WorldWideScience

Sample records for br-2 zero power mock-up reactor

  1. Analysis of the BN-600 fast-spectrum core mock-up at BFS-2 zero-power facility using MCNPX

    International Nuclear Information System (INIS)

    Highlights: ► We model the BFS-62-3A experiment with the MCNPX code and four nuclear libraries. ► We show the impact on reactivity of heterogeneous structures in the reactor. ► We model experimental uncertainties, e.g. in materials dimension and density. ► The model agrees with experiments on k-eff, CR worth, Na voids and fission rates. ► The analysis questions experimental data measured in the reflector region. - Abstract: A 3D full-core heterogeneous model of the BFS-62-3A critical benchmark experiment was developed and validated using the Monte Carlo MCNPX-2.4.0 code. The BFS-2 critical facility at the Institute of Physics and Power Engineering (IPPE) was designed for simulation of fast reactor core neutronics, and for the validation of codes and nuclear data. The BFS-62-3A critical benchmark experiment was set-up as a mock-up of the BN-600 reactor core with (U, Pu) O2 fuel of 17% Pu content and stainless-steel reflectors. It was operated to measure the effective multiplication factor, spectral indices, radial fission rate distributions, control rod worths and sodium void effects. In the present study, special care was taken to run the MCNPX model to make Monte-Carlo confidence intervals comparable with uncertainties reported in the experiments; such as in material dimensions, number densities and isotopic compositions. In addition to the effective multiplication factor, sodium void effect, fission rate distributions and control rod worth were calculated. Simulations were carried out with four different modern nuclear data libraries; the primary aim being to estimate sensitivity of the results to the nuclear data. This task, besides being a library comparison, is also meant as a first step towards a code-to-code verification with deterministic methods. Results agree well with experimental values on most of the nuclear characteristics, even though a discrepancy up to more than 20% was found on the flux distribution in the stainless-steel reflector

  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. 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

  6. 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)

  7. 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

  8. 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)

  9. Effect of inhomogeneities inserted into fuel assembly in the VVER-1000 mock-up on the LR-0 research reactor

    International Nuclear Information System (INIS)

    Highlights: • Repairs in fuel assembly. • Effect of control rod. • Calculations and experiment comparison of inhomogeneities effect. • Influence of nuclear data libraries on inhomogeneities effect. - Abstract: The work presents a detailed comparison of the calculated and experimentally determined effect of inhomogeneities inserted into a fuel assembly on power density in the adjoining fuel pins. The power density is determined by means of fission density. The fission density was determined using measurements of photons emitted by the decay of the 92Sr fission product, which has been produced during a 2.5 h irradiation at a power level of 9.41 W. Experiments were performed in the VVER-1000 mock-up placed in LR-0 reactor. The calculations were made using the Monte Carlo approach. The effect of different data libraries on results are discussed as well

  10. High-power tests of a remote steering launcher mock-up at 140 GHz

    International Nuclear Information System (INIS)

    This paper reports the results of the high-power test of a remote steering launcher mock-up at 140 GHz, which were performed at the ECRH installation for the future stellarator W7-X at IPP Greifswald. The mock-up test system consists of a 6.62 m long corrugated square waveguide with a steerable optic at the entrance and various diagnostics at the exit of the waveguide. A straight and a dog-leg version of the launcher were investigated. The high-power tests of the straight setup have been performed with powers up to P0 = 700 kW (typically 500 kW) and pulse lengths of up to 10 seconds. For both polarizations (parallel and perpendicular to the steering plane), no arcing was observed in spite of the fact, that the experiments were performed under ambient atmospheric conditions. After the integration of 2 mitre bends in the setup, arcing limited the usable parameter range. The ohmic loss Px of the waveguide was measured via the temperature increase of the waveguide wall, and was used to calibrate the calculated angular dependence of the total ohmic losses of the waveguide. Short-pulse radiation pattern measurements with thermographic recording show high beam quality and confirm the steering range of -12 deg. < φ < 12deg

  11. Numerical evaluation of weld overlay applied to a pressurized water reactor nozzle mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Rabello, Emerson G.; Silva, Luiz L.; Gomes, Paulo T.V., E-mail: egr@cdtn.b, E-mail: silvall@cdtn.b, E-mail: gomespt@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Integridade Estrutural

    2011-07-01

    The primary water stress corrosion cracking (PWSCC) is a major mechanism of failure in the primary circuit of PWR type nuclear power plants. The PWSCC is associated with the presence of corrosive environment, the susceptibility to corrosion cracking of the materials involved and the tensile stresses presence. Residual stresses generated during dissimilar materials welding can contribute to PWSCC. An alternative to the PWSCC mitigation is the application of external weld layers in the regions of greatest susceptibility to corrosion cracking. This process, called Weld Overlay (WOL), has been widely used in regions of dissimilar weld (low alloy steel and stainless steel with nickel alloy addition) of nozzles and pipes on the primary circuit in order to promote internal compressive stresses on the wall of these components. This paper presents the steps required to the numerical stress evaluation (by finite element method) during the dissimilar materials welding as well as application of Weld Overlay process in a nozzle mock-up. Thus, one can evaluate the effectiveness of the application of weld overlay process to internal compressive stress generation on the wall nozzle. (author)

  12. Control-room mock-up for the Philippsburg nuclear power plant, Unit 2

    International Nuclear Information System (INIS)

    The paper describes the major aspects of the construction of a full-scale control-room mock-up for Unit 2 (1300 MW PWR) of the Philippsburg nuclear power plant. The multitude of monitoring and control systems, co-operative control-room modelling by the planning staff as well as the operating staff, including the feedback of experience gained during operation, and the practice-oriented application of human factors engineering, especially the optimization of the man/machine interface, are emphasized. The control-room complex is subdivided into three functional areas: the entrance area, the central shift supervisor desk and, as a focal point, the process instrumentation and control area which includes the master control console (operator main control console and associated information board) and the system control consoles. The most important improvements in the application of human factors engineering in process instrumentation and control are listed. The consequent structuring of mimic diagrams and instruments both by colour and shape is a fundamental step. A computerized operator support system has been installed with the aim of improving the man/machine interface. (author)

  13. 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

  14. Calculations of partial LOCA in a swimming-pool-reactor with MTR-elements and planned mock-up experiment

    International Nuclear Information System (INIS)

    A partial uncovering of the MTR fuel plates of the swimming pool reactor SAPHIR located at the Swiss Federal Reactor Research Institute (E.I.R.) could be caused by a loss of coolant accident due to a beam tube break. The transient temperature excursions of the fuel plates during the LOCA have been predicted with computer simulations. Because a reliable prediction of the flow regime and hence the heat transfer in the uncovered part of the plate is not possible with current knowledge, a parametric study employing different heat transfer models is presented in this paper. The results show, that the heat transfer model in the uncovered part of the fuel plate has an important influence on the predicted temperatures. A mock-up experimental facility, which will supply data for the heat transfer occuring in the uncovered part, will also be described at the end of the paper. (author)

  15. Fast Neutron Transport in the Biological Shielding Model and Other Regions of the VVER-1000 Mock-Up on the LR-0 Research Reactor

    Science.gov (United States)

    Košťál, Michal; Milčák, Ján; Cvachovec, František; Jánský, Bohumil; Rypar, Vojtěch; Juříček, Vlastimil; Novák, Evžen; Egorov, Alexander; Zaritskiy, Sergey

    2016-02-01

    A set of benchmark experiments was carried out in the full scale VVER-1000 mock-up on the reactor LR-0 in order to validate neutron transport calculation methodologies and to perform the optimization of the shape and locations of neutron flux operation monitors channels inside the shielding of the new VVER-1000 type reactors. Compared with previous experiments on the VVER-1000 mock-up on the reactor LR-0, the fast neutron spectra were measured in the extended neutron energy interval (0.1-10 MeV) and new calculations were carried out with the MCNPX code using various nuclear data libraries (ENDF/B VII.0, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND 2009, and CENDL 3.1). Measurements and calculations were carried out at different points in the mock-up. The calculation and experimental data are compared.

  16. 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

  17. Neutron and gamma field investigations in the VVER-1000 mock-up concrete shielding on the reactor LR-0

    International Nuclear Information System (INIS)

    Two sets of neutron and gamma field investigations were carried out in the dismountable model of radiation shielding of the VVER-1000 mock-up on the LR-0 reactor. First, measurements and calculations of the 3He(n,p)T reaction rate and fast neutrons and gamma flux spectra in the operational neutron monitor channel inside a concrete shielding for different shapes and locations of the channel (cylindrical channel in a concrete, channels with collimator in a concrete, cylindrical channel in a graphite). In all cases measurements and calculations of the 3He(n,p)T reaction rate were done with and without an additional moderator-polyethylene insert inside the channel. Second, measurements and calculations of the 3He(n,p)T reaction rate spatial distribution inside a concrete. The 3He(n,p)T reaction rate measurements and calculations were carried out exploring the relative thermal neutron density in the channels and its space distribution in the concrete. Fast neutrons and gamma measurements were carried out with a stilbene (45 x 45 mm) scintillation spectrometer in the energy regions 0.5-10 MeV (neutrons) and 0.2-10 MeV (gammas). (authors)

  18. 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

  19. The Swedish Zero Power Reactor R0

    International Nuclear Information System (INIS)

    The reactor R0 is a critical facility built for heavy water and natural uranium or fuel of low enrichment,, The first criticality was achieved September 25, 1959. During a first period of more than two years the R0 will be operated as a bare reactor in order to simplify interpretation of results. The reactor tank is 3. 2 m high and 2. 25 m in diameter. The fuel suspension system is quite flexible in order to facilitate fuel exchange and lattice variations. The temperature of the water can be varied between about 10 and 90 C by means of a heater and a cooler placed in the external circulating system. The instrumentation of the reactor has to meet the safety requirements not only during operation but also during rearrangements of the core in the shut-down state. Therefore, the shut-down state is always defined by a certain low 'safe' moderator level in the reactor tank. A number of safety rods are normally kept above the moderator ready for action. For manual or automatic control of the reactor power a specially designed piston pump is needed, by which the moderator level is varied. The pump speed is controlled from the reactor power error by means of a Ward-Leonard system. Moderator level measurement is made by means of a water gauge with an accuracy of ± 0. 1 mm

  20. The Swedish Zero Power Reactor R0

    Energy Technology Data Exchange (ETDEWEB)

    Landergaard, Olof; Cavallin, Kaj; Jonsson, Georg

    1961-05-15

    The reactor R0 is a critical facility built for heavy water and natural uranium or fuel of low enrichment,, The first criticality was achieved September 25, 1959. During a first period of more than two years the R0 will be operated as a bare reactor in order to simplify interpretation of results. The reactor tank is 3. 2 m high and 2. 25 m in diameter. The fuel suspension system is quite flexible in order to facilitate fuel exchange and lattice variations. The temperature of the water can be varied between about 10 and 90 C by means of a heater and a cooler placed in the external circulating system. The instrumentation of the reactor has to meet the safety requirements not only during operation but also during rearrangements of the core in the shut-down state. Therefore, the shut-down state is always defined by a certain low 'safe' moderator level in the reactor tank. A number of safety rods are normally kept above the moderator ready for action. For manual or automatic control of the reactor power a specially designed piston pump is needed, by which the moderator level is varied. The pump speed is controlled from the reactor power error by means of a Ward-Leonard system. Moderator level measurement is made by means of a water gauge with an accuracy of {+-} 0. 1 mm.

  1. Temperature field downstream of an heated bundle mock-up results for different power distribution

    International Nuclear Information System (INIS)

    The aim of these peculiar experiments performed on the ML4 loop in ISPRA is to evaluate the characteristics of the temperature field over a length of 20 to 30 dias downstream of a rod bundle for different temperatures profiles at the bundle outlet. The final purpose of this work will be to establish either directly or through models whether it is possible or not to detect subassembly failures using suitable of the subassembly outlet temperature signal. 15 hours of digital and analog recording were taped for five different power distributions in the bundle. The total power dissipation remained constant during the whole run. Two flow rates and seven axial location were investigated. It is shown that the different temperature profiles produce slight differences in the variance and skewness of the temperature signal measured along the axis of the pipe over 20 dias

  2. 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

  3. For zero maintenance in nuclear reactors and power plants

    International Nuclear Information System (INIS)

    The goals of 'zero maintenance' in nuclear reactors and power plants can be achieved by implementing on line condition monitoring along with concepts of predictive maintenance, quality control, maintainability, reliability, human engineering, zero defect, operations research, tribology and good maintenance management system. Instruments for condition monitoring are also listed. (author)

  4. A study of 239Pu production rate in a water cooled natural uranium blanket mock-up of a fusion-fission hybrid reactor

    Science.gov (United States)

    Feng, Song; Liu, Rong; Lu, Xinxin; Yang, Yiwei; Xu, Kun; Wang, Mei; Zhu, Tonghua; Jiang, Li; Qin, Jianguo; Jiang, Jieqiong; Han, Zijie; Lai, Caifeng; Wen, Zhongwei

    2016-03-01

    The 239Pu production rate is important data in neutronics design for a natural uranium blanket of a fusion-fission hybrid reactor, and the accuracy and reliability should be validated by integral experiments. The distribution of 239Pu production rates in a subcritical natural uranium blanket mock-up was obtained for the first time with a D-T neutron generator by using an activation technique. Natural uranium foils were placed in different spatial locations of the mock-up, the counts of 277.6 keV γ-rays emitted from 239Np generated by 238U capture reaction were measured by an HPGe γ spectrometer, and the self-absorption of natural uranium foils was corrected. The experiment was analyzed using the Super Monte Carlo neutron transport code SuperMC2.0 with recent nuclear data of 238U from the ENDF/B-VII.0, ENDF/B-VII.1, JENDL-4.0u2, JEFF-3.2 and CENDL-3.1 libraries. Calculation results with the JEFF-3.2 library agree with the experimental ones best, and they agree within the experimental uncertainty in general with the average ratios of calculation results to experimental results (C/E) in the range of 0.93 to 1.01.

  5. Siloette, Siloe mock-up

    International Nuclear Information System (INIS)

    Siloette is the Siloe mock-up. The main installations are described: various tanks, building, auxiliaries, control systems... Precis ions are given about precautions taken for using spent fuel elements. (authors)

  6. 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.

  7. 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.

  8. Recent activities at the zero-power teaching reactor CROCUS

    International Nuclear Information System (INIS)

    CROCUS is a zero-power critical facility used mainly for educational purposes at the Swiss Federal Institute of Technology (EPFL) in Lausanne, Switzerland. It is a low-enriched-uranium fuelled, light-water moderated reactor, with the fission power limited to 100 W. The presentation will discuss the crucial role of CROCUS in teaching -- both as framework for reactor practicals offered to physics students at EPFL and as key educational tool in the recently established Swiss Master of Science in Nuclear Engineering. Regular development work is needed for the various instruments and components associated with the facility. As illustration, the recently completed refurbishment of the control rod system and the related calibration experiments will also be discussed.

  9. 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)

  10. Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor

    International Nuclear Information System (INIS)

    In addition to the safety analysis for the zero power RB reactor, this report contains a general description of the reactor, reactor components, auxiliary equipment and the reactor building. Reactor Rb has been reconstructed during 1961-1962 and supplied with new safety-control system as well as with a complete dosimetry instrumentation. Since RB reactor was constructed without shielding special attention is devoted to safety and protection of the staff performing experiments. Due to changed circumstances in the Institute ( start-up of the RA 7 MW power reactor) the role of the RB reactor was redefined

  11. 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)

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

    International Nuclear Information System (INIS)

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

  13. Strength of Mock-up Trial Grout

    DEFF Research Database (Denmark)

    Sørensen, Eigil V.

    The present report describes tests carried out on samples taken and cast during the execution of a mock-up trial placement of the high performance grout MASTERFLOW 9500 on January 21, 2009.......The present report describes tests carried out on samples taken and cast during the execution of a mock-up trial placement of the high performance grout MASTERFLOW 9500 on January 21, 2009....

  14. Experimental results from the BNL zero power reactor HITREX

    Energy Technology Data Exchange (ETDEWEB)

    Kitching, S.J.; Lewis, T.A.; Playle, T.S.

    1973-10-15

    This report presents experimental results obtained with the BNL reactor Hitrex. Measurements of reactivity, and of thermal and fast neutron reaction rate distributions have been made with various experimental control rod configurations.

  15. Hot zero power reactor calculations using the Insilico code

    Science.gov (United States)

    Hamilton, Steven P.; Evans, Thomas M.; Davidson, Gregory G.; Johnson, Seth R.; Pandya, Tara M.; Godfrey, Andrew T.

    2016-06-01

    In this paper we describe the reactor physics simulation capabilities of the Insilico code. A description of the various capabilities of the code is provided, including detailed discussion of the geometry, meshing, cross section processing, and neutron transport options. Numerical results demonstrate that Insilico using an SPN solver with pin-homogenized cross section generation is capable of delivering highly accurate full-core simulation of various pressurized water reactor problems. Comparison to both Monte Carlo calculations and measured plant data is provided.

  16. Zirconium-hydride solid zero power reactor and its application research

    International Nuclear Information System (INIS)

    The Zirconium Hydride Solid Zero Power Reactor built at China Institute of Atomic Energy is introduced. In the reactor Zirconium-hydride is used as moderator, plexiglass as reflector and U3O8 with enrichment of 20% as the fuel, Since its initial criticality, the physical characteristics and safety features have been measured with the result showing that the reactor has sound stability and high sensitivity, etc. It has been successfully used for the personnel training and for the testing of reactor control instruments and experiment devices. It also presents the special advantage for the pre-research of some applications

  17. Feasibility study of the thermo-siphon mock-up test

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jung Woon; Kim, Young Jin; Lee, Kye Hong; Kim, Young Ki; Jeong, Sang Kwon

    2004-09-01

    Described is the feasibility of the thermo-siphon mock-up test for the HANARO-CNS facility. The purposes of the mock-up tests are discussed in detail as the three concepts: for the detailed design, for the operation of the CNS facility, for the safety assurance of itself. This report considers the two stages of mock-up tests in terms of the experimental schedule and plan. As the first stage, the small-size mock-up test using Argon will be implemented to obtain the experience in the cryogenic fluid and to understand the basic concept of the CNS thermo-siphon. In the second stage, two kinds of mock-up tests are discussed: the full-scale mock-up test using liquid hydrogen or the integrated final test using hydrogen outside the reactor after the full-scale mock-up test using Freon gas. The contents discussed in this report will be the basis or the guide lines for the mock-up test. In addition, the results of the mock-up test will be the foundation for the safe operation of the HANARO-CNS facility.

  18. Feasibility study of the thermo-siphon mock-up test

    International Nuclear Information System (INIS)

    Described is the feasibility of the thermo-siphon mock-up test for the HANARO-CNS facility. The purposes of the mock-up tests are discussed in detail as the three concepts: for the detailed design, for the operation of the CNS facility, for the safety assurance of itself. This report considers the two stages of mock-up tests in terms of the experimental schedule and plan. As the first stage, the small-size mock-up test using Argon will be implemented to obtain the experience in the cryogenic fluid and to understand the basic concept of the CNS thermo-siphon. In the second stage, two kinds of mock-up tests are discussed: the full-scale mock-up test using liquid hydrogen or the integrated final test using hydrogen outside the reactor after the full-scale mock-up test using Freon gas. The contents discussed in this report will be the basis or the guide lines for the mock-up test. In addition, the results of the mock-up test will be the foundation for the safe operation of the HANARO-CNS facility

  19. The control-and-instrumentation system of the IEA zero power reactor and its reliability calculation

    International Nuclear Information System (INIS)

    The control-and instrumentation system for the Instituto de Energia Atomica Zero Power Reactor is described and the design criteria are presented and discussed. The reliability analysis for the reactor protection system was performed using the fault tree method. This was done using a computer code based on the Monte Carlo simulation. That code is an adaptation of the SAFTE-I, for the IBM 360/155 IEA Computer. (Author)

  20. 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

  1. A simulated test of physical starting and reactor physics on zero power facility of PWR

    International Nuclear Information System (INIS)

    The core neutron economics has been verified through experiments conducted at a zero power reactor with baffles of various thickness. A simulated test of physical starting of Qinshan PWR has been introduced. The feasibility and safety of the programme are verified. The research provides a valuable foundation for developing physical starting programme

  2. Design of the zero power reactor core of Instituto de Energia Atomica, SP, Brazil

    International Nuclear Information System (INIS)

    The main characteristics of a graphite moderated core of a critical assembly to be installed in the zero power reactor of the Instituto de Energia Atomica have been defined. Several simple geometric configurations have been selected and criticality studies have been made. The necessary quantity of fissile uranium has been calculated. (author)

  3. Measures of the zero power nuclear reactor's kinetic parameters with application of noise analysis

    International Nuclear Information System (INIS)

    The purpose of this work was to establish an experimental technique based on noise analysis for measuring the ratio of kinetic parameters β/ Λ and the power of the Zero Power Nuclear Reactor IPEN-MB 01. A through study of the microscopic and macroscopic noise analysis techniques has been carried out. The Langevin technique and the point kinetic model were chosen to describe the stochastic phenomena that occur in the zero power reactor. Measurements have been made using two compensated ionization chambers localized in the water reflector at symmetric positions in order to minimize spatial effects on the neutron flux fluctuation. Power calibrations based on the low frequency plateau of the cross-power spectral density has also been carried out. (author)

  4. Thermal fatigue cycling of Be/Cu joining mock-ups

    International Nuclear Information System (INIS)

    To evaluate beryllium-to-copper joining techniques for potential use by US manufacturers in making first wall components for International Thermonuclear Experimental Reactor (ITER), we tested two mock-ups with S65C beryllium (Be) tiles Hot Isostatic Pressing (HIP) bonded to CuCrZr heat sinks. Under the aegis of the US ITER Project Office, Sandia prepared the mock-ups working with industrial vendors and performed high heat flux testing at Sandia's Plasma Material Test Facility (PMTF) to ascertain the robustness of the Be/Cu joints to 1000 thermal fatigue cycles at a heat flux level of 1.5 MW/m2. Thermal stress analysis provided insight into choosing the heat flux and flow conditions required for accelerated fatigue testing at 1000 cycles and 1.5 MW/m2 that is comparable to the 12,000 cycles and 0.875 MW/m2 required for the ITER First Wall Qualification Mock-ups. Each mock-up had three Be tiles, 35.5 mm square and 10 mm thick, bonded to a CuCrZr heat sink 134.5 mm x 36 mm x 25 mm with a single bored 12.7 mm (dia.) cooling channel. The bonding techniques included various interlayer metallizations and HIPping at 100 MPa pressure and temperature of 580 or 560 deg. C for 2 h. Each tile had a thermocouple (TC) in the center 1 mm below the Be/Cu interface. The test arrangement allowed for both mock-ups to be tested at the same time with alternate heating and cooling cycles of equal duration of 30 s. A total power of 12.7 kW was absorbed by the heated area of 4000 mm2 during the on-cycle. The mock-up was cooled by water at 2.3 m/s (0.27 kg/s), 1 MPa and 20 deg. C inlet temperature. These operating conditions did not permit the mock-ups to cool down to their initial temperature state during the off-cycle. Both mock-ups survived 1000 cycles with no significant changes. The temperature of the top surface on each reached 254 deg. C; while the center TCs reached 136 and 139 deg. C, respectively. Despite localized changes observed in the surface emissivity, the corrected

  5. The key role of critical mock-up facilities for neutronic physics assessment of advanced reactors: an overview of Cea Cadarache tools

    International Nuclear Information System (INIS)

    The Experimental Physics section of CEA Cadarache operates three critical facilities devoted to neutronic studies of advanced reactors (EOLE, MINERVE and MASURCA) covering a large scope of interests. These include 100% MOX core in ABWR qualification, knowledge improvement of basic nuclear data for heavy nuclides for new options of the fuel cycle - especially the multi-recycling of plutonium - and accelerator-driven systems neutronic behaviour for transmutation studies. The paper describes these facilities, the scientific programmes associated and the progressive improvement of experimental techniques, the aim being to significantly reduce the uncertainties regarding the evaluation of the physical parameters. (authors)

  6. Development and trial manufacturing of 1/2-scale partial mock-up of blanket box structure for fusion experimental reactor

    International Nuclear Information System (INIS)

    Conceptual design of breeding blanket has been discussed during the CDA (Conceptual Design Activities) of ITER (International Thermonuclear Experimental Reactor). Structural concept of breeding blanket is based on box structure integrated with first wall and shield, which consists of three coolant manifolds for first wall, breeding and shield regions. The first wall must have cooling channels to remove surface heat flux and nuclear heating. The box structure includes plates to form the manifolds and stiffening ribs to withstand enormous electromagnetic load, coolant pressure and blanket internal (purge gas) pressure. A 1/2-scale partial model of the blanket box structure for the outboard side module near midplane is manufactured to estimate the fabrication technology, i.e. diffusion bonding by HIP (Hot Isostatic Pressing) and EBW (Electron Beam Welding) procedure. Fabrication accuracy is a key issue to manufacture first wall panel because bending deformation during HIP may not be small for a large size structure. Data on bending deformation during HIP was obtained by preliminary manufacturing of HIP elements. For the shield structure, it is necessary to reduce the welding strain and residual stress of the weldment to establish the fabrication procedure. Optimal shape of the parts forming the manifolds, welding locations and welding sequence have been investigated. In addition, preliminary EBW tests have been performed in order to select the EBW conditions, and fundamental data on built-up shield have been obtained. Especially, welding deformation by joining the first wall panel to the shield has been measured, and total deformation to build-up shield by EBW has been found to be smaller than 2 mm. Consequently, the feasibility of fabrication technologies has been successfully demonstrated for a 1m-scaled box structure including the first wall with cooling channels by means of HIP, EBW and TIG (Tungsten Inert Gas arc)-welding. (author)

  7. Major accident analyses for experimental zero-power fast reactor assemblies

    International Nuclear Information System (INIS)

    A study has been made of the possibility, mechanism, and consequence of melt-down and other major nuclear accidents for a ZPR-III type experimental zero-power fast reactor of the two-half type. This study has been supplemented by an evaluation of the importance of the Doppler effect for a wide range of nuclear reactor assemblies for such a reactor. A melt-down event is highly improbable because of the restricted sequence of events which must be postulated. A discussion of the mechanism of the collapse is followed by the results of coupled neutronics-hydrodynamic s calculations for two zero-power assemblies. A 1200-l core has been examined because it represents a relatively large reactor of common core composition. A smaller core with a high-void fraction has been examined as a potentially more dangerous system. Very different time-wise behaviour has been found for the two systems. For sharp accidents in zero-power assemblies, the U235-atoms, separated as plates of enriched uranium, will heat very rapidly while the remainder of the core remains essentially cold, so that a gas of U235-vapour will provide the disassembly pressure. The adaption of the neutronics-hydrodynamic s code AX-I to the use of a Van der Waals gas is described. Another important change in the equation of state used in the code is to employ a Mie-Griineisen type equation derivable from solid state theory. This change provides a more satisfactory way to evaluate the pressure term for cores of variable composition. Because the highly enriched U235 plates of a zero-power assembly will heat much more rapidly than the depleted uranium plates, the possibility of a net positive Doppler effect is much larger for an experimental assembly than for the equivalent power breeder reactor. This hazard has been examined for a range of possible assemblies. These calculations indicate that the Doppler coefficient for a zero-power assembly does not become important as a hazard until one approaches systems with the

  8. 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)

  9. Bentonite THM behaviour mock-up studies

    International Nuclear Information System (INIS)

    Nuclear waste disposal rely on multi-barrier system. Engineered Barrier Systems make use of swelling clay buffer set in place unsaturated in deposition hole. After waste emplacement, buffer hydrates and swells, being submitted to the heat from nuclear waste decay. In order to characterize the THM behaviour of swelling clays under temperatures exceeding 100 C, Andra has gathered research laboratories, CEA/LECBA, Eurogeomat and EDF/CPM to conduct experimental a d modelling studies. The analysis of the state of art led to the definition of mock-up tests to evaluate effects of high temperatures and high thermal gradients on the heat and mass transfer, and the stress-strain behaviour of initially unsaturated MX80 bentonite during a thermal loading at constant volume for both closed and opened system. This paper presents the first mock-up tests and their experimental result. (authors)

  10. 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)

  11. Design and Analysis of the Power Control System of the Fast Zero Energy Reactor FR-0

    International Nuclear Information System (INIS)

    This report describes the power control by means of the fine-control rod and the design of the control system of the fast zero energy reactor FR-0 located in Studsvik, Sweden. System requirements and some operational conditions were used as design criteria. Manual and automatic control is possible. Variable electronic end-stops for the control rod have been designed, because of the special construction of the reactor and control rod. Noise in the control system caused by the reactor, detector and electronics caused disturbances of the control system at the lower power levels. The noise power-spectrum was measured. Statistical design methods, using the measured noise power spectrum, were used to design filters, which will reduce the influence of the noise at the lower power levels. Root Loci sketches and Bode diagrams were used for stability analyses. The system was simulated on an analogue computer, taking into account even nonlinearities of the control system and noise. Typical cases of reactor operation were simulated and stability analysis performed

  12. The making of a mock-up

    DEFF Research Database (Denmark)

    Rosenqvist, Tanja Schultz; Heimdal, Elisabeth Jacobsen

    As part of a research project about user involvement in textile design we have carried out two Design:Labs (Binder & Brandt 2008) engaging different stakeholders in designing textile products for Danish hospital environments. In this paper we follow a mock-up session done as part of the second...... important element engaging in the hypothetical space of the Design:Lab, as it can function as a scaffold for ideas, ease the communication within the group, as well as help communicating ideas to people who have not participated in the Design:Lab....

  13. Modeling of a double fission chamber using MCNPX for power calibration at the zero-power teaching reactor CROCUS

    International Nuclear Information System (INIS)

    MCNPX-2.5 simulations and experiments were performed to improve the power prediction of the zero-power teaching reactor CROCUS at the Ecole Polytechnique Federale de Lausanne (EPFL) using a calibrated double fission chamber (DFC). The CROCUS facility is a zero-power critical reactor used for educational purposes. Traditionally, the core power is determined by irradiating thin gold foils placed along the core centre and by measuring the 411 keV γ-rays on HPGe detectors. The average 197Au(n,γ) self-shielded macroscopic cross-section obtained with the deterministic BOXER code (1σ - 10%) is employed to determine the flux and the reactor power. To benchmark the BOXER calculations, a DFC containing known amounts of enriched 235U and 239Pu deposits was installed within the reflector core and simulated with MCNPX-2.5/JEF-2.2. Particular care was taken to model the fissile deposits allowing to reduce the power uncertainty to 2% compared to the gold foil technique. A code-to-code comparison (BOXER vs. MCNPX) was performed and the results have shown a good agreement (2 to 5%) for most of the quantities calculated (flux, reaction rates). However, the normalization factor differed by 17% (flux-to-power ratio). Consequently, the core power was overestimated by 17% until now. Finally, the current investigations lead to an improved fission power determination and contribute to better core safety standard. (author)

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

    International Nuclear Information System (INIS)

    Oak Ridge National Laboratory, under contract with Argonne National Laboratory, completed the fabrication of 1743 fuel elements for use in their Zero Power Reactor. The contract also included recovery of 20 kg of 233U from rejected elements. This report describes the steps associated with conversion of purified uranyl nitrate (as solution) to U3O8 powder (suitable for fuel) and subsequent charging, sealing, decontamination, and testing of the fuel elements (packets) preparatory to shipment. The nuclear safety, radiation exposures, and quality assurance aspects of the program are discussed

  15. Development and Optimization of Nuclear Heating Measurement Techniques in Zero Power Experimental Reactors

    International Nuclear Information System (INIS)

    The objective of this study is to develop nuclear heating measurement methods in Zero Power experimental reactors. This paper presents the analysis of Thermo- Luminescent Detectors (TLDs) and Optically Stimulated Luminescent Detectors (OSLDs) experiments in the MINERVE research reactors at the French Atomic Energy and Alternative Energies Commission center in Cadarache. The experimental sources of uncertainties on the dose have been reduced by using the optimum conditions of charged particle equilibrium (CPE) of the calibration step and reactor measurement for each detector types; by improving the process of the TLDs/OSLDs reading and calibration processes. The interpretation of these measurements needs to take into account several correction factors related to both the environment of calibration step and the type of detectors used. Similarly, the correction due to the neutrons contributions to the total dose integrated by the detectors is evaluated with Monte Carlo calculation methods. These calculations are based on MCNP simulations of neutron-gamma and gamma-electron transport coupled particles using ENDF/B-VI nuclear data library. TLDs and OSLDs are positioned inside aluminum or hafnium pillboxes. Comparisons between calculated and measured integral gamma-ray absorbed doses by TLDs in these new experiments are carried out in the MINERVE reactor in the surrounding aluminum material. They show that calculations slightly overestimate the measurements by about 8 %. By using OSLDs, the calculation slightly overestimates the measurement by about 6 %. (authors)

  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. Passive and Active Radiation Measurements Capability at the INL Zero Power Physics Reactor (ZPPR) Facility

    International Nuclear Information System (INIS)

    The Zero Power Physics Reactor (ZPPR) facility is a Department of Energy facility located in the Idaho National Laboratory's (INL) Materials and Fuels Complex. It contains various nuclear and non-nuclear materials that are available to support many radiation measurement assessments. User-selected, single material, nuclear and non-nuclear materials can be readily utilized with ZPPR clamshell containers with almost no criticality concerns. If custom, multi-material configurations are desired, the ZPPR clamshell or an approved aluminum Inspection Object (IO) Box container may be utilized, yet each specific material configuration will require a criticality assessment. As an example of the specialized material configurations possible, the National Nuclear Security Agency's Office of Nuclear Verification (NNSA/NA 243) has sponsored the assembly of six material configurations. These are shown in the Appendixes and have been designated for semi-permanent storage that can be available to support various radiation measurement applications.

  18. 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)

  19. Detaching test of an irradiated mock-up containing with tritium from the core of JMTR

    International Nuclear Information System (INIS)

    The second in-situ irradiation experiment using a mock-up (ORIENT-II, JMTR capsule Number: 99M-54J) with a tritium breeder (Li2TiO3) pebble bed in the Japan Materials Testing Reactor (JMTR) was finished on Aug. 1, 2006. Correspondingly an investigation on the detaching procedure of the irradiated mock-up containing with tritium was carried out, followed by the actual detaching test of this mock-up. Firstly, tritium removal characteristics were studied for the irradiated mock-up, the sweep gas tube, the protective tube and the junction box, Out-of pile melting/enclosing tests of the sealing plug were also carried out for prevention of tritium leakage from sweep gas lines of Li2TiO3 pebble bed. From the results, tritium release amount were estimated during the detaching test of the real irradiated mock-up was estimated, and the melting/enclosing procedures of sealing plug were fixed. Then, the actual detaching test of the Li2TiO3 pebble bed was carried out. The tritium release to the area of detaching test was favorably suppressed, decreased, and the irradiated mock-up was safely detached from the core of JMTR as planned. This report describes the results of 1) tritium removal tests for the sweep gas line and the protective tube, 2) out-of pile melting/enclosing test of the sealing plug, 3) examination of the detaching procedure before the detaching test of the irradiated mock-up, and 4) the actual detaching test, as well as knowledge obtained from these tests and works. (author)

  20. 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)

  1. Measurements of the Reactivity Properties of the Aagesta Nuclear Power Reactor at Zero Power

    International Nuclear Information System (INIS)

    The moderator level and temperature coefficients of reactivity and control rod differential reactivity worths have been determined at zero power by means of period measurements. The moderator level coefficient and the corresponding critical level have been measured for the 32, 68 and 136 fuel assembly cores at room temperature for cores with and without control rods. From these results the worths of control rods have been derived. HETERO calculations give up to 15 % lower values than the experimental results. The cold fresh core has an excess reactivity of 9.0 ± 0.2 %. The temperature coefficient and differential control rod worths were measured for the fully loaded core with filled tank in the temperature range between 30 and 210 deg C. Critical positions as a function of temperature were obtained for the corresponding control rod groups. No relevant calculations of the temperature coefficient for comparison with the experimental values have yet been made, but the experimental results together with measured critical control rod positions give good opportunities to check calculational programs. HETERO has been shown in these cases to reproduce differential control rod worths and critical positions fairly well. However, a certain underestimation of the rod effectiveness is quite noticeable. The relative increase in control rod effectiveness with a temperature change from 20 to 220 deg C has been estimated to be 0.29 ± 0.06

  2. Two-dimensional heterogeneous transport theory hot zero-power benchmarks for the WWER-1000 reactors

    International Nuclear Information System (INIS)

    The Mariko code, based on the method of characteristics, has been used to calculate several two-dimensional full core heterogeneous 23-group transport theory solutions for hot zero power states of the WWER-1000 reactor. The initial loading for the three-year fuel cycle is considered. Helios-1.5 has been used to prepare 23-group cross-section data. The benchmarks differ by the positions (up or down) of the control rods groups. The asymptotic assembly-averaged and cell-averaged two-group diffusion parameters for all assembly types, the group-to-group albedo son the radial reflector boundary, and the effective diffusion parameters, including reference discontinuity factors, for the radial reflector nodes are all calculated by Mariko. The accuracy of the SPPS-1.6 and DYN3D nodal diffusion codes and the HEX2DA pin-by-pin diffusion code have been tested. The benchmark with all control rods down poses a great problem for the modal diffusion codes, the maximum error in the relative assembly-wise power distribution reaching 18% for SPPS-1.6. The pin-by-pin code performs well in all cases (Authors)

  3. How tangible mock-ups support design collaboration

    DEFF Research Database (Denmark)

    Brandt, Eva

    2007-01-01

    differences in these affected the dialogue. Reflective conversations were established by using tangible mock-ups as 'things-to-think with'. They served as boundary objects that spanned the gap between the different competencies and interests of participants in design. The design mock-ups evoked different...

