Science.gov (United States)

    Thom, K.; Schneider, R. T.

    1977-01-01

    The paper reviews studies dealing with the concept of a gaseous fuel reactor and describes the structure and plans of the current NASA research program of experiments on uranium hexafluoride systems and uranium plasma systems. Results of research into the basic properties of uranium plasmas and fissioning gases are reported. The nuclear pumped laser is described, and the main results of experiments with these devices are summarized.

  15. Neutrino physics at reactors

    International Nuclear Information System (INIS)

    Reviewing experiments with neutrinos from reactors seems at first to be a simple task, since there were only few. But almost all of them address fundamental questions in particle physics and are of great relevance. This paper reports on these experiments which made use of some of the most sophisticated techniques available at the time they were designed. In these two respects new proposed experiments re in the tradition of the older ones

  16. Decommissioning of research reactors

    International Nuclear Information System (INIS)

    Research reactors of WWR-S type were built in countries under Soviet influence in '60, last century and consequently reached their service life. Decommissioning implies removal of all radioactive components, processing, conditioning and final disposal in full safety of all sources on site of radiological pollution. The WWR-S reactor at Bucuresti-Magurele was put into function in 1957 and operated until 1997 when it was stopped and put into conservation in view of decommissioning. Presented are three decommissioning variants: 1. Reactor shut-down for a long period (30-50 years) what would entail a substantial decrease of contamination with lower costs in dismantling, mechanical, chemical and physical processing followed by final disposal of the radioactive wastes. The drawback of this solution is the life prolongation of a non-productive nuclear unit requiring funds for personnel, control, maintenance, etc; 2. Decommissioning in a single stage what implies large funds for a immediate investment; 3. Extending the operation on a series of stages rather phased in time to allow a more convenient flow of funds and also to gather technical solutions, better than the present ones. This latter option seems to be optimal for the case of the WWR-S Research at Bucharest-Magurele Reactor. Equipment and technologies should be developed in order to ensure the technical background of the first operations of decommissioning: equipment for scarification, dismantling, dismemberment in a highly radioactive environment; cutting-to-pieces and disassembling technologies; decontamination modern technologies. Concomitantly, nuclear safety and quality assurance regulations and programmes, specific to decommissioning projects should be implemented, as well as a modern, coherent and reliable system of data acquisition, recording and storing. Also the impact of decommissioning must be thoroughly evaluated. The national team of specialists will be assisted by IAEA experts to ensure the

  17. Nuclear reactor constructions

    International Nuclear Information System (INIS)

    An improvement in the construction of liquid metal cooled nuclear reactors of the kind in which the fuel assembly is submerged in a pool of coolant contained by a primary vessel housed in a concrete vault, is described. In this modification the roof of the vault carries heat exchangers immersed in the pool of coolant, the lower ends of which are hydraulically damped against oscillation caused by seismic disturbances. (U.K.)

  18. Reactor pressure vessel steels

    International Nuclear Information System (INIS)

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

  19. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    SCK-CEN's programme on fusion reactor materials includes studies (1) to investigate fracture mechanics of neutron-irradiated beryllium; (2) to describe the helium behaviour in irradiated beryllium at atomic scale; (3) to define the kinetics of beryllium reacting with air or steam; (3) to perform a feasibility study for the testing of integrated blanket modules under neutron irradiation. Progress and achievements in 1997 are reported

  20. Halden reactor project

    International Nuclear Information System (INIS)

    The research programme at the Halden Project is focused on the following three areas: 1. In-core behavior of reactor fuel, particularly reliability and safety aspects, which is studied through irradiation of test fuel elements. 2. Prediction, surveillance and control of fuel and core performance for which models of fuel and core behavior are developed. 3. Applications of process computers to power plant control, for which prototype software systems and hardware arrangements are developed