  4. Pulsed Neutron Measurements on a Heavy Water Power Reactor (MZFR) at Zero Energy

    International Nuclear Information System (INIS)

    The pulsed neutron method was used for zero-power measurements in the core of a heavy water reactor. Various methods were used for the evaluation of the pulsed measurements. The so-called ''integral'' evaluation methods are based on theories published by Sjöstrand and Gozani; so far they have been applied mainly to light water reactors. These methods use not only the prompt neutron decay constant but also the information contained in the delayed neutron tails to determine the reactivity. For measurements on the heavy water reactor, however, the methods had to be modified so as to adequately take into account the time dependence of the delayed neutrons. The fraction of the delayed neutrons was calculated using a reasonable assumption for its time dependence. All the information needed could be obtained from the measurements. These methods are well suited for hand calculations to yield the reactivity with proper accuracy. An analytical procedure was applied to check the results of the integral methods. This essentially involves the exact calculation of the time dependence of the delayed neutron fraction by an iteration procedure. The results of the different evaluation methods mentioned above are compared by plotting them as functions of the D2O level and of the boron concentration. Due to the inclined control rods the flux distribution is distorted in a rather complicated manner when the rods are inserted. Therefore the time dependence of this distribution was measured for different positions of the pulsed neutron source. It was possible to find one position for which the influence of higher modes on the measurements of the shutdown reactivity was sufficiently small. Finally it is shown that the values of (δρ(H, ci)/δ(l/H2)) H = Hi and (δρ(Hi, c)/δc) c = ci (ρ reactivity, Hi critical D2O level for boron concentration c1) obtained by period measurements in the slightly supercritical state and pulsed measurements in the subcritical state are in excellent

  5. Experimental estimation of the delayed neutron fraction βeff of the MAESTRO core in the MINERVE zero power reactor

    International Nuclear Information System (INIS)

    A method for determining the effective delayed neutron fraction βeff using in-pile reactivity oscillations and Fourier analysis is presented. This method is based on measurements of the reactor's power response to small periodic in-pile reactivity perturbations and utilizes Fourier analysis for reconstruction of the reactor zero power transfer function. This approach enables the estimation of βeff using multi-parameter nonlinear weighted least-squares fit. The method extends previous works by accounting for higher harmonics excitation in the frequency domain by the trapezoidal reactivity signal, both in the reactivity perturbation and in the reactor power response. We show that by using this new approach it is possible to obtain the reactor transfer function in a wide range of frequencies, using only a single oscillation frequency. This method is applied to a set of measurements of the MAESTRO core configuration in the MINERVE zero power reactor (ZPR) located at the Cadarache Research Center. The derived value of βeff, using this method, is 711 ± 17 pcm. (author)

  6. 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

  7. Preparation of mandatory documentation before the start up of the RA-0 'zero power' nuclear reactor at Cordoba National University

    International Nuclear Information System (INIS)

    Before the start up of the RA-0 'zero power' nuclear reactor installed at Cordoba National University, it was necessary to send to the Regulatory Authority the mandatory documentation which is required in the licensing process. With the previous papers existing for the operation in the first years of the '70, a work program for the future operational training personnel was elaborated. Based on the Authority's applicable rules and the recommendations and with particular criteria originated in the working university conditions, the SAFETY report of RA-0 nuclear reactor was prepared. This paper describes the principal contents, items and documents involved in the safety report. (Author)

  8. Neutronics Experiment on A HCPB Breeder Blanket Mock-Up

    International Nuclear Information System (INIS)

    A neutronics experiment has been performed in the frame of European Fusion Technology Program on a mock-up of the EU Test Blanket Module (TBM), Helium Cooled Pebble Bed (HCPB) concept, with the objective to validate the capability of nuclear data to predict nuclear responses, such as the tritium production rate (TPR), with qualified uncertainties. The experiment has been carried out at the FNG 14-MeV neutron source in collaboration between ENEA, Technische Universitaet Dresden, Forschungszentrum Karlsruhe, J. Stefan Institute Ljubljana and with the participation of JAEA. The mock-up, designed in such a way to replicate all relevant nuclear features of the TBM-HCPB, consisted of a steel box containing beryllium block and two intermediate steel cassettes, filled with of Li2CO3 powder, replicating the breeder insert main characteristics: radial thickness, distance between ceramic layers, thickness of ceramic layers and of steel walls. In the experiment, the TPR has been measured using Li2CO3 pellets at various depths at two symmetrical positions at each depth, one in the upper and one in the lower cassette. Twelve pellets were used at each position to determine the TPR profile through the cassette. Three independent measurements were performed by ENEA, TUD/VKTA and JAEA. The neutron flux in the beryllium layer was measured as well using activation foils. The measured tritium production in the TBM (E) was compared with the same quantity (C) calculated by the MCNP.4c using a very detailed model of the experimental set up, and using neutron cross sections from the European Fusion File (EFF ver.3.1) and from the Fusion Evaluated Nuclear Data Library (FENDL ver. 2.1, ITER reference neutron library). C/E ratios were obtained with a total uncertainty on the C/E comparison less than 9% (2 s). A sensitivity and uncertainty analysis has also been performed to evaluate the calculation uncertainty due to the uncertainty on neutron cross sections. The results of such analysis

  9. 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.

  10. Mock-up experiments for the project of high dose irradiation on the RPV concrete

    International Nuclear Information System (INIS)

    Aging of NPP's concrete structures comes into growing interest in connection with solution of life extension programmes of operated units. Securing continued safe operation of NPPs calls for additional proofs of suitable long term behaviour of loaded reinforced concrete structures. An irradiation test of concrete samples was performed in the core of the LVR-15 reactor. The irradiation capsule was hung in the irradiation channel and the cooling of the capsule was ensured through direct contact of the capsule wall with the primary circuit water. Cylindrical, serpentine concrete samples (50 mm in diameter and 100 mm in length), representing composition of WWER RPV cavity, was chosen as a compromise of mechanical properties testing needs and dimension limitations of reactor irradiation channel. Heating during irradiation test was maintained under 93 Celsius degrees by cooling and was controlled by embedded thermocouple. Design of the cooling management was supported by computational analysis. The dependencies of heated concrete samples to the neutron fluence and the gamma heating were obtained by changing the thermal power of the reactor and by changing the vertical position of the sample in the irradiation channel. The irradiation capsule was filled with inert gas (helium) to allow the measurement of generated gas. The determination of concrete samples activity for long-term irradiation was performed on the principles of the Neutron Activation Analysis. Preliminary mock-up tests have proved the ability to fulfill technical needs for planned high dose irradiation experiment

  11. Millimeter wave experiment of ITER equatorial EC launcher mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, K.; Oda, Y.; Kajiwara, K.; Kobayashi, N.; Isozaki, M.; Sakamoto, K. [Japan Atomic Energy Agency, Naka (Japan); Omori, T.; Henderson, M. [ITER Organization, St. Paul lez Durance (France)

    2014-02-12

    The full-scale mock-up of the equatorial launcher was fabricated in basis of the baseline design to investigate the mm-wave propagation properties of the launcher, the manufacturability, the cooling line management, how to assemble the components and so on. The mock-up consists of one of three mm-wave transmission sets and one of eight waveguide lines can deliver the mm-wave power. The mock-up was connected to the ITER compatible transmission line and the 170GHz gyrotron and the high power experiment was carried out. The measured radiation pattern of the beam at the location of 2.5m away from the EL mock-up shows the successful steering capability of 20°∼40°. It was also revealed that the radiated profile at both steering and fixed focusing mirror agreed with the calculation. The result also suggests that some unwanted modes are included in the radiated beam. Transmission of 0.5MW-0.4sec and of 0.12MW-50sec were also demonstrated.

  12. 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.

  13. 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)

  14. Verification Survey of the Building 315 Zero Power Reactor-6 Facility, Argonne National Laboratory-East, Argonne, Illinois

    Energy Technology Data Exchange (ETDEWEB)

    W. C. Adams

    2007-05-25

    Oak Ridge Institute for Science and Education (ORISE) conducted independent verification radiological survey activities at Argonne National Laboratory’s Building 315, Zero Power Reactor-6 facility in Argonne, Illinois. Independent verification survey activities included document and data reviews, alpha plus beta and gamma surface scans, alpha and beta surface activity measurements, and instrumentation comparisons. An interim letter report and a draft report, documenting the verification survey findings, were submitted to the DOE on November 8, 2006 and February 22, 2007, respectively (ORISE 2006b and 2007).

  15. Siloette, Siloe mock-up; Siloette, modele nucleaire de siloe

    Energy Technology Data Exchange (ETDEWEB)

    Delcroix, V.; Jeanne, G.; Mitault, G.; Schulhof, P. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1964-07-01

    Siloette is the Siloe mock-up. The main installations are described: various tanks, building, auxiliaries, control systems... Precis ions are given about precautions taken for using spent fuel elements. (authors) [French] Siloette est le modele nucleaire de SILOE. On decrit ses diverses installations: bassins, batiments, auxiliaires, controle... Des precisions sont donnees sur les precautions prises pour y utiliser des elements uses. (auteurs)

  16. Digital Mock-up Technology in Product Development and Research

    Institute of Scientific and Technical Information of China (English)

    CHEN Xu; ZHANG Pandeng; YANG Cheng; XU Zhongming

    2006-01-01

    After introducing the present status of digital mock-up (DMU) technology in product development and research, the modeling and its key technologies in product design are described. The architecture of digital design platform system for main DMU model is developed. Based on the architecture, a method of skeleton design has been applied to the development of digital design system.

  17. Demonstration of the LHC Safety Training Tunnel Mock-Up

    CERN Multimedia

    Brice, Maximilien

    2014-01-01

    Members of CERN's management visit the LHC tunnel mock-up at the Safety Training Centre on the Prévessin site. The facility is used to train personnel in emergency responses including the use of masks and safe evacuation.

  18. 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)

  19. 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)

  20. Verification Survey of the Building 315 Zero Power Reactor-6 Facility, Argonne National Laboratory-East, Argonne, Illinois

    International Nuclear Information System (INIS)

    Oak Ridge Institute for Science and Education (ORISE) conducted independent verification radiological survey activities at Argonne National Laboratory's Building 315, Zero Power Reactor-6 facility in Argonne, Illinois. Independent verification survey activities included document and data reviews, alpha plus beta and gamma surface scans, alpha and beta surface activity measurements, and instrumentation comparisons. An interim letter report and a draft report, documenting the verification survey findings, were submitted to the DOE on November 8, 2006 and February 22, 2007, respectively (ORISE 2006b and 2007). Argonne National Laboratory-East (ANL-E) is owned by the U.S. Department of Energy (DOE) and is operated under a contract with the University of Chicago. Fundamental and applied research in the physical, biomedical, and environmental sciences are conducted at ANL-E and the laboratory serves as a major center of energy research and development. Building 315, which was completed in 1962, contained two cells, Cells 5 and 4, for holding Zero Power Reactor (ZPR)-6 and ZPR-9, respectively. These reactors were built to increase the knowledge and understanding of fast reactor technology. ZPR-6 was also referred to as the Fast Critical Facility and focused on fast reactor studies for civilian power production. ZPR-9 was used for nuclear rocket and fast reactor studies. In 1967, the reactors were converted for plutonium use. The reactors operated from the mid-1960's until 1982 when they were both shut down. Low levels of radioactivity were expected to be present due to the operating power levels of the ZPR's being restricted to well below 1,000 watts. To evaluate the presence of radiological contamination, DOE characterized the ZPRs in 2001. Currently, the Melt Attack and Coolability Experiments (MACE) and Melt Coolability and Concrete Interaction (MCCI) Experiments are being conducted in Cell 4 where the ZPR-9 is located (ANL 2002 and 2006). ANL has performed final

  1. Analysis of Rod Removal Transient Experiments in VVER Reactors at Zero Power

    International Nuclear Information System (INIS)

    Within the context of the Fissile Materials Disposition Program of the U.S. Department of Energy we analyzed rod removal transient experiments performed at the Kurchatov Institute in a full-scale mockup of WWER reactors. The transients were started (via water inlet) in slightly (few cents) supercritical configurations with all the control rods withdrawn. After a few minutes, control rods banks or individual control rods were f and t inserted and later withdrawn (returning to the initial state). Available experimental data include the relative time profiles of nine incore and excore detectors. Because of the mild nature of the transients (very low power and no more than 2 $ reactivities) we decided to use a quasistatic approach. The time-dependent flux is factorized into two terms: a function of phase space, given by the solution of the static equation with parametric excitation; and a function of time, given by the solution of the point kinetic equations with time-dependent kinetics para meters. Due to the nature of the experiment, cold conditions, control rods withdrawn and critical state with water level, the power distributions, measured and calculated, are quite unusual, with the inner part of the core heavily shielded. Measured power levels at the center of the reactor are almost 20 times smaller than similar regions at the periphery. Transport and diffusion calculations of the power distributions are in reasonable agreement, so the division code BOLD-VENTURE was used to calculate the kinetics parameters and the relative changes of the detector field of view. The numerical integration of the time-dependent part of the solution was made with the LSODE package using ENDFIB-V and VI delayed neutron data. Very good results were obtained for the nine lime profiles

  2. EVALUATION OF ZERO-POWER, ELEVATED-TEMPERATURE MEASUREMENTS AT JAPAN’S HIGH TEMPERATURE ENGINEERING TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Nozomu Fujimoto; James W. Sterbentz; Luka Snoj; Atsushi Zukeran

    2011-03-01

    The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The experimental benchmark performed and currently evaluated in this report pertains to the data available for two zero-power, warm-critical measurements with the fully-loaded HTTR core. Six isothermal temperature coefficients for the fully-loaded core from approximately 340 to 740 K have also been evaluated. These experiments were performed as part of the power-up tests (References 1 and 2). Evaluation of the start-up core physics tests specific to the fully-loaded core (HTTR-GCR-RESR-001) and annular start-up core loadings (HTTR-GCR-RESR-002) have been previously evaluated.

  3. 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...

  4. 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)

  5. A new magnet for the LHC mock-up

    CERN Multimedia

    HSE Unit

    2013-01-01

    This year, the safety training centre on the Prévessin site acquired a mock-up of the LHC, which simulates the work and safety conditions in the tunnel.   Photo: Christoph Balle. A new dummy quadrupole has just been added to the magnet chain, making the mock-up even more realistic. The new facility, which was a joint endeavour by the TE, GS, BE and EN Departments, will significantly improve the quality of the various training courses held at the centre, particularly the course on the use of self-rescue masks. To consult the safety training catalogue and/or sign up for radiation protection training, please go to: https://cta.cern.ch. For further information, please contact the Safety Training and Awareness service by telephone on 73811 or 79935 or by e-mail to safety-training@cern.ch.  

  6. Mock-up test of the high level liquid waste solidification rpcess (1st campaign)

    International Nuclear Information System (INIS)

    In the Power Reator and Nuclear Fuel Development Corporation (PNC), plans are in progress for the completion of a pilot plant for the glass solidification of liquid-wastes in fiscal 1987, intended for the high-level liquid wastes from the fuel reprocessing plant of PNC. The mock-up test facility for this pilot plant is for grasping the operating characteristics of the process and the development of the remote operation and maintenance techniques. The test facility is composed of the stages of liquid waste pretreatment, glass material feed, glass melting, off-gas treatment, canister handling, and secondary-waste treatment. The following matters are described: mock-up test building, process constitution, machinery arrangement, operation control, test plans, and operation test results (1st campaign: pretreatment/off-gas treatment, and glass melting). (Mori, K.)

  7. China Overseas Plaza Mock-up Floor Grand Openning

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    @@ Global Leasing launched on Full Scale Beijing, November 6, 2008- China Overseas Plaza, located at the heart of the Beijing CBD with Chang'an Avenue to the south and China World Trade Center PhasenⅢto the east, is constructed including two international Grade-A office towers with a commercial podium. The developer held a grand opening ceremony on the completed mock-up floor the sixth floor of the South Tower.

  8. Evaluation on sweep gas pressure drop in fusion blanket mock-up for in-pile test

    International Nuclear Information System (INIS)

    In the ITER/CDA (Conceptual Design Activity) of a tritium breeding blanket, Japan have proposed the pebble-typed blanket. The in-pile mock-up test will be preparing in JMTR (Japan Materials Testing Reactor) for Japanese engineering design with the pebble-typed blanket. Therefore, the He sweep gas pressure drop in the pebble bed was measured for the design of the mock-up used on in-pile test. From the results of this test, it was clear that the pressure drop was predicted on Kozeny- Carman's equation within +25 ∼ -60 %, and that the pressure drop was not affected by moisture concentration (< 100 ppm). (author)

  9. A Coupled THMC model of FEBEX mock-up test

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, Liange; Samper, Javier

    2008-09-15

    FEBEX (Full-scale Engineered Barrier EXperiment) is a demonstration and research project for the engineered barrier system (EBS) of a radioactive waste repository in granite. It includes two full-scale heating and hydration tests: the in situ test performed at Grimsel (Switzerland) and a mock-up test operating at CIEMAT facilities in Madrid (Spain). The mock-up test provides valuable insight on thermal, hydrodynamic, mechanical and chemical (THMC) behavior of EBS because its hydration is controlled better than that of in situ test in which the buffer is saturated with water from the surrounding granitic rock. Here we present a coupled THMC model of the mock-up test which accounts for thermal and chemical osmosis and bentonite swelling with a state-surface approach. The THMC model reproduces measured temperature and cumulative water inflow data. It fits also relative humidity data at the outer part of the buffer, but underestimates relative humidities near the heater. Dilution due to hydration and evaporation near the heater are the main processes controlling the concentration of conservative species while surface complexation, mineral dissolution/precipitation and cation exchanges affect significantly reactive species as well. Results of sensitivity analyses to chemical processes show that pH is mostly controlled by surface complexation while dissolved cations concentrations are controlled by cation exchange reactions.

  10. Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

    International Nuclear Information System (INIS)

    The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of 232Th and 237Np, measured in GFR-like lattices. (authors)

  11. Preparation and properties of CVD-W coated W/Cu FGM mock-ups

    Energy Technology Data Exchange (ETDEWEB)

    Lian, Youyun, E-mail: lianyy@swip.ac.cn [Southwestern Institute of Physics, P.O. Box 432, 610041 Chengdu (China); Liu, Xiang; Xu, Zengyu [Southwestern Institute of Physics, P.O. Box 432, 610041 Chengdu (China); Song, Jiupeng; Yu, Yang [Xiamen Honglu Tungsten and Molybdenum Industry Co. Ltd., 361021 Xiamen (China)

    2013-10-15

    Highlights: • CVD-W coating was deposited at high deposition rate about 0.7 mm/h. • CVD-W coating has high density, purity and thermal conductivity. • Graded W/Cu composite was used as a transition layer between W coating and CuCrZr. • CVD-W mock-ups have good thermal–mechanical properties. -- Abstract: Tungsten was coated on a W/Cu functionally graded material (FGM) by chemical vapor deposition technique (CVD), and then the tungsten coated tile was brazed on the CuCrZr heat sink with a cooling channel. The thickness of CVD-W was 2 mm deposited by a fast rate of about 0.7 mm/h. The features of the CVD-W coating including morphology, element composition and thermal properties were characterized. A tungsten coating with high density, purity and thermal conductivity is achieved. The bonding strength between the CVD-W layer and FGM was measured using bonding tensile tests. Thermal screening and fatigue tests were performed on the CVD-W mock-ups under fusion relevant conditions using an electron beam device. Experimental results showed that the CVD-W mock-up failed by melting of Cu beneath the tungsten layer under a high heat load of 14.5 MW/m{sup 2} and 30 s pulse duration. Thermal fatigue tests showed that the CVD-W mock-up could sustain 1000 cycles at a heat load of 11.7 MW/m{sup 2} absorbed power density and 15 s pulse duration without visible failure.

  12. Preparation and properties of CVD-W coated W/Cu FGM mock-ups

    International Nuclear Information System (INIS)

    Highlights: • CVD-W coating was deposited at high deposition rate about 0.7 mm/h. • CVD-W coating has high density, purity and thermal conductivity. • Graded W/Cu composite was used as a transition layer between W coating and CuCrZr. • CVD-W mock-ups have good thermal–mechanical properties. -- Abstract: Tungsten was coated on a W/Cu functionally graded material (FGM) by chemical vapor deposition technique (CVD), and then the tungsten coated tile was brazed on the CuCrZr heat sink with a cooling channel. The thickness of CVD-W was 2 mm deposited by a fast rate of about 0.7 mm/h. The features of the CVD-W coating including morphology, element composition and thermal properties were characterized. A tungsten coating with high density, purity and thermal conductivity is achieved. The bonding strength between the CVD-W layer and FGM was measured using bonding tensile tests. Thermal screening and fatigue tests were performed on the CVD-W mock-ups under fusion relevant conditions using an electron beam device. Experimental results showed that the CVD-W mock-up failed by melting of Cu beneath the tungsten layer under a high heat load of 14.5 MW/m2 and 30 s pulse duration. Thermal fatigue tests showed that the CVD-W mock-up could sustain 1000 cycles at a heat load of 11.7 MW/m2 absorbed power density and 15 s pulse duration without visible failure

  13. Theoretical Work for the Fast Zero-Power Reactor FR-0

    International Nuclear Information System (INIS)

    The theoretical part of the fast reactor physics work in Sweden, has mainly been connected with the FR-0 reactor. The report describes the principal features of this reactor, evaluation of cross sections, calculations of critical masses, reactivity of the air gap and of control rods and calculations of neutron generation time and effective beta values. Carlson codes in spherical and in cylindrical geometry are used to evaluate critical masses and fluxes. In cases when reactivity changes are calculated, complementary methods are perturbation theory and variational calculus. The agreement with experiments is in some cases good, especially the determination of critical mass, but in other cases discrepancies are observed, e.g. the activation of U-238 in the reflector is much larger than the theoretical spectrum predicts

  14. Calculation and measurement of neutron flux in internal parts of the VVER-1000 mock-up

    International Nuclear Information System (INIS)

    Highlights: • Measurement of reaction rates in reactor baffle simulator. • Comparison of calculated and measured reaction rates in reactor baffle. • Measurement of fast neutron spectra in reactor baffle cooling channel. • Determination of 3He attenuation in lateral reflector model. - Abstract: The radiation situation in the reactor’s internal stainless steel parts is an important parameter during reactor operation. They are the most radiation-stressed structures because they are very close to the fuel. Knowledge of neutron flux distribution is important both for estimation of radiation-induced swelling of internal parts, radiation heating, and internal part activation. This paper aims to compare the experimental and calculation reaction rate distribution in a VVER-1000 mock-up placed in reactor LR-0

  15. Reanalysis of the Gas-cooled fast reactor experiments at the zero power facility Proteus - Spectral indices

    Science.gov (United States)

    Perret, G.; Pattupara, R. M.; Girardin, G.; Chawla, R.

    2013-03-01

    PROTEUS is a zero power reactor at the Paul Scherrer Institute which has been employed during the 1970's to study experimentally the physics of the gas-cooled fast reactor. Reaction rate distributions, flux spectrum and reactivity effects have been measured in several configurations featuring PuO2/UO2 fuel, absorbers, large iron shields, and thorium oxide and thorium metal fuel either distributed quasihomogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. This papers focus on the spectral indices - including fission and capture in 232Th and 237Np - measured in the reference PuO2/UO2 lattices and their predictions with an MCNPX model specially developed for the PROTEUS-GCFR core. Predictions were obtained with JEFF-3.1 and -3.11, ENDF/B-VII.0 and VII.1, and JENDL-3.3 and -4.0. A general good agreement was demonstrated. The ratio of 232Th fission to 239Pu fission, however, was under-predicted by 8.7±2.1% and 6.5±2.1% using ENDF/B-VII.0 and VII.1, respectively. Finally, the capture rates in 237Np tended to be underpredicted by the JEFF and JENDL libraries, although the new cross section in JEFF-3.1.1 slightly improved the 237Np capture to 239Pu fission results (3.4±2.4%).

  16. Benchmark Evaluation of Start-Up and Zero-Power Measurements at the High-Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Benchmark models were developed to evaluate six cold-critical and two warm-critical, zero-power measurements of the HTTR. Additional measurements of a fully-loaded subcritical configuration, core excess reactivity, shutdown margins, six isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as acceptable benchmark experiments. Insufficient information is publicly available to develop finely-detailed models of the HTTR as much of the design information is still proprietary. However, the uncertainties in the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias uncertainties incurred through the simplification process used to develop the benchmark models. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the impurity content of the various graphite blocks that comprise the HTTR. Monte Carlo calculations of keff are between approximately 0.9 % and 2.7 % greater than the benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulation of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments

  17. The Texts of the Instruments relating to a Project for a Joint Agency-Norwegian Program of Research with the Zero Power Reactor 'NORA'

    International Nuclear Information System (INIS)

    The text of the Supply Agreement between the Agency and the Governments of Norway and of the United States of America, and the text of the related Project Agreement between the Agency and the Government of Norway concerning an Agency project for cooperation in carrying out a joint program of research in reactor physics with the zero power reactor 'NORA', are reproduced in this document for the information of all Members of the Agency

  18. 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.)

  19. 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.

  20. Flow test for the full scale core mock-up to the KUHFR, (2)

    International Nuclear Information System (INIS)

    The Research Reactor Institute, Kyoto University, has carried out a variety of research and development in support of the high flux reactor (KUHFR) project. As for the thermal-hydraulic design of the reactor core, the flow test with a full scale mock-up of the core was performed in order to verify the design calculation. This report shows the result of measurement of the vibration of the core vessel and core itself obtained during the flow test. The flow rate through the core mock-up reached up to 1920 m3/h, which is approximately 1.3 times as much as the normal flow rate. Non-contact displacement sensors and piezoelectric accelerometers were used to measure the vibration of the core vessel, core components and outer fuel elements. The traces of the vibration were reproduced on charts to read the maximum amplitude. The data were analyzed by FFT method to find the characteristics of the vibration. The observations of the corrosion and deformation of the components were made. The results obtained are as follows. The vibration of the core vessel was excited by coolant flow. The predominant frequency was about 7 Hz, which is nearly equal to that of the free vibration of the core vessel. The maximum displacement was 300 mu m, and the maximum acceleration was 1.8 g. (Kako, I.)

  1. Fabrication of an instrumented fuel rod mock-up using a precise drilling machine

    International Nuclear Information System (INIS)

    When a new nuclear fuel is developed, Irradiation test needs to carried out in the research reactor to analyze the performance of the new nuclear fuel. In addition, to check the performance of the nuclear fuel during the burn up test in the test loop, it is necessary to attach sensors near the fuel rod and connect them with instrumentation cables to the measuring device located outside of the reactor pool. In particular, to check the temperature fluctuation of a nuclear fuel during the irradiation test, a thermocouple should be instrumented at the center of the nuclear fuel. Therefore, A hole needs to be made at the center of a fuel pellet to put in the thermocouple. However, because the hardness and density of a sintered UO2 pellet are very high, it is difficult to make a small fine hole on the sintered UO2 pellet with a simple drilling machine. In this study, an instrumented fuel rod mock-up was fabricated using an automated precise drilling machine. Four sintered alumina were drilled off and assembled into the zircaloy tube and a K-type thermocouple was instrumented in the fuel rod mock-up

  2. Reanalysis of the Gas-cooled fast reactor experiments at the zero power facility Proteus – Spectral indices

    Directory of Open Access Journals (Sweden)

    Girardin G.

    2013-03-01

    Full Text Available PROTEUS is a zero power reactor at the Paul Scherrer Institute which has been employed during the 1970’s to study experimentally the physics of the gas-cooled fast reactor. Reaction rate distributions, flux spectrum and reactivity effects have been measured in several configurations featuring PuO2/UO2 fuel, absorbers, large iron shields, and thorium oxide and thorium metal fuel either distributed quasihomogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. This papers focus on the spectral indices – including fission and capture in 232Th and 237Np - measured in the reference PuO2/UO2 lattices and their predictions with an MCNPX model specially developed for the PROTEUS-GCFR core. Predictions were obtained with JEFF-3.1 and -3.11, ENDF/B-VII.0 and VII.1, and JENDL-3.3 and -4.0. A general good agreement was demonstrated. The ratio of 232Th fission to 239Pu fission, however, was under-predicted by 8.7±2.1% and 6.5±2.1% using ENDF/B-VII.0 and VII.1, respectively. Finally, the capture rates in 237Np tended to be underpredicted by the JEFF and JENDL libraries, although the new cross section in JEFF-3.1.1 slightly improved the 237Np capture to 239Pu fission results (3.4±2.4%.

  3. Reanalysis of the gas-cooled fast reactor experiments at the zero power facility Proteus - Spectral indices

    International Nuclear Information System (INIS)

    PROTEUS is a zero power reactor at the Paul Scherrer Institute which has been employed during the 1970's to study experimentally the physics of the gas-cooled fast reactor. Reaction rate distributions, flux spectrum and reactivity effects have been measured in several configurations featuring PuO2/UO2 fuel, absorbers, large iron shields, and thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. This papers focuses on the spectral indices - including fission and capture in 232Th and 237Np - measured in the reference PuO2/UO2 lattices and their predictions with an MCNPX model specially developed for the PROTEUS-GCFR core. Predictions were obtained with JEFF-3.1 and -3.11, ENDF/B-VII.0 and VII.1, and JENDL-3.3 and -4.0. A general good agreement was demonstrated. The ratio of 232Th fission to 239Pu fission, however, was under-predicted by (8.7±2.1)% and (6.5±2.1)% using ENDF/B-VII.0 and VII.1, respectively. Finally, the capture rates in 237Np tended to be under-predicted by the JEFF and JENDL libraries, although the new cross section in JEFF-3.1.1 slightly improved the 237Np capture to 239Pu fission results (3.4±2.4)%. (authors)

  4. Pressure tests of two KBS-3 canister mock-ups

    International Nuclear Information System (INIS)

    The Swedish concept for geological disposal of spent nuclear fuel, the so-called KBS-3 concept, relies on a multibarrier system with the copper/cast iron canister as the first barrier. The canister is designed to retain its integrity for at least 100,000 years, which means that future glaciations need to be considered. A 3 km thick ice block together with hydrostatic pressure from groundwater and swelling of the buffer material would produce hydrostatic compressive stresses of maximum 44 MPa (440 bar). Although the canister is loaded globally in compression, tensile stresses develop at fuel channel surface with increasing load. Tensile tests of the insert material in the development phase of the KBS-3 canister indicated a large scatter and relatively low values of the inserts' ductility. An important issue was whether this could lead to mechanical failure of canisters at the 44 MPa iso-static load either by plastic collapse or fracture from the defects in the regions with tensile stresses. SKB therefore initiated a project together with the European commission's Joint Research Centre (JRC) Institute of Energy in Petten and a number of Swedish partners to evaluate the probability of mechanical failure during glaciation. Three inserts manufactured by different Swedish foundries and referred to as 1, 125 and 126 were used in the project. A large statistical test programme was developed to determine statistical distributions of various material parameters and defect distributions. These data were subsequently used in probabilistic analysis to determine the probability for local plastic collapse or fracture. The main conclusion was that the failure probability is extremely low at the design load (44 MPa) provided some basic geometrical requirements are fulfilled. In parallel to the statistical test programme and the associated analysis, the group decided also to perform two pressure tests of canister mock-ups to demonstrate the actual safety margins. The fractographic

  5. Engineering and manufacturing of ITER first mirror mock-ups

    International Nuclear Information System (INIS)

    Most of the ITER optical diagnostics aiming at viewing and monitoring plasma facing components will use in-vessel metallic mirrors. These mirrors will be exposed to a severe plasma environment and lead to an important tradeoff on their design and manufacturing. As a consequence, investigations are carried out on diagnostic mirrors toward the development of optimal and reliable solutions. The goals are to assess the manufacturing feasibility of the mirror coatings, evaluate the manufacturing capability and associated performances for the mirrors cooling and polishing, and finally determine the costs and delivery time of the first prototypes with a diameter of 200 and 500 mm. Three kinds of ITER candidate mock-ups are being designed and manufactured: rhodium films on stainless steel substrate, molybdenum on TZM substrate, and silver films on stainless steel substrate. The status of the project is presented in this paper.

  6. Mock-up tests for a hot vitrification system

    International Nuclear Information System (INIS)

    A vitrification system for high-level wastes has been designed to prepare actural glass products used for the safety evaluation. For the final design making, mock-up tests of the system were performed. Heating ability and temperature location, prevention method of blocking phenomenum in the melter, homogenization of the glass products, mechanics of freeze valve, durability of used materials, and remote handling of the system were tested on the engineering points. The prevention of blocking by the heating of the upper zone, the homogenization of the glass products by the bubbling with N2 gas and its caused volatilization of the waste components, and new mechanics of freeze valve were found as the results. (author)

  7. Advanced smile diagnostics using CAD/CAM mock-ups.

    Science.gov (United States)

    Sancho-Puchades, Manuel; Fehmer, Vincent; Hämmerle, Christoph; Sailer, Irena

    2015-01-01

    Diagnostics are essential for predictable restorative dentistry. Both patient and clinician must agree on a treatment goal before the final restorations are delivered to avoid future disappointments. However, fully understanding the patient's desires is difficult. A useful tool to overcome this problem is the diagnostic wax-up and mock-up. A potential treatment outcome is modeled in wax prior to treatment and transferred into the patient's mouth using silicon indexes and autopolymerizing resin to obtain the patient's approval. Yet, this time-consuming procedure only produces a single version of the possible treatment outcome, which can be unsatisfactory for both the patient and the restorative team. Contemporary digital technologies may provide advantageous features to aid in this diagnostic treatment step. This article reviews opportunities digital technologies offer in the diagnostic phase, and presents clinical cases to illustrate the procedures. PMID:26171442

  8. 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.

  9. Results of the mock-up experiment on partial LOCA

    International Nuclear Information System (INIS)

    A mockup experiment has been performed to verify the heat transfer model for a partial loss of coolant accident in the swimming pool reactor SAPHIR. Three coolant channels with the same dimensions as in a SAPHIR fuel element were simulated using four electrically heated plates. For a water level such that the heated plates are partially submerged, plate temperatures remain below 160 deg. C for plate powers of up to 650 W. For water levels low enough to just block the channels, plate temperatures of 400 deg. C are reached for plate powers as low as 60 W. Details of the experiment and further results are discussed. (author)

  10. Safety evaluation report related to the renewal of the operating license for the Zero-Power Reactor at Cornell University, Docket No. 50-97

    International Nuclear Information System (INIS)

    This Safety Evaluation Report for the application filed by Cornell University (CU) for a renewal of Operating License R-80 to continue to operate a zero-power reactor (ZPR) has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by Cornell University and is located on the Cornell campus in Ithaca, New York. The staff concludes that the ZPR facility can continue to be operated by CU without endangering the health and safety of the public

  11. 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)

  12. Integral experiments for verification of tritium production on the beryllium/lithium titanate blanket mock-up with a one-breeder layer

    International Nuclear Information System (INIS)

    The first series of integral experiments on the blanket mock-up with a one breeder layer was performed in support of the concept of the solid breeding blanket cooled with water, proposed by JAERI for application in the DEMO reactor. The mock-up for the first series of experiments was designed to be as simple as possible within the proposed blanket concept. Key objectives of the experiments were: to check how correctly the tritium production rate can be predicted in the breeder layer closest to the first wall, since this particular location is greatly affected by changes of incoming neutron spectra; to validate the modified experimental techniques for measurements of tritium production rate in conditions of quick gradient thermal neutron field inside the lithium titanate layer. The mock-up contains F82H, lithium titanate and beryllium layers, with respective thicknesses of 16 mm, 12 mm and 203 mm. An additional tungsten layer was installed in front of the first layer in order to simulate armor material. The mock-up, being placed inside the SS316 cylindrical enclosure, is shaped as a pseudo-cylindrical slab with an area-equivalent diameter of 628 mm. Integral experiments on the blanket mock-up irradiated by neutrons from the D-T source with and without the source reflector were executed. A detailed description of experimental results and an example of calculation analysis are presented. (author)

  13. Characterization of flaws in a tube bundle mock-up for reliability studies

    Energy Technology Data Exchange (ETDEWEB)

    Kupperman, D.S.; Bakhtiari, S. [Argonne National Lab., IL (United States)

    1997-02-01

    As part of an assessment of in-service inspection of steam generator tubes, the authors will assemble a steam generator mock-up for round robin studies and use as a test bed in evaluating emerging technologies. Progress is reported on the characterization of flaws that will be part of the mock-up. Eddy current and ultrasonic techniques are being evaluated as a means to characterize the flaws in the mock-up tubes before final assembly. Twenty Inconel 600 tubes with laboratory-grown cracks, typical of those to be used in the mock-up, were provided by Pacific Northwest National Laboratory for laboratory testing. After the tubes were inspected with eddy current and ultrasonic techniques, they were destructively analyzed to establish the actual depths, lengths, and profiles of the cracks. The analysis of the results will allow the best techniques to be used for characterizing the flaws in the mock-up tubes.

  14. Some particular aspects of control in nuclear power reactors

    International Nuclear Information System (INIS)

    This paper reviews the experience acquired in France on the question, of reactor safety. Since a special paper is being presented on reactors of the graphite gas type, the safety of the other types studied in France is discussed here: - heavy water-gas reactors, - fast neutron reactors, - water research reactors of the swimming-pool and tank types. The safety rules peculiar to the different types are explained, with emphasis on their influence on the reactor designs and on the power limits they impose. The corresponding safety studies are presented, particular stress being placed on the original work developed in these fields. Special mention is made of the experimental systems constructed for these studies: the reactor CABRI, pile loop for depressurization tests, loops outside the pile, mock-ups etc. (authors)

  15. The Design Concept of a Steam Generator Cassette Mock-Up for ISI of Helical Tubes in SMART Steam Generator

    International Nuclear Information System (INIS)

    The SMART reactor steam generator is composed of 8 Steam Generator Cassettes (SGC) and each the SGC has a once-through-type, helical-coil-tube bundle structure using INCONEL alloy 690 tubes. The SGC installed in reactor vessel is a kind of heat exchanger made of INCONEL alloy 690 tubes. This paper introduces the design concepts of an SGC mock-up for the test probe insertion ability of In- Service Inspection (ISI). The backgrounds of selected tube material, size and tube composition are described

  16. The Design Concept of a Steam Generator Cassette Mock-Up for ISI of Helical Tubes in SMART Steam Generator

    Energy Technology Data Exchange (ETDEWEB)

    Choung, Yun Hang; Kim, Dong Ok; Park, Jin Seok; Lee, Won Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    The SMART reactor steam generator is composed of 8 Steam Generator Cassettes (SGC) and each the SGC has a once-through-type, helical-coil-tube bundle structure using INCONEL alloy 690 tubes. The SGC installed in reactor vessel is a kind of heat exchanger made of INCONEL alloy 690 tubes. This paper introduces the design concepts of an SGC mock-up for the test probe insertion ability of In- Service Inspection (ISI). The backgrounds of selected tube material, size and tube composition are described.

  17. Residual stress measurement inside a dissimilar metal weld mock-up of the pressurizer safety and relief nozzle

    International Nuclear Information System (INIS)

    Residual stresses are present in materials or structural component in the absence of external loads or changes in temperatures. The most common causes of residual stresses being present are the manufacturing or assembling processes. All manufacturing processes, such as casting, welding, machining, molding, heat treatment, among others, introduces residual stresses into the manufactured object. The residual stresses effects could be beneficial or detrimental, depending on its distribution related to the component or structure, its load service and if it is compressive or tensile. In this work, the residual strains and stresses inside a mock-up that simulates the safety and relief nozzle of Angra 1 Nuclear Power Plant pressurizer were studied. The current paper presents a blind hole-drilling method residual stress measurements both at the inner surface of dissimilar metal welds of dissimilar metal weld nozzle mock-up. (author)

  18. Determination of the decay constants and relative abundances of delayed neutrons by noise analysis in zero-power reactors

    International Nuclear Information System (INIS)

    A reactor noise approach has been employed at the IPEN/MB-01 research reactor facility in order to determine experimentally the effective delayed neutron parameters βi and λi in a six group model and assuming the point reactor. The method can be considered a novice one because exploits the very low frequency domain of the spectral densities. The proposed method has some advantages to other in-pile methods since it does not disturb the reactor system and consequently does not 'excite' any sort of harmonic modes. As a byproduct and a consistency check, the βeff parameter was obtained without the need of the Diven factor and the power normalization and it is in excellent agreement with independent measurements. The theory/experiment comparison shows that for the abundances the JENDL 3.3 presents the best performance while for the decay constants the revised version of ENDF/B-VI.8 shows the best agreement. The best performance for the βeff determination is obtained with JENDL3.3. In contrast, ENDF/B-VI.8 and its revised version performed at LANL overestimate βeff by as much as 4%. The βeff results of this work support totally the proposal of reducing the thermal delayed neutron number for 235U fission as made by Sakurai and Okajima. A new observed effect related to the correlation between the fluctuations of both measurement channels is also presented and discussed. This effect can be considered as an indirect evidence for the use of the point reactor model in this work as well as a possible useful tool in the understanding of reactor dynamics. (author)

  19. 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

  20. Status report about the works for the start up of the RA-0 'zero power' nuclear reactor at the Cordoba National University

    International Nuclear Information System (INIS)

    After two years of works at the Cordoba National University for the new start-up of the RA-0 'zero power' nuclear reactor, the results obtained are herein presented. Starting with practically null infrastructure at the beginning, specially in human resources and instrumentation of the reactor, the objectives can be considered satisfactory. The training in work of the future operational staff, the design and the construction of the instrumentation and the fitting of the installations are the principal items described in this paper. An special attention is devoted to the insertion of this type of installation in the university organization, usually not prepared for the quality and control activities, which is necessarily considered in these type of works. (Author)

  1. 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

  2. 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)

  3. 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)

  4. 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)

  5. Preparing ITER ICRF: Test of the load resilient matching systems on an antenna mock-up

    International Nuclear Information System (INIS)

    The ICRF system of ITER must couple 20 MW to the plasma in the frequency band 40 - 55 MHz through a surface of ∼ 1.5 m x 1.9 m, with a wave spectrum appropriate for both heating and current drive. Its matching system must furthermore be resilient to the large load variations that are typically for ELMy discharges. A compact ICRH antenna concept where 24 radiating straps are grouped in 8 triplets by passive junctions has been selected. In view of its complexity the matching systems are studied and tested on a scaled down mock-up. When decreasing the lengths and increasing the frequency by the same scale factor the impedance matrix of the array remains identical realistic simulation of plasma-like load conditions can be obtained by facing the strap array by means of a large dielectric constant medium, such as water. The two selected matching network options to provide the needed load resilience are (i) 4 'conjugate T' (CT) circuits, or (ii) 4 hybrid junctions with reflected power dumped in dummy loads. Two challenges have to be faced hereby: (i) Due to the compactness of the array the mutual coupling effects between the radiating triplets are important and lead to a coupling between all matching actuators and power sources and to an asymmetry in the radiated power distribution. (ii) The low range of antenna loading resistance to be expected in ITER because of the large antenna-LCMS distance renders the adjustment of the matching circuits very critical and amplifies the mutual coupling effects. From the present study the following results and conclusions can be drawn: (i) Decouplers neutralizing the dominant coupling terms of the input admittance matrix are mandatory; (ii) Both matching options, if well adjusted, have good performances. The advantages of the hybrid option are its potential resilience for any value of the coupling resistance and its larger insensitivity to reactive load changes. The load resilience domain for the CT option becomes small if the mean

  6. The VVER Core Physics, Reactor Dosimetry, and Shielding Researches in the LR-0 Reactor

    International Nuclear Information System (INIS)

    Zero-power water reactor LR-0 was created by the Nuclear Research Institute Rez, Nuclear Machinery Skoda, and RRC 'Kurchatov Institute' for researches of neutron parameters of the WWER type power reactors core, fuel storages, and-first of all-for researches in the reactor pressure vessel and internals dosimetry. Suitable geometrical conditions and flexible technical arrangements of the LR-0 facility enabled to carry out the wide experimental program on several full-scale models (mock-ups) of the WWER-440 and WWER-1000 reactors. The tasks of that experiments were the measurements of the neutron (from thermal energy up to 10 MeV) and gamma (from 0.1 up to 10 MeV) spectra and integral parameters of neutron and gamma fields in the different representative points of the mock-ups from the core to the outer pressure vessel surface and the biological shielding (including channel for ex-reactor ionizing chamber), as well as the measurement of spatial power distribution in the core. Fast neutron (energy from 0.5 to 10 MeV) and gamma spectra were measured in several representative points of the mock-ups by the two-parameter spectrometer with the cylindrical stilbene scintillation detectors. Measurements in the thermal and epithermal neutron region were carried out with the activation method using a broad set of activation monitors and with the 3He(n,p) counter. Activation measurements with threshold fast neutron detectors enlarge also the proton-recoil spectra measurements, such activation measurements were carried out especially in cases, when a spectrometer couldn't be put in the necessary position. The core fission rate distribution was obtained by means of gamma-scanning of the fuel pins. The calculations were carried out by different methods (deterministic and Monte Carlo). Experimental and calculation results in the core, internals, pressure vessel and shielding are reviewed and compared. (Authors)

  7. Report on material tests and mock-up tests for fabrication of the KUHFR reflector tank

    International Nuclear Information System (INIS)

    The Kyoto University high flux reactor (KUHFR) is designed to have a pair of cylindrical core vessels made of aluminum A 6061 and a spherical reflector tank made of stainless steel SUS 316 L containing the core vessels. The cooling and moderating light water flows downward. The reflector tank is filled with heavy water which is a good neutron moderator to offer a proper thermal neutron field for various irradiation and beam experiments. Many kinds of the test on SUS 316 L specimens treated under various conditions were carried out, such as metallographic microstructure observation, mechanical test and corrosion test. The results are summarized as follows. Most of the specimens showed good microstructure and excellent Huey corrosion resistance. The zone with reduced Cr at grain boundary was not observed in the specimens treated above 920 deg C. The specimens prepared after a commercial plate was treated according to the normal fabrication procedure showed good mechanical properties. A mock-up of actual size with typical nozzles was made of SUS 316 L plates and pipes. After the actual heat treatment, the change of the size and shape, residual stress, microstructure and corrosion resistance were examined to establish the fabrication procedure and heat treatment. (Kako, I.)

  8. Set-up of a pre-test mock-up experiment in preparation for the HCPB Breeder Unit mock-up experimental campaign

    International Nuclear Information System (INIS)

    Highlights: ► As preparation for the HCPB-TBM Breeder Unit out-of-pile testing campaign, a pre-test experiment (PREMUX) has been prepared and described. ► A new heater system based on a wire heater matrix has been developed for imitating the neutronic volumetric heating and it is compared with the conventional plate heaters. ► The test section is described and preliminary thermal results with the available models are presented and are to be benchmarked with PREMUX. ► The PREMUX integration in the air cooling loop L-STAR/LL in the Karlsruhe Institute for Technology is shown and future steps are discussed. -- Abstract: The complexity of the experimental set-up for testing a full-scaled Breeder Unit (BU) mock-up for the European Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) has motivated to build a pre-test mock-up experiment (PREMUX) consisting of a slice of the BU in the Li4SiO4 region. This pre-test aims at verifying the feasibility of the methods to be used for the subsequent testing of the full-scaled BU mock-up. Key parameters needed for the modeling of the breeder material is also to be determined by the Hot Wire Method (HWM). The modeling tools for the thermo-mechanics of the pebble beds and for the mock-up structure are to be calibrated and validated as well. This paper presents the setting-up of PREMUX in the L-STAR/LL facility at the Karlsruhe Institute of Technology. A key requirement of the experiments is to mimic the neutronic volumetric heating. A new heater concept is discussed and compared to several conventional heater configurations with respect to the estimated temperature distribution in the pebble beds. The design and integration of the thermocouple system in the heater matrix and pebble beds is also described, as well as other key aspects of the mock-up (dimensions, layout, cooling system, purge gas line, boundary conditions and integration in the test facility). The adequacy of these methods for the full-scaled BU mock-up is

  9. Eddy-Current analysis round robin using the NRC steam generator mock-up

    International Nuclear Information System (INIS)

    This paper discusses round robin exercises to assess inspection reliability using the NRC steam generator (SG) mock-up at Argonne National Laboratory. The purpose of the round robins is to assess the current reliability of SG tubing in-service inspection, determine the probability of detection (POD) as a function of flaw size or severity, and assess the capability for sizing of flaws. The mock-up contains hundreds of cracks and simulations of artifacts, such as corrosion deposits and tube support plates (TSPs), that make detection and characterization of cracks more difficult in operating steam generators than in most laboratory situations. Eddy current signals from the laboratory-grown cracks used in the mock-up have been reviewed to ensure that they provide reasonable simulations of those obtained in the field. The mock-up contains 400 tube openings. Each tube contains nine 22.2-mm (7/8-in.) diameter, 30.5-cm (1-ft) long Alloy 600 test sections. The flaws are located in the tube sheet near the roll transition zone, in the TSP and in the freespan. The flaws are primarily intergranular stress corrosion cracks (axial and circumferential, ID and OD). In addition to the simulated tube sheet and TSP the mock-up contains simulated sludge and magnetite deposits. A validated multiparameter eddy current algorithm, which provided a detailed isometric plot for every flaw, was used to establish the reference state of defects in the mock-up. The detection results of the 11 teams were used to develop POD curves as a function of maximum depth, voltage, and the parameter mp for the various types of flaws. The 95% one-sided confidence limits which include errors in maximum depth estimates, are presented along with the POD curves. For the second round robin a reconfigured mock-up is being used to evaluate the effectiveness of eddy current arrays. (author)

  10. 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.

  11. Tests of load resilient matching procedure for the ITER ICRH system on a mock-up and layout proposal

    International Nuclear Information System (INIS)

    An external matching option of the ITER ICRH system has been successfully tested by means of a mock-up of the antenna plug loaded by a movable water load. The problem of coupling between the straps of the antenna array is solved by the use of decouplers and passive power distribution. High load resilience is obtained for different toroidal phasings needed for heating or current drive. A straightforward matching procedure for which the decouplers and half of the matching parameters are pre-adjusted in vacuum condition has been developed

  12. MASURCA, a Fast-Neutron Critical Mock-Up: Operation and Uses

    International Nuclear Information System (INIS)

    Under the EURATOMCEA Association project a fast-neutron critical mock-up, Masurca, is now being built at the Cadarache Nuclear Research Centre. The main purpose of this extremely versatile facility is the study of non-moderated, plutonium critical assemblies of large volume and hence having a relatively soft neutron spectrum. The paper explains what these studies are for. The facility must satisfy certain conditions and, in essence, combine great versatility with almost absolute operational safety. The safety problem was dealt with by: (1) Seeking inherent safety: with simulated fuel elements it was possible to obtain (a) a negative reactivity coefficient from the cumulative longitudinal expansion of these elements: (b) a negative Doppler coefficient; (2) Using a set of shim-safety rods which can be placed in a square lattice with spacings of about 30cm; (3) A pressure vessel, containing reserves of argon in case of fire: and (4) Strict administrative supervision. A U-Pu-Fe metallic alloy being chosen as the basic element in the fuel simulation, provision for cooling large-volume critical assemblies must be incorporated in the facility. Sodium, the coolant used in simulated reactors, will be represented by sodium strips clad in stainless steel. The facility is designed as a vertical single-block unit in view of the maximum volume of the cores to be simulated (about 5000 1). The simulated elements are shaped like a right prism with a square base (except in the case of fuel elements which have a circular base) with an outer side (or diameter) of 12.7 mm and a height of 102 mm. They are placed in tubes having an over- all length of about 4 m and square sections whose outer side is 10.6 mm. These tubes are placed side by side and suspended. Smaller tubes can be placed in the central area of the suspension plate so that smaller cores can be made. A special heating loop can also be placed in the central part of the facility to measure the Doppler coefficient. The paper

  13. Mock-up experiment of radiation streaming through coolant pipe penetration

    International Nuclear Information System (INIS)

    The pressure vessel of Mutsu is enclosed by a primary shield consisting of a laminated iron-water shield padded with lead (lower primary shield), atop which is placed concrete shields (middle and upper primary shield). Between the primary shield and the pressure vessel, there is an air gap through which coolant piping of 23.1 cm outside diameter padded with thermal insulation penetrates from the pressure vessel into and through the concrete shield. The radiation from the reactor core streams into the air gap, and then into the coolant pipe. A mock-up experiment to examine the radiation attenuation along the coolant pipe penetration through the pressure vessel wall and concrete shield was performed in the JRR-4 swimming pool reactor. The radiation was measured along the air gap and along the coolant pipe. Neutron measurements were made in terms of the reaction rates of 58Ni(n,p) 58Co and 197Au(n,γ)198Au. Lithium fluoride thermoluminescence dosimeters were also used to obtain neutron flux and gamma-ray dose rates. For analyzing the experimental results, radiation transport calculations were performed by means of the S/sub n/ code TWOTRAN. The annular air gap around the pressure vessel was treated in two dimensions. The resulting angular flux was adapted to calculations covering the coolant pipe, through conversion of coordinates accompanied to bootstrap treatment. Comparisons made between calculated results and experimental data indicated that the preset data and design method are adequate for estimating the radiation leakage with satisfactory accuracy

  14. SCC behavior of alloy 690 from a CDRM mock-up

    International Nuclear Information System (INIS)

    Stress corrosion cracking (SCC) response of Alloy 690 when the material has been subjected to nonuniform cold working is of interest to understand the behavior of the weld heat affected zone (HAZ) of Alloy 690 in which localised plastic strain exists due to weld shrinkage. This has a special interest in the case of control-rod-drive mechanisms (CRDM) of vessel head. To simulate these conditions during last years many crack growth rate (CGR) data were obtained in deformed material by cold work (rolling, forging or tensile straining), up to 40% of cold working. However, it is unclear to what extent this simulation procedure reproduces the conditions of the material in a CRDM. A research project is being carried out in order to obtain CGR data in realistic situations existing in operating power plants, by the use of CT specimens extracted from CRDMs. This presentation shows the characterization and some results of crack growth rate data on Alloy 690 TT base metal/HAZ/weld metal using specimens made from a CRDM mock-up. It has been fabricated following the usual procedures used for the RPV head fabrication for the Spanish PWR NPP. (authors)

  15. Tests of load resilient matching procedures for the ITER ICRH system on a mock-up and layout proposals

    International Nuclear Information System (INIS)

    The ICRH antenna of ITER consists of an array of 24 radiating straps and must radiate 20 MW with resilience to load variations due to the ELMs. Because of its compactness the mutual coupling effects between the straps are far from negligible. Moreover they considerably increase the difficulty of matching and lead to coupling between the generators. Different external matching system layouts are under consideration. A reduced scale (1/5) mock-up loaded by a movable water tank is used for their experimental investigation. A first layout using full passive power distribution among the straps and a single matching circuit with one '' Conjugate-T '' (CT) or one hybrid has already been successfully tested. Its drawbacks are the difficulty of changing the toroidal phasing and the use of a single 20 MW feeding line section. In this paper we describe the mock-up tests of a second layout based on two 10 MW CT circuits, and allowing switching between heating or current drive phasings without any hardware modification. Two decouplers are used to minimize the effect of mutual coupling on matching. A robust four-parameter CT matching procedure has been developed based on adjusting the two first parameters - the positions of the line stretchers in the CT branches - of each CT in vacuum conditions (this is done once for all for each frequency). High load resilience, i.e. a VSWR remaining < 1.5 for an 8-fold increase of antenna resistance, can be obtained for the 4 toroidal phasing configurations considered: (0π/2π3π/2), (0-π/2-π-3π/2), (00ππ) and (0ππ0). The change of phasing only requires the adjustment of the phase difference between the two power sources and of the two last parameters (stub and line stretcher in the common line) of each of the two CT circuits. These properties have first been derived from the experimental scattering matrix of the antenna array and are verified by reflection measurements on the mock-up. Feedback control of the phasing and the last two

  16. Neutronics experiments on HCPB and HCLL TBM mock-ups in preparation of nuclear measurements in ITER

    International Nuclear Information System (INIS)

    In support of the breeder blanket development program, the EU is conducting a dedicated neutronics R and D effort to provide the basis for the design of nuclear tests to be performed in ITER on the Test Blanket Modules (TBMs). It includes the development of computational tools comprising both Monte-Carlo and deterministic transport, sensitivity and uncertainty codes, the generation of high quality neutron cross-section and covariance data libraries. These are validated experimentally in view of their application in the ITER TBM and the DEMO design. To this purpose, two neutronics experiments have been carried out on mock-ups of both the Helium Cooled Pebble Bed (HCPB) and the Helium Cooled Lithium Lead (HCLL) variants of ITER TBMs, at 14-MeV neutron sources. Redundant experimental techniques have been used to measure the resulting tritium production rate and the neutron and gamma ray spectra which are needed to predict the blanket shielding performance, nuclear power production and all nuclear loads. The comparison of experiment and corresponding calculation is obtained with the associated uncertainty margin based on experimental as well as calculational uncertainties. At the same time, suitable nuclear measuring techniques for TBMs in ITER, in particular for the tritium production, are being developed, optimised, and tested in the mock-up experiments. In particular, the present paper summarizes the preliminary results of the latest of such experiments, i.e. the one conducted on a mock-up of the HCLL blanket. It describes also the dedicated neutronics R and D activities, including the efforts to specify the neutronics tests and objectives in TBMs in ITER, considering the various design concepts.

  17. Synthesis and characterisation of SiCf/Cu matrix composites and their application in a divertor flat-tile mock-up

    International Nuclear Information System (INIS)

    In future fusion reactors materials are operating under extreme conditions. The fusion plasma leads to heat fluxes to the plasma facing materials (PFM) of 15-20 MW/m2 in the divertor region. To increase the thermal efficiency of future fusion power plants like DEMO, a higher coolant temperature of 300 C compared to 150 C at ITER is necessary which leads to an increased temperature of 550 C at the interface between tungsten as PFM and CuCrZr as heat sink material. This will cause high stresses as a result of the temperature gradient and the different coefficients of thermal expansion (CTE). Due to insufficient mechanical properties of the precipitation-hardened CuCrZr at this temperature, SiCf/Cu composites are considered to strengthen the critical zone with an interlayer between the PFM and the heat sink, as they combine high mechanical strength and a good thermal conductivity. The aim of this investigation is the preparation and mechanical as well as thermal characterisation (in particular the mechanical strength and thermal conductivity) of SiCf/Cu composites, and in addition, the manufacture of a metal matrix composite (MMC), as well as the assembly of a flat-tile mock-up to investigate the performance of an MMC interlayer under heat loads. A new method was developed to synthesise an appropriate MMC for the flat-tile mock-up and in addition to enable the measurement of the thermal conductivity perpendicular to fibre direction. Unidirectional (UD) layers were prepared by two subsequent electroplating processes which allow adjusting various fibre volume fractions. The UD layers were stacked with different fibre orientations (0 /0 , 0 /90 ) and consolidated by vacuum hot pressing to form a multilayer MMC. In addition, MMC specimens were prepared by hot isostatic pressing (HIP) in order to measure the mechanical properties. To improve the bonding between fibre and matrix, the fibres were coated with thin titanium (SCS6) and Ti-TaC (SCS0) interlayers. Tensile tests

  18. Vacuum tests of a beamline front-end mock-up at the Advanced Photon Source

    International Nuclear Information System (INIS)

    A-mock-up has been constructed to test the functioning and performance of the Advanced Photon Source (APS) front ends. The mock-up consists of all components of the APS insertion-device beamline front end with a differential pumping system. Primary vacuum tests have been performed and compared with finite element vacuum calculations. Pressure distribution measurements using controlled leaks demonstrate a better than four decades of pressure difference between the two ends of the mock-up. The measured pressure profiles are consistent with results of finite element analyses of the system. The safety-control systems are also being tested. A closing time of ∼20 ms for the photon shutter and ∼7 ms for the fast closing valve have been obtained. Experiments on vacuum protection systems indicate that the front end is well protected in case of a vacuum breach

  19. Vacuum tests of a beamline front-end mock-up at the Advanced Photon Source

    International Nuclear Information System (INIS)

    A mock-up has been constructed to test the functioning and performance of the Advanced Photon Source (APS) front ends. The mock-up consists of all components of the APS insertion-device beamline front end with a differential pumping system. Primary vacuum tests have been performed and compared with finite element vacuum calculations. Pressure distribution measurements using controlled leaks demonstrate a better than four decades of pressure difference between the two ends of the mock-up. The measured pressure profiles are consistent with results of finite element analyses of the system. The safety-control systems are also being tested. A closing time of ∼20 ms for the photon shutter and ∼7 ms for the fast closing valve have been obtained. Experiments on vacuum protection systems indicate that the front end is well protected in case of a vacuum breach

  20. Delayed neutron measurements of induced fission rates in burnt LWR fuel samples at the Proteus zero-power reactor facility - 125

    International Nuclear Information System (INIS)

    The LIFE'at'PROTEUS program at the Paul Scherrer Institut is being undertaken to characterize the interfaces between burnt and fresh fuel assemblies in modern LWRs. Techniques are being developed to measure fission rates in burnt fuel, following re-irradiation in the zero-power PROTEUS research reactor. In the presented approach, the fission rates are estimated by measuring delayed neutrons emitted by re-irradiated fuel. To demonstrate the feasibility of this technique, fresh and burnt fuel samples (with burnup varying from 36 to 64 GWd/MTU) were irradiated in the PROTEUS reactor, and their neutron outputs were recorded shortly after irradiation. Relative fission rates between different core lattice positions were derived for a fresh sample as well as for the three burnt samples. The measured fission rate ratios have 1-σ uncertainties between 2% and 3.5%, with the larger uncertainties corresponding to the more highly burnt fuel. Results obtained by Monte Carlo simulations agree with the experimentally determined values within these limits. With further development of the technique, the experimental uncertainties can be further reduced. Continuing effort is being directed towards accurate comparison of fission rates between fuel samples of different burn-up. (authors)

  1. Intercomparison of liquid metal fast reactor seismic analysis codes. V. 2: Verification and improvement of reactor core seismic analysis codes using core mock-up experiments. Proceedings of a research co-ordination meeting held in Vienna, 26-28 September 1994

    International Nuclear Information System (INIS)

    This report (Volume II) contains the papers summarizing the verification of and improvement to the codes on the basis of the French and Japanese data. Volume I: ''Validation of the Seismic Analysis Codes Using the Reactor Code Experiments'' (IAEA-TECDOC-798) included the Italian PEC reactor data. Refs, figs and tabs

  2. Fabrication results of full scale mock-up for ITER VV port in Korea

    International Nuclear Information System (INIS)

    After a contract with Hyundai Heavy Industries Co. Ltd. (HHI) on January 2010 for the manufacture of the ITER equatorial and lower ports, manufacturing preparation activities have been performed. As part of the preparation activities, a full scale mock-up of the lower port stub extension (LPSE) #16 was fabricated by HHI in order to verify fabrication feasibility and set up the fabrication sequence for the LPSE. In this paper, major technical results from fabrication of a full scale mock-up will be presented with emphasis on the main manufacturing procedure, welding, nondestructive examination (NDE) and 3D dimensional inspection. Afterwards, progress of real product manufacturing is introduced

  3. Effective use of plant simulators and mock-up facilities for cultivation and training of younger regulators

    International Nuclear Information System (INIS)

    In order to achieve effective safety regulation, the staff members of a regulatory body who are engaged in regulatory work are requested to be well familiar with the characteristics, operations and maintenances of nuclear power plants at a practical level as far as possible. Although the regulators are not always required to have the same level of skills as those of plant designers or operators, the skills of the regulatory staff are essential elements to achieve high quality of the national nuclear safety regulation. Especially understanding of fundamentals such as operations, transient behaviors, trouble responses and plant inspections is indispensable not only to practical regulatory work but also to the establishment of the trust and confidence in safety regulation. To acquire these skills, the use of facilities such as plant simulators and inspection mock-up facilities is very effective to back up classroom lectures on theories and procedures. Practical training using these facilities under the guidance of well-experienced instructors inspires motivations and enhances capabilities of younger regulators. To support the countries newly embarking on nuclear power programs, JNES will continue to cooperate with those countries in cultivating and training younger regulators, by focusing on the training by veteran instructors using full-scale plant simulators and inspection mock-up facilities to give the trainees more practical skills and knowledge difficult to obtain through classroom lectures or textbooks. (author)

  4. Design study of a fast spectrum zero-power reactor dedicated to source driven sub-critical experiments

    Energy Technology Data Exchange (ETDEWEB)

    Mercatali, L.; Serikov, A. [Forschungszentrum Karlsruhe, Institute for Neutron Physics and Reactor Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Baeten, P.; Uyttenhove, W. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Lafuente, A. [Univerisdad Politecnica de Madrid, 28006 Madrid (Spain); Teles, P. [Instituto Tecnologico e Nuclear, EN 10, 2680-953 Sacavem (Portugal)

    2010-09-15

    In the framework of the European P and T program (IFP6-EUROTRANS), the Generation of Uninterrupted Intense NEutrons pulses at the lead VEnus REactor (GUINEVERE) project consists of an Accelerator Driven System (ADS) that is composed by a fast lead simulated-cooled reactor operated in sub-critical conditions, coupled with an updated version of the GENEPI neutron generator previously used for the MUSE experiments. The GUINEVERE facility aims at developing and improving different techniques for the reactivity monitoring of sub-critical ADS's. As such, the GUINEVERE project will comprise a series of major experiments that will be performed in the near future. The GUINEVERE facility will be located at the VENUS light water moderated research reactor at the SCK-CEN site of Mol (Belgium), which needs to be modified in order to accommodate a completely different and new type of core. A series of constraints were taken into account in the technical design of the GUINEVERE core, in order to properly conjugate the technical feasibility of this facility and the necessity to comply with the envisioned experimental program and its associated scientific outcome. The complete design study of the GUINEVERE core is the subject of this paper. The final design of the fuel assemblies, safety and control rods is provided. Also, the critical core configuration, to be used as reference for absolute reactivity measurements, is presented along with its associated reactor physics parameters, calculated by means of Monte Carlo methodologies. Finally, for licensing purposes, the GUINEVERE facility must satisfy the required nuclear safety criteria of the Belgian safety authorities, and in this paper, an overview of the safety analysis that has been performed with regard to the core physics, thermal assessment and shielding issues is also provided. (author)

  5. Thermal-hydraulics analysis of thermal mock-up test for JCO precipitation vessel using α-FLOW code system

    International Nuclear Information System (INIS)

    In the criticality accident occurred in the JCO Tokai plant, a significant power level in the plateau part covering about 20-hour duration continued following the initial-burst power. In order to reproduce the thermal characteristics of the plateau part, a mock-up device was made. A series of power-tracking tests had been performed to estimate the amount of water evaporation and the solution temperature change by changing the power of electric heaters. Based on the experimental data of the power-tracking test on the JCO precipitation vessel, a series of thermal-hydraulics analyses in the plateau part was performed using α-FLOW code system. Assuming a heat transfer coefficient at outer surface of the vessel, two-dimensional calculations were performed, and the calculated solution temperatures and the outlet temperature of the cooling-water were reproduced the measured value within 3-5degC. (author)

  6. Towards zero-power ICT

    Science.gov (United States)

    Gammaitoni, Luca; Chiuchiú, D.; Madami, M.; Carlotti, G.

    2015-06-01

    Is it possible to operate a computing device with zero energy expenditure? This question, once considered just an academic dilemma, has recently become strategic for the future of information and communication technology. In fact, in the last forty years the semiconductor industry has been driven by its ability to scale down the size of the complementary metal-oxide semiconductor-field-effect transistor, the building block of present computing devices, and to increase computing capability density up to a point where the power dissipated in heat during computation has become a serious limitation. To overcome such a limitation, since 2004 the Nanoelectronics Research Initiative has launched a grand challenge to address the fundamental limits of the physics of switches. In Europe, the European Commission has recently funded a set of projects with the aim of minimizing the energy consumption of computing. In this article we briefly review state-of-the-art zero-power computing, with special attention paid to the aspects of energy dissipation at the micro- and nanoscales.

  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

    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.

  8. 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

  9. 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

  10. Qualification of APOLLO2 BWR calculation scheme on the BASALA mock-up

    International Nuclear Information System (INIS)

    A new neutronic APOLLO2/MOC/SHEM/CEA2005 calculation scheme for BWR applications has been developed by the French 'Commissariat a l'Energie Atomique'. This scheme is based on the latest calculation methodology (accurate mutual and self-shielding formalism, MOC treatment of the transport equation) and the recent JEFF3.1 nuclear data library. This paper presents the experimental validation of this new calculation scheme on the BASALA BWR mock-up The BASALA programme is devoted to the measurements of the physical parameters of high moderation 100% MOX BWR cores, in hot and cold conditions. The experimental validation of the calculation scheme deals with core reactivity, fission rate maps, reactivity worth of void and absorbers (cruciform control blades and Gd pins), as well as temperature coefficient. Results of the analysis using APOLLO2/MOC/SHEM/CEA2005 show an overestimation of the core reactivity by 600 pcm for BASALA-Hot and 750 pcm for BASALA-Cold. Reactivity worth of gadolinium poison pins and hafnium or B4C control blades are predicted by APOLLO2 calculation within 2% accuracy. Furthermore, the radial power map is well predicted for every core configuration, including Void configuration and Hf / B4C configurations: fission rates in the central assembly are calculated within the ±2% experimental uncertainty for the reference cores. The C/E bias on the isothermal Moderator Temperature Coefficient, using the CEA2005 library based on JEFF3.1 file, amounts to -1.7±03 pcm/ deg. C on the range 10 deg. C-80 deg. C. (authors)

  11. Sensitivity and Uncertainty Analyses of the Tritium Production in the HCPB Breeder Blanket Mock-up Experiment

    International Nuclear Information System (INIS)

    Dedicated computational methods, tools and data have been recently developed in the framework of the European Fusion Technology Programme to enable sensitivity and uncertainty analyses of fusion neutronics experiments. severely limited due to these two requirements. (author)er productgeneration and the associated uncertainties against the experimental data provided in the neutronics experiment at the Frascati Neutron Generator on a mock-up of the HCPB (Helium-Cooled Pebble Bed) breeder test blanket. This work is devoted to the computational analyses of this experiment comprising the following steps: (i) Calculation of the Tritium production rates (TPR) in the Li2CO3 pellets using a detailed 3D model of the experimental set-up; the Monte Carlo code MCNP and the discrete ordinates code TORT were applied for these calculations with EFF-3 and FENDL-2.0/2.1 nuclear data. (ii) Sensitivity calculations for the Li2CO3 pellets stacks to assess the sensitivity of the Tritium production to the reactions cross-sections of the involved nuclides Be, 6,7Li, C and O; the calculations were performed with the MCSEN Monte Carlo code using the track length estimator and, in parallel, with the deterministic SUSD3D code using neutron fluxes calculated by TORT in forward and adjoint mode. (iii) Calculations of the data related uncertainties of the TPR using co-variance data from EFF (9Be, 6Li, 12C), FENDL-2 (7Li) and JENDL-3.3 (16O); both probabilistic (MCNP/MCSEN) and deterministic (TORT/SUSD3D) approaches were applied. (iv) Assessment of the total uncertainties for the TPR including uncertainties of the measurements, the nuclear data and the calculations. The data related uncertainties of the calculated Tritium generation are in the order of 4 - 5 % (2 sigma). The main uncertainties are due to the Be cross-section data. The total uncertainties of the predicted TPR including data uncertainties, statistical uncertainties of the Monte Carlo calculation and the experimental uncertainties

  12. Recent tritium breeding experiments with lithium titanate mock-ups irradiated with DT neutrons

    International Nuclear Information System (INIS)

    Experiments with breeding blanket mock-ups composed of layers of beryllium, ferritic steel F82H and 6Li enriched lithium titanate, Li2TiO3 are currently under investigation at Fusion Neutronics Source (FNS) of JAERI. Pellets of Li2TiO3 were used as detectors for the tritium production rate inside the tritium breeding layer. A method for direct measurement of low concentrations of tritium in Li2TiO3 was developed. After irradiation, the pellets were dissolved in concentrated hydrochloric acid and the tritium activity in the sample solution was measured by liquid scintillation counting. This method was applied to several mock-up experiments and the experiments analyzed with three-dimensional Monte Carlo calculations. In some cases the tritium production rate was found to be overestimated by the calculations. (author)

  13. Status report on active stabilisation of a linear collider final focus quadrupole mock-up

    International Nuclear Information System (INIS)

    The measurements done with the sensors available in our laboratories used for ground motion analysis are presented. The first sensors studied are seismic sensors measuring ground velocity, other sensors are accelerometers available for measuring ground acceleration. The first step has been to characterize the sensors, the second step has been to model and simulate the acceleration in order to identify Eigen frequencies and to display mode shapes. The third step has been to assess the performances of a new algorithm for disturbance rejection. In order to facilitate the analysis, a reduced-size mock-up has been used. The goal was to eliminate or at least to reduce as much as possible the main frequencies of the disturbance. A new mock-up is currently being developed that will have a geometry closer to a final focus quadrupole. Measurements will be done to validate the whole system in view of active stabilization for a future linear collider

  14. Development and manufacture of a PF coil-tail mock-up

    International Nuclear Information System (INIS)

    The ITER poloidal field (PF) winding pack design consists of a stack of double pancakes made of NbTi CIC conductor. A structural element is required, at each electrical joint, to react the operation hoop load. In the present design, this is provided by a hollow profiled steel part welded to the conductor jacket and bonded to the adjacent turns, named coil-tail. As part of the present development, prototype coil tails have been manufactured and welded to the PF conductor jacket. A full-size mechanical mock-up, in the shape of a straight beam has been manufactured. It is made of coil-tails and steel plates simulating the adjacent conductors, insulated and impregnated. The mock-up is to be subjected soon to fatigue tests at ENEA-Brasimone laboratory at liquid N2 temperature. (authors)

  15. Manufacturing and testing W/Cu functionally graded material mock-ups for plasma facing components

    International Nuclear Information System (INIS)

    The W/Cu functionally graded material (FGM) mock-ups were manufactured by resistance sintering under ultrahigh pressure or three times hot pressing. The bonding strength of W/Cu FGM was determined by tensile and shearing tests. A thermodiffusion experiment was used for testing thermal conductivity of the region containing W, the first and second W-Cu alloy layers. High heat flux and thermal fatigue tests have been carried out using electron beam or laser. The results are that the specimens with higher density in the W layer have better performances in high heat flux and thermal fatigue tests. Using the above sintering techniques, W/Cu FGM mock-ups for plasma facing components have been successfully manufactured at less cost

  16. Advanced Spacesuit Portable Life Support System Packaging Concept Mock-Up Design & Development

    Science.gov (United States)

    O''Connell, Mary K.; Slade, Howard G.; Stinson, Richard G.

    1998-01-01

    A concentrated development effort was begun at NASA Johnson Space Center to create an advanced Portable Life Support System (PLSS) packaging concept. Ease of maintenance, technological flexibility, low weight, and minimal volume are targeted in the design of future micro-gravity and planetary PLSS configurations. Three main design concepts emerged from conceptual design techniques and were carried forth into detailed design, then full scale mock-up creation. "Foam", "Motherboard", and "LEGOtm" packaging design concepts are described in detail. Results of the evaluation process targeted maintenance, robustness, mass properties, and flexibility as key aspects to a new PLSS packaging configuration. The various design tools used to evolve concepts into high fidelity mock ups revealed that no single tool was all encompassing, several combinations were complimentary, the devil is in the details, and, despite efforts, many lessons were learned only after working with hardware.

  17. Thermal response for divertor mock-up using surface-modified CFC tile

    International Nuclear Information System (INIS)

    It is necessary for the divertor plate to be actively cooled in order to remove the extremely high heat load from the fusion plasma. CFC material has been considered as one of the candidate plasma-facing materials because of its high thermal shock resistance. However, CFC causes several problems, such as the enhancement of hydrogen recycling, large erosion due to oxygen and radiation enhanced sublimation where the temperature exceeds about 1000 C. In this study the surface of CFC, CX-3002U, was converted to B4C and SiC by using a chemical vapor reaction, CVR. The thermal response properties of divertor mock-ups made by these materials and CFC were examined. These mock-ups were irradiated by electron beams with heat flux up to 15 MW/m2. The surface temperature rise of B4C-converted CFC tile was the highest and that of CFC the lowest. This difference was consistent with the value of the thermal conductivity e.g. B4C-converted CFC has the lowest thermal conductivity, 200 W/m K and CFC has the highest one, 450 W/m K. The heat flux that increases the surface temperature to 1000 C was approximately 8, 10 or 11 MW/m2 for B4C-converted CFC, SiC-converted CFC or CFC, respectively. Thermal cycling tests with more than 2000 shots were also conducted for these mock-ups. No deterioration in the heat transfer for each mock-up was found for the heat flux which increased the surface temperature to 1000 C. (orig.)

  18. Measurements in the Functional Mock Up Test of the NAL QSTOL Aircraft Control System

    OpenAIRE

    TADA, Akira; Ogawa,Toshio; YAMATO, Hiroyuki; Uchida, Tadao; Okada, Noriaki; 多田, 章; 小川, 敏雄; 大和, 裕幸; 内田, 忠夫; 岡田, 典秋

    1987-01-01

    In the functional mock up test of NAL QSTOL Research Aircraft control system, measurements were planned and conducted with the intention of obtaining both real time results to support the development immediately, and reserved data suitable for academically rigorous and detailed analyses from various points of view. The physical quantities of 208 system variables were converted to analogue voltage signals, and supplied from junction boxes to devices for recordings and analyses. The system char...

  19. Mock-up of a support structure of the ITER vacuum vessel

    International Nuclear Information System (INIS)

    The ITER vacuum vessel support systems located in the lower level sustain loads in radial and vertical direction. The support system consists of various sub-components like a linkage system, a pot type bearing, a vertical rope, a toroidal constraint, and dampers. In order to examine performance of the mechanism of the system, a mock-up of the linkage system which is comparatively complicated has been manufactured. Various fabrication methods were studied through the mock-up fabrication, and also several tests have been done using the mock-up. Those include assembly study, stroke test, static load test and fatigue test. In the full stroke test, the functional mechanism of the support structure has been demonstrated. In the structural test, the strength of the all components is evaluated by measuring reaction and strain of each component. In order to investigate the effect of tolerances and the damage due to the tests, the performance tests were conducted before and after the static and fatigue tests. The backlash for each stage is found from measured displacement hysteresis. As results of those tests, the performance of the ITER vacuum vessel support structure as well as its structural integrity has been evaluated in this study.

  20. 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.)

  1. Performance test results of helium gas circulator of mock-up test facility with full-scale reaction tube for HTTR hydrogen production system. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Akira; Kato, Michio; Hayashi, Koji [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2003-03-01

    Hydrogen production system by steam reforming of methane will be connected to the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Research Institute (JAERI) against development of nuclear heat utilization system. To obtain design and safety database of the HTTR hydrogen production system, mock-up test facility with full-scale reaction was constructed in FY 2001 and hydrogen of 120m{sup 3}N{sub /}h was successfully produced in overall performance test. This report describes performance test results of a helium gas circulator in this facility. The circulator performance curves regarding to pressure-rise, input power and adiabatic thermal efficiency at standard revolution number were made based on the measured flow-rate, temperature and pressure data in overall performance test. The circulator performance prediction code was made based on these performance curves. The code can calculate revolution number, electric power and temperature-rise of the circulator using flow-rate, inlet temperature, inlet pressure and pressure-rise data. The verification of the code was carried out with the test data in FY 2002. Total pressure loss of the helium gas circulation loop was also evaluated. The circulator should be operated in conditions such as pressure from 2.7MPa to 4.0MPa and flow-rate from 250g/s to 400g/s and at maximum pressure-rise of 250 kPa in test operation. It was confirmed in above verification and evaluations that the circulator had performance to satisfy above conditions within operation limitation of the circulator such as maximum input-power of 150 kW and maximum revolution number of 12,000 rpm. (author)

  2. 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

  3. Experimental results of start-up and shutdown test on HTTR hydrogen production system using mock-up test facility

    International Nuclear Information System (INIS)

    Japan Atomic Energy Research Institute has been developing a hydrogen production system using the High Temperature Engineering Test Reactor (HTTR) as a heat source. A system integration technology is newly required to couple the hydrogen production system with the HTTR; a steam generator with be installed as thermal buffer at downstream of chemical reactor in the secondary helium gas loop, and a pressure control system was installed to keep the pressure difference between the secondary helium and the process gases in constant to assure the structural integrity of reaction tube. In order to establish the operational procedure, the normal start-up and shutdown tests were performed using a mock-up test facility by steam reforming of methane. From the test results, it was confirmed that the thermal and the pressure disturbances caused by the hydrogen production were sufficiently small compared with allowable ranges. Next, in order to avoid the fluctuations of secondary helium gas temperature caused by the lack of steam supply, the initial and final conditions of control variables for hydrogen production were investigated and cleared through the experimental tests and the evaluation using the analysis code. (author)

  4. Non destructive examination of primary wall small scale mock-up DS-1F

    International Nuclear Information System (INIS)

    Ultrasonic examination of primary wall small scale mock up DS-1F before thermal testing showed no major defects on studied interfaces. However, some small indications were found on copper to copper and copper to steel interfaces and surface roughness of the outer surface of copper layer gave clear indications on ultrasonic images. After thermal test a curved 50 mm long crack along the Y- direction in the middle of the heated surface of the mock up and a 220 mm long crack along the copper to copper interface on the side surface of the mock up were detected. Small cracks, less than 60-80 μm in depth, were observed on copper surface. After thermal test the corresponding ultrasonic examination showed a strong effect on ultrasonic attenuation properties and on leaky Rayleigh waves on outer surface of copper layer. A major indication was found on copper to copper interface. About 50% of the copper to copper interface was delaminated. However, some small indications found already before thermal test were also found after thermal test and they were not grown in size. No indications were observed on copper to stainless steel interfaces. Additionally, major indications were found on stainless steel tube to copper interfaces. Tubes No. 1 and 2 were almost completely whereas tube No. 3 only partly separated from copper. No indications were found on stainless steel tube to copper interface on tube No. 4. Eddy current measurements showed no volumetric or crack like flaws in the stainless steel tubes, however, delamination of the copper to copper interface along the tubes No. 1, 2 and 3 was observed. (orig.)

  5. Import and Export of Functional Mock-up Units in JModelica.org

    OpenAIRE

    Andersson, Christian; Åkesson, Johan; Führer, Claus; Gäfvert, Magnus

    2011-01-01

    Different simulation and modeling tools often use their own definition of how a model is represented and how model data is stored. Complications arise when trying to model parts in one tool and importing the resulting model in another tool or when trying to verify a result by using a different simulation tool. The Functional Mock-up Interface (FMI) is a standard to provide a unified model execution interface. In this paper we present an implementation of the FMI specification in the JModelica...

  6. NDE of explosion welded copper stainless steel first wall mock-up

    International Nuclear Information System (INIS)

    The study showed that reflection type C-mode scanning acoustic microscope (C-SAM) and internal ultrasonic inspection (IRIS) equipment can be applied for ultrasonic examination of copper stainless steel compound structures of ITER first wall mock-ups. Explosive welding can be applied to manufacture fully bonded copper stainless steel compound plates. However, explosives can be applied only for mechanical tightening of stainless steel cooling tubes within copper plate. If metallurgical bonding between stainless steel tubes and copper plate is required Hot Isostatic Pressing (HIP) method can be applied. (orig.)

  7. The digital mock-up system to simulate and evaluate the dismantling scenarios for decommissioning of a NPP

    International Nuclear Information System (INIS)

    Highlights: • This paper is to develop the DMU system to simulate the dismantling scenarios for decommissioning of a NPP. • Features of the DMU system are kinematic simulation and human simulation of decommissioning scenarios of a NPP. • The DMU system could evaluate and optimize the decommissioning scenarios of a NPP. - Abstract: This paper is to develop the digital mock-up system to simulate the dismantling scenarios for decommissioning of a NPP. Features of the digital mock-up system are kinematic simulation and human simulation of decommissioning scenarios of a NPP. The digital mock-up consists of major components, internals, support facilities, and equipments. The digital mock-up system could evaluate and optimize the decommissioning scenarios of a NPP

  8. Design, construction, monitoring & control of a mock-up building module for testing new components and systems

    OpenAIRE

    Sánchez Labrador, Raúl

    2011-01-01

    In view of the difficulties with implementing the innovative components and systems conceived in the I3CON project on a dwelled building (because of their early stage of development), one of the main demonstration activities was building a Mock-up module to test the feasibility (in terms of physical integration and logical interoperability) of these components and systems, and evaluate their overall performance. The design of all the systems involved in the Mock-up has the aim to develop new ...

  9. Program mock-up of a system for NPP prompt monitoring

    International Nuclear Information System (INIS)

    The increase of safety of NPP operation requires the promotion of automation means for the plant control. The system of operator support are the means intensively developed today to increase monitoring quality. The program mockup is designed to check and debug the principles of designing the system for prompt monitoring, being one of a variant of the operator support system. The mock-up is based on the DVK-2 microcomputer or SM-4 minicomputer and is the program package operating under the control of a special monitor. The system can operate in two regimes: tuning and diagnosis. During tuning the system preparation to operation and adoptation to the certain installation occurs. Data acquisition and correction, file record, primary processing, state identification, diagnostic information and recommendation output are made during diagnostics. All mock-up programs are connected through the data base and a general region of immediate access store. Such a structure of the system permits to change the system composition flexibly when the programs are independent of data and each other

  10. Preparation of W/CuCrZr monoblock test mock-up using vacuum brazing technique

    International Nuclear Information System (INIS)

    Development of the joining for W/CuCrZr monoblock PFC test mock-up is an interest area in Fusion R and D. W/Cu bimetallic material has prepared using OFHC copper casting approach on the radial surface of W monoblock tile surface. The W/Cu bimetallic material has been joined with CuCrZr tube (heat sink) material with the vacuum brazing route. Vacuum brazing of W/Cu-CuCrZr has been performed @ 970 °C for 10 mins using NiCuMn-37 filler material under deep vacuum environment (10-6 mbar). Graphite fixtures were used for OFHC copper casting and vacuum brazing experiments. The joint integrity of W/Cu-CuCrZr monoblock mock-up on W/Cu and Cu-CuCrZr has been checked using ultrasonic immersion technique. Micro-structural examination and Spot-wise elemental analysis have been carried out using HR-SEM and EDAX. The results of the experimental work will be discussed in the paper. (author)

  11. Manufacturing and testing of a copper/CFC divertor mock-up for JET

    International Nuclear Information System (INIS)

    An actively cooled divertor is a possible option for future developments at The Joint European Torus (JET). A proof of principle actively cooled tile has been produced in order to qualify the relevant manufacturing technologies and the non destructive control processes. In this frame Ansaldo Ricerche (ARI) has been involved in the construction of a mock-up comprising 6 OFHC copper tubes for water cooling that are brazed to a plate made out of carbon fibre composite (CFC). The final objective was the high heat flux testing of the mock-up at JET in order to evaluate the general behaviour of the component under relevant operating conditions. The key point of the work was the realisation of a sound joint by adapting the expertise gained in ARI in previous R and D activities on brazing heterogeneous materials. Reliable methods for ultrasonic examinations of the pieces were also set up. For successful application to the JET pumped divertor a water-cooled CFC target plate must show surface temperatures of 2. Furthermore, global hydraulic considerations specific to JET limit the system pressure to 0.7 MPa. In such a design, critical heat flux is not the key limit, rather the reliability of the CFC-copper joint in terms of extent of wetting. First tests in the neutral beam test bed at JET show an adequate response for fluxes up to 15 MW/m2. (orig.)

  12. ORNL mock-up tests of inside launch pellet injection on JET and LHD

    International Nuclear Information System (INIS)

    In experiments on ASDEX-Upgrade and DIII-D tokamaks, the injection of D2 pellets from the magnetic high-field side of the plasma resulted in deeper pellet penetration and improved fueling efficiency. Based on those successful experiments, fusion researchers at the Joint European Torus and the Large Helical Device decided to implement inside launch pellet injection. These injection schemes require the use of curved guide tubes to route the pellets from the acceleration devices to the inside launch locations, and the pellets are subjected to stresses from centrifugal and impact forces in traversing the tubes. Before the installations on the large experimental fusion devices, mock-ups of the guide tubes were constructed and tested at the Oak Ridge National Laboratory to determine the pellet speed limit for reliable operation without pellet fracturing. In laboratory testing of the mock-ups, it was found that the pellet speed had to be limited to a few hundreds of meters per second for intact pellets. In this paper, the test equipment and experimental results are described

  13. Implementation of an Architecture for the Dismantling Digital Mock-up System

    International Nuclear Information System (INIS)

    It is necessary to forecast the various dismantling activities prior to dismantling nuclear facilities by using various software instead of a physical mock-up system because the dismantling in a contaminated with radioactivity cause the results of an unexpected situation. The component that needs to develop a dismantling mock-up system was examined. There are many component systems such as a decommissioning database system, 3D dosimetric mapping that represents a distribution of a radionuclide contamination, a component of modeling for nuclear facility and devices include the decontamination and decommissioning. The research of software architecture about these components was carried out because these component systems that have been independently doesn't describe not only to visual an activities of Decontamination and Decommissioning(D and D) but also to evaluate it. The result was established an architecture that consist of an visualization module which could be visualized an D and D activities and a simulation module which can be evaluated a dismantling schedule and decommissioning cost.

  14. Parallel Execution of Functional Mock-up Units in Buildings Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Ozmen, Ozgur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Nutaro, James J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); New, Joshua Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-06-30

    A Functional Mock-up Interface (FMI) defines a standardized interface to be used in computer simulations to develop complex cyber-physical systems. FMI implementation by a software modeling tool enables the creation of a simulation model that can be interconnected, or the creation of a software library called a Functional Mock-up Unit (FMU). This report describes an FMU wrapper implementation that imports FMUs into a C++ environment and uses an Euler solver that executes FMUs in parallel using Open Multi-Processing (OpenMP). The purpose of this report is to elucidate the runtime performance of the solver when a multi-component system is imported as a single FMU (for the whole system) or as multiple FMUs (for different groups of components as sub-systems). This performance comparison is conducted using two test cases: (1) a simple, multi-tank problem; and (2) a more realistic use case based on the Modelica Buildings Library. In both test cases, the performance gains are promising when each FMU consists of a large number of states and state events that are wrapped in a single FMU. Load balancing is demonstrated to be a critical factor in speeding up parallel execution of multiple FMUs.

  15. 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

  16. TANK 18 AND 19-F TIER 1A EQUIPMENT FILL MOCK UP TEST SUMMARY

    Energy Technology Data Exchange (ETDEWEB)

    Stefanko, D.; Langton, C.

    2011-11-04

    The United States Department of Energy (US DOE) has determined that Tanks 18-F and 19-F have met the F-Tank Farm (FTF) General Closure Plan Requirements and are ready to be permanently closed. The high-level waste (HLW) tanks have been isolated from FTF facilities. To complete operational closure they will be filled with grout for the purpose of: (1) physically stabilizing the tanks, (2) limiting/eliminating vertical pathways to residual waste, (3) discouraging future intrusion, and (4) providing an alkaline, chemical reducing environment within the closure boundary to control speciation and solubility of select radionuclides. Bulk waste removal and heel removal equipment remain in Tanks 18-F and 19-F. This equipment includes the Advance Design Mixer Pump (ADMP), transfer pumps, transfer jets, standard slurry mixer pumps, equipment-support masts, sampling masts, dip tube assemblies and robotic crawlers. The present Tank 18 and 19-F closure strategy is to grout the equipment in place and eliminate vertical pathways by filling voids in the equipment to vertical fast pathways and water infiltration. The mock-up tests described in this report were intended to address placement issues identified for grouting the equipment that will be left in Tank 18-F and Tank 19-F. The Tank 18-F and 19-F closure strategy document states that one of the Performance Assessment (PA) requirements for a closed tank is that equipment remaining in the tank be filled to the extent practical and that vertical flow paths 1 inch and larger be grouted. The specific objectives of the Tier 1A equipment grout mock-up testing include: (1) Identifying the most limiting equipment configurations with respect to internal void space filling; (2) Specifying and constructing initial test geometries and forms that represent scaled boundary conditions; (3) Identifying a target grout rheology for evaluation in the scaled mock-up configurations; (4) Scaling-up production of a grout mix with the target rheology

  17. FST-formation of cryogenic layer inside spherical shells of HiPER-class. Results of mathematical modeling and mock-ups testing

    International Nuclear Information System (INIS)

    Complete text of publication follows. Current stage in the IFE research has passed to a closing stage: creation of the experimental reactor and realization of electric power generation. HiPER is a proposed European High Power laser Energy Research facility dedicated to demonstrating the feasibility of laser driven fusion for IFE reactor. The HiPER facility operation requires the formation and delivery of spherical shock ignition cryogenic targets with a rate of several Hz. The targets must be free-standing, or un-mounted. At the Lebedev Physical Institute (LPI), significant progress has been made in the technology development based on rapid fuel layering inside moving free-standing targets which refers to as FST layering method. It allows one to form cryogenic targets with a required rate. In this report, we present the results of a feasibility study on high rep-rate formation of HiPER-class targets by FST. We consider two types of the baseline target for shock ignition. The first one (BT-2) is a 2.094-mm diameter compact polymer shell with a 3 μm thick wall. The solid layer thickness is 211 μm. The second (BT-2a) consists of a 2.046-mm diameter compact polymer shell (3 μm thick also) having a DT-filled CH foam (70 μm) on its inner surface, and then a 120 μm thick solid layer of pure DT. The work addresses the physical concept, and the modeling results of the major stages of FST technologies for different shell materials: Filling stage optimization (computation): optimal filling of a target batch up to ∼ 1000 atm at 300 K requires minimizing the diffusion fill time due to using the ramp filling method for both BT-2 and BT-2a; Depressurization stage optimization (computation and experiments): it requires providing the shell container leak proofness during the process of its cooling down to a depressurization temperature. This allows one to fulfill the technical requirements on the risks minimization associated with the damage of the HiPER-class targets

  18. Achievements on engineering and manufacturing of ITER first mirrors mock-ups

    International Nuclear Information System (INIS)

    Most of ITER optical diagnostics will be equipped with in-vessel metallic mirrors as plasma viewing components. These mirrors will be exposed to severe plasma environment which implies important research and developments on their design and manufacturing. Therefore investigations on engineering and manufacturing have been carried out on diagnostic mirrors towards the development of full-scale stainless steel and TZM (Mo-based alloy) ITER mirrors. Several micrometers in thickness of rhodium and molybdenum reflective coating layers have been deposited on the components to insure long-lasting of the mirrors exposed to an environment that could be dominated by neutral flux (charge-exchange). Three major issues have been addressed and reported in this paper: First, investigations have been performed on the design and manufacturing of the mirror integrated cooling system, so that the optical surface deformation due to radiations from the plasma and nuclear heating is limited. For the thermo mechanical design of the mock-ups, plasma radiation flux of 0,5 MW/m2 and neutron head load of 7 MW/m3 have been considered. Secondly, the polishing capability of full-scale (109 mm in diameter) metallic mirrors has been demonstrated: the mock ups Surface Front Error is lower than 0,1 μm Root Mean Square, and the mirrors exhibit low roughness (Ra ≤ 2 nm) and low surface defects (scratch width lower than 0,02 mm) after polishing. Thirdly, the manufacturing feasibility of molybdenum and rhodium thick coating layers deposited by magnetron sputtering has been evaluated. The objective of depositing layers up to 3 μm to 5 μm thick has been achieved on the mock-ups, with spectral performances reaching the theoretical values and showing high reflectivity over a large spectral range (from 400 nm to 11 μm). Finally the test campaign of the manufactured mirrors, which is being prepared in several European facilities to expose the mirrors to deuterium plasma, ELMs, neutrons, erosion and

  19. DWPF Sample Vial Insert Study-Statistical Analysis of DWPF Mock-Up Test Data

    International Nuclear Information System (INIS)

    This report is prepared as part of Technical/QA Task Plan WSRC-RP-97-351 which was issued in response to Technical Task Request HLW/DWPF/TTR-970132 submitted by DWPF. Presented in this report is a statistical analysis of DWPF Mock-up test data for evaluation of two new analytical methods which use insert samples from the existing HydragardTM sampler. The first is a new hydrofluoric acid based method called the Cold Chemical Method (Cold Chem) and the second is a modified fusion method.Both new methods use the existing HydragardTM sampler to collect a smaller insert sample from the process sampling system. The insert testing methodology applies to the DWPF Slurry Mix Evaporator (SME) and the Melter Feed Tank (MFT) samples. Samples in small 3 ml containers (Inserts) are analyzed by either the cold chemical method or a modified fusion method. The current analytical method uses a HydragardTM sample station to obtain nearly full 15 ml peanut vials. The samples are prepared by a multi-step process for Inductively Coupled Plasma (ICP) analysis by drying, vitrification, grinding and finally dissolution by either mixed acid or fusion. In contrast, the insert sample is placed directly in the dissolution vessel, thus eliminating the drying, vitrification and grinding operations for the Cold chem method. Although the modified fusion still requires drying and calcine conversion, the process is rapid due to the decreased sample size and that no vitrification step is required.A slurry feed simulant material was acquired from the TNX pilot facility from the test run designated as PX-7.The Mock-up test data were gathered on the basis of a statistical design presented in SRT-SCS-97004 (Rev. 0). Simulant PX-7 samples were taken in the DWPF Analytical Cell Mock-up Facility using 3 ml inserts and 15 ml peanut vials. A number of the insert samples were analyzed by Cold Chem and compared with full peanut vial samples analyzed by the current methods. The remaining inserts were analyzed by

  20. 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.)

  1. 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.)

  2. XML-based assembly visualization for a multi-CAD digital mock-up system

    International Nuclear Information System (INIS)

    Using a virtual assembly tool, engineers are able to design accurate and interference free parts without making physical mock-ups. Instead of a single CAD source, several CAD systems are used to design a complex product in a distributed design environment. In this paper, a multi-CAD assembly method is proposed through an XML and the lightweight CAD file. XML data contains a hierarchy of the multi-CAD assembly. The lightweight CAD file produced from various CAD files through the ACIS kemel and InterOp includes not only mesh and B-Rep data, but also topological data. It is used to visualize CAD data and to verify dimensions of the parts. The developed system is executed on desktop computers. It does not require commercial CAD systems to visualize 3D assembly data. Multi-CAD models have been assembled to verify the effectiveness of the developed DMU system on the Internet

  3. Mock-up stands for a rotating target for CSNS project

    International Nuclear Information System (INIS)

    This paper summarises pre-conceptual solutions for all sub-units of a potential rotating target system for the CSNS project. In order to test the validity of this concept and to gain first experience with a rotating target, the CSNS project has decided to embark on the construction of a mock-up test stand. The purpose is to provide first demonstration of the viability of the above concept by using a full model of the target head components and shaft and a dummy target with the right diameter and weight; confirm that acceptance criteria can be reached; gain experience in running a rotating target; verify certain parameters obtained by calculations. By carrying out a development program, it should be possible to produce a sound basis for a decision as to whether or not CSNS wants to adopt a rotating target as the preferred solution.

  4. Seismic tests on a reduced scale mock-up of a reprocessing plant cooling pond

    International Nuclear Information System (INIS)

    In conjunction with COGEMA and SGN, CEA has launched an important research program to validate the reprocessing plant cooling pond calculation mainly for the effect of the racks on the fluid-pond interaction. The paper presents the tests performed on a reduced scale mock-up (scale 1/5). The tests are composed by: -random excitations at very low excitation level to measure the natural frequencies, especially the first sloshing mode frequency; -sinusoidal tests to measure the damping; -seismic tests performed with 3 different time reduction scales (1, 1/5, 1/√5) and 3 different synthetic accelerograms. Two types of simplified model with added masses and finite element model were developed. Comparisons of measured and calculated pressure fields against the panels will be presented. The measured frequencies, obtained during tests, are in good agreement with Housner's results. (authors). 2 refs., 4 figs., 5 tabs

  5. Extension of incompressible algorithms to compressible flows: validation on a governing valve mock up

    International Nuclear Information System (INIS)

    The capacity of turbogenerators in PWR is regulated with governing valves located at the admission of the high-pressure turbine. In this paper we present a comparison between measurements and a numerical simulation of the flow in a 2D mock up of this governing valve. To predict and simulate transonic flow at low Mach numbers, we present a new extension of two codes initially devoted to incompressible and unsteady flows (pressure based method). The codes use either FInite Difference Method or, for complex geometry, Finite Element Method. Predicting those kinds of flows is difficult due to strong coupling between physical phenomena like turbulence on one hand, and the complexity of industrial geometry on the other hand. The comparison of numerical results with pressure measurements and also with Schlieren photographs confirms the validation of this approach. The results show clearly how the method correctly captures the structure of the jet. (authors). 10 figs., 11 refs

  6. Status of the neutronics experiments with MOCK-UPS of the two European TBM for ITER irradiated with 14 MeV neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Klix, A.; Fischer, U.; Leichtle, D. [Forschungszentrum Karlsruhe, Association Euratom-FZK (Germany); Freiesleben, H.; Henniger, J.; Seidel, K.; Sommer, M.; Unholzer, S. [Technische Univ. Dresden (Germany). Inst. fuer Kern- und Teilchenphysik; Batistoni, P. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy)

    2009-07-01

    The breeding blanket in fusion reactor has to produce enough tritium to maintain the fusion reaction, convert the fusion-produced energy into heat and shield the surrounding areas. The blanket design are optimized by radiation transport calculations based on Monte-Carlo codes. The neutron spectra inside the blanket reaches from thermal neutrons to 14 MeV, neutron scattering, neutron absorption, neutron multiplication reactions and particle emissive nuclear reactions have to be considered. For validation of the available neutron transport codes and the data libraries an experiment has been performed with a neutronics mock-up of one of the two lines of the test HCPB (helium cooled pebble bed) test blanket modules as proposed for ITER. The tritium production rate in the breeding layer was measured. A second test blanket module (TBM) is focused on a eutectic lithium and lead as breeder material and neutron multiplier in the HCLL (helium-cooled lithium lead). The HCLL-TBM mock-up experiment is underway.

  7. ZERO EMISSION POWER GENERATION TECHNOLOGY DEVELOPMENT

    Energy Technology Data Exchange (ETDEWEB)

    Ronald Bischoff; Stephen Doyle

    2005-01-20

    Clean Energy Systems (CES) was previously funded by DOE's ''Vision 21'' program. This program provided a proof-of-concept demonstration that CES' novel gas generator (combustor) enabled production of electrical power from fossil fuels without pollution. CES has used current DOE funding for additional design study exercises which established the utility of the CES-cycle for retrofitting existing power plants for zero-emission operations and for incorporation in zero-emission, ''green field'' power plant concepts. DOE funding also helped define the suitability of existing steam turbine designs for use in the CES-cycle and explored the use of aero-derivative turbines for advanced power plant designs. This work is of interest to the California Energy Commission (CEC) and the Norwegian Ministry of Petroleum & Energy. California's air quality districts have significant non-attainment areas in which CES technology can help. CEC is currently funding a CES-cycle technology demonstration near Bakersfield, CA. The Norwegian government is supporting conceptual studies for a proposed 40 MW zero-emission power plant in Stavager, Norway which would use the CES-cycle. The latter project is called Zero-Emission Norwegian Gas (ZENG). In summary, current engineering studies: (1) supported engineering design of plant subsystems applicable for use with CES-cycle zero-emission power plants, and (2) documented the suitability and availability of steam turbines for use in CES-cycle power plants, with particular relevance to the Norwegian ZENG Project.

  8. Design and testing of a 5 GHz TE10–TE30 mode converter mock-up for the lower hybrid antenna proposed for ITER

    International Nuclear Information System (INIS)

    Highlights: ► Design and validation of a 5 GHz TE10–TE30 mode converter. ► This mode converter is a RF element of a 20 MW CW LH system proposed for ITER. ► A low power mock-up has been manufactured at CEA/IRFM. ► RF measurements indicate a return loss of 40 dB and a transmission loss of 4.78 dB ± 0.03 dB for the three outputs. ► The forward conversion efficiency has been measured from electric field probing to 99.9%. - Abstract: The design and overall dimensions of a 5 GHz TE10–TE30 mode converter are presented. This mode converter is a RF element of a 20 MW CW lower hybrid system proposed for ITER. A low power mock-up of this device has been manufactured at CEA/IRFM and measured at low power. RF measurements indicate a return loss of 40 dB and a transmission loss of 4.78 dB ± 0.03 dB for the three outputs. The forward conversion efficiency from TE10 mode to TE30 has been measured from electric field probing to 99.9%. The good RF performances obtained validate the RF design of this element.

  9. ITER first mirror mock-ups exposed in Magnum-PSI

    Science.gov (United States)

    Marot, L.; De Temmerman, G.; van den Berg, M. A.; Renault, P.-O.; Covarel, G.; Joanny, M.; Travère, J. M.; Steiner, R.; Mathys, D.; Meyer, E.

    2016-06-01

    The goal of this work was to investigate coated first mirrors under very harsh erosion conditions. Mock-up mirrors were exposed to high-flux hydrogen/argon plasma in the linear plasma facility Magnum-PSI. Rhodium (Rh) and molybdenum (Mo) coated mirrors of different coating thicknesses, with or without water cooling, exhibited different responses to this exposure. Failures of Rh films were demonstrated for 5 micron thick film, 1 micron film revealed 10% decrease in the specular reflectivity only in the exposed area. In comparison, water cooled Mo mock-ups showed a significant diffuse reflectivity on the entire surface leading to more than 50% specular reflectivity losses in the visible range. The losses for non-cooled Mo samples did not exceed 7% in the whole studied wavelength range of 250–2500 nm. Three phenomena were proposed to explain these results. First the mechanical properties of the films as characterized by scratch and hardness measurements as well as residual stress analysis measured by x-ray diffraction. Rh films showed a high compressive stress value of 2.5  ±  0.4 GPa leading to poor adhesion of the thick films deposited on stainless steel substrate due to the high amount of available energy per area stored in the unbuckled film i.e. {{G}0}>30 J m‑2. It was confirmed by ANSYS simulation that the von Mises stress for the Rh coating was twice as high as that for the Mo coating due to different mechanical properties. Moreover, the maximum stress for thick Rh film (261 MPa) was higher than the critical buckling stress calculated with a buckle clamped Euler column model demonstrating the failure mode of the film. The second phenomenon was roughening of the mirror surface which was flux and temperature dependent, i.e. at low temperatures the surface would roughen randomly without any oriented surface morphology and at higher temperatures the surface diffusion constants would dominate the process and smoothen the surface. The last

  10. 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

  11. AP1000® PWR reactor physics analysis with VERA-CS and KENO-VI. Part 1. Zero power physics tests

    International Nuclear Information System (INIS)

    Westinghouse has applied the Core Simulator of the Virtual Environment for Reactor Applications, VERA-CS, under development by the Consortium for Advanced Simulation of LWRs (CASL) to the core physics analysis of the AP1000® PWR. The AP1000 PWR features and advanced first core with radial and axial heterogeneities, including enrichment zoning, multiple burnable absorbers, and a combination of light and heavy control banks to enable the MSHIMTM advanced operational strategy. These advanced features make application of VERA-CS to the AP1000 PWR first core especially relevant to qualify VERA performance. A companion paper at this conference describes the power distribution analysis of the AP1000 PWR with VERA-CS and the KENO Monte-Carlo code. This paper describes the results obtained for the startup physics tests simulations of the AP1000 PWR first core (critical boron, rod worth and reactivity coefficients), supporting the excellent numerical agreement reported in the companion paper for the power distribution. (author)

  12. Use of hydraulic and aerial mock up to study atmospheric pollution

    International Nuclear Information System (INIS)

    Fundamental studies on turbulent atmospheric diffusion of finely divided particles, cannot remain on a purely theoretical basis. Further experimental studies must be considered. - In full scale, from accidental and induced releases. - On a reduced scale, in aerodynamic wind tunnels or hydraulic water tunnels. A first set of studies on reduced scale models has been worked out according to a contract between French 'Meteorologie Nationale' and French 'Commissariat a l'Energie Atomique' and with the Collaboration of Saint-Cyr 'Institut Aerotechnique'. Essentially two kinds of results have been obtained: - The mathematical model of SUTTON for the turbulent diffusion in the atmosphere, deduced from the SUTTON theory, generally used by us, has been correctly verified, qualitatively and quantitatively whenever experiments were consistent with the theory conditions. - The quantitative assays of photographic and cinematographic visualization have given precise details on the phenomena inaccessible to calculations, due to the influence of obstacles and release conditions. - Generally, it can be asserted, that the atmospheric pollution studies are worked out by mock up experimentations and that, in some cases these experiments never can be replaced by mathematically pure models. (authors)

  13. Divided listening in noise in a mock-up of a military command post.

    Science.gov (United States)

    Abel, Sharon M; Nakashima, Ann; Smith, Ingrid

    2012-04-01

    This study investigated divided listening in noise in a mock-up of a vehicular command post. The effects of background noise from the vehicle, unattended speech of coworkers on speech understanding, and a visual cue that directed attention to the message source were examined. Sixteen normal-hearing males participated in sixteen listening conditions, defined by combinations of the absence/presence of vehicle and speech babble noises, availability of a vision cue, and number of channels (2 or 3, diotic or dichotic, and loudspeakers) over which concurrent series of call sign, color, and number phrases were presented. All wore a communications headset with integrated hearing protection. A computer keyboard was used to encode phrases beginning with an assigned call sign. Subjects achieved close to 100% correct phrase identification when presented over the headset (with or without vehicle noise) or over the loudspeakers, without vehicle noise. In contrast, the percentage correct phrase identification was significantly less by 30 to 35% when presented over loudspeakers with vehicle noise. Vehicle noise combined with babble noise decreased the accuracy by an additional 12% for dichotic listening. Vision cues increased phrase identification accuracy by 7% for diotic listening. Outcomes could be explained by the at-ear energy spectra of the speech and noise. PMID:22594135

  14. Ultimate tensile strength testing campaign on ITER pre-compression ring mock-ups

    Energy Technology Data Exchange (ETDEWEB)

    Rossi, Paolo, E-mail: paolo.rossi@enea.it [Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, C.P. 65, 00044 Frascati (Rome) (Italy); Capobianchi, Mario; Crescenzi, Fabio; Massimi, Alberto; Mugnaini, Giampiero; Nardi, Claudio; Pizzuto, Aldo [Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, C.P. 65, 00044 Frascati (Rome) (Italy); Bettinali, Livio [Consorzio CREATE, Via Claudio 21, 80125 Napoli (Italy); Knaster, Juan [ITER, Route de Vinon-sur-Verdon CS 90 046, 13067, St. Paul-lez-Durance Cedex (France); Rajainmaki, Hannu [FUSION FOR ENERGY, Josep Pla no 2, Torres Diagonal Litoral Edificio B3, 08019 Barcelona (Spain); Evans, David [Advanced Cryogenic Materials, Abingdon, Oxon (United Kingdom)

    2011-10-15

    ENEA has developed and characterized a high strength glass fibre-epoxy composite as reference material for the manufacture of the two sets of 3 pre-compression rings located at top and bottom of the inner straight leg region of the ITER Toroidal Field (TF) coils. These rings will provide a radial force of about 70 MN/coil at cryogenic temperature pulling the TF coils into contact and reducing toroidal tension in the four outer intercoil structures. The paper describes the ultimate tensile strength (UTS) testing campaign carried out at ENEA Frascati laboratories on six different rings manufactured winding S2 glass fibers on a diameter of 1 m (1/5 of the full scale) by both vacuum pressure epoxy impregnation and filament wet winding techniques. The volumetric glass content was around 70%. The rings were expanded with radial steps of 0.1 mm into a dedicated hydraulic testing machine consisting of 18 radial actuators working in position control with a total capability of 1000 tons. All the mock-ups showed very high tensile strength (1550 MPa is the average of the mean hoop stresses at failure) and a practically constant tensile modulus. The test results are reported and discussed.

  15. Non destructive examination of primary wall small scale mock-up PHS-1F

    International Nuclear Information System (INIS)

    Ultrasonic and eddy current examination of primary wall small scale mock up PHS-1F before thermal testing showed no major defects on studied interfaces. However, some small indications were found on copper to copper interface. After thermal test numerous small cracks on copper surface were observed in visual inspection. Crack depth was about 0.6 mm. The corresponding ultrasonic examination showed a strong effect on ultrasonic attenuation properties and on leaky Rayleigh waves on outer surface of copper layer. This strong attenuation caused by the high density of cracks on the copper surface disturbed the examination of interfaces below the heat treated surface. However, some small indications were found on copper to copper interface. No indication were found on copper to stainless steel interface. Clear indications were found on stainless steel tube to copper interfaces. Eddy current measurements showed no volumetric or crack like defects on the inner surfaces of stainless steel tubes but some indications were found corresponding to copper to copper interface around stainless steel tubes. (orig.)

  16. Thermo-siphon Mock-up Test for the HANARO-CNS

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jungwoon; Lee, Kye Hong; Kim, Hark Rho; Kim, Youngki; Kim, Myong Seop; Wu, Sang Ik; Kim, Bong Su

    2006-04-15

    In order to moderate thermal neutrons into cold neutrons, the liquid hydrogen is selected as a moderator for the HANARO CNS. By the non-nuclear heat load and nuclear heat load induced from collision of gamma-ray, beta-ray, and thermal neutrons, the liquid hydrogen in the moderator cell evaporates and flows into the heat exchanger. This evaporated hydrogen gas is liquefied by the cryogenic helium supplied from the helium refrigeration system,, then flows back to the moderator cell. This is so-called two-phase thermo-siphon. The most important point in the stable thermo-siphon is to have the good balance between the cooling capacity of the HRS and the heat load on the moderator cell so as to maintain the stable two-phase liquid level in the moderator cell. Accordingly, for not only the experience of the cryogenic two-phase thermo-siphon but also setup of the operation procedure, the full-scaled mock-up test has been performed using the liquid hydrogen. Through the test, the stable thermo-siphon establishment is confirmed at the cold normal operation; furthermore, the detail design parameter is validated. On top of the normal operation procedure setup, the abnormal operation procedure is settled based on the understanding the abnormal pressure and temperature transient dynamics in the hydrogen system.

  17. Structural analysis, design and evaluation of mock-up platform, monorail, and tank plate cut-out

    International Nuclear Information System (INIS)

    Platform - Structural analyses were performed for design seismic, live and dead load combinations for the freestanding platform over the partial DST mock-up section. The platform is to be used for Robotic ultrasonic inspection of the tank wall. It is a free standing structure anchored to floor slab with Hilti Kwik bolts

  18. 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.)

  19. Critical experiments in support of the CNPS [Compact Nuclear Power Source] program

    International Nuclear Information System (INIS)

    Zero-power static and kinetic measurements have been made on a mock-up of the Compact Nuclear Power Source (CNPS), a graphite moderated, graphite reflected, U(19.9% 235U) fueled reactor design. Critical configurations were tracked from a first clean configuration (184 most central fuel channels filled and all control rod and heat pipe channels empty) to a fully loaded configuration (all 492 fuel channels filled, core-length stainless steel pipe in the twelve heat-pipe channels, and approximately half-core-length boron carbide in the outer 4 control rod channels. Reactor physics data such as material worths and neutron lifetime are presented only for the clean and fully loaded configurations

  20. 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

  1. Local, zero-power void coefficient measurements in the ACPR

    International Nuclear Information System (INIS)

    Changes in reactivity may be stimulated in the ACPR by the local introduction of voids into the reactor coolant. The local void coefficients of reactivity which describe this effect are of interest from a reactor safety point-of-view, and their determination is the subject of this presentation. Bottled nitrogen gas was used to produce the voids. The gas was forced out of a small diameter tube which was positioned vertically in the core lattice with its open end below the fuel. The gas was passed through a pressure regulator, a valve, and a flowmeter to establish a steady flow condition, following which a delayed-critical (zero-power) reactor state was established. Correlation of the average volume of core void created by the nitrogen flow with the reactivity worth of the delayed-critical control-rod bank position produced the values of the zero-power void coefficients of reactivity. The void coefficients were determined at various core positions from ∼6 mm to 142 mm beyond the central irradiation space and for three different flow rates. For the range of void fractions investigated, these coefficients are negative, with values ranging between -$0.02 and -$0.12. Tabular and graphical results of the measurements are presented, and details of the coefficient determination are explained. (author)

  2. 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)

  3. Novel methods of tritium production rate measurements in HCLL TBM mock-up experiment with liquid scintillation technique

    International Nuclear Information System (INIS)

    Two novel methods of tritium production rate (TPR) measurements applied in HCLL TBM mock-up neutronic experiment are described and discussed. In the first method LiF TLD detectors used for gamma radiation dose measurements are applied as detectors of -β decay of tritium (T) produced by neutrons. TL signal expressed as mGy/h is proportional to energy accumulated in self-irradiation process (SI). However, the TL signal is proportional to tritium activity generated in detector material, recalculation to Bq/mg requires calibration by independent measurement of T activity in the TLD. In our experiments determination of this activity was performed by LSC technique. Good correlation between the 3H activity and TLD signal was observed. The second presented method is based on direct measurement of T in LiPb material with the use of LSC spectrometer. Dissolution of LiPb in nitric acid, and than composing aliquot containing well known mass of irradiated LiPb mixed than with scintillation cocktail, is measured in LS spectrometer providing signal proportional to the T activity. It was proved by series of experiments that procedure performed at controlled conditions ensures no loss of tritium while preparation. Optimization measurement conditions with respect of type of the scintillator used and composition of aliquot/scintillator mixture provide counting efficiency above 22%. Sample material for testing the both techniques was irradiated by neutron flux either at nuclear research reactor 'Maria', Swierk, Poland or at the Frascati Neutron Generator Laboratory in HCLL TBM neutronic experiment. Li2CO3 pellets used as reference were irradiated in the same experiment with TLDs, and samples of LiPb-eutectic and then analysed by modified Dierckx method, a well established technique. Measurement results obtained applying the both techniques described above as well as the control Li2CO3 pellets were in very good consistency. Direct measurements of T produced in LiPb at different depth

  4. Intercomparison of analysis methods for seismically isolated nuclear structures (KAERI HLRB and CRIEPI Isolated Rigid Mass Mock-Up)

    International Nuclear Information System (INIS)

    The combined shear and compression behaviors of the KAERI HLRB made of MRPRA rubber and the shaking responses of the CRIEPI isolated rigid mass mock-up are analyzed. For FEM analyses of KAERI HLRB, three kinds of strain energy density functions of the ABAQUS program are used as constitutive law for rubber with hyperelastic characteristics. The analysis results are compared with test results, depending on the constitutive models. The simulation results for the shaking table tests of the CRIEPI rigid mass mock-up supported by scaled lead rubber bearings are obtained by ABAQUS time history analyses. In the analysis, the linear and bilinear hysterisis models simulating the behaviors of the rubber bearing are used. (author)

  5. Inspection of heat transfer tubes after mock-up tests of miniaturized apparatus for the acid recovery evaporator. Contract research

    International Nuclear Information System (INIS)

    The demonstration test for the acid recovery evaporator and the dissolver used in the major equipment of Rokkasho Reprocessing Plant (RRP), has been carried out. The mock-up miniature equipment has been employed to it. This test had been performed from April in 1998. The total time of demonstration test using the mock-up equipment is about two and half years, which corresponds to about 20,000 hours. After that, four of the seven heat transfer tubes used in the evaporator were drawn out and the corrosion level and the mechanical properties were evaluated for one of them. As a result, intergranular corrosion was recognized in the inner surface of the heat transfer tube and the corrosion depth at the grain boundary was statistically shown to be about one grain from the inner surface. Further, no change in mechanical properties was observed and growth of intergranular cracks in the inner surface of the specimen was found after flattering test. (author)

  6. Development of method to measure vibrational stress with multiple displacement sensors. Examination of applicability by vibration test using mock-up piping

    International Nuclear Information System (INIS)

    In the nuclear power plants, many small-bore piping are frequently oscillated by vibrational sources like pumps, and it is important to evaluate vibration stress of the piping efficiently to prevent the fatigue failure. This study proposes a new method developed to calculate vibration stress using displacements measured by multiple contactless displacement sensors. In this study, the applicability of the proposed method was verified with vibration experiments by using an actual mock-up piping. The experiments were conducted under sweep operations or steady operations of rotation frequency of pump. The stress calculated by the proposed method was compared with the stress measured by strain gauge. As the results, the calculated stress was in good agreement with the measured stress though having some error which was negligible comparing to endurance limit of failure. Therefore, the method developed was applicable to measure vibration stress of small-bore piping in the actual plants. (author)

  7. Re-analysis of HCPB/HCLL Blanket Mock-up Experiments Using Recent Nuclear Data Libraries

    Science.gov (United States)

    Kondo, K.; Fischer, U.; Klix, A.; Pereslavtsev, P.; Serikov, A.; Villari, R.

    2014-06-01

    We have re-analysed the two breeding blankets experiments performed previously in the frame of the European fusion program on two mock-ups of the European Helium-Cooled-Lithiium Lead (HCLL) and Helium-Cooled-Pebble-Bed (HCPB) test blanket modules for ITER. The tritium production rate and the neutron and photon spectra measured in these mock-ups were compared with calculations using FENDL-3 Starter Library, release 4 and state-of-the-art nuclear data evaluations, JEFF-3.1.2, JENDL-4.0 and ENDF/B-VII.0. The tritium production calculated for the HCPB mock-up underestimates the experimental result by about 10%. The result calculated with FENDL-3/SLIB4 gives slightly smaller tritium production by 2% than the one with FENDL-2.1. The difference attributes to the slight modification of the total and elastic scattering cross section of Be. For the HCLL experiment, all libraries reproduce the experimental results well. FENDL-3/SLIB4 gives better result both for the measured spectra and the tritium production compared to FENDL-2.1.

  8. Experiments of multichannel least-square methods for sound field reproduction inside aircraft mock-up: Objective evaluations

    Science.gov (United States)

    Gauthier, P.-A.; Camier, C.; Lebel, F.-A.; Pasco, Y.; Berry, A.; Langlois, J.; Verron, C.; Guastavino, C.

    2016-08-01

    Sound environment reproduction of various flight conditions in aircraft mock-ups is a valuable tool for the study, prediction, demonstration and jury testing of interior aircraft sound quality and annoyance. To provide a faithful reproduced sound environment, time, frequency and spatial characteristics should be preserved. Physical sound field reproduction methods for spatial sound reproduction are mandatory to immerse the listener's body in the proper sound fields so that localization cues are recreated at the listener's ears. Vehicle mock-ups pose specific problems for sound field reproduction. Confined spaces, needs for invisible sound sources and very specific acoustical environment make the use of open-loop sound field reproduction technologies such as wave field synthesis (based on free-field models of monopole sources) not ideal. In this paper, experiments in an aircraft mock-up with multichannel least-square methods and equalization are reported. The novelty is the actual implementation of sound field reproduction with 3180 transfer paths and trim panel reproduction sources in laboratory conditions with a synthetic target sound field. The paper presents objective evaluations of reproduced sound fields using various metrics as well as sound field extrapolation and sound field characterization.

  9. Manufacture of first wall mock-ups with calibrated defects for fabrication control methods: Development of UT detectable defects

    International Nuclear Information System (INIS)

    A Research and Development program for the ITER Blanket-First Wall has been implemented in Europe to provide input data for the manufacture of the full-scale production components. In this frame, FW mock-ups have been fabricated according to ITER FW design requirements. In order to define acceptance criteria for non-destructive examination (NDE) for the series production, FW mock-ups (FWMU) representative of ITER FW are manufactured with calibrated defects to be validated by heat flux tests to assess the critical defect dimensions able to degrade fatigue performance and lifetime, when located at Be/CuCrZr joint corners and beryllium tile edges, and at the CuCrZr/CuCrZr and CuCrZr/316L SS joints. In order to create the defects of given dimensions, two techniques were studied: alumina and zirconia coating using a PVD technique in one hand; and on the another hand alumina and quartz thicker inserts. The paper describes the different approaches used to manufacture test samples with calibrated defects, before applying on FW mock-ups, and related non-destructive examination (NDE) by ultrasonic examination (UT). High heat flux (HHF) testing is not part of this work.

  10. Re-analysis of HCPB/HCLL Blanket Mock-up Experiments Using Recent Nuclear Data Libraries

    Energy Technology Data Exchange (ETDEWEB)

    Kondo, K., E-mail: keitaro.kondo@kit.edu [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen, 76344 (Germany); Fischer, U.; Klix, A.; Pereslavtsev, P.; Serikov, A. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen, 76344 (Germany); Villari, R. [ENEA C.R. Frascati, Via E. Fermi 45, I00044 Frascati (Italy)

    2014-06-15

    We have re-analysed the two breeding blankets experiments performed previously in the frame of the European fusion program on two mock-ups of the European Helium-Cooled-Lithiium Lead (HCLL) and Helium-Cooled-Pebble-Bed (HCPB) test blanket modules for ITER. The tritium production rate and the neutron and photon spectra measured in these mock-ups were compared with calculations using FENDL-3 Starter Library, release 4 and state-of-the-art nuclear data evaluations, JEFF-3.1.2, JENDL-4.0 and ENDF/B-VII.0. The tritium production calculated for the HCPB mock-up underestimates the experimental result by about 10%. The result calculated with FENDL-3/SLIB4 gives slightly smaller tritium production by 2% than the one with FENDL-2.1. The difference attributes to the slight modification of the total and elastic scattering cross section of Be. For the HCLL experiment, all libraries reproduce the experimental results well. FENDL-3/SLIB4 gives better result both for the measured spectra and the tritium production compared to FENDL-2.1.

  11. Thermal cycling tests of 1st wall mock-ups with beryllium/CuCrZr bonding

    International Nuclear Information System (INIS)

    The innovative bonding technology between beryllium and CuCrZr with Hot Isostatic Pressing (HIP) has been proposed for the manufacturing of the ITER first wall. In the next step, thermal cycling test of first wall mock-ups manufactured with the bonding technology, were carried out under the ITER heat load condition. The test condition is 1000 cycles of On and Off under 5 MW/m2, and two types of the mock-up were manufactured for evaluation of the effects on HIP temperature (520 degree C and 610 degree C). The tensile properties of the bonding were also evaluated in room temperature and 200 degree C. As for the results of the thermal cycling tests, the temperature near the bonding interface were scarcely any change up to 1000 cycles, and obvious damage of the mock-up was not detected under the tests. As for the results of the tensile tests in 200 degree C, the test pieces of the HIP bonding at 610 degree C were broken in parent CuCrZr material, not broken in the bonding interface. (author)

  12. Re-analysis of HCPB/HCLL Blanket Mock-up Experiments Using Recent Nuclear Data Libraries

    International Nuclear Information System (INIS)

    We have re-analysed the two breeding blankets experiments performed previously in the frame of the European fusion program on two mock-ups of the European Helium-Cooled-Lithiium Lead (HCLL) and Helium-Cooled-Pebble-Bed (HCPB) test blanket modules for ITER. The tritium production rate and the neutron and photon spectra measured in these mock-ups were compared with calculations using FENDL-3 Starter Library, release 4 and state-of-the-art nuclear data evaluations, JEFF-3.1.2, JENDL-4.0 and ENDF/B-VII.0. The tritium production calculated for the HCPB mock-up underestimates the experimental result by about 10%. The result calculated with FENDL-3/SLIB4 gives slightly smaller tritium production by 2% than the one with FENDL-2.1. The difference attributes to the slight modification of the total and elastic scattering cross section of Be. For the HCLL experiment, all libraries reproduce the experimental results well. FENDL-3/SLIB4 gives better result both for the measured spectra and the tritium production compared to FENDL-2.1

  13. Mock-up facilities for the development of an advanced spent fuel management process using molten salt technology

    International Nuclear Information System (INIS)

    The Korea Atomic Energy Research Institute (KAERI) has investigated a new approach to spent fuel storage technology that would reduce the total storage volume and the amount of decay heat. The technology utilizes the reduction of oxide fuel to a metal to reduce the volume and preferentially removing the fission products to reduce the decay heat. The uranium oxide is reduced to uranium metal by Li metal in a molten LiCl salt bath. During the reduction process, fission products are dissolved into the LiCl bath and some of the highly radioactive elements, such as Sr and Cs, are preferentially removed from the bath. The reduced uranium metal is cast into an ingot, put into a storage capsule, and stored using conventional storage methods. The fission products are treated as high level radioactive wastes. Each process of the technology has been studied and analyzed for technical feasibility, and has come to the point for designing and constructing of the mock-up for a demonstration of the technology. This paper presents the detailed design of the mock-up of the system and operational characteristics, along with all the details of the equipment for the system. KAERI plans to use the mock-up for the demonstration using an in-active spent fuel specimen. (authors)

  14. Application of integral transport theory with linearly-anisotropic scattering to the calculation of the neutron distribution in finite plate arrays of fast reactors

    International Nuclear Information System (INIS)

    A method was derived to calculate from integral transport theory the neutron flux density close to boundaries between plate lattice regions of fast zero power reactors. Leakage parallel to the plate planes is treated by a momentum method. Linearly anisotropic scattering and the space dependence of effective cross sections in the resonance region are taken into account by suitable approximations. The method was applied to evaluate reaction rate measurements in the vicinity of the core-blanket-boundary of a SNEAK-assembly mocking up parts of the reactor SNR 300. Using the cross section set KFKINR agreement with the experiment was achieved for the space dependence of the 239Pu fission rate and the 238U capture rate. Further improvements of the nuclear data of 238U are required for solving discrepancies found in the outer blanket regions and in a U-metal blanket. (Orig.)

  15. 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

  16. 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.)

  17. In-cell crane and repair hoist for power reactor and nuclear fuel development corporation

    International Nuclear Information System (INIS)

    Transformation of the high-level radioactive liquid waste into stable vitrification has been proceeding as a national project by PNC (Power Reactor and Nuclear Fuel Development Corporation). IHI has participated in this project as a main member and has undertaken development and designing of the key technology and facilities of the demonstrational plant. The in-cell crane and repair hoise are the mock up that had been developed for simulating remote technology and remote handling other facilities in the mock vitrification cell that were constructed at the PNC Tokai Works. The in-cell crane installed in the radioactive atmospheric cell has the ability of remote maintenance itself in addition to remote handling other facilities and repair hoist installed on the ceiling of the cell is used for remote maintenance of other machines. (author)

  18. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

  19. 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)

  20. Zero CO2 emission SOLRGT power system

    International Nuclear Information System (INIS)

    A novel hybrid power system with zero CO2 emission (ZE-SOLRGT) has been proposed and analyzed in this paper. It consists of a high temperature Brayton-like topping cycle and a high pressure-ratio Rankine-like bottoming cycle, integrated with methane-steam reforming, solar heat-assisted steam generation and CO2 capture and compression. Water is selected to be the working fluid. Solar heat input enhances the steam generation and power output, and reduces fossil fuel consumption. Besides CO2 capture with oxy-fuel combustion and cascade recuperation of turbine exhaust heat, the system is featured with indirect upgrading of low-mid temperature solar heat and cascade release of fossil fuel chemical exergy, which is described by the energy level concept. With nearly 100% CO2 capture, the system attains a net energy efficiency of 50.7% (including consideration of the energy needed for oxygen separation). The cost of generated electricity and the payback period of ZE-SOLRGT are found to be $0.056/kWh and 11.3 years, respectively. The system integration accomplishes the complementary utilization of fossil fuel and solar heat, and attains their high efficiency conversion into electricity. -- Highlights: ► A novel hybrid power system ZE-SOLRGT has been proposed and analyzed. ► The system integrates power generation with methane-steam reforming, solar heat driven steam generation and CO2 capture. ► The system is featured with indirect upgrading of solar heat and cascade release of fossil fuel chemical exergy. ► The system thermodynamic and economic performances have been investigated.

  1. Development of the cooling technology on TRU fuel pin bundle during fuel fabrication process (4). Steady state cooling test of full mock up fuel pin bundle

    International Nuclear Information System (INIS)

    The development of the fast reactor cycle is being preceded in Japan to utilize plutonium and trans-uranium materials which come from the simplified PUREX reprocessing. But the TRU fuel bundle generates heat due to fission of TRU during the fabrication process of the wire wrapped Fast Breeder Reactor (FBR) fuel pin bundle. Then it is a big issue to develop an efficient cooling system for the horizontally laid bundle and to clarify its thermal behavior. Then in this paper the steady state full mock up test results are described. Inlet air velocity and heat generation rate were varied in the tests as the parameter. Then it is ascertained that the fuel can be cooled under the 473 K which is the criterion for the steady state cooling of this study to keep cladding soundness. The temperature and velocity fields of the bundle upper side were also measured by moving thermocouples to vertical and horizontal directions, by the infrared thermometer and by PIV (Particle Image Velocimetry). Then the temperature and velocity fields at outlet region are clarified. (author)

  2. 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

  3. Tritium breeding mock-up experiments containing lithium titanate ceramic pebbles and lead irradiated with DT neutrons

    International Nuclear Information System (INIS)

    Highlights: • Breeding benchmark experiment on LLCB TBM in ITER was performed. • Nuclear responses measured are TPR and reaction rate of 115In(n, n′)115mIn reaction. • Measured responses are compared with calculations by MCNP and FENDL 2.1 library. • TPR measurements agree with calculations in the estimated error bar. • Measured 115In(n, n′)115mIn reaction rates are underestimated by the calculations. - Abstract: Experiments were conducted with breeding blanket mock-up consisting of two layers of breeder material lithium titanate pebbles and three layers of pure lead as neutron multiplier. The radial dimensions of breeder, neutron multiplier and structural material layers are similar to the current design of the Indian Lead–Lithium cooled Ceramic Breeder (LLCB) blanket. The mock-up assembly was irradiated with 14 MeV neutrons from DT neutron generator. The local tritium production rates (TPR) from 6Li and 7Li in breeder layers were measured with the help of two different compositions of Li isotopes (60.69% 6Li and 7.54% 6Li) in Li2CO3. Tritium production in the multiplication layers were also measured with above mentioned two types of pellets to compare the experimental tritium production with calculations. TPR from 6Li at one location in the breeder layer was also measured by direct online measurement of tritons from 6Li(n, t)4He reaction using silicon surface barrier detector and 6Li to triton converter. Additional verification of neutron spectra (En > 0.35 MeV) in the mock-up zones were obtained by measuring 115In(n, n′)115mIn reaction rate and comparing it with calculated values in all five layers of mock-up. All the measured nuclear responses were compared with transport calculations using code MCNP with FENDL2.1 and FENDL3.0 cross-section libraries. The average C/E ratio for tritium production in enriched Li2CO3 pellets was 1.11 in first breeder zone and 1.09 in second breeder zone with uncertainty 8.3% at 1σ level. The experimental details

  4. Tritium breeding mock-up experiments containing lithium titanate ceramic pebbles and lead irradiated with DT neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Jakhar, Shrichand; Abhangi, M.; Tiwari, S. [Institute for Plasma Research, Bhat, Gandhinagar 382 428 (India); Makwana, R. [Department of Physics, MS University, Vadodara (India); Chaudhari, V.; Swami, H.L.; Danani, C.; Rao, C.V.S.; Basu, T.K. [Institute for Plasma Research, Bhat, Gandhinagar 382 428 (India); Mandal, D.; Bhade, Sonali; Kolekar, R.V.; Reddy, P.J. [Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Bhattacharyay, R.; Chaudhuri, P. [Institute for Plasma Research, Bhat, Gandhinagar 382 428 (India)

    2015-06-15

    Highlights: • Breeding benchmark experiment on LLCB TBM in ITER was performed. • Nuclear responses measured are TPR and reaction rate of {sup 115}In(n, n′){sup 115m}In reaction. • Measured responses are compared with calculations by MCNP and FENDL 2.1 library. • TPR measurements agree with calculations in the estimated error bar. • Measured {sup 115}In(n, n′){sup 115m}In reaction rates are underestimated by the calculations. - Abstract: Experiments were conducted with breeding blanket mock-up consisting of two layers of breeder material lithium titanate pebbles and three layers of pure lead as neutron multiplier. The radial dimensions of breeder, neutron multiplier and structural material layers are similar to the current design of the Indian Lead–Lithium cooled Ceramic Breeder (LLCB) blanket. The mock-up assembly was irradiated with 14 MeV neutrons from DT neutron generator. The local tritium production rates (TPR) from {sup 6}Li and {sup 7}Li in breeder layers were measured with the help of two different compositions of Li isotopes (60.69% {sup 6}Li and 7.54% {sup 6}Li) in Li{sub 2}CO{sub 3}. Tritium production in the multiplication layers were also measured with above mentioned two types of pellets to compare the experimental tritium production with calculations. TPR from {sup 6}Li at one location in the breeder layer was also measured by direct online measurement of tritons from {sup 6}Li(n, t){sup 4}He reaction using silicon surface barrier detector and {sup 6}Li to triton converter. Additional verification of neutron spectra (E{sub n} > 0.35 MeV) in the mock-up zones were obtained by measuring {sup 115}In(n, n′){sup 115m}In reaction rate and comparing it with calculated values in all five layers of mock-up. All the measured nuclear responses were compared with transport calculations using code MCNP with FENDL2.1 and FENDL3.0 cross-section libraries. The average C/E ratio for tritium production in enriched Li{sub 2}CO{sub 3} pellets was 1

  5. Power Reactors. Appendix VIII

    International Nuclear Information System (INIS)

    Decommissioning of nuclear facilities in many countries has evolved into a mature industry that has benefited from experience gained from previous projects and decommissioning costs can now be estimated to a good degree of accuracy. As a result of lessons learned, future decommissioning projects can be performed with higher levels of efficiency. Decommissioning of old power reactors is in progress in several countries. In some cases, decommissioning has been completed (i.e. plant sites have been released from regulatory control), while in other countries decommissioning is still in progress. Several large power reactors have been successfully decommissioned since 1995. The key areas of particular importance for decommissioning are decontamination, radiation protection, dismantling and demolition. The technologies which can be used for these tasks are commonly available on the market, but effective decommissioning still depends on an optimal choice of technologies, including site specific developments. It is not possible to recommend the use of a single specific technology for dismantling, demolition, segmentation or decontamination; rather, it is good practice to take into account as much information as possible from other decommissioning projects and to draw comparisons between various techniques in order to choose the one with the best performance in a particular situation. The exchange of information on all types of decommissioning experience, including decommissioning techniques and their applicability as well as disadvantages for specific tasks, is taking place on various levels, such as: — Collaborative working groups established by international organizations such as the IAEA, the OECD Nuclear Energy Agency and the European Commission and the publication of technical reports by such organizations; — National and international conferences; — Bilateral or multilateral cooperation and information exchange between organizations with responsibilities for

  6. The GUINEVERE-project: the first zero-power fast lead reactor coupled to a 14 MeV neutron generator (GENEPI)

    International Nuclear Information System (INIS)

    The GUINEVERE project is an European project in the framework of FP6 IP-EUROTRANS. The IP-EUROTRANS project aims at addressing the main issues for ADS development in the framework of partitioning and transmutation for nuclear waste volume and radio toxicity reduction. The GUINEVERE-project is carried out in the context of domain 2 of IP-EUROTRANS, ECATS, devoted to specific experiments for the coupling of an accelerator, a target and a subcritical core. These experiments should provide an answer to the questions of on-line reactivity monitoring, sub-criticality determination and operational procedures (loading, start-up, shut-down) in an ADS by 2009-2010. During the definition of the experimental programme ECATS, it was judged that there was a strong need for a European managed experiment in the line of the FP5 MUSE-project. Reanalyzing the outcome of MUSE, two points were left open for significant improvement. To validate the methodology for reactivity monitoring, a continuous beam is needed, which was not present in the MUSE-project. In the definition of the MUSE-project, from the beginning a strong request was made for a lead core in order to have representative conditions of a lead-cooled ADS which was only partially answered by the MUSE-programme. Therefore, there is a need for a lead fast critical facility connected to a continuous beam accelerator. Since such a programme/installation is not present at the European nor at the international level, SCK-CEN has proposed to use a modified VENUS critical facility located at its Mol-site and to couple it to a modified GENEPI deuteron accelerator (used in MUSE) working in current mode delivering 14 MeV neutrons by bombardment of deuterons on a tritium-target: the GUINEVERE-project (Generator of Uninterrupted Intense NEutrons at the lead VEnus REactor). This proposal was formally accepted by the Governing Council of IP-Eurotrans in December 2006. This project represents a close collaboration between SCK-CEN, CEA and

  7. Pickering B nuclear dose reduction through innovative shielding and mock-up training

    International Nuclear Information System (INIS)

    An innovative temporary shielding tool design has made a significant contribution to the success of the P761 Single Fuel Channel Replacement (SFCR). Significantly reducing collective radiation dose and project expenses, the tool (referred to as REACTORshield) will lower production costs and has proven to be an effective tool that can be implemented industry-wide in the future. Aligning around the nuclear values of teamwork, respect, integrity and commitment, the team responsible for this success overcame administrative obstacles and time constraints to get the REACTORshield pieces in place before the first project evolution. Prior to unit shutdown, the team worked diligently to source the right product provider and ensure just-in-time delivery of the tool. Contributing to this phase of the success story was the solid support of line management, who worked side by side with the team. As the project gained momentum, technical and management staff from across Ontario Power Generation (OPG) assisted in the design, development and procurement processes. Compared to previous campaigns at various CANDU stations, the collective dose rates received during the Pickering B SFCR were well-below industry norms. Previous campaigns at both Darlington and the Bruce site resulted in 39 and 38 person-rem (0.39 and 0.38 p- Sv), respectively. The collective dose for this project was 26.5 person-rem (0.265 p-Sv) versus a target of 31.3 person-rem (0.313 p-Sv). This translates approximately into a five person-rem (0.05 p-Sv) and $200 000 cost savings. The capital expense for acquiring these REACTORshield shielding tools was $150 000. The gain is obvious and will be significant when future reactor face projects are taken into consideration. For instance, this tool was further utilized for the Feeder Replacement project during the Pickering A P711 outage. In summary, the tool has consistently demonstrated a reduction of general gamma fields by 40 to 60 per cent. (author)

  8. 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.

  9. Calibration of RB reactor power

    International Nuclear Information System (INIS)

    The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8√2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation

  10. 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)

  11. Design and validation of the THMC China-Mock-Up test on buffer material for HLW disposal

    Directory of Open Access Journals (Sweden)

    Yuemiao Liu

    2014-04-01

    Full Text Available According to the preliminary concept of the high-level radioactive waste (HLW repository in China, a large-scale mock-up facility, named China-Mock-Up was constructed in the laboratory of Beijing Research Institute of Uranium Geology (BRIUG. A heater, which simulates a container of radioactive waste, is placed inside the compacted Gaomiaozi (GMZ-Na-bentonite blocks and pellets. Water inflow through the barrier from its outer surface is used to simulate the intake of groundwater. The numbers of water injection pipes, injection pressure and the insulation layer were determined based on the numerical modeling simulations. The current experimental data of the facility are herein analyzed. The experiment is intended to evaluate the thermo-hydro-mechano-chemical (THMC processes occurring in the compacted bentonite-buffer during the early stage of HLW disposal and to provide a reliable database for numerical modeling and further investigation of engineered barrier system (EBS, and the design of HLW repository.

  12. Development of 3D full-core ERANOS-2.2/MCNPX-2.7.0 models and neutronic analysis of the BFS-2 zero-power facility

    International Nuclear Information System (INIS)

    The present paper is addressing the development and validation against experimental data of 3D full-core models of the BFS-2 zero-power fast-reactor using both the deterministic system code ERANOS-2.2 and the stochastic code MCNPX-2.7.0. The model configuration of BFS considered for analysis is the BFS-62-3A benchmark. To extend the - deterministic/stochastic - code-to-code comparison, neutronic parameters, i.e. reactivity, neutron spectrum and reaction rates, were also simulated at the cell level with the Monte Carlo code SERPENT-1.1.7 with two modern data libraries, ENDF-B/VII and JEFF-3.1.1. The BFS-2 critical zero-power facility at the Inst. of Physics and Power Engineering (IPPE) was designed for simulations of the core and shielding of sodium-cooled, fast reactors, for neutron data validation and comparison with experimental results. At the BFS-2 facility, the BFS-62-3A critical benchmark experiment was set-up as a mock-up of the BN-600 reactor, with hybrid MOX fuel and stainless steel reflectors. A UO2 blanket and a large non-homogeneous stainless-steel reflector surround the core. The lattice is hexagonal of pitch 5.1 cm and metallic dowels are used to keep in central position cylindrical rods made of different types of material (fissile, fertile, blanket, plenum, shielding and absorber). A typical subassembly is formed in piling up various pellets of about 1 cm in height and 4.6 cm in diameter, conferring large heterogeneity in the axial direction. The full-core model development was a complex task due to the large number of subassemblies and the axial subassembly heterogeneity. In ERANOS-2.2, it was necessary to homogenize axially per region the pellets used to form the subassembly. The self-shielded macroscopic cross-sections were calculated using the cell code ECCO in association with JEFF-3.1 and ENDF/B-VI.8 data libraries. The core calculations were performed with broad cross-sections data in 33 neutron energy groups with the solver AVNM in the

  13. Development of 3D full-core ERANOS-2.2/MCNPX-2.7.0 models and neutronic analysis of the BFS-2 zero-power facility

    Energy Technology Data Exchange (ETDEWEB)

    Girardin, G.; Alonso, M. [Ecole Polytechnique Federale de Lausanne EPFL, CH-1015 Lausanne (Switzerland); Mikityuk, K. [Paul Scherrer Institut PSI, CH-5232 Villigen-PSI (Switzerland)

    2012-07-01

    The present paper is addressing the development and validation against experimental data of 3D full-core models of the BFS-2 zero-power fast-reactor using both the deterministic system code ERANOS-2.2 and the stochastic code MCNPX-2.7.0. The model configuration of BFS considered for analysis is the BFS-62-3A benchmark. To extend the - deterministic/stochastic - code-to-code comparison, neutronic parameters, i.e. reactivity, neutron spectrum and reaction rates, were also simulated at the cell level with the Monte Carlo code SERPENT-1.1.7 with two modern data libraries, ENDF-B/VII and JEFF-3.1.1. The BFS-2 critical zero-power facility at the Inst. of Physics and Power Engineering (IPPE) was designed for simulations of the core and shielding of sodium-cooled, fast reactors, for neutron data validation and comparison with experimental results. At the BFS-2 facility, the BFS-62-3A critical benchmark experiment was set-up as a mock-up of the BN-600 reactor, with hybrid MOX fuel and stainless steel reflectors. A UO{sub 2} blanket and a large non-homogeneous stainless-steel reflector surround the core. The lattice is hexagonal of pitch 5.1 cm and metallic dowels are used to keep in central position cylindrical rods made of different types of material (fissile, fertile, blanket, plenum, shielding and absorber). A typical subassembly is formed in piling up various pellets of about 1 cm in height and 4.6 cm in diameter, conferring large heterogeneity in the axial direction. The full-core model development was a complex task due to the large number of subassemblies and the axial subassembly heterogeneity. In ERANOS-2.2, it was necessary to homogenize axially per region the pellets used to form the subassembly. The self-shielded macroscopic cross-sections were calculated using the cell code ECCO in association with JEFF-3.1 and ENDF/B-VI.8 data libraries. The core calculations were performed with broad cross-sections data in 33 neutron energy groups with the solver AVNM in the

  14. Heat transfer characteristics evaluation of heat exchangers of mock-up test facility with full-scale reaction tube for HTTR hydrogen production system (Contract research)

    International Nuclear Information System (INIS)

    Connection of hydrogen production system by steam reforming of methane to the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Research Institute (JAERI) has been surveyed until now. Mock-up test facility of this steam reforming system with full-scale reaction tube was constructed in FY 2001 and hydrogen of 120 Nm3/h was successfully produced in overall performance test. Totally 7 times operational tests were performed from March 2002 to December 2004. A lot of operational test data on heat exchanges were obtained in these tests. In this report specifications and structures of steam reformer, steam superheater, steam generator, condenser, helium gas cooler, feed gas heater and feed gas superheater were described. Heat transfer correlation equations for inside and outside tube were chosen from references. Spreadsheet programs were newly made to evaluate heat transfer characteristics from measured test data such as inlet and outlet temperature pressure and flow-rate. Overall heat-transfer coefficients obtained from the experimental data were compared and evaluated with the calculated values with heat transfer correlation equation. As a result, actual measurement values of all heat exchangers gave close agreement with the calculated values with correlation equations. Thermal efficiencies of the heat exchangers were adequate as they were well accorded with design value. (author)

  15. Automatic reactor power control device

    International Nuclear Information System (INIS)

    Anticipated transient without scram (ATWS) of a BWR type reactor is judged to generate a signal based on a reactor power signal and a scram actuation demand signal. The ATWS signal and a predetermined water level signal to be generated upon occurrence of ATWS are inputted, and an injection water flow rate signal exhibiting injection water flow rate optimum to reactor flooding and power suppression is outputted. In addition, a reactor pressure setting signal is outputted based on injection performance of a high pressure water injection system or a lower pressure water injection system upon occurrence of ATWS. Further, the reactor pressure setting signal is inputted to calculate opening/closing setting pressure of a main steam relief valve and output an opening setting pressure signal and a closure setting pressure signal for the main steam relief valve. As a result, the reactor power and the reactor water level can be automatically controlled even upon occurrence of ATWS due to failure of insertion of all of the control rods thereby enabling to maintain integrity and safety of the reactor, the reactor pressure vessel and the reactor container. (N.H.)

  16. Qualification of the numerical simulation of a core disruptive accident on the mars mock-up

    International Nuclear Information System (INIS)

    In case of a Hypothetical Core Disruptive Accident (HCDA) in a Liquid Metal Reactor, the interaction between fuel and liquid sodium creates a high pressure gas bubble in the core. The violent expansion of this bubble loads the vessel and the internal structures, whose deformation is important. A simulation was undertaken using the fluid-structure improvements and the description of the peripheral structures (heat exchangers and pumps) by means of the porosity model. This paper presents the comparison of the results of the third numerical simulation with the experimental results and the numerical results of the previous simulations, as well as a synthesis of all the results of the simulation. (authors)

  17. Construction of PREMUX and preliminary experimental results, as preparation for the HCPB breeder unit mock-up testing

    International Nuclear Information System (INIS)

    Highlights: • PREMUX has been constructed as preparation for a future out-of-pile thermo-mechanical qualification of a HCPB breeder unit mock-up. • The rationale and constructive details of PREMUX are reported in this paper. • PREMUX serves as a test rig for the new heater system developed for the HCPB-BU mock-up. • PREMUX will be used as benchmark for the thermal and thermo-mechanical models developed in ANSYS for the pebble beds of the HCPB-BU. • Preliminary results show the functionality of PREMUX and the good agreement of the measured temperatures with the thermal model developed in ANSYS. - Abstract: One of the European blanket designs for ITER is the Helium Cooled Pebble Bed (HCPB) blanket. The core of the HCPB-TBM consists of so-called breeder units (BUs), which encloses beryllium as neutron multiplier and lithium orthosilicate (Li4SiO4) as tritium breeder in form of pebble beds. After the design phase of the HCPB-BU, a non-nuclear thermal and thermo-mechanical qualification program for this device is running at the Karlsruhe Institute of Technology. Before the complex full scale BU testing, a pre-test mock-up experiment (PREMUX) has been constructed, which consists of a slice of the BU containing the Li4SiO4 pebble bed. PREMUX is going to be operated under highly ITER-relevant conditions and has the following goals: (1) as a testing rig of new heater concept based on a matrix of wire heaters, (2) as benchmark for the existing finite element method (FEM) codes used for the thermo-mechanical assessment of the Li4SiO4 pebble bed, and (3) in situ measurement of thermal conductivity of the Li4SiO4 pebble bed during the tests. This paper describes the construction of PREMUX, its rationale and the experimental campaign planned with the device. Preliminary results testing the algorithm used for the temperature reconstruction of the pebble bed are reported and compared qualitatively with first analyses completed with the FEM codes

  18. Cell heterogeneity problems in the analysis of zero power experiments

    International Nuclear Information System (INIS)

    Methods are described for treating plate and pin cell heterogeneity in the preparation of broad group cross-sections used in the analysis of zero power fast reactor experiments. Methods used at Karlsruhe and Winfrith are summarised and compared, with particular reference to the treatment of resonance shielding, the calculation of broad group spatial fine structure, the treatment of leakage and the calculation of anisotropic diffusion coefficients. The problems of cells near boundaries such as core-breeder interfaces and of singularities such as control rods are also considered briefly. Numerical studies carried out to investigate approximations in the methods are described. These include tests of the accuracy of one-dimensional cell modelling techniques, and the validation by Monte Carlo of methods for treating streaming in the calculation of diffusion coefficients. Comparisons are shown between the heterogeneity effects calculated by the Karlsruhe and Winfrith methods for typical pin and plate cells used in the BIZET experimental programme, and their effect in a whole reactor calculation is indicated. Comparisons are given with measurements which provide tests of the heterogeneity calculations. These include reaction rate scans within pin and plate cells, and reaction rate measurements across sectors of pin and plate fuel, where the flux tilt is determined by the relative reactivity of the pin and plate cells. Finally, the heterogeneity problems arising in the interpretation of reaction rate measurements are discussed. (author)

  19. The Use of Prestressed Concrete Vessels in the French Power Reactor Programme

    International Nuclear Information System (INIS)

    This paper deals with the use of pre-stressed concrete for the G2 and G3 reactors at Marcoule and for the EDF3 reactor now under construction at Chinon. The first two reactors have been operating at power since 1959 and 1960 respectively. Messrs. Conte and Dambrine discuss the problems that arose during construction of the vessels for G2 and G3 and also deal with the experience gained in operation - experience which suggests that they are extremely safe- Work on the EDF3 vessel, begun at Chinon in the second half of 1961, is still under way and should be finished towards the end of 1963. Mr. Gaussot discusses the reasons for choosing this type of vessel, the results of calculations and mock-up tests, and the problems presented by the construction itself. A number of studies have been devoted to the future prospects of prestressed concrete structures for reactors. It would seem that working pressures could be increased, if desired, and, in any case, that dimensions could be considerably enlarged, thus offering the chance of integral-type solutions. (author)

  20. 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.

  1. Fractals in Power Reactor Noise

    International Nuclear Information System (INIS)

    In this work the non- lineal dynamic problem of power reactor is analyzed using classic concepts of fractal analysis as: attractors, Hausdorff-Besikovics dimension, phase space, etc. A new non-linear problem is also analyzed: the discrimination of chaotic signals from random neutron noise signals and processing for diagnosis purposes. The advantages of a fractal analysis approach in the power reactor noise are commented in details

  2. Power reactor information system (PRIS)

    International Nuclear Information System (INIS)

    Since the very beginning of commercial operation of nuclear power plants, the nuclear power industry worldwide has accumulated more than 5000 reactor years of experience. The IAEA has been collecting Operating Experience data for Nuclear Power Plants since 1970 which were computerized in 1980. The Agency has undertaken to make Power Reactor Information System (PRIS) available on-line to its Member States. The aim of this publication is to provide the users of PRIS from their terminals with description of data base and communication systems and to show the methods of accessing the data

  3. Experimental results of passive vibro-acoustic leak detection in SFR steam generator mock-up

    International Nuclear Information System (INIS)

    Regarding to GEN 4 context, it is necessary to fulfil the high safety standards for sodium fast reactors (SFR), particularly against water-sodium reaction which may occur in the steam generator units (SGU) in case of leak. This reaction can cause severe damages in the component in a short time. Detecting such a leak by visual in-sodium inspection is impossible because of sodium opacity. Hydrogen detection is then used but the time response of this method can be high in certain operating conditions. Active and passive acoustic leak detection methods were studied before SUPERPHENIX plant shutdown in 1997 to detect a water-into-sodium leak with a short time response. In the context of the new R and D studies for SFR, an innovative passive vibro-acoustic method is developed in the framework of a Ph.D. thesis to match with GEN 4 safety requirements. The method consists in assuming that a small leak emits spherical acoustic waves in a broadband frequency domain, which propagate in the liquid sodium and excite the SGU cylindrical shell. These spatially coherent waves are supposed to be buried by a spatially incoherent background noise. The radial velocities of the shell is measured by an array of accelerometers positioned on the external envelop of the SGU and a beam forming treatment is applied to increase the signal-to-noise ratio (SNR) and to detect and localize the acoustic source. Previous numerical experiments were achieved and promising results were obtained. In this paper, experimental results of the proposed passive vibro-acoustic leak detection are presented. The experiment consists in a cylindrical water-filled steel pipe representing a model of SGU shell without tube bundle. A hydro-phone emitting an acoustic signal is used to simulate an acoustic monopole. Spatially uncorrelated noise or water-flow induced shell vibrations are considered as the background noise. The beam-forming method is applied to vibration signals measured by a linear array of

  4. Development and material testing of OF-Cu/DS-Cu/OF-Cu triplex tube (dispersion strengthened copper clad with oxygen free-copper) and trial fabrication of a vertical target mock-up for ITER divertor

    International Nuclear Information System (INIS)

    For the divertor target of the international thermonuclear experimental reactor (ITER), an OF-Cu/DS-Cu/OF-Cu triplex-structured cooling tube has been newly fabricated through powder metallurgy and drawing. The triplex structure comprised an aluminium oxide (0.5 mass%)-dispersion strengthened copper core (DS-Cu) clad with oxygen free copper (OF-Cu), a compliant layer for joining to the carbon fiber composite (CFC) tiles, and with an inner skin which tightly grasps a twisted INCONEL tape to assist heat transfer. Physical and mechanical properties of the DS-Cu core after heat treatment at 850 C for 600 s were investigated. Also, CFC brazability, fabricability and feasibility of the triplex tube for cooling channels for the divertor target were studied: A large scale vertical target mock-up of a 1500 mm length, 35 mm width and 3000 mm radius curved front face, has been fabricated with nearly 50 pieces of ''saddle''-shaped one-dimensional (1D)-CFC tiles were brazed on to 1500 mm long triplex tubes set in grooves of OF-Cu heat sink blocks joined to a stainless-steel back plate. The mock-up was tested under 20 MW/m2 for 15 s for 1000 cycle thermal loadings, which simulated transient heat loadings of a vertical target of an ITER divertor. (orig.)

  5. High heat flux tests of small-scale Be/Cu mock-ups for ITER

    International Nuclear Information System (INIS)

    Several kinds of Be/Cu joints have been made by hot isostatic press (HIP) in China in order to develop the ITER-FW blanket fabrication technology. At the first stage, high temperature HIP technology was investigated, and both Ti film and PVD (physical vapor deposition)-coating were adopted as intermediate layers between high purity beryllium made by HIP and CuCrZr alloy. The average bonding strength of Be/CuCrZr joints is larger than 60 Mpa and a good metallurgical bonding was formed. The Be/CuCrZr joints at optimized technology can sustain 1000 cycles under an absorbed power density of about 2.5 MW·m-2, which shows relatively good thermal fatigue properties. Temperature and stress distributions were also calculated by 2D ANSYS, showing a good accord with experimental results. Low temperature HIP joining is being developed and the heat load evaluation is also under way. (author)

  6. Detection of Signals of Mock-up Pipes of Carbon Steel and Stainless Steel using Guided Ultrasonic Waves due to Magnetostrictive Sensors

    International Nuclear Information System (INIS)

    A piping mock-up with a diameter of 6 inch and schedule number 80 of carbon steel and stainless steel were fabricated. The signals of weldments of these pipes were detected with a torsional vibration mode of frequency of 32 kHz using sensors, such as a pure Ni or a 49Fe-49Co-2V alloy strip. The signals from the 49Fe-49Co-2V alloy strip sensor were more detectable than those from the Ni strip sensor. The signals of 49Fe-49Co-2V alloy strip sensor of tile stainless steel piping mock-up were more detectable than those of 49Fe-49Co-2V alloy strip sensor of the carbon steel piping mock-up.

  7. HHF test with 80x80x1 Be/Cu/SS Mock-ups for verifying the joining technology of the ITER blanket First Wall

    International Nuclear Information System (INIS)

    Through the fabrication of the Cu/SS and Be/Cu joint specimens, fabrication procedure such as material preparation, canning, degassing, HIP (Hot Isostatic Pressing), PHHT (Post HIP heat treatment) was established. The HIP conditions (1050 .deg. C, 100 MPa 2 hr for Cu/SS, 580 .deg. C 100 MPa 2 hr for Be/Cu) were developed through the investigation on joint specimen fabricated with the various HIP conditions; the destructive tests of joint include the microstructure observation of the interface with the examination of the elemental distribution, tension test, bend test, Charpy impact test and fracture toughness test. However, since the joint should be tested under the High Heat Flux (HHF) conditions like the ITER operation for verifying its joint integrity, several HHF tests were performed like the previous HHF test with the Cu/SS, Be/Cu, Be/Cu/SS Mock-ups. In the present study, the HHF test with Be/Cu/SS Mock-ups, which have 80 mm x 80 mm single Be tile and each material depths were kept to be the same as the ITER blanket FW. The Mock-ups fabricated with three kinds of interlayers such as Cr/Ti/Cu, Ti/Cr/Cu, Ti/Cu, which were different from the developed interlayer (Cr/Cu), total 6 Mock-ups were fabricated. Preliminary analysis were performed to decide the test conditions; they were tested with up to 2.5 MW/m2 of heat fluxes and 20 cycles for each Mock-up in a given heat flux. They were tested with JUDITH-1 at FZJ in Germany. During tests, all Mock-ups showed delamination or full detachment of Be tile and it can be concluded that the joints with these interlayers have a bad joining but it can be used as a good data for developing the Be/Cu joint with HIP

  8. HHF test with 80x80x1 Be/Cu/SS Mock-ups for verifying the joining technology of the ITER blanket First Wall

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Bae, Young Dug; Kim, Suk Kwon; Hong, Bong Guen; Jeong, Yong Hwan; Park, Jeong Yong; Choi, Byung Kwon; Jung, Hyun Kyu

    2008-11-15

    Through the fabrication of the Cu/SS and Be/Cu joint specimens, fabrication procedure such as material preparation, canning, degassing, HIP (Hot Isostatic Pressing), PHHT (Post HIP heat treatment) was established. The HIP conditions (1050 .deg. C, 100 MPa 2 hr for Cu/SS, 580 .deg. C 100 MPa 2 hr for Be/Cu) were developed through the investigation on joint specimen fabricated with the various HIP conditions; the destructive tests of joint include the microstructure observation of the interface with the examination of the elemental distribution, tension test, bend test, Charpy impact test and fracture toughness test. However, since the joint should be tested under the High Heat Flux (HHF) conditions like the ITER operation for verifying its joint integrity, several HHF tests were performed like the previous HHF test with the Cu/SS, Be/Cu, Be/Cu/SS Mock-ups. In the present study, the HHF test with Be/Cu/SS Mock-ups, which have 80 mm x 80 mm single Be tile and each material depths were kept to be the same as the ITER blanket FW. The Mock-ups fabricated with three kinds of interlayers such as Cr/Ti/Cu, Ti/Cr/Cu, Ti/Cu, which were different from the developed interlayer (Cr/Cu), total 6 Mock-ups were fabricated. Preliminary analysis were performed to decide the test conditions; they were tested with up to 2.5 MW/m2 of heat fluxes and 20 cycles for each Mock-up in a given heat flux. They were tested with JUDITH-1 at FZJ in Germany. During tests, all Mock-ups showed delamination or full detachment of Be tile and it can be concluded that the joints with these interlayers have a bad joining but it can be used as a good data for developing the Be/Cu joint with HIP.

  9. Low-power nuclear reactors

    International Nuclear Information System (INIS)

    A brief development history of low-power nuclear reactors is presented in this paper. Nowadays, some countries have plans to build a series of small nuclear power plants (also floating ones) for use in remote regions. Present constructions of such NPP are presented in this paper. (author)

  10. 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

  11. Power Reactor Information System (PRIS)

    International Nuclear Information System (INIS)

    The IAEA has been collecting Operating Experience data for Nuclear Power Plants of the IAEA Member States since 1970. In order to facilitate an analysis of nuclear power plant performance as well as to produce relevant publications, all previously collected data supplied from the questionnaires were computerized in 1980 and the Power Reactor Information System was implemented. PRIS currently contains production records for the years up to and including 1990 and about 98% of the reactors-years operating experience in the world is contained in PRIS. (orig.)

  12. 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.

  13. Zero Power Physics Test by using Core Simulator(Co Si) for OPR1000

    International Nuclear Information System (INIS)

    Core management staff performs ZPPT at least every one and half year. So periodic training for ZPPT should be required to minimize human error in that test and to be well acquainted with test procedures. Co Si could provide environments for readiness of all test procedure of initial OPR1000 core. Co Si has been developed for the training of core management staff in the area of zero power physics test in order to enhance the capacity of core management personnel in OPR1000. Through this training, reactor parameters in zero power physics test are determined as the same way as commercial nuclear power plant's test procedure before real reactor physics test. Training test consists of confirming initial critical approach, point of adding heat, boron end point and measuring isothermal temperature coefficient and rod bank worth in a various way

  14. Cascade ICF power reactor

    International Nuclear Information System (INIS)

    The double-cone-shaped Cascade reaction chamber rotates at 50 rpm to keep a blanket of ceramic granules in place against the wall as they slide from the poles to the exit slots at the equator. The 1 m-thick blanket consists of layers of carbon, beryllium oxide, and lithium aluminate granules about 1 mm in diameter. The x rays and debris are stopped in the carbon granules; the neutrons are multiplied and moderated in the BeO and breed tritium in the LiAlO2. The chamber wall is made up of SiO tiles held in compression by a network of composite SiC/Al tendons. Cascade operates at a 5 Hz pulse rate with 300 MJ in each pulse. The temperature in the blanket reaches 1600 K on the inner surface and 1350 K at the outer edge. The granules are automatically thrown into three separate vacuum heat exchangers where they give up their energy to high pressure helium. The helium is used in a Brayton cycle to obtain a thermal-to-electric conversion efficiency of 55%. Studies have been done on neutron activation, debris recovery, vaporization and recondensation of blanket material, tritium control and recovery, fire safety, and cost. These studies indicate that Cascade appears to be a promising ICF reactor candidate from all standpoints. At the 1000 MWe size, electricity could be made for about the same cost as in a future fission reactor

  15. Preparing ITER ICRF: development and analysis of the load resilient matching systems based on antenna mock-up measurements

    International Nuclear Information System (INIS)

    The reference design for the ICRF antenna of ITER is constituted by a tight array of 24 straps grouped in eight triplets. The matching network must be load resilient for operation in ELMy discharges and must have antenna spectrum control for heating or current drive operation. The load resilience is based on the use of either hybrid couplers or conjugate-T circuits. However, the mutual coupling between the triplets at the low expected loading strongly counteracts the load resilience and the spectrum control. Using a mock-up of the ITER antenna array with adjustable water load matching solutions are designed. These solutions are derived from transmission line modelling based on the measured scattering matrix and are finally tested. We show that the array current spectrum can be controlled by the anti-node voltage distribution and that suitable decoupler circuits can not only neutralize the adverse mutual coupling effects but also monitor this anti-node voltage distribution. A matching solution using four 3 dB hybrids and the antenna current spectrum feedback control by the decouplers provides outstanding performance if each pair of poloidal triplets undergoes a same load variation. Finally, it is verified by modelling that this matching scenario has the same antenna spectrum and load resilience performances as the antenna array loaded by plasma as described by the TOPICA simulation. This is true for any phasing and frequency in the ITER frequency band. The conjugate-T solution is presently considered as a back-up option.

  16. Enhancement of the use of digital mock-ups in the verification and validation process for ITER remote handling systems

    Energy Technology Data Exchange (ETDEWEB)

    Sibois, R., E-mail: romain.sibois@vtt.fi [VTT Technical Research Centre of Finland, P.O. Box 1300, 33101 Tampere (Finland); Salminen, K.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, 33101 Tampere (Finland); Mattila, J. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T. [VTT Technical Research Centre of Finland, P.O. Box 1300, 33101 Tampere (Finland)

    2013-10-15

    Highlights: • Verification and validation process for ITER remote handling system. • Verification and validation framework for complex engineering systems. • Verification and validation roadmap for digital modelling phase. • Importance of the product life-cycle management in the verification and validation framework. -- Abstract: The paper is part of the EFDA's programme of European Goal Oriented Training programme on remote handling (RH) “GOT-RH”. The programme aims to train engineers for activities supporting the ITER project and the long-term fusion programme. This paper is written based on the results of a project “verification and validation (V and V) of ITER RH system using digital mock-ups (DMUs)”. The purpose of this project is to study efficient approach of using DMU for the V and V of the ITER RH system design utilizing a system engineering (SE) framework. This paper reviews the definitions of DMU and virtual prototype and overviews the current trends of using virtual prototyping in the industry during the early design phase. Based on the survey of best industrial practices, this paper proposes ways to improve the V and V process for ITER RH system utilizing DMUs.

  17. Enhancement of the use of digital mock-ups in the verification and validation process for ITER remote handling systems

    International Nuclear Information System (INIS)

    Highlights: • Verification and validation process for ITER remote handling system. • Verification and validation framework for complex engineering systems. • Verification and validation roadmap for digital modelling phase. • Importance of the product life-cycle management in the verification and validation framework. -- Abstract: The paper is part of the EFDA's programme of European Goal Oriented Training programme on remote handling (RH) “GOT-RH”. The programme aims to train engineers for activities supporting the ITER project and the long-term fusion programme. This paper is written based on the results of a project “verification and validation (V and V) of ITER RH system using digital mock-ups (DMUs)”. The purpose of this project is to study efficient approach of using DMU for the V and V of the ITER RH system design utilizing a system engineering (SE) framework. This paper reviews the definitions of DMU and virtual prototype and overviews the current trends of using virtual prototyping in the industry during the early design phase. Based on the survey of best industrial practices, this paper proposes ways to improve the V and V process for ITER RH system utilizing DMUs

  18. Thermal fatigue testing of a diffusion-bonded beryllium divertor mock-up under ITER relevant conditions

    International Nuclear Information System (INIS)

    Thermal response and thermal fatigue tests of four 5 mm thick beryllium tiles on a Russian divertor mock-up were completed on the Electron Beam Test System at Sandia National Laboratories. The beryllium tiles were diffusion bonded onto an OFHC copper saddleblock and a DSCu (MAGT) tube containing a porous coating. Thermal response tests were performed on the tiles to an absorbed heat flux of 5 MW/m2 and surface temperatures near 300 degrees C using 1.4 MPa water at 5.0 m/s flow velocity and an inlet temperature of 8-15 degrees C. One tile was exposed to incrementally increasing heat fluxes up to 9.5 MW/m2 and surface temperatures up to 690 degrees C before debonding at 10 MW/m2. A third tile debonded after 9200 thermal fatigue cycles at 5 MW/m2, while another debonded after 6800 cycles. In all cases, fatigue failure occurred in the intermetallic layers between the beryllium and copper. No fatigue cracking of the bulk beryllium was observed. During thermal cycling, a gradual loss of porous coating produced increasing sample temperatures. These experiments indicate that diffusion-bonded beryllium tiles can survive several thousand thermal cycles under ITER relevant conditions without failure. However, the reliability of the diffusion bonded Joint remains a serious issue

  19. Nuclear reactor power control device

    International Nuclear Information System (INIS)

    When occurrence of earthquakes is judged in a BWR type reactor, the power is decreased by inserting a portion of control rods, reducing a speed of recycling pumps, stopping recycling pumps, increasing the opening degree of a main steam control valve and opening a main steam relief valve. The reactor scram can be avoided by bypassing neutron flux high signal, settling a filter to neutron flux signals and setting a reactor scram set value by neutron flux signals, for example, to 120%. There is constituted an interlock for performing reactor scram when both of a neutron flux high signal and a signal outputted if a surface heat flux corresponding signal formed by applying calculation to the neutron flux high signal exceeds a set value are valid, to avoid unnecessary reactor scram. As a measuring means, not only an acceleration meter in the power plant, but also acceleration meters at remote places, acceleration meters or displacement meters for various kinds of equipments in the power plant are used, and when signals from them exceed set values, earthquake judgement is conducted. (N.H.)

  20. Low power unattended defense reactor

    International Nuclear Information System (INIS)

    A small, low power, passive, nuclear reactor electric power supply has been designed for unattended defense applications. Through innovative utilization of existing proven technologies and components, a highly reliable, walk-away safe design has been obtained. Operating at a thermal power level of 200 kWt, the reactor uses low enrichment uranium fuel in a graphite block core to generate heat that is transferred through heat pipes to a thermoelectric (TE) converter. Waste heat is removed from the TEs by circulation of ambient air. Because such a power supply offers the promise of minimal operation and maintenance (O and M) costs as well as no fuel logistics, it is particularly attractive for remote, unattended applications such as the North Warning System

  1. Power Reactor Embrittlement Data Base

    International Nuclear Information System (INIS)

    Regulatory and research evaluations of embrittlement predication models and of pressure vessel integrity can be greatly expedited by the use of a well-designed, computerized data base. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The Nuclear Regulatory Commission (NRC) has provided financial support, and the Electric Power Research Institute (EPRI) has provided technical assistance in the quality assurance (QA) of the data to establish an industry-wide data base that will be maintained and updated on a long-term basis. Successful applications of the data base to several of NRC's evaluations have received favorable response and support for its continuation. The future direction of the data base has been designed to include the test reactor and other types of data of interest to the regulators and the researchers. 1 ref

  2. Experimental power reactor

    International Nuclear Information System (INIS)

    The following five topics are discussed using figures and diagrams: (1) energy storage and transfer program, (2) thermomechanical analysis, (3) a steam dual-cycle power conversion system for the EPR, (4) EPR tritium facility scoping studies, and (5) vacuum systems

  3. Tokamak experimental power reactor studies

    International Nuclear Information System (INIS)

    The principal results of a scoping and project definition study for the Tokamak Experimental Power Reactor are presented. Objectives are discussed; a preliminary conceptual design is described; detailed parametric, survey and sensitivity studies are presented; and research and development requirements are outlined. (U.S.)

  4. Response to high heat fluxes and metallurgical examination of a brazed carbon-fiber-composite/refractory-metal divertor mock-up

    International Nuclear Information System (INIS)

    As a feasibility-study an actively cooled divertor mock-up has been subjected to high heat flux loading in electron beam simulation. The divertor design concept is based on a carbon-fiber-composite material (Aerolor 05) brazed onto a TZM/Mo41Re heat sink. The plasma facing carbon armor is divided in seven tiles to allow variable loading parameters - and repeated destructive tests. The mock-up has survived high heat flux loading up to about 12 MW/m2 surface heat flux in steady-state conditions. One armor tile showed no change in the thermal response even after 500 s at ∝14 MW/m2. To estimate the general thermal response of the mock-up design, numerical methods were applied. The predicted behavior was confirmed by the experimental results. The loading experiments were followed by a detailed metallurgical investigation of the loaded sample regions and the braze joints. The typical damages after high heat flux testing and cycling were failure (i.e. detachment) in the Zr brazed carbon/TZM joint, and failure in the CuPd bonded TZM/TZM joint due to an excess of the melting temperature of the brazes. The microstructural changes in the braze regions and the recrystallization behavior of the refractory alloys are discussed. Only in one case the loaded surface of the carbon armor shows considerable erosion, caused by a partial detachment along a braze joint and thus loss of the good thermal contact during the last applied loading shots. The thermal analyses and high heat flux performance of the Aerolor-05 armored mock-up are compared to the thermal response of a previously tested mock-up of corresponding geometry with armor tiles of isotropic graphite. (orig.)

  5. Design and setup of a testing device to investigate a reduced sized attachment system mock up for the ITER EU HCPB-TBM under different mechanical loading conditions

    International Nuclear Information System (INIS)

    Highlights: • A measurement system is developed to measure the electro-magnetic forces acting on the ITER EU HCPB TBM during operation. • A system based on the measurement of strain signals in combination with two force reconstruction methods is proposed. • Two mock ups have been designed to test and calibrate the measurement system. • A suitable testing device has been built to test the measuring system. • First results with simulated strain data are presented. - Abstract: In order to determine the forces acting on the EU-Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) during operation, a measurement system is developed. Therefore, two force reconstruction (FR) methods using measured strain signals are selected that are suitable for the application to the TBM. The first one, the augmented Kalman filter is a combined deterministic-stochastic approach. A second FR method based on the concept of a model predictive controller is proposed in this paper, which uses an optimization algorithm. In order to test the selected methods a testing device has been built which can be used to apply different force excitations on a reduced sized TBM mock up and measure the resulting strain signals of 16 strain gages. A simple tube mock up has been designed and manufactured to test and calibrate the FR algorithms. In addition, a second TBM mock up with attachment system is described. Finally, first results of the FR of a worst-case test case from simulated strain data of the simple tube mock up are presented

  6. Tokamak power reactor ignition and time dependent fractional power operation

    International Nuclear Information System (INIS)

    A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transport power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve

  7. Magnet powering with zero downtime - a dream?

    CERN Document Server

    Zerlauth, Markus

    2012-01-01

    Despite a number of improvements already applied in the course of the year, the magnet powering system of the LHC still accounts for around 50% of the premature beam dumps. This number might even further increase when moving to higher beam energies in the next years. With mitigations of radiation effects and the prospects for beam induced magnet quenches being discussed elsewhere, we aim at identifying possible mid- and long-term improvements within the various equipment systems to further reduce the number of equipment failures leading to a loss of the particle beams. Amongst others, this includes the sensitivity of equipment to external causes such as electromagnetic perturbations or perturbations on the electrical network. To conclude, the gain of the identified mitigations will have to be balanced against the potential impact on schedule and cost.

  8. Measurement of residual stresses in the dissimilar metal weld joint of a safe-end nozzle mock-up

    International Nuclear Information System (INIS)

    Knowledge of the origin, magnitude and distribution of residual stresses generated during the manufacture of nuclear power plants is of vital importance to their structural integrity assessment. The overall aim of this work was to measure welding residual stresses in components prone to primary water stress corrosion cracking in nuclear reactor pressure vessels. This paper describes the on-site application of the Deep-Hole Drilling (DHD) technique to measure the through-thickness residual stress distributions through a safe-end nozzle component containing a dissimilar metal weld joint at different stages of manufacture

  9. TRIGA Reactor Power Upgrading Analysis

    International Nuclear Information System (INIS)

    Reactor physics safety analysis supporting the power upgrading from 1MW to 2MW of a typical TRIGA Mark II reactor is presented for steady state and pulse operation. The analysis is performed for mixed core configuration consisting of two types of fuel elements: standard 8,5% or 12% stainless-steel clad fuel elements and LEU fuel elements (20% uranium concentration). The following reactor physics codes are applied: WIMS, TRIGAC, EXTERMINATOR, PULSTRI and TRISTAN. Results of the calculations are compared to experiments for steady state operation at 1 MW. The analysis shows that besides technical modifications of the core (installation of an additional control rod) also some strict administrative limitations have to be imposed on operational parameters (excess reactivity, pulse reactivity, core composition) to assure safe operation within design limits. (author)

  10. Measurement of prompt neutron multiplication on the zero power

    International Nuclear Information System (INIS)

    Using a set of ST-PMT detector, the prompt neutron multiplication on the zero power is measured by the reliable and effective shielding methods. The results are compared with the results of Rossi-α method and 252Cf fission chamber technique. (authors)

  11. Neutronics of nuclear power reactors

    International Nuclear Information System (INIS)

    This review, prepared on the occasion of 25th ETAN Conference describes the research activities in the field of neutronics which started in 1947. A number of researchers in Yugoslav Institutes was engaged in development of neutronics theory and calculation methods related to power reactors since 1960. To illustrate the activities of Yugoslav authors, this review contains the list of the most important relevant papers published in international journals

  12. 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.)

  13. 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

  14. ZERO SET OF SOBOLEV FUNCTIONS WITH NEGATIVE POWER OF INTEGRABILITY

    Institute of Scientific and Technical Information of China (English)

    姜惠强; 林芳华

    2004-01-01

    Here the authors are interested in the zero set of Sobolev functions and functions of bounded variation with negative power of integrability. The main result is a general Hausdorff dimension estimate on the size of zero set. The research is motivated by the model on van der waal force driven thin film, which is a singular elliptic equation.After obtaining some basic regularity result, the authors get an estimate on the size of singular set; such set corresponds to the thin film rupture set in the thin film model.

  15. Use of mock-up training to reduce personnel exposure at the North Anna Unit 1 Steam Generator Replacement Project

    International Nuclear Information System (INIS)

    The North Anna Power Station is located on the southern shore of Lake Anna in Louisa County, approximately forty miles northwest of Richmond, Virginia. The two 910 Mw nuclear units located on this site are owned by Virginia Electric and Power Company (Virginia Power) and Old Dominion Electric Cooperative and operated by Virginia Power. Fuel was loaded into Unit 1 in December 1977, and it began commercial operation in June 1978. Fuel was loaded into Unit 2 in April 1980 and began commercial operation in December 1980. Each nuclear unit includes a three-coolant-loop pressurized light water reactor nuclear steam supply system that was furnished by Westinghouse Electric Corporation. Included within each system were three Westinghouse Model 51 steam generators with alloy 600, mill-annealed tubing material. Over the years of operation of Unit 1, various corrosion-related phenomena had occurred that affected the steam generators tubing and degraded their ability to fulfill their heat transfer function. Advanced inspection and repair techniques helped extend the useful life of the steam generators, but projections based on the results of the inspections indicated that the existing steam generators tubing and degraded their ability to fullfill their heat transfer function. Advanced inspection and repair techniques helped extend the useful life of the steam generators, but projections based on the results of the inspections indicated that the existing steam generators would not last their design life and must be repaired. To this end Virginia Power determined that a steam generator replacement (SGR) program was necessary to remove the old steam generator tube bundles and lower shell sections, including the channel heads (collectively called the lower assemblies), and replace them with new lower assemblies incorporating design features that will prevent the degradation problems that the old steam generators had experienced

  16. Reactor power for space exploration

    International Nuclear Information System (INIS)

    Potential 21st century missions envisioned by National Aeronautics and Space Administration (NASA) planners encompass ambitious, wide-ranging human and robotic solar system exploration objectives and scenarios. A critical common element in many of these future civil space mission initiatives is the ability to generate, with a very high degree of reliability, the considerable amounts of power needed to realize the mission goals. The extended duration and/or high power level requirements for many missions and, in instances, the lack of adequate solar energy flux for others, render the use of versatile nuclear power sources as either missions-enabling or very advantageous. Further, the use of high-performance reactor systems, when coupled with very high impulse electric propulsion systems, can enable or significantly enhance both human near-planets operations and robotic scientific missions to the very farthest reaches of the solar system. It is important that this nation continue to develop the means of acquiring a space reactor power source to ensure availability at such time that approved missions and possibly political considerations warrant its use

  17. Use of nuclear data sensitivity and uncertainty analysis for the design preparation of the HCLL breeder blanket mock-up experiment for ITER

    International Nuclear Information System (INIS)

    An experiment on a mock-up of the Test Blanket module based on Helium Cooled Lithium Lead (HCLL) concept will be performed in 2007 in the FNG utility in Frascati in order to study neutronics characteristics of the module and the performance of the computational tools in the accurate prediction of the neutron transport. With the objective to prepare and optimise the design of the mock-up in the sense to provide maximum information on the state-of-the-art of the cross section data the mock-up was pre-analysed using the deterministic codes for the sensitivity/uncertainty analysis. The neutron fluxes and tritium production rate (TPR), their sensitivity to the underlying basic cross sections, as well as the corresponding uncertainty estimations were calculated using the deterministic transport codes (DOORS package), the sensitivity/uncertainty code package SUSD3D and the VITAMIN-J/COVA covariance matrix libraries. The cross section reactions with largest contribution to the uncertainty in the calculation of the TPR were identified to be (n,2n) and (n,3n) reactions on plumb. The conclusions of this work support the main benchmark design and suggest some modifications and improvements. In particular this study recommends the use, as far as possible, of both natural and enriched lithium pellets for the TRP measurements. The combined use is expected to provide additional and complementary information on the sensitive cross sections. (author)

  18. Tool coupling for the design and operation of building energy and control systems based on the Functional Mock-up Interface standard

    Energy Technology Data Exchange (ETDEWEB)

    Nouidui, Thierry Stephane; Wetter, Michael

    2014-03-01

    This paper describes software tools developed at the Lawrence Berkeley National Laboratory (LBNL) that can be coupled through the Functional Mock-up Interface standard in support of the design and operation of building energy and control systems. These tools have been developed to address the gaps and limitations encountered in legacy simulation tools. These tools were originally designed for the analysis of individual domains of buildings, and have been difficult to integrate with other tools for runtime data exchange. The coupling has been realized by use of the Functional Mock-up Interface for co-simulation, which standardizes an application programming interface for simulator interoperability that has been adopted in a variety of industrial domains. As a variety of coupling scenarios are possible, this paper provides users with guidance on what coupling may be best suited for their application. Furthermore, the paper illustrates how tools can be integrated into a building management system to support the operation of buildings. These tools may be a design model that is used for real-time performance monitoring, a fault detection and diagnostics algorithm, or a control sequence, each of which may be exported as a Functional Mock-up Unit and made available in a building management system as an input/output block. We anticipate that this capability can contribute to bridging the observed performance gap between design and operational energy use of buildings.

  19. Containment Evaluation under Severe Accidents (CESA): synthesis of the predictive calculations and analysis of the first experimental results obtained on the Civaux mock-up

    International Nuclear Information System (INIS)

    In 1996, EDF decided to build a containment model at the scale 1:3, the MAEVA mock-up, in order to check and study the behaviour of a pre-stressed concrete containment vessel without a liner in terms of mechanical strength and leaktightness, for loadings corresponding to its design and beyond design conditions. In parallel with the construction and testing of the mock-up, a cost-shared R and D action supported by the European Union, the CESA project, is dealing with quantification of leak rates through concrete cracks and porosity, predictive calculations of the behaviour of the mock-up and analysis of the experimental results. In this paper, we propose a synthesis of the main theoretical and experimental results, obtained after 2.5 years. It should however be noted that, due to some unexpected delays in the experimental programme, quite natural with such a huge and unique experimental set-up, only the design-basis accident sequences, already performed, have been reported in this paper. The first results are nevertheless very interesting, both from a scientific and nuclear utility point of view

  20. Manufacturing of small-scale mock-ups and of a semi-prototype of the ITER Normal Heat Flux First Wall

    International Nuclear Information System (INIS)

    This paper describes the manufacturing development and fabrication of reduced scale ITER First Wall (FW) mock-ups of the Normal Heat Flux (NHF) design, including a “semi-prototype” with a dimension of 305 mm × 660 mm, corresponding to about 1/6 of a full-scale panel. The activity was carried out in the framework of the pre-qualification of the European Domestic Agency (EU-DA or F4E) for the supply of the European share of the ITER First Wall. The hardware consists of three Upgraded (2 MW/m2) Normal Heat Flux (U-NHF) small-scale mock-ups, bearing 3 beryllium tiles each, and of one Semi-Prototype, representing six full-scale fingers and bearing a total of 84 beryllium tiles. The manufacturing process makes extensive use of Hot Isostatic Pressing, which was developed over more than a decade during ITER Engineering Design Activity phase. The main manufacturing steps for the semi-prototype are described, with special reference to the lessons learned and the implications impacting the future fabrication of the full-scale prototype and the series which consists of 218 panels plus spares. In addition, a “tile-size” mock-up was manufactured in order to assess the performance of larger tiles. The use of larger tiles would be highly beneficial since it would allow a significant reduction of the panel assembly time

  1. 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)

  2. 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)

  3. Initial RattleSnake Calculations of the Hot Zero Power BEAVRS

    Energy Technology Data Exchange (ETDEWEB)

    M. Ellis; J. Ortensi; Y. Wang; K. Smith; R.C. Martineau

    2014-01-01

    The validation of the Idaho National Laboratory's next generation of reactor physics analysis codes is an essential and ongoing task. The validation process requires a large undertaking and includes detailed, realistic models that can accurately predict the behavior of an operational nuclear reactor. Over the past few years the INL has developed the RattleSnake application and supporting tools on the MOOSE framework to perform these reactor physics calculations. RattleSnake solves the linearized Boltzmann transport equation with a variety of solution meth­ ods. Various traditional reactor physics benchmarks have already been performed, but a more realistic light water reactor comparison was needed to solidify the status of the code and deter­ mine its fidelity. The INL team decided to use the Benchmark for Evaluation and Validation of Reactor Simulations, which was made available in early 2013. This benchmark is a one­ of-a-kind document assembled by the Massachusetts Institute of Technology, which includes two cycles of detailed, measured PWR operational data. The results from this initial study of the hot zero power conditions show the current INL analysis procedure with DRAGON4 cross section preparation and using the low order diffusion solver in RattleSnake for the whole core calculations yield very encouraging results for PWR analysis. The radial assembly power distributions, radial detector measurements and control rod worths were computed with good accuracy. The computation of the isothermal temperature coefficients of reactivity require further study.

  4. Prospects for small and medium power reactors

    International Nuclear Information System (INIS)

    A searching examination of the present status of nuclear power technology and economics was made in 64 papers presented to the Conference on Small and Medium Power Reactors held by the IAEA in Vienna during the week 5 - 9 September 1960. The IAEA Conference concentrated on small and medium power reactors because these are the sizes of primary interest to less-developed countries around the world. The Conference brought forward information on a wide range of subjects related to power reactors, including power costs, summaries of national programs, applications in less-developed countries, process heat reactors, reactor safety, results of experience in the actual construction and operation of power reactors and technical appraisals of various reactor types

  5. Estudos etnoictiológicos sobre o pirarucu Arapaima gigas na Amazônia Central Ethnoictiology studies on Pirarucu (Arapaima mock-ups in Central Amazon

    Directory of Open Access Journals (Sweden)

    Liane Galvão de Lima

    2012-09-01

    Full Text Available O presente estudo visou identificar saberes comuns entre o conhecimento científico e o conhecimento local sobre a ecologia e biologia do pirarucu (Arapaima gigas, contribuindo com informações úteis para a implementação e consolidação de projetos de manejo participativo pesqueiro na região. Foram realizadas 57 entrevistas semi-estruturadas, com pescadores profissionais de Manaus e pescadores de subsistência de Manacapuru durante o período de junho a dezembro do ano de 2002. Foi observado que os pescadores profissionais possuem informações igualmente precisas e abrangentes em relação aos saberes dos pescadores ribeirinhos de subsistência nos aspectos de reprodução, predação, migração, crescimento e mortalidade. Os aspectos que não são equivalentes entre os pescadores profissionais comerciais citadinos e ribeirinhos de subsistência são nos aspectos de tipo de alimentação e no tamanho de recrutamento pesqueiro. Concluímos que os pescadores da Amazônia central possuem os conhecimentos necessários que possibilitam o manejo participativo do pirarucu, como um profundo saber nos aspectos comportamentais, biológicos e ecológicos desta espécie, podendo assim contribuir de fato com a participação de gestão nos recursos pesqueiros locais.Present study it aimed at to identify to know common between scientific knowledge and local knowledge on ecology and biology of pirarucu (Arapaima mock-ups, contributing with useful information for implementation and consolidation of projects of participative handling fishing boat in region. 57 half-structuralized interviews had been carried through, with fishing of Manaus and Manacapuru during period of June to December of year 2002. It was observed that professional fishermen also have accurate and comprehensive information in relation to knowledge of subsistence fishermen in coastal aspects of reproduction, predation, migration, growth and mortality. Aspects that are not equivalent

  6. 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.)

  7. 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.)

  8. Analysis of higher power research reactors' parameters

    International Nuclear Information System (INIS)

    The objective of this monograph was to analyze and compare parameters of different types of research reactors having higher power. This analysis could be used for decision making and choice of a reactor which could possibly replace the existing ageing RA reactor in Vinca. Present experimental and irradiation needs are taken into account together with the existing reactors operated in our country, RB and TRIGA reactor

  9. 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

  10. Mathematical game type optimization of powerful fast reactors

    International Nuclear Information System (INIS)

    To obtain maximum speed of putting into operation fast breeders it is recommended on the initial stage of putting into operation these reactors to apply lower power which needs less fission materials. That is why there is an attempt to find a configuration of a high-power reactor providing maximum power for minimum mass of fission material. This problem has a structure of the mathematical game with two partners of non-zero-order total and is solved by means of specific aids of theory of games. Optimal distribution of fission and breeding materials in a multizone reactor first is determined by solution of competitive game and then, on its base, by solution of the cooperation game. The second problem the solution for which is searched is developed from remark on the fact that a reactor with minimum coefficient of flux heterogenity has a configuration different from the reactor with power coefficient heterogenity. Maximum burn-up of fuel needs minimum heterogenity of the flux coefficient and the highest power level needs minimum coefficient of power heterogenity. That is why it is possible to put a problem of finding of the reactor configuration having both coefficients with minimum value. This problem has a structure of a mathematical game with two partners of non-zero-order total and is solved analogously giving optimal distribution of fuel from the new point of view. In the report is shown that both these solutions are independent which is a result of the aim put in the problem of optimization. (author)

  11. Reactor technology: power conversion systems and reactor operation and maintenance

    International Nuclear Information System (INIS)

    The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

  12. Microstructure and nano-hardness analyses of stress corrosion cracking, utilizing 316L core shroud of BWR power reactors

    International Nuclear Information System (INIS)

    The water cooled shield blanket made of Type 316L SS for the international thermonuclear experimental reactor (ITER) has potential issues related to stress corrosion cracking (SCC). Shroud mock-ups and boat samples taken from the core shroud of the boiling water reactor (BWR) with SCC were investigated from the viewpoint of microstructures and nano-hardness. Fine grains and deformation bands were observed in the hardened surface thin layers of the shroud mock-up, where hardness profiles in the ground portion was different from those in the milled portion. In the fine grain region, crevices were found only in the ground surface. In the core shroud, hardened surface regions were also found. Results showed that the crevices found on the ground surface could be one possible factor for SCC initiation

  13. Performance indicators for power reactors

    International Nuclear Information System (INIS)

    A review of Canadian and worldwide performance indicator definitions and data was performed to identify a set of indicators that could be used for comparison of performance among nuclear power plants. The results of this review are to be used as input to an AECB team developing a consistent set of performance indicators for measuring Canadian power reactor safety performance. To support the identification of performance indicators, a set of criteria was developed to assess the effectiveness of each indicator for meaningful comparison of performance information. The project identified a recommended set of performance indicators that could be used by AECB staff to compare the performance of Canadian nuclear power plants among themselves, and with international performance. The basis for selection of the recommended set and exclusion of others is provided. This report provides definitions and calculation methods for each recommended performance indicator. In addition, a spreadsheet has been developed for comparison and trending for the recommended set of indicators. Example trend graphs are included to demonstrate the use of the spreadsheet. (author). 50 refs., 11 tabs., 3 figs

  14. The surface mock-up KENTEX: on the thermal-hydro-mechanical behaviors in the buffer of a Korean HLW repository

    International Nuclear Information System (INIS)

    The concept for a disposal of high-level wastes (HLW) in Korea is based upon a multi barrier system composed of engineered barriers and its surrounding plutonic rock (Kang et. al., 2002). A repository is constructed in a bedrock of several hundred meters in depth below the ground surface. The engineered barrier system (EBS), which is similar to the configuration considered by many other countries, consists of the HLW-encapsulating disposal container, the buffer between the container and the wall of a borehole, and the backfill in the inside space of the emplacement room, to isolate the HLW from the surrounding rock masses. The engineering performance of a HLW repository is dependent, to a large extent, upon the thermal-hydro-mechanical (THM) behaviors in the buffer which are complicated by the processes such as the decay heat generated from the HLW, the ground water flowing in from the surrounding host rock, and the swelling pressure exerted by compacted bentonite. For this reason, the Korea Atomic Energy Research Institute (KAERI), to investigate the THM behaviors in the buffer of the Korean reference disposal system (KRS), planned large-scale tests to be conducted in two stages: a surface mock-up and then a full-scale 'in situ' test. This paper deals with the surface mock-up called as 'KENTEX' and presents the THM behaviors in the buffer which have been investigated from the KENTEX test. The KENTEX is a third scale of the KRS. It consists of five major components: a heating system, a confining cylinder, a hydration tank, bentonite blocks, and sensors and instruments. The heating system measures 0.41 m in diameter and 0.68 m in length, which includes three heating elements in its inside, capable of supplying a thermal power of 1 kW each. The confining cylinder, which plays a role of the wall of a borehole excavated in the host rock, is a steel body with a length of 1.36 m and an inner diameter of 0.75 m, the inside wall of which is lined with layers of geotextile

  15. Evaluation of residual stresses in electron-beam welded Zr2.5Nb0.9Hf Zircadyne flange mock-up of a reflector vessel beam tube flange

    Energy Technology Data Exchange (ETDEWEB)

    Muránsky, O., E-mail: ondrej.muransky@ansto.gov.au [Institute of Material Engineering, ANSTO, Locked Bag 2001, Kirrawee DC, 2234 NSW (Australia); Holden, T.M. [Northern Stress Technologies, Deep River, Ontario, Canada K0J 1P0 (Canada); Kirstein, O. [European Spallation Source, EES AB, Tunavagen 24, SE-211 00 Lund (Sweden); James, J.A. [Open University, Materials Engineering, Milton Keynes MK7 6BJ (United Kingdom); Paradowska, A.M. [Bragg Institute, ANSTO, Locked Bag 2001, Kirrawee DC, 2234 NSW (Australia); Edwards, L. [Institute of Material Engineering, ANSTO, Locked Bag 2001, Kirrawee DC, 2234 NSW (Australia)

    2013-07-15

    The dual-phase alloy Zr2.5Nb alloy is an important nuclear material, because of its use in current and possible use in future nuclear reactors. It is, however, well-known that Zr2.5Nb weldments can fail through a time-dependent mechanism called delayed hydride cracking which is typically driven by the presence of tensile residual stresses. With a view to understanding the development of residual stresses associated with Zr2.5Nb welds the current study focuses on the evaluation of the residual stresses in a mock-up of a reactor beam tube flange made from Zr2.5Nb0.9Hf. The present results suggests that, like ferritic welds which undergo a solid-state phase transformation upon welding, Zr2.5Nb0.9Hf welds also develop high tensile residual stresses in the heat-affected zone whereas the stresses closer to the weld tip are reduced by the effects of the β → α solid-state phase transformation.

  16. Evaluation of residual stresses in electron-beam welded Zr2.5Nb0.9Hf Zircadyne flange mock-up of a reflector vessel beam tube flange

    Science.gov (United States)

    Muránsky, O.; Holden, T. M.; Kirstein, O.; James, J. A.; Paradowska, A. M.; Edwards, L.

    2013-07-01

    The dual-phase alloy Zr2.5Nb alloy is an important nuclear material, because of its use in current and possible use in future nuclear reactors. It is, however, well-known that Zr2.5Nb weldments can fail through a time-dependent mechanism called delayed hydride cracking which is typically driven by the presence of tensile residual stresses. With a view to understanding the development of residual stresses associated with Zr2.5Nb welds the current study focuses on the evaluation of the residual stresses in a mock-up of a reactor beam tube flange made from Zr2.5Nb0.9Hf. The present results suggests that, like ferritic welds which undergo a solid-state phase transformation upon welding, Zr2.5Nb0.9Hf welds also develop high tensile residual stresses in the heat-affected zone whereas the stresses closer to the weld tip are reduced by the effects of the β → α solid-state phase transformation.

  17. Evaluation of residual stresses in electron-beam welded Zr2.5Nb0.9Hf Zircadyne flange mock-up of a reflector vessel beam tube flange

    International Nuclear Information System (INIS)

    The dual-phase alloy Zr2.5Nb alloy is an important nuclear material, because of its use in current and possible use in future nuclear reactors. It is, however, well-known that Zr2.5Nb weldments can fail through a time-dependent mechanism called delayed hydride cracking which is typically driven by the presence of tensile residual stresses. With a view to understanding the development of residual stresses associated with Zr2.5Nb welds the current study focuses on the evaluation of the residual stresses in a mock-up of a reactor beam tube flange made from Zr2.5Nb0.9Hf. The present results suggests that, like ferritic welds which undergo a solid-state phase transformation upon welding, Zr2.5Nb0.9Hf welds also develop high tensile residual stresses in the heat-affected zone whereas the stresses closer to the weld tip are reduced by the effects of the β → α solid-state phase transformation

  18. Safety analysis for non-power reactors

    International Nuclear Information System (INIS)

    Non-power reactors have been operating in Canada since 1945, with NRU (National Research Universal, 1957) being the oldest operating non-power reactor. Presently, there are five generic 'types' of non-power reactors: NRU, ZED-2, SLOWPOKE, MNR and MAPLE, the latter undergoing commissioning as the MDS Medical Isotope Reactor. These reactors range in thermal power from 200 Watts to more than 100 MW. Other non-power reactors are likely to be built for new applications and to replace older reactors. The uniqueness of each reactor, the wide range of power levels and the evolution of safety philosophy over time have lead to non-uniform practices for safety analysis. This non-uniformity may be a problem for the preparation by the licensee and review by the regulator of the safety analysis report required for licensing of the reactor facility. Clearly, there is no universally applicable practice, while at the same time, expectations for safety analyses have evolved in order to demonstrate higher levels of overall safety. This paper examines a new 'graded approach' to preparing the safety analysis report for reactors of diverse features but with a common standard of safety. It discusses necessary content, methods and the training and qualification of the safety analyst. (author)

  19. Measurement and analysis of neutron and gamma-ray flux spectra in a neutronics mock-up of the HCPB Test Blanket Module

    International Nuclear Information System (INIS)

    Neutron and γ-ray flux spectra were measured with a NE 213 spectrometer in the rear block of a mock-up of the HCPB Test Blanket Module. The flux of the slow neutrons was investigated by time-of-arrival spectroscopy with a pulsed D-T neutron source. The experimental results were compared versus calculations performed with the Monte Carlo code MCNP and the data libraries EFF-3, FENDL-2.0 and FENDL-2.1, and are discussed with respect to the shielding capability of the TBM and to tritium breeding

  20. Benchmarking of MCNPX Results with Measured Tritium Production Rate and Neutron Flux at the Mock-up of EU TBM (HCPB concept)

    International Nuclear Information System (INIS)

    In order to reassesses the available design results of Test Breeder Modules (TBMs) a framework contract agreement between F4E and IDOM-Spain has been signed. SEA SL-Spain and UNED-Spain participate as sub-contractors of IDOM. In this study, a qualification of MCNPX code and nuclear data libraries are performed with benchmarking of measured tritium production and neutron flux at the mock-up of the EU TBM, HCPB concept. The irradiation and measurements had been performed in the frame of European Fusion Technology Program by ENEA-Italy, TUD-Germany and JAERI -Japan.

  1. Reactor power measuring method and device therefor

    International Nuclear Information System (INIS)

    The present invention concerns measurement of a BWR type reactor power and provides a method of and a device for ensuring accuracy of calibration of sensitivity of neutron detectors and measurement of reactor power even if γ-ray thermometers are failed. Namely, the output signals of the γ-ray thermometers are compared with previously determined judging values to detect failures. The reactor power is measured based on the signals of neutron detectors calibrated by integral thermometers except for neutron detectors calibrated by γ-ray thermometers detected as failed. Calibration for sensitivity of neutron detectors as objects of γ-ray thermometers detected as failed is preferably prohibited. Accuracy of measurement of the reactor power can be ensured by the method described above. If axial power distribution of the reactor core is measured while eliminating the signals of γ-ray thermometers detected as failed, accuracy of the measurement of axial power distribution can be ensured. (N.H.)

  2. Applications of a low power nuclear reactor

    International Nuclear Information System (INIS)

    The Omaha Nebraska Veterans Administration Medical Center (OVAMC) TRIGA reactor is a research reactor designed and fabricated by General Atomic. The reactor first achieved criticality on June 30, 1959. It is a below grade, open-tank-type, ligh water moderated, cooled, and shielded reactor that currently is authorized to operate in the steady-state mode at thermal power levels up to 18KW with an excess reactivity limitation of 0.79% Delta K/K. (author)

  3. Automatic power control system for 235 MWe atomic power reactor

    International Nuclear Information System (INIS)

    The paper highlights the essential features of the design, fabrication and testing of microprocessor based reactor power regulating system of Narora Atomic Power Plant (NAPP) and Kakrapar Atomic Power Plant (KAPP). The improved system design at KAPP employs the reactor power control based on neutron flux signal after correction. The control system responses have been presented and compared with the responses using a reactor functional simulator. A new fault tolerant reactor regulating system has been designed using a dual active and hot stand-by microprocessor system to improve operational reliability. (author). 1 ref., 8 figs

  4. Power Control Method for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baang, Dane; Suh, Yongsuk; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Considering safety-oriented design concept and other control environment, we developed a simple controller that provides limiting function of power change- rate as well as fine tracking performance. The design result has been well-proven via simulation and actual application to a TRIGA-II type research reactor. The proposed controller is designed to track the PDM(Power Demand) from operator input as long as maintaining the power change rate lower than a certain value for stable reactor operation. A power control method for a TRIGA-II type research reactor has been designed, simulated, and applied to actual reactor. The control performance during commissioning test shows that the proposed controller provides fine control performance for various changes in reference values (PDM), even though there is large measurement noise from neutron detectors. The overshoot at low power level is acceptable in a sense of reactor operation.

  5. 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.)

  6. Fabrication of a small-scaled first wall mock-up with beryllium armor hip bonded to DSCu/SS structure

    International Nuclear Information System (INIS)

    For the bonding of beryllium (Be) and alumna dispersion strengthened copper (DSCu), Hot Isostatic Pressing (HIP) method has been tried. A preliminary study on optimal HIP conditions for the joining of Be/DSCu had been previously performed with various interlayer materials. Based on this previous study, an additional screening test was performed focusing on lower HIP temperatures to avoid stainless steel (SS) sensitization during HIP process. From the results of bending tests and microscopic observation of the HIPed interfaces, two options were selected as follows: 1) 580 deg.C HIP temperature with Ti/Cu interlayer and 2) 550 deg.C HIP temperature with A1/Ti/Cu interlayer. Based on the selection of above HIP conditions and interlayers, a small-scale first wall mock-up was fabricated. The mock-up consists of four 10 mm thick Be armor tiles, 20 mm thick DSCu heat sink with a built-in SS316L coolant tube and SS316L backing plate. After the DSCu heat sink and SS316L coolant tube and backing plate were joined by HIP, Be armor tiles were HIPed onto DSCu. Castellation of the Be armor was then carried out with 1.5 mm gap width between armor tiles. Among four Be tiles, two tiles were bonded with Ti/Cu interlayer and the other tiles with A1/Ti/Cu interlayer. (author)

  7. Higher power density TRIGA research reactors

    International Nuclear Information System (INIS)

    The uranium zirconium hydride (U-ZrH) fuel is the fundamental feature of the TRIGA family of reactors that accounts for its widely recognized safety, good performance, economy of operation, and its acceptance worldwide. Of the 65 TRIGA reactors or TRIGA fueled reactors, several are located in hospitals or hospital complexes and in buildings that house university classrooms. These examples are a tribute to the high degree of safety of the operating TRIGA reactor. In the early days, the majority of the TRIGA reactors had power levels in the range from 10 to 250 kW, many with pulsing capability. An additional number had power levels up to 1 MW. By the late 1970's, seven TRIGA reactors with power levels up to 2 MW had been installed. A reduction in the rate of worldwide construction of new research reactors set in during the mid 1970's but construction of occasional research reactors has continued until the present. Performance of higher power TRIGA reactors are presented as well as the operation of higher power density reactor cores. The extremely safe TRIGA fuel, including the more recent TRIGA LEU fuel, offers a wide range of possible reactor configurations. A long core life is assured through the use of a burnable poison in the TRIGA LEU fuel. In those instances where large neutron fluxes are desired but relatively low power levels are also desired, the 19-rod hexagonal array of small diameter fuel rods offers exciting possibilities. The small diameter fuel rods have provided extremely long and trouble-free operation in the Romanian 14 MW TRIGA reactor

  8. Power source device for reactor recycling pump

    International Nuclear Information System (INIS)

    The device of the present invention prevents occurrence of an accident of a reactor forecast upon spontaneous power stoppage, loss of power source or trip of the reactor. Namely, a AC/DC converter and a DC/AC connector having an AC voltage frequency controller are connected in series between an AC (bus) in the plant and reactor recycling pumps. A DC voltage controller, a superconductive energy storing device and an excitation power source are connected to the input of the DC/AC converter. The control device receives signals of the spontaneous power stoppage, loss of power source or trip of the reactor to maintain the output voltage of the superconductive energy storing device to a predetermined value. Further, the ratio of AC power voltage and the frequency of AC voltage to be supplied to the reactor recycling pumps is constantly varied to control the flow rate of the pump to a predetermined value. With such procedures, a power source device for the reactor recycling pumps compact in size, easy for maintenance and having high reliability can be realized by adopting a static-type superconductive energy storing device as an auxiliary power source for the reactor recycling pumps. (I.S.)

  9. Specific power of liquid-metal-cooled reactors

    International Nuclear Information System (INIS)

    Calculations of the core specific power for conceptual space-based liquid-metal-cooled reactors, based on heat transfer considerations, are presented for three different fuel types: (1) pin-type fuel; (2) cermet fuel; and (3) thermionic fuel. The calculations are based on simple models and are intended to provide preliminary comparative results. The specific power is of interest because it is a measure of the core mass required to produce a given amount of power. Potential problems concerning zero-g critical heat flux and loss-of-coolant accidents are also discussed because these concerns may limit the core specific power. Insufficient experimental data exists to accurately determine the critical heat flux of liquid-metal-cooled reactors in space; however, preliminary calculations indicate that it may be a concern. Results also indicate that the specific power of the pin-type fuels can be increased significantly if the gap between the fuel and the clad is eliminated. Cermet reactors offer the highest specific power because of the excellent thermal conductivity of the core matrix material. However, it may not be possible to take fuel advantage of this characteristic when loss-of-coolant accidents are considered in the final core design. The specific power of the thermionic fuels is dependent mainly on the emitter temperature. The small diameter thermionic fuels have specific powers comparable to those of pin-type fuels. 11 refs., 12 figs, 2 tabs

  10. Thyristor Controlled Reactor for Power Factor Improvement

    Directory of Open Access Journals (Sweden)

    Sheila Mahapatra

    2014-04-01

    Full Text Available Power factor improvement is the essence of any power sector for reliable operation. This paper provides Thyristor Controlled Reactor regulated by programmed microcontroller which aids in improving power factor and retaining it close to unity under various loading conditions. The implementation is done on 8051 microcontrollerwhich isprogrammed using Keil software. To determine time lag between current and voltage PSpice softwareis used and to display power factor according tothe variation in loadProteus software is used. Whenever a capacitive load is connected to the transmission linea shunt reactor is connected which injects lagging reactive VARs to the power system. As a result the power factor is improved. The results given in this paper provides suitable microcontroller based reactive power compensation and power factor improvement technique using a Thyristor Controlled Reactor module.

  11. Small and medium power reactors 1987

    International Nuclear Information System (INIS)

    This TECDOC follows the publication of TECDOC-347 Small and Medium Power Reactors Project Initiation Study - Phase I published in 1985 and TECDOC-376 Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power programme. It consists of two parts: 1) Guidelines for the Introduction of Small and Medium Power Reactors in Developing Countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of Small and Medium Power Reactors in developing countries; 2) Up-dated Information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex I of the above mentioned TECDOC-347. Figs

  12. First measurements of the kinetic response of the muse-4 fast Ads mock-up to fast neutron pulse

    International Nuclear Information System (INIS)

    The MUSE-4 experiment has started its first commissioning measurements at the beginning of the year 2001 at CEA/Cadarache (France). This international experiment co-ordinated by CEA, included in the 5FWP of the European Union and GEDEON, is intended to study the physics of fast sub-critical assemblies coupled with a pulsed external source. To achieve this objective, the GENEPI accelerator, a (d,d) or (d,t) neutron source developed at CRNS/IN2P3/ISN (Grenoble), has been coupled with the MASURCA reactor, a uranium-plutonium MOX-based fast reactor, with solid sodium simulating a liquid metal coolant and a lead buffer to simulate a spallation target. The very short neutron pulse (1 μs) provided by GENEPI, together with the possibility to change the pulse repetition rate up to 5 kHz and the different levels of sub-criticality available will facilitate a study of the reactor kinetic parameters in situations close to most of the proposed accelerator-driven Systems (ADS). The paper presents the first experimental results for dynamic measurements performed in MUSE-4 configurations. Several pulsed neutron source experiments have been carried out using the (d,d) GENEPI neutron source in configurations going from USD 1,33 to USD 12,6. In addition, noise techniques (Rossi and Feynman-alpha) have been applied to stationary states in the same range of sub-criticalities. Reactivity levels obtained by these techniques have been compared with more classic rod drop/source multiplication measurements. The kinetic parameters, β(which ranges between 330 and 360 pcm) and β/Λ (with a value of approximately 6270 s-1), have been determined by Monte Carlo and/or deterministic codes. (author)

  13. Ion Beam Analysis methods applied to the examination of Be//Cu joints in hipped Be tiles for ITER first wall mock- ups

    International Nuclear Information System (INIS)

    A proposed fabrication route for ITER first wall components implies a diffusion welding step of Be tiles onto a Cu-based substrate. However, Be has a tendency to form particularly brittle intermetallics with Cu and a lot of other elements. Insertion of interlayers may be a solution to increase bond quality. Applying traditional analyses to this study can be problematic because of Be toxicity and low atomic number Z. Ion Beam Analysis methods have thus been considered together with scanning electron microscopy (SEM) and electron back-scattering diffraction (EBSD) as complementary techniques. The following work aims at demonstrating how such techniques (used in micro-beam mode), and in particular NRA (Nuclear Reaction Analysis) and PIXE (Particle Induced X-ray Emission) techniques, coupled with SEM/EBSD data, can bring valuable information in this area. Quantification of data allow to obtain concentration values (provided the hypotheses on the initial junction composition are valuable), then phase diagrams give clues about the composition and structure of the junction. SEM retro-diffused electrons chemical contrast images and EBSD allow to characterize the presence of the awaited intermetallics, and finally confirm or refine the conclusions of Ion Beam Analysis data quantification. A series of reference first wall mock-ups have been analysed. Interlayer-free mock-ups reveal intermetallics which are mainly BeCu (apparently mixed with lower quantities of BeCu2 compound). While Cr or Ti interlayers seem to behave as good Be diffusion barriers in the sense that they prevent the formation of BeCu, they strongly interact with Cu to form CuTi2 or Cr2Ti intermetallics. In the case of Cr, Be seems to be incorporated into the Cr layer. PIXE analysis has however been unable to characterize Al-based interlayers (Z=13, close to the lower PIXE sensibility limit) and emphasizes one limitation of Ion Beam Analysis methods for lighter metals, justifying the use of other complementary

  14. Measurement and Analysis of the Neutron and Gamma-Ray Flux Spectra in a Neutronics Mock-Up of the HCPB Test Blanket Module

    International Nuclear Information System (INIS)

    The nuclear parameters of a breeding blanket, such as tritium production rate, nuclear heating, activation and dose rate, are calculated by integral folding of an energy dependent cross section (or coefficient) with the neutron (or gamma-ray) flux energy spectra. The uncertainties of the designed parameters are determined by the uncertainties of both the cross section data and the flux spectra obtained by transport calculations. Also the analysis of possible discrepancies between measured and calculated integral nuclear parameter represents a two-step procedure. First, the energy region and the amount of flux discrepancies has to be found out and second, the cross section data have to be checked. To this end, neutron and gamma-ray flux spectra in a mock-up of the EU Helium-Cooled Pebble Bed (HCPB) breeder Test Blanket Module (TBM), irradiated with 14 MeV neutrons, were measured and analysed by means of Monte Carlo transport calculations. The flux spectra were determined for the energy ranges that are relevant for the most important nuclear parameters of the TBM, which are the tritium production rate and the shielding capability. The fast neutron flux which determines the tritium production on 7Li and dominates the shield design was measured by the pulse-height distribution obtained from an organic liquid scintillation detector. Simultaneously, the gamma-ray flux spectra were measured. The neutron flux at lower energies, down to thermal, which determines the tritium production on 6Li, was measured with time-of-arrival spectroscopy. For this purpose, the TUD neutron generator was operated in pulsed mode (pulse width 10 μs, frequency 1 kHz) and the neutrons arriving at a 3He proportional counter in the mock-up were recorded as a function of time after the source neutron pulse. The spectral distributions for the two positions in the mock-up, where measurements were carried out, were calculated with the Monte Carlo code MCNP, version 5, and nuclear data from the

  15. MIT research reactor. Power uprate and utilization

    International Nuclear Information System (INIS)

    The MIT Research Reactor (MITR) is a university research reactor located on MIT campus. and has a long history in supporting research and education. Recent accomplishments include a 20% power rate to 6 MW and expanding advanced materials fuel testing program. Another important ongoing initiative is the conversion to high density low enrichment uranium (LEU) monolithic U-Mo fuel, which will consist of a new fuel element design and power increase to 7 MW. (author)

  16. Contribution of Italy to the activities on intercomparison of analysis methods for seismically isolated nuclear structures: Shake table tests on a steel frame structure mock-up

    International Nuclear Information System (INIS)

    This report describes a series of tests performed during a wide ranging experimental campaign on a steel frame structure mock-up subjected to shake table excitations, in both fixed and base isolated configurations. Preliminary numerical simulations of experimental results, carried out by the ENEA and ENEL, are also reported. The main features of a simplified model of rubber bearings, developed and proposed by ENEA and ENEL, used in the above mentioned numerical analysis, are presented together with example of validations. Activities are being carried out in the framework of the four years' Co-ordinated Research Programme (CRP) of the International Atomic Energy Agency (IAEA) on Intercomparison of Analysis Methods for Seismically Isolated Nuclear Structures. (author)

  17. Characterization of the TRIGA Mark II reactor full-power steady state

    OpenAIRE

    Cammi, Antonio; Zanetti, Matteo; Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica; Magrotti, Giovanni; Prata, Michele; Salvini, Andrea

    2015-01-01

    In this work, the characterization of the full-power steady state of the TRIGA Mark II nuclear reactor of the University of Pavia is performed by coupling Monte Carlo (MC) simulation for neutronics with "Multiphysics" model for thermal-hydraulics. Neutronic analyses have been performed starting from a MC model of the entire reactor system, based on the MCNP5 code, that was already validated in fresh fuel and zero-power configuration (in which thermal effects are negligible) using the availabl...

  18. reactor power control using fuzzy logic

    International Nuclear Information System (INIS)

    power stabilization is a critical issue in nuclear reactors. convention pd- controller is currently used in egypt second testing research reactor (ETRR-2). two fuzzy controllers are proposed to control the reactor power of ETRR-2 reactor. the design of the first one is based on a set of linguistic rules that were adopted from the human operators experience. after off-line fuzzy computations, the controller is a lookup table, and thus, real time controller is achieved. comparing this f lc response with the pd-controller response, which already exists in the system, through studying the expected transients during the normal operation of ETRR-2 reactor, the simulation results show that, fl s has the better response, the second controller is adaptive fuzzy controller, which is proposed to deal with system non-linearity . The simulation results show that the proposed adaptive fuzzy controller gives a better integral square error (i se) index than the existing conventional od controller

  19. Numerical Study of High Heat Flux Performances of Flat-Tile Divertor Mock-ups with Hypervapotron Cooling Concept

    Science.gov (United States)

    Chen, Lei; Liu, Xiang; Lian, Youyun; Cai, Laizhong

    2015-09-01

    The hypervapotron (HV), as an enhanced heat transfer technique, will be used for ITER divertor components in the dome region as well as the enhanced heat flux first wall panels. W-Cu brazing technology has been developed at SWIP (Southwestern Institute of Physics), and one W/CuCrZr/316LN component of 450 mm×52 mm×166 mm with HV cooling channels will be fabricated for high heat flux (HHF) tests. Before that a relevant analysis was carried out to optimize the structure of divertor component elements. ANSYS-CFX was used in CFD analysis and ABAQUS was adopted for thermal-mechanical calculations. Commercial code FE-SAFE was adopted to compute the fatigue life of the component. The tile size, thickness of tungsten tiles and the slit width among tungsten tiles were optimized and its HHF performances under International Thermonuclear Experimental Reactor (ITER) loading conditions were simulated. One brand new tokamak HL-2M with advanced divertor configuration is under construction in SWIP, where ITER-like flat-tile divertor components are adopted. This optimized design is expected to supply valuable data for HL-2M tokamak. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2011GB110001 and 2011GB110004)

  20. New generation of reactors for space power

    International Nuclear Information System (INIS)

    Space nuclear reactor power is expected to enable many new space missions that will require several times to several orders of magnitude anything flown in space to date. Power in the 100-kW range may be required in high earth orbit spacecraft and planetary exploration. The technology for this power system range is under development for the Department of Energy with the Los Alamos National Laboratory responsible for the critical components in the nuclear subsystem. The baseline design for this particular nuclear sybsystem technology is described in this paper; additionally, reactor technology is reviewed from previous space power programs, a preliminary assessment is made of technology candidates covering an extended power spectrum, and the status is given of other reactor technologies

  1. BN-1200 reactor power unit design development

    International Nuclear Information System (INIS)

    In February 2010, the RF Government has approved the Federal Target Program “New Generation Nuclear Power Technologies for the Period of 2010-2015 and for the long-term up to 2020”. Within this Program, the R&D work for new generation 1200MWe sodium fast reactor is provided. The BN-1200 design is based on the combination of approved and innovative technical decisions, which allow: – reliable power unit with large BN reactor to be developed in a short period of time for commercial construction as a part of closed nuclear fuel cycle; – qualitatively new technical level of power unit to be provided according to generation 4 NPP requirements. The paper characterizes the activities performed now for the power unit design in various areas: – power unit design; – reactor plant (RP) detailed design development; – R&D work to validate the RP system and equipment; – code upgrading and verification; – safety validation. (author)

  2. Neutron Diffraction Residual Strain Tensor Measurements Within The Phase IA Weld Mock-up Plate P-5

    Energy Technology Data Exchange (ETDEWEB)

    Hubbard, Camden R [ORNL

    2011-09-01

    Oak Ridge National Laboratory (ORNL) has worked with NRC and EPRI to apply neutron and X-ray diffraction methods to characterize the residual stresses in a number of dissimilar metal weld mockups and samples. The design of the Phase IA specimens aimed to enable stress measurements by several methods and computational modeling of the weld residual stresses. The partial groove in the 304L stainless steel plate was filled with weld beads of Alloy 82. A summary of the weld conditions for each plate is provided in Table 1. The plates were constrained along the long edges during and after welding by bolts with spring-loaded washers attached to the 1-inch thick Al backing plate. The purpose was to avoid stress relief due to bending of the welded stainless steel plate. The neutron diffraction method was one of the methods selected by EPRI for non-destructive through thickness strain and stress measurement. Four different plates (P-3 to P-6) were studied by neutron diffraction strain mapping, representing four different welding conditions. Through thickness neutron diffraction strain mappings at NRSF2 for the four plates and associated strain-free d-zero specimens involved measurement along seven lines across the weld and at six to seven depths. The mountings of each plate for neutron diffraction measurements were such that the diffraction vector was parallel to each of the three primary orthogonal directions of the plate: two in-plane directions, longitudinal and transverse, and the direction normal to the plate (shown in left figure within Table 1). From the three orthogonal strains for each location, the residual stresses along the three plate directions were calculated. The principal axes of the strain and stress tensors, however, need not necessarily align with the plate coordinate system. To explore this, plate P-5 was selected for examination of the possibility that the principal axes of strain are not along the sample coordinate system axes. If adequate data could

  3. Extended Cooling System for High Power Reactors

    International Nuclear Information System (INIS)

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants (NPPs) and proposed for advanced light water reactors (LWRs). However, it is not clear that currently proposed external reactor vessel cooling (ERVC) could provide sufficient heat removal for higher power reactors. This paper proposes a dual retention strategy to realize fail-proof defense-in-depth in the APR1400 (Advanced Power Reactor 1400 MWe) and the OPR 1000 (Optimized Power Reactor 1000 MWe). The dual retention has the advantage of IVR-ERVC as well as ex-vessel cooling (EVC) strategies. The multilateral, multidisciplinary project calls for national and international cutting-edge technologies to research and produce (R and P) the D2R2 (Duel Retention Demonstration Reactor) equipped with OASIS (Optimized Advanced Safety Injection System) and ROSIS (Reactor Outer Safety Injection System) to cope with design-basis accidents and beyond in a coherent, continual, comprehensive manner. The enterprise aims to develop the design-basis and severe accident engineering solutions. The enterprise aims to develop the design-basis and severe accident engineering solutions. The former embraces ISAIAH (Injection System Annular Interactive Aero Hydrodynamics) and MESIAH (Methodical Evaluation System Interactive Aero Hydrodynamics). The latter comprises GODIVA (Geo metrics of Direct Injection Versatile Arrangement), SONATA (Simulation of Narrow Annular Thermomechanical Arrest or), TOCATA (Termination of Corium Ablation Thermal Attack) and STRADA (Solution to Reactor Advanced Design Alternatives). D2R2 will contribute to enhancement of both safety and economics for an advanced high power particular and nuclear power in general

  4. Power calibration study at the Musashi reactor

    International Nuclear Information System (INIS)

    The Musashi reactor (TRIGA-II,100 kW) initially went critical in January of 1963. The reactor had been used for training, isotope production and medical irradiation for boron neutron capture therapy (1). The initial power calibration was based on the use of a calibrated electrical heater in a calorimetric procedure where the rate of rise of the bulk pit water temperature was measured using 2 kW heaters x 6 pieces. The rate of rise of water temperature was determined to be 0.0474 C/kWh. The reactor was then operated to give the same rate of rise of water temperature. Thus the reactor power was established at the value produced by the electrical heaters. A stirrer for tank water mixing was not used. Recent communications (2)(3) indicated that power calibrations using a stirrer provided a much more uniform mixing, and heating in the reactor tank water which was essential for an accurate calibration. In this paper, the effect of mixing using a stirrer was investigated considering the physical factors such as room temperature, humidity, tank water temperature and it's distributions. The room temperature and humidity around the reactor varies 6-30 and 30-80 %, respectively, depending on four seasons. The heat flow through the surface of the pool was also evaluated because the reactor usually operates without cover on the surface of the pool. (orig.)

  5. Thorium utilization in power reactors

    International Nuclear Information System (INIS)

    In this work the recent (prior to Aug, 1976) literature on thorium utilization is reviewed briefly and the available information is updated. After reviewing the nuclear properties relevant to the thorium fuel cycle we describe briefly the reactor systems that have been proposed using thorium as a fertile material. (author)

  6. Analysis of gamma-ray dosimetry experiments in the zero power MINERVE facility

    International Nuclear Information System (INIS)

    The objective of this study is to develop nuclear heating measurement methods in zero power experimental reactors. These developments contribute to the qualification of photonics calculation schemes for the assessment of gamma heating in the future Jules Horowitz Material Testing Reactor. This paper presents the analysis of thermoluminescent detector (TLD) experiments in the UO2 core of the MINERVE Research Reactor at the French Alternative Energies and Atomic Energy Commission center in Cadarache. The experimental sources of uncertainty in the gamma dose have been reduced by improving the measurement conditions and the repeatability of the calibration step for each individual TLD. The interpretation of these measurements needs to take into account the calculation of cavity correction factors related to calibration and irradiation configurations, as well as neutron correction calculations. These calculations are based on Monte Carlo simulations of neutron-gamma and gamma-electron transport coupled particles. The comparison between calculated and measured integral gamma-ray absorbed doses in the aluminum material surrounding the TLD shows that calculations slightly overestimate the measurement, with a calculated versus experimental ratio equal to 1.04 ± 5.7 % (k=2). (authors)

  7. Analysis of gamma-ray dosimetry experiments in the zero power MINERVE facility

    Energy Technology Data Exchange (ETDEWEB)

    Amharrak, H.; Di Salvo, J.; Lyoussi, A.; Roche, A.; Masson-Fauchier, M.; Bosq, J. C. [CEA, DEN, DER, F-13108 Saint-Paul-lez-Durance (France); Carette, M. [Aix-Marseille Univ., LCP UMR 6264, 13397, Marseille (France)

    2011-07-01

    The objective of this study is to develop nuclear heating measurement methods in zero power experimental reactors. These developments contribute to the qualification of photonics calculation schemes for the assessment of gamma heating in the future Jules Horowitz Material Testing Reactor. This paper presents the analysis of thermoluminescent detector (TLD) experiments in the UO{sub 2} core of the MINERVE Research Reactor at the French Alternative Energies and Atomic Energy Commission center in Cadarache. The experimental sources of uncertainty in the gamma dose have been reduced by improving the measurement conditions and the repeatability of the calibration step for each individual TLD. The interpretation of these measurements needs to take into account the calculation of cavity correction factors related to calibration and irradiation configurations, as well as neutron correction calculations. These calculations are based on Monte Carlo simulations of neutron-gamma and gamma-electron transport coupled particles. The comparison between calculated and measured integral gamma-ray absorbed doses in the aluminum material surrounding the TLD shows that calculations slightly overestimate the measurement, with a calculated versus experimental ratio equal to 1.04 {+-} 5.7 % (k=2). (authors)

  8. Enhancement of aerobic granulation by zero-valent iron in sequencing batch airlift reactor.

    Science.gov (United States)

    Kong, Qiang; Ngo, Huu Hao; Shu, Li; Fu, Rong-Shu; Jiang, Chun-Hui; Miao, Ming-sheng

    2014-08-30

    This study elucidates the enhancement of aerobic granulation by zero-valent iron (ZVI). A reactor augmented with ZVI had a start-up time of aerobic granulation (43 days) that was notably less than that for a reactor without augmentation (64 days). The former reactor also had better removal efficiencies for chemical oxygen demand and ammonium. Moreover, the mature granules augmented with ZVI had better physical characteristics and produced more extracellular polymeric substances (especially of protein). Three-dimensional-excitation emission matrix fluorescence showed that ZVI enhanced organic material diversity. Additionally, ZVI enhanced the diversity of the microbial community. Fe(2+) dissolution from ZVI helped reduce the start-up time of aerobic granulation and increased the extracellular polymeric substance content. Conclusively, the use of ZVI effectively enhanced aerobic granulation. PMID:25108827

  9. 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.)

  10. Introduction to the neutron kinetics of nuclear power reactors

    CERN Document Server

    Tyror, J G; Grant, P J

    2013-01-01

    An Introduction to the Neutron Kinetics of Nuclear Power Reactors introduces the reader to the neutron kinetics of nuclear power reactors. Topics covered include the neutron physics of reactor kinetics, feedback effects, water-moderated reactors, fast reactors, and methods of plant control. The reactor transients following faults are also discussed, along with the use of computers in the study of power reactor kinetics. This book is comprised of eight chapters and begins with an overview of the reactor physics characteristics of a nuclear power reactor and their influence on system design and

  11. RELAP5-3D calculation of steam outlet header rupture of WWER-1000 NPP at hot zero power

    International Nuclear Information System (INIS)

    This presentation is devoted to one of the transients from the spectrum of steam line ruptures, which are analyzed for safety report purposes in Czech Republic. It is focused on the SGI steam outlet header rupture of WWER-1000 at hot zero power conditions analysis with advanced thermal-hydraulic code RELAP5-3D. The attention is addressed a reactor vessel and reactor core nodalization. The steam line rupture followed by steam release results in a strong drain oaf energy from primary circuit, what presents significant decreasing of primary coolant temperature and pressure. Then the feedback reactivity coefficients in connection with reduction of coolant temperature on the reactor inlet would cause introduction of positive reactivity and they could result in reactor restart (after its shutdown). Such transients are characterised by an unsymmetrical cool down of reactor and a strongly non-uniform distribution of power increase in the core. Safety analyses of such transients, where substantial changes oaf power distribution could occur, need apply thermal-hydraulic computational programs containing a 3D neutronic and thermal-hydraulic model of the reactor. The used computer codes, initial conditions for thermal-hydraulic calculations and basic results are presented. The DNBR value was monitored from the point of view of fuel integrity (Authors)

  12. Heat pipe reactors for space power applications

    Science.gov (United States)

    Koenig, D. R.; Ranken, W. A.; Salmi, E. W.

    1977-01-01

    A family of heat pipe reactors design concepts has been developed to provide heat to a variety of electrical conversion systems. Three power plants are described that span the power range 1-500 kWe and operate in the temperature range 1200-1700 K. The reactors are fast, compact, heat-pipe cooled, high-temperature nuclear reactors fueled with fully enriched refractory fuels, UC-ZrC or UO2. Each fuel element is cooled by an axially located molybdenum heat pipe containing either sodium or lithium vapor. Virtues of the reactor designs are the avoidance of single-point failure mechanisms, the relatively high operating temperature, and the expected long lifetimes of the fuel element components.

  13. Nuclear power reactor development in Japan

    International Nuclear Information System (INIS)

    Based on the current situation in Japan that nearly 40 LWRs (PWR/BWR nearly 50/50) have been operating with excellent performance; the number of un-intended shutdown is the smallest, leaker fuel is practically zero and thus coolant activity is the lowest in the world, they proceed to introduce advanced LWR (ALWR) of 1,300 MWe in early '90s, followed by the further improvement into higher duty LWR (so-called next generation LWR). National program for advanced thermal reactor (ATR) and FBR development has been steadily performed; a demonstration reactor of ATR and a prototype of FBR are under construction, which are to be put into operation by middle 1990s. Lately-revised high temperature gas cooled reactor program for process heat generation is put into practice in 1989 and its experimental reactor to be used also for irradiation facility will be completed 7 years later. Smaller-size LWR and so-called 'inherent safety type reactor' have also been interested. Overview on those activities are presented. (author)

  14. Space nuclear reactor power plants

    International Nuclear Information System (INIS)

    Requirements for electrical and propulsion power for space are expected to increase dramatically in the 1980s. Nuclear power is probably the only source for some deep space missions and a major competitor for many orbital missions, especially those at geosynchronous orbit. Because of the potential requirements, a technology program on space nuclear power plant components has been initiated by the Department of Energy. The missions that are foreseen, the current power plant concept, the technology program plan, and early key results are described

  15. Power Nuclear Reactors: technology and innovation for development in future

    International Nuclear Information System (INIS)

    The conference is about some historicals task of the fission technology as well as many types of Nuclear Reactors. Enrichment of fuel, wastes, research reactors and power reactors, a brief advertisment about Uruguay electric siystem and power generation, energetic worldwide, proliferation, safety reactors, incidents, accidents, Three-Mile Island accident, Chernobil accident, damages, risks, classification and description of Power reactors steam generation, nuclear reactor cooling systems, future view

  16. Enhancement of aerobic granulation by zero-valent iron in sequencing batch airlift reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kong, Qiang, E-mail: kongqiang0531@hotmail.com [College of Life Science, Shandong Normal University, 88 Wenhua Donglu, Jinan 250014, Shandong (China); Ngo, Huu Hao [School of Civil and Environmental Engineering, University of Technology Sydney, Broadway, NSW 2007 (Australia); Shu, Li [School of Engineering, Faculty of Science, Engineering and Built Environment, Deakin University, Geelong, Victoria 3216 (Australia); Fu, Rong-shu; Jiang, Chun-hui [College of Life Science, Shandong Normal University, 88 Wenhua Donglu, Jinan 250014, Shandong (China); Miao, Ming-sheng, E-mail: mingshengmiao@163.com [College of Life Science, Shandong Normal University, 88 Wenhua Donglu, Jinan 250014, Shandong (China)

    2014-08-30

    Highlights: • Zero-valent iron (ZVI) was used firstly to enhance the aerobic granulation. • ZVI significantly decreased the start-up time of the aerobic granulation. • ZVI had the function of enhancing organic material diversity identified by 3-D EEM. • ZVI could enhance the diversity of microbial community. - Abstract: This study elucidates the enhancement of aerobic granulation by zero-valent iron (ZVI). A reactor augmented with ZVI had a start-up time of aerobic granulation (43 days) that was notably less than that for a reactor without augmentation (64 days). The former reactor also had better removal efficiencies for chemical oxygen demand and ammonium. Moreover, the mature granules augmented with ZVI had better physical characteristics and produced more extracellular polymeric substances (especially of protein). Three-dimensional-excitation emission matrix fluorescence showed that ZVI enhanced organic material diversity. Additionally, ZVI enhanced the diversity of the microbial community. Fe{sup 2+} dissolution from ZVI helped reduce the start-up time of aerobic granulation and increased the extracellular polymeric substance content. Conclusively, the use of ZVI effectively enhanced aerobic granulation.

  17. Enhancement of aerobic granulation by zero-valent iron in sequencing batch airlift reactor

    International Nuclear Information System (INIS)

    Highlights: • Zero-valent iron (ZVI) was used firstly to enhance the aerobic granulation. • ZVI significantly decreased the start-up time of the aerobic granulation. • ZVI had the function of enhancing organic material diversity identified by 3-D EEM. • ZVI could enhance the diversity of microbial community. - Abstract: This study elucidates the enhancement of aerobic granulation by zero-valent iron (ZVI). A reactor augmented with ZVI had a start-up time of aerobic granulation (43 days) that was notably less than that for a reactor without augmentation (64 days). The former reactor also had better removal efficiencies for chemical oxygen demand and ammonium. Moreover, the mature granules augmented with ZVI had better physical characteristics and produced more extracellular polymeric substances (especially of protein). Three-dimensional-excitation emission matrix fluorescence showed that ZVI enhanced organic material diversity. Additionally, ZVI enhanced the diversity of the microbial community. Fe2+ dissolution from ZVI helped reduce the start-up time of aerobic granulation and increased the extracellular polymeric substance content. Conclusively, the use of ZVI effectively enhanced aerobic granulation

  18. Power Reactors in Small Packages

    Energy Technology Data Exchange (ETDEWEB)

    Corliss, William R.

    1968-10-01

    This booklet discusses the introduction of nuclear power to remote places on earth where the resources of civilization are almost scarce. It also discusses nuclear power plants designed for use when warfare or natural catastrophes have wiped out the usual sources of energy, and in places beyond the reach of oil pipelines and coal trains. It also discusses how nuclear power may one day be used to manufacture chemical fuels for the world's vehicles when fossil fuels begin to run out.

  19. CNSC power reactor operating licence reform

    International Nuclear Information System (INIS)

    CNSC staff introduced a new Power Reactor Operating Licence (PROL) in order to strengthen the regulatory oversight of power reactor operation, while increasing regulatory effectiveness and efficiency by focusing on risk-significant issues and reducing purely administrative efforts. The PROLs have been simplified by incorporating a more risk-informed approach and by eliminating cascading references to working level licensee documentation and regulatory expectations. To ensure that there is a common understanding for each requirement specified in the PROL, CNSC staff prepared a Licence Conditions Handbook (LCH), which provides technical details and compliance verification criteria on how licence conditions are to be met. (author)

  20. The AMMON experiment in EOLE zero power facility. A challenging program devoted to the neutron and photon physics

    International Nuclear Information System (INIS)

    The AMMON integral experiment was carried out in the EOLE critical mock-up at CEA Cadarache. It has been dedicated to the analysis of the Jules Horowitz Reactor (JHR) neutron and photon physics and has a very specific design in comparison with previous cores (MISTRAL, FUBILA, etc.) consisting in the loading of UOX or MOX fuel pins in EOLE. The core configurations consisted of an experimental zone of six or seven JHR-type assemblies with U3Si2–Al 27% 235U enriched curved fuel plates, surrounded by a driver zone with enough standard pressurized water reactor (PWR) UOX fuel pins to reach criticality. Five main configurations have been studied during this three-year ambitious program. The purpose of this paper is to synthesize the main features of this program in terms of experimental methods and associated uncertainties. Almost all the measurements have been analyzed with TRIPOLI-4® Monte Carlo reference calculations, using the JEFF-3.1.1 nuclear data library; this paper gives examples of the analysis of the critical states studied in the experiment as well as the analysis of the isothermal temperature coefficient. (author)

  1. Comparison of zero-dimensional and one-dimensional thermonuclear burn computations for the reversed-field pinch reactor (RFPR)

    International Nuclear Information System (INIS)

    Conceptual fusion reactor designs of the Reversed-Field Pinch Reactor (RFPR) have been based on profile-averaged zero-dimensional (point) plasma models. The plasma response/performance that has been predicted by the point plasma model is re-examined by a comprehensive one-dimensional (radial) burn code that has been developed and parametrically evaluated for the RFPR. Agreement is good between the zero-dimensional and one-dimensional models, giving more confidence in the RFPR design point reported previously from the zero-dimensional analysis

  2. Reactor power system deployment and startup

    Science.gov (United States)

    Wetch, J. R.; Nelin, C. J.; Britt, E. J.; Klein, G.

    1985-01-01

    This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems.

  3. Residual stress measurements in the dissimilar metal weld in pressurizer safety nozzle of nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Campos, Wagner R.C.; Rabello, Emerson G.; Mansur, Tanius R.; Scaldaferri, Denis H.B.; Paula, Raphael G., E-mail: wrcc@cdtn.br, E-mail: egr@cdtn.br, E-mail: tanius@cdtn.br, E-mail: dhbs@cdtn.br, E-mail: tanius@cdtn.br, E-mail: raphaelmecanica@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Souto, Joao P.R.S.; Carvalho Junior, Ideir T., E-mail: joprocha@yahoo.com.br, E-mail: ideir_engenharia@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Metalurgica

    2013-07-01

    Weld residual stresses have a large influence on the behavior of cracking that could possibly occur under normal operation of components. In case of an unfavorable environment, both stainless steel and nickel-based weld materials can be susceptible to stress-corrosion cracking (SCC). Stress corrosion cracks were found in dissimilar metal welds of some pressurized water reactor (PWR) nuclear plants. In the nuclear reactor primary circuit the presence of tensile residual stress and corrosive environment leads to so-called Primary Water Stress Corrosion Cracking (PWSCC). The PWSCC is a major safety concern in the nuclear power industry worldwide. PWSCC usually occurs on the inner surface of weld regions which come into contact with pressurized high temperature water coolant. However, it is very difficult to measure the residual stress on the inner surfaces of pipes or nozzles because of inaccessibility. A mock-up of weld parts of a pressurizer safety nozzle was fabricated. The mock-up was composed of three parts: an ASTM A508 C13 nozzle, an ASTM A276 F316L stainless steel safe-end, an AISI 316L stainless steel pipe and different filler metals of nickel alloy 82/182 and AISI 316L. This work presents the results of measurements of residual strain from the outer surface of the mock-up welded in base metals and filler metals by hole-drilling strain-gage method of stress relaxation. (author)

  4. Fusion experimental power reactor (EPR) design tasks

    International Nuclear Information System (INIS)

    Several key physics and technology problem areas which were identified in a previous Experimental Power Reactor study were investigated. These were plasma confinement, plasma heating, reactor refueling, and reactor first wall regeneration. The plasma confinement experimental studies showed no instabilities or enhanced transport in the trapped ion regime. The RF heating experiments indicated that RF could produce highly efficient plasma heating. Two reactor refueling schemes were considered in a theoretical analysis: the first was the convective transport from the cold plasma blanket to the plasma interior and the second was the use of high speed frozen pellets to carry the fuel to the plasma interior. Both schemes were shown to be feasible. Finally, the ''in-situ'' replacement of first walls using atomic coating processes were considered. The vapor deposition of carbon was shown to be promising

  5. Fusion experimental power reactor (EPR) design tasks

    International Nuclear Information System (INIS)

    Several key physics and technology problem areas which were identified in the previous Experimental Power Reactor study were investigated. These were plasma confinement, plasma heating, reactor refueling, and reactor first wall regeneration. The plasma confinement experimental studies showed no instabilities or enhanced transport in the trapped ion regime. The RF heating experiments indicated that RF could produce highly efficient plasma heating. Two reactor refueling schemes was considered in a theoretical analysis: the first was the convective transport from the cold plasma blanket to the plasma interior and the second was the use of high speed frozen pellets to carry the fuel to the plasma interior. Both schemes were shown to be feasible. Finally, the in-situ replacement of first walls using atomic coating processes was considered. The vapor deposition of carbon was shown to be promising

  6. Renewal of safety circuitry on a zero-energy research reactor using microprocessor units

    International Nuclear Information System (INIS)

    The conventional hard-wired safety-circuitry of the zero-energy research reactor at the Central Electricity Generating Board's Berkeley Nuclear Laboratories is being replaced by microprocessor-based units. The Paper describes how levels of reliability that are necessary for safety circuitry have been achieved by the use of two entirely different guard line systems based on a Motorola 6800 microprocessor and an Intel 8085A microprocessor. The two systems operate in parallel and either will trip the reactor. Each has been programmed by a different programmer using different philosophies. The two units and the test programme involving over 106 simulated guard line trips are described. An overall reliability of better than 10-6 per annum is claimed. (author)

  7. Numerical benchmarks TRIPOLI - MCNP with use of MCAM on FNG ITER bulk shield and FNG HCLL TBM mock-up experiments

    International Nuclear Information System (INIS)

    3D Monte Carlo (MC) transport codes are of first importance for the assessment of breeding blankets neutronic performances. This article supported by the EFDA Goal Oriented Training Program Eurobreed presents the difference in results between the CEA MC code TRIPOLI-4 and MCNP on two fusion neutronics benchmarks, assessing therefore TRIPOLI-4 calculation capabilities on shielding and tritium production rate (TPR). The first selected benchmark, assessing the shielding capability, is the Frascati neutron generator (FNG) ITER bulk shield experiment whereas the second benchmark, assessing the TPR calculation, is the preliminary design of the FNG helium cooled lithium-lead (HCLL) test blanket module (TBM) mock-up. To ensure the consistency of the geometry description, MCAM tool is used for automatic TRIPOLI - MCNP geometry conversions and check. A good coherence between TRIPOLI-4 and MCNP for neutron flux, reaction rates and TPR calculations is obtained. Moreover, it appears that MCAM performs fast, automatic and appropriate TRIPOLI - MCNP geometry conversions and finally that the tabulated FNG neutron source model from KIT is appropriate for TRIPOLI-4 calculations.

  8. Performance test results of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Inagaki, Yoshiyuki; Hayashi, Koji; Kato, Michio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2003-03-01

    Research on a hydrogen production system by steam reforming of methane, chemical reaction; CH{sub 4} + H{sub 2}O {yields} 3H{sub 2}O + CO, has been carried out to couple with the HTTR for establishment of high-temperature nuclear heat utilization technology and contribution to hydrogen energy society in future. The mock-up test facility with a full-scale reaction tube test facility, a model simulating one reaction tube of a steam reformer of the HTTR hydrogen production system in full scale, was fabricated to perform tests on controllability, hydrogen production performance etc. under the same pressure and temperature conditions as those of the HTTR hydrogen production system. The design and fabrication of the test facility started from 1997, and the all components were installed until September in 2001. In a performance test conducted from October in 2001 to February in 2002, performance of each component was examined and hydrogen of 120m{sup 3}{sub N}/h was successfully produced with high-temperature helium gas. This report describes the performance test results on components performance, hydrogen production characteristics etc., and main troubles and countermeasures. (author)

  9. CFD study of natural convection mixing in a steam generator mock-up: Comparison between full geometry and porous media approaches

    International Nuclear Information System (INIS)

    In CFD simulations of flow mixing in a steam generator (SG) during natural circulation, one is faced with the problem of representing the thousands of SG U-tubes. Typically simplifications are made to render the problem computationally tractable. In particular, one or a number of tubes are lumped in one volume which is treated as a single porous medium. This approach dramatically reduces the computational size of the problem and hence simulation time. In this work, we endeavor to investigate the adequacy of this approach by performing two separate simulations of flow in a mock-up with 262 U-tubes, i.e. one in which the porous media model is used for the tube bundle, and another in which the full geometry is represented. In both simulations, the Reynolds Stress (RMS) model of turbulence is used. We show that in steady state conditions, the porous media treatment yields results which are comparable to those of the full geometry representation (temperature distribution, recirculation ratio, hot plume spread, etc). Hence, the porous media approach can be extended with a good degree of confidence to the full scale SG. (authors)

  10. Numerical benchmarks TRIPOLI - MCNP with use of MCAM on FNG ITER bulk shield and FNG HCLL TBM mock-up experiments

    Energy Technology Data Exchange (ETDEWEB)

    Fausser, Clement, E-mail: clement.fausser@cea.fr [CEA, DEN, Saclay, DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette (France); Lee, Yi-Kang [CEA, DEN, Saclay, DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette (France); Villari, Rosaria [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Zeng Qin; Zhang Junjun [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Serikov, Arkady [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology (Germany); Trama, Jean-Christophe; Gabriel, Franck [CEA, DEN, Saclay, DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette (France)

    2011-10-15

    3D Monte Carlo (MC) transport codes are of first importance for the assessment of breeding blankets neutronic performances. This article supported by the EFDA Goal Oriented Training Program Eurobreed presents the difference in results between the CEA MC code TRIPOLI-4 and MCNP on two fusion neutronics benchmarks, assessing therefore TRIPOLI-4 calculation capabilities on shielding and tritium production rate (TPR). The first selected benchmark, assessing the shielding capability, is the Frascati neutron generator (FNG) ITER bulk shield experiment whereas the second benchmark, assessing the TPR calculation, is the preliminary design of the FNG helium cooled lithium-lead (HCLL) test blanket module (TBM) mock-up. To ensure the consistency of the geometry description, MCAM tool is used for automatic TRIPOLI - MCNP geometry conversions and check. A good coherence between TRIPOLI-4 and MCNP for neutron flux, reaction rates and TPR calculations is obtained. Moreover, it appears that MCAM performs fast, automatic and appropriate TRIPOLI - MCNP geometry conversions and finally that the tabulated FNG neutron source model from KIT is appropriate for TRIPOLI-4 calculations.

  11. Electrolysers as a load management mechanism for power systems with wind power and zero-carbon thermal power plant

    Energy Technology Data Exchange (ETDEWEB)

    Troncoso, E. [School of Industrial Engineering, Universidad Las Palmas de Gran Canaria (Spain); Newborough, M. [ITM Power Research Ltd., Mill House, Royston Road, Wendens Ambo, Saffron Walden CB11 4JX (United Kingdom)

    2010-01-15

    For an isolated power system the deployment of a large stock of electrolysers is investigated as a means for increasing the penetrations of wind power plant and zero-carbon thermal power plant. Consideration is given to the sizing and utilization of an electrolyser stock for three electrolyser implementation cases and three operational strategies, installed capacity ranges of 20-100% for wind power and 10-35% for zero-carbon thermal power plant (as proportions of the power system's maximum electrical demand) were investigated. Relative to wind-hydrogen alone, hydrogen yields are substantially increased especially on low-wind days. The average load placed on fossil-fuelled power plant is substantially decreased (while achieving a virtually flat load profile) and the carbon intensity of electricity can be reduced to values of <0.1 kg CO{sub 2}/kWh{sub e}. The trade-offs between the carbon intensity of the electricity delivered, the carbon intensity of the hydrogen produced and the daily hydrogen yield are explored. For example (on the variable wind day for Strategy C with respective wind power and zero-carbon thermal power penetrations of 100% and 35%), if the carbon intensity of hydrogen is relaxed from 0 to 3 kg CO{sub 2}/kg H{sub 2}, the hydrogen yield can be increased from 435 tonnes to 1115 tonnes (which is the energy equivalent of 120% of consumer demand for electricity on that day). The findings suggest that the deployment of electrolysers on both the supply and demand-side of the power system can contribute nationally-significant amounts of zero or low-carbon hydrogen without exceeding the power system's current maximum system demand. (author)

  12. An Innovative Test Platform for Hydrogen Production and Zero Emission Power Generation from Coal

    International Nuclear Information System (INIS)

    The ZECOMIX project, conceived by ENEA in the framework of Italian National Hydrogen Project, is aimed at studying an integrated process that produces both hydrogen and electricity from coal, with zero emissions and very high efficiency. The key element is the integration of a gasification process, characterized by coal hydro-gasification technology and carbon dioxide sequestration, with the power island, where an oxy-combustion occurs. Many optimization analysis and simulations have been carried out demonstrating the possibility to achieve very high net efficiencies (higher than 50% LHV) and very low (quasi-zero) emissions. The project schedule consists of the design, already started, the construction and the operation of an experimental facility finalized to demonstrate the feasibility of the described reference process. The facility will be realized in the ENEA Research Center of Casaccia, near Rome. It consists of a very flexible plant, in which more components can be tested separately or connected together. The plant is provided with a 50 kg/h coal atmospheric fluid bed gasifier, a fluid bed decarbonator/calcinator reactor filled with calcium oxide pellets, a pressurized hydro-gasifier reactor characterized by a pressure variable from 30 to 100 bar, a 100 kWe micro-turbine test bench, with the combustor chamber modified because of de-carbonized syngas fuelling and finally an oxygen/hydrogen combustor test bench, for experimental activities about the definition of stability limits, operative conditions (dilution, temperature pattern, chemicals) and combustion control. Other auxiliary components are mixing station for hydrogen-based syngas production, and an ordinary steam generator. The first part of the research project is aimed at testing the single component, in particular the main preliminary design criteria adopted for hydro-gasification reactor and carbonator reactor are presented in this paper. The second part of the Project is focused on the integration

  13. Simulation of power excursions - Osiris reactor

    International Nuclear Information System (INIS)

    Following the experimental work accomplished in the U.S.A. on Borax 1 and SPERT 1 and the accident of SL 1, the 'Commissariat a l'Energie Atomique' started a research program about the safety of its own swimming Pool reactors, with regard to power excursions. The first research work led to the design of programmed explosive charges, adapted to the simulation of a power excursion. This report describes the application of these methods to the investigation of Osiris safety. (author)

  14. Cell heterogeneity problems in the analysis of zero-power experiments

    International Nuclear Information System (INIS)

    Methods are described for treating plate and pin cell heterogeneity in the preparation of broad group cross-sections used in the analysis of zero-power fast reactor experiments. Methods used at Karlsruhe and Winfrith are summarized and compared, with particular reference to the treatment of resonance shielding, the calculation of broad group spatial fine structure, the treatment leakage and the calculation of anisotropic diffusion coefficients. The problems of cells near boundaries such as core-breeder interfaces and of singularities such as control rods are also considered briefly. Numerical studies carried out to investigate approximations in the methods are described. These include tests of the accuracy of one-dimensional cell-modelling techniques, and the validation by Monte Carlo of methods for treating streaming in the calculation of diffusion coefficients. Comparisons are shown between the heterogeneity effects calculated by the Karlsruhe and Winfrith methods for typical pin and plate cells used in the BIZET experimental programme, and their effect in a whole reactor calculation is indicated. Comparisons are given with measurements which provide tests of the heterogeneity calculations. These include reaction rate scans within pin and plate cells, and reaction rate measurements across sectors of pin and plate fuel, where the flux tilt is determined by the relative reactivity of the pin and plate cells. Finally, the heterogeneity problems arising in the interpretation of reaction rate measurements are discussed. (author)

  15. Utilization of thorium in power reactors

    International Nuclear Information System (INIS)

    The IAEA convened a Panel on the utilization of thorium in power reactors from 14 to 18 June 1965. 45 scientists from 14 countries and two international organizations took part in it. The proceedings of the Panel include 23 survey papers and brief reviews which stress the importance of utilizing thorium. A separate abstract was prepared for each of these papers. Refs, tabs, figs

  16. The program of reactors and nuclear power plants

    International Nuclear Information System (INIS)

    Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined

  17. Reactor power distribution pattern judging device

    International Nuclear Information System (INIS)

    The judging device of the present invention comprises a power distribution readout system for intaking a power value from a fuel segment, a neural network having an experience learning function for receiving a power distribution value as an input variant, mapping it into a desirable property and self-organizing the map, and a learning date base storing a plurality of learnt samples. The read power distribution is classified depending on the similarity thereof with any one of representative learnt power distribution, and the corresponding state of the reactor core is outputted as a result of the judgement. When an error is found in the classified judging operation, erroneous cases are additionally learnt by using the experience and learning function, thereby improving the accuracy of the reactor core characteristic estimation operation. Since the device is mainly based on the neural network having a self-learning function and a pattern classification and judging function, a judging device having a human's intuitive pattern recognition performance and a pattern experience and learning performance is obtainable, thereby enabling to judge the state of the reactor core accurately. (N.H.)

  18. Experimental development of power reactor intelligent control

    International Nuclear Information System (INIS)

    The US nuclear utility industry initiated an ambitious program to modernize the control systems at a minimum of ten existing nuclear power plants by the year 2000. That program addresses urgent needs to replace obsolete instrumentation and analog controls with highly reliable state-of-the-art computer-based digital systems. Large increases in functionality that could theoretically be achieved in a distributed digital control system are not an initial priority in the industry program but could be logically considered in later phases. This paper discusses the initial development of an experimental sequence for developing, testing, and verifying intelligent fault-accommodating control for commercial nuclear power plant application. The sequence includes an ultra-safe university research reactor (TRIGA) and a passively safe experimental power plant (Experimental Breeder Reactor 2)

  19. Utilization of stable isotopes in power reactor

    International Nuclear Information System (INIS)

    The stable isotopes, besides uranium, used in EDF power nuclear reactors are mainly the boron 10 and the lithium 7. Boron is used in reactors as a neutrophagous agent for core reactivity control, and lithium, and more especially lithium 7, is extensively used as a solution in PWR moderators for primary fluid pH control. Boron and lithium ore reserves and producers are presented; industrial isotopic separation techniques are described: for the boron 10, they include dissociative distillation (Sulzer process) and separation on anionic resins, and for lithium 7, ion exchange columns (Cogema). 1 tab

  20. TRIGA research reactors with higher power density

    International Nuclear Information System (INIS)

    The recent trend in new or upgraded research reactors is to higher power densities (hence higher neutron flux levels) but not necessarily to higher power levels. The TRIGA LEU fuel with burnable poison is available in small diameter fuel rods capable of high power per rod (∼48 kW/rod) with acceptable peak fuel temperatures. The performance of a 10-MW research reactor with a compact core of hexagonal TRIGA fuel clusters has been calculated in detail. With its light water coolant, beryllium and D2O reflector regions, this reactor can provide in-core experiments with thermal fluxes in excess of 3 x 1014 n/cm2·s and fast fluxes (> 0.1 MeV) of 2 x 1014 n/cm2·s. The core centerline thermal neutron flux in the D2O reflector is about 2 x 1014 n/cm2·s and the average core power density is about 230 kW/liter. Using other TRIGA fuel developed for 25-MW test reactors but arranged in hexagonal arrays, power densities in excess of 300 kW/liter are readily available. A core with TRIGA fuel operating at 15-MW and generating such a power density is capable of producing thermal neutron fluxes in a D2O reflector of 3 x 1014 n/cm2·s. A beryllium-filled central region of the core can further enhance the core leakage and hence the neutron flux in the reflector. (author)

  1. Small power sodium cooled fast nuclear reactors

    International Nuclear Information System (INIS)

    1.5 MW(e), 12 MW(e) and 170 MW(e) small power sodium cooled fast reactors have been developed. The reactor plants were developed as universal power units for economically effective energy and industrial steam generation and heat supply. The main features increasing the power unit economic efficiency are: serial fabrication of standard RPs at the factory and delivery of reactor vessels in ready made form; realization of self-protection principles and use of passive systems in RP; use of standard machine room equipment, fabricated in accordance with the rules of conventional heat power engineering; use of turbine plant with thermodynamic coefficient, exceeding the corresponding value for the plants of PWR type. For MBRU-1.5 and MBRU-12 RPs it is proposed to use a core without FA replacement during the whole service life (30 years) and for BMN-170 RP it is proposed to use a core with a 4 year operating period and 1 year between the refueling shutdowns. During the whole service life a minimal number of operating personnel will be needed for the plant servicing. The personnel functions will be periodically to observe the parameters of technological process. Passive principles are used in the main RP safety systems: a passive type system of emergency residual heat removal system provides heat removal directly through the reactor vessel forced air cooling due to the natural air chimney effect; an emergency reactor shut-down system is provided by emergency protection rods with active-passive action. (author)

  2. WWER safety investigations on LR-0 reactor

    International Nuclear Information System (INIS)

    A set of the measurement needed for the WWER-440 and WWER-1000 reactor lifetime assessment, verification of the methods, codes and input cross section libraries for the WWER reactor pressure vessel exposure evaluation has been performed on the LR-0 experimental reactor. The WWER Mock-ups (engineering benchmarks) has been carried out on the reactor, with the aim to investigate differential neutron spectra for reactor dosimetry purposes. Critical experiments have also been performed to determine the perturbation of the fission density distribution caused by the WWER-440 control assembly. Such assembly, partially inserted in the core, has significant influence on the space power distribution. A wide research program for sub-criticality investigations of the spent nuclear fuel storage has been realized on the LR-0 reactor. A benchmark experiment is realized on the reactor in corresponding geometry for CASTOR 440/84 container for storage and transportation of spent fuel. Critical experiments with new fuel assemblies including various burnable absorbers and different enrichments are performed. A set of critical experiments is performed using the fuel assemblies with 3,6% and 4,4% enrichment, arranged in the WWER-440 type cores with various lattice pitch. The critical high of the moderator level and the moderator level coefficient of reactivity are measured and the effect of the fuel assembly, placed in a hexagonal tube of stainless steel containing boron absorber (ATABOR - STANDARD) is investigated. The obtained results are used for the validation of the codes (MCNP, KENO and SCALE) in the frame of the contract 'Burn-up credit implementation for the storage and transport containers of the spent fuel'. Combined neutron-gamma spectra measurements in the WWER-1000 Mock-up are carried out during 2001

  3. Safety Analysis for Power Reactor Protection System

    International Nuclear Information System (INIS)

    The main function of a Reactor Protection System (RPS) is to safely shutdown the reactor and prevents the release of radioactive materials. The purpose of this paper is to present a technique and its application for used in the analysis of safety system of the Nuclear Power Plant (NPP). A more advanced technique has been presented to accurately study such problems as the plant availability assessments and Technical Specifications evaluations that are becoming increasingly important. The paper provides the Markov model for the Reactor Protection System of the NPP and presents results of model evaluations for two testing policies in technical specifications. The quantification of the Markov model provides the probability values that the system will occupy each of the possible states as a function of time.

  4. Neutron fluctuations in accelerator driven and power reactors via backward master equations

    International Nuclear Information System (INIS)

    The transport of neutrons in a reactor is a random process, and thus the number of neutrons in a reactor is a random variable. Fluctuations in the number of neutrons in a reactor can be divided into two categories, namely zero noise and power reactor noise. As the name indicates, they dominate (i.e. are observable) at different power levels. The reasons for their occurrences and utilization are also different. In addition, they are described via different mathematical tools, namely master equations and the Langevin equation, respectively. Zero noise carries information about some nuclear properties such as reactor reactivity. Hence methods such as Feynman- and Rossi-alpha methods have been established to determine the subcritical reactivity of a subcritical system. Such methods received a renewed interest recently with the advent of the so-called accelerator driven systems (ADS). Such systems, intended to be used either for energy production or transuranium transmutation, will use a subcritical core with a strong spallation source. A spallation source has statistical properties that are different from those of the traditionally used radioactive sources which were also assumed in the derivation of the Feynman- and Rossi-alpha formulae. Therefore it is necessary to re-derive the Feynman- and Rossi-alpha formulae. Such formulae for ADS have been derived recently but in simpler neutronic models. One subject of this thesis is the extension of such formulae to a more general case in which six groups of delayed neutron precursors are taken into account, and the full joint statistics of the prompt and all delayed groups is included. The involved complexity problems are solved with a combination of effective analytical techniques and symbolic algebra codes. Power reactor noise carries information about parametric perturbation of the system. Langevin technique has been used to extract such information. In such a treatment, zero noise has been neglected. This is a pragmatic

  5. Static and dynamic performance tests of nuclear powered ship Mutsu reactor (report on nuclear ship Mutsu power-up tests)

    International Nuclear Information System (INIS)

    The power-up tests of the Mutsu reactor were performed from March 29th 1990 to December 14th. The tests were divided into six phases: The tests Phase 0 and Phase 1 were done in the state that the ship was moored at the quay of Sekinehama port in March and April; The tests Phase 2, Phase 3, Phase 4, and Phase 5 were done on the Pacific Ocean from July to December. Present report describes the test results on the static and dynamic plant performance. On static plant performance tests, there are 13 test items including measurements of primary system heat balance at low and high power levels, a virgin run of feed water pump with SG steam, a change-over test of steam supply of auxiliary boiler to SG. On the dynamic plant performance, there are 11 test items including a test of reactor power auto-control system, a test of main feed water auto-control system, a test of small load variation, a load increasing test, a turbine trip test, tests of ahead and astern maneuvering, a test of single loop operation, and a reactor scram test. The reactor power for each item's test was increased step by step from zero power to the goal of rated power of 100 %, 36 MWt. In order to confirm proper reactor system performance, criteria were laid down for the static and dynamic tests: for example, (1) reactor scram shall not occur, (2) pressurizer relief valve and steam generator safety valve shall not work, and (3) after the transients reactor systems shall become the steady state without manual adjustment of the reactor control system. The test results satisfied these criteria and some of test data showed that reactor had much more margin in any performance for design. It is verified, therefore, that the Mutsu reactor systems have adequate performances as a marine reactor and that one is capable to respond smoothly and safely to the load of ship's demand. (author)

  6. Damage and fatigue crack growth of Eurofer steel first wall mock-up under cyclic heat flux loads. Part 1: Electron beam irradiation tests

    International Nuclear Information System (INIS)

    Highlights: • Clear evidence of microscopic damage and crack formation at the notch root in the early stage of the fatigue loading (50–100 load cycles). • Propagation of fatigue crack at the notch root in the course of subsequent cyclic heat-flux loading followed by saturation after roughly 600 load cycles. • No sign of damage on the notch-free surface up to 800 load cycles. • No obvious effect of the pulse time duration on the crack extension. • Slight change in the grain microstructure due to the formation of sub-grain boundaries by plastic deformation. - Abstract: Recently, the idea of bare steel first wall (FW) is drawing attention, where the surface of the steel is to be directly exposed to high heat flux loads. Hence, the thermo-mechanical impacts on the bare steel FW will be different from those of the tungsten-coated one. There are several previous works on the thermal fatigue tests of bare steel FW made of austenitic steel with regard to the ITER application. In the case of reduced-activation steel Eurofer97, a candidate structural material for the DEMO FW, there is no report on high heat flux tests yet. The aim of the present study is to investigate the thermal fatigue behavior of the Eurofer-based bare steel FW under cyclic heat flux loads relevant to DEMO operation. To this end, we conducted a series of electron beam irradiation tests with heat flux load of 3.5 MW/m2 on water-cooled mock-ups with an engraved thin notch on the surface. It was found that the notch root region exhibited a marked development of damage and fatigue cracks whereas the notch-free surface manifested no sign of crack formation up to 800 load cycles. Results of extensive microscopic investigation are reported

  7. High Power Zero-Voltage and Zero-Current Switching DC-DC Converters

    Directory of Open Access Journals (Sweden)

    Jaroslav Dudrik

    2005-01-01

    Full Text Available The paper presents principles and properties of the soft switching PWM DC-DC converters. The attention is focused mainly on high power applications and thus the full-bridge inverters are used in DC-DC converters. Considerations are also given to the control methods and principles of the switching and conduction losses reduction.

  8. 78 FR 64028 - Decommissioning of Nuclear Power Reactors

    Science.gov (United States)

    2013-10-25

    ... COMMISSION Decommissioning of Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION... regulatory guide (RG) 1.184 ``Decommissioning of Nuclear Power Reactors.'' This guide describes a method NRC... decommissioning process for nuclear power reactors. The revision takes advantage of the 13 years...

  9. Modular stellarator reactor: a fusion power plant

    International Nuclear Information System (INIS)

    A comparative analysis of the modular stellarator and the torsatron concepts is made based upon a steady-state ignited, DT-fueled, reactor embodiment of each concept for use as a central electric-power station. Parametric tradeoff calculations lead to the selection of four design points for an approx. 4-GWt plant based upon Alcator transport scaling in l = 2 systems of moderate aspect ratio. The four design points represent high-aspect ratio. The four design points represent high-(0.08) and low-(0.04) beta versions of the modular stellarator and torsatron concepts. The physics basis of each design point is described together with supporting engineering and economic analyses. The primary intent of this study is the elucidation of key physics and engineering tradeoffs, constraints, and uncertainties with respect to the ultimate power reactor embodiment

  10. A series of lectures on operational physics of power reactors

    International Nuclear Information System (INIS)

    This report discusses certain aspects of operational physics of power reactors. These form a lecture series at the Winter College on Nuclear Physics and Reactors, Jan. - March 1980, conducted at the International Centre for Theoretical Physics, Trieste, Italy. The topics covered are (a) the reactor physics aspects of fuel burnup (b) theoretical methods applied for burnup prediction in power reactors (c) interpretation of neutron detector readings in terms of adjacent fuel assembly powers (d) refuelling schemes used in power reactors. The reactor types chosen for the discussion are BWR, PWR and PHWR. (author)

  11. Compact approach to fusion power reactors

    International Nuclear Information System (INIS)

    The potential of the Reversed-Field Pinch (RFP) for development into an efficient, compact, copper-coil fusion reactor has been quantified by comprehensive parametric tradeoff studies. These compact systems promise to be competitive in size, power density, and cost to alternative energy sources. Conceptual engineering designs that largely substantiate these promising results have since been completed. This 1000-MWe(net) design is described along with a detailed rationale and physics/technology assessment for the compact approach to fusion

  12. Leaching of nuclear power reactor wastes forms

    International Nuclear Information System (INIS)

    The leaching tests for power reactor wastes carried out at IPEN/CNEN-SP are described. These waste forms consist mainly of spent resins and boric acid concentrates solidified in ordinary Portland cement. All tests were conducted according to the ISO and IAEA recommendations. 3 years leaching results are reported, determining cesium and strontium diffusivity coefficients for boric acid waste form and ion-exchange resins. (Author)

  13. Small and medium power reactors 1985

    International Nuclear Information System (INIS)

    This report is intended for designers and planners concerned with Small and Medium Power Reactors. It provides a record of the presentations during the meetings held on this subject at the Agency's General Conference in September 1985. This information should be useful as it indicates the principal findings and main conclusions and recommendations resulting from these meetings. A separate abstract was prepared for each of the 10 presentations in this report

  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. Nuclear power reactors and hydrogen storage systems

    International Nuclear Information System (INIS)

    Among conclusions and results come by, a nuclear-electric-hydrogen integrated power system was suggested as a way to prevent the energy crisis. It was shown that the hydrogen power system using nuclear power as a leading energy resource would hold an advantage in the current international situation as well as for the long-term future. Results reported provide designers of integrated nuclear-electric-hydrogen systems with computation models and routines which will allow them to explore the optimal solution in coupling power reactors to hydrogen producing systems, taking into account the specific characters of hydrogen storage systems. The models were meant for average computers of a type easily available in developing countries. (author)

  16. Zero-Net Power, Low-Cost Sensor Platform

    Energy Technology Data Exchange (ETDEWEB)

    Hardy, J.E.

    2005-04-15

    Numerous national studies and working groups have identified very low-power, low-cost sensors as a critical technology for increasing energy efficiency, reducing waste, and optimizing processes. This research addressed that need by developing an ultra low-power, low-cost sensor platform based on microsensor (MS) arrays that includes MS sensors, very low-power electronics, signal processing, and two-way data communications, all integrated into a single package. MSs were developed to measure carbon dioxide and room occupancy. Advances were made in developing a coating for detecting carbon dioxide and sensing thermal energy with MSs with a low power electrical readout. In addition, robust algorithms were developed for communications within buildings over power lines and an integrated platform was realized that included gas sensing, temperature, humidity, and room occupancy with on-board communications.

  17. Power reactor noise studies and applications

    International Nuclear Information System (INIS)

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  18. Power reactor noise studies and applications

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, V

    2002-03-01

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  19. Operating U.S. 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 United States which have been issued operating licenses. Table I shows the number of such reactors and their net capacities as of June 30, 1992, 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 covered in each report and the cumulative values of these parameters at the end of the covered quarter since the beginning of commercial operation. The information for this table was obtained from the Nuclear Regulatory Commission (NRC) Office of Information Resources Management. The Maximum Dependable Capacity (MDC) Unit Capacity (in percent) is defined as follows: (Net electrical energy generated during the reporting period x 100) divided by the product of the number of hours in the reporting period and the MDC of the reactor in question. The forced outage rate (in percent) is defined as: (The total number of hours in the reporting period during which the unit was inoperable as the result of a forced outage x 100) divided by the sum (forced outage hours + operating hours)

  20. Operating U.S. 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 United States which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of Sept. 30, 1990, the end of the three-month period covered in this report. Table 2 used to list the unit capacity and forced outage rate for each licensed reactor for each of the three months covered in each report and the cummulative values of these parameters since the beginning of commercial operation. Since, however, Nuclear Regulatory Commission (NRC) publication NUREG-0200, known as 'The Gray Book,' which has been the source of this information, has been discontinued, in this issue, and henceforth until other data again become available from the NRC, this table will contain maximum dependable capacity (MDC) factors for each month and the year-to-date MDC capacity as of the end of the quarter covered, in this issue the end of September 1990. The information for this table is derived from World Nuclear Performance, published by McGraw-Hill Nuclear Publications and used with their permission. The MDC Unit Capacity (in percent) is defined as follows: (Net electrical energy generated during the reporting period x 100) divided by the product of the number of hours in the reporting period and the MDC of the reactor in question

  1. FEBEX Full-Scalle Engineered Barriers Experiment in Crystalline Host Rock Preoperational Thermo-Hydro-Mechanical (THM) Modelling of the Mock Up Test

    International Nuclear Information System (INIS)

    The object of this report is to present and discuss the results of a series of 1-D and 2-D coupled thermo-hydro-mechanical (THM) and 2-D coupled thermo-hydro-mechanical (THM) analyses modelling the FEBEX mock-up test. The analyses have been carried out during the preoperational storage of the test and attempt to incorporate all available information obtained from laboratory characterisation work. The aim is not only to offer the best estimate of test performance using current models and information but also to provide a basis for future model improvements. Both the theoretical framework adopted in the analysis and the computer code employed are briefly described. The set of parameters used in the computation is then presented with particular reference to the source from which they have been derived. Initial and boundary condition are also defined. The results of a 1-D radially symmetric analysis are used to examine the basic patterns of thermal, hydraulic and mechanical behaviour of the test. A set of sensitivity analyses has been carried out in order to check the effects that the variation of a number of important parameters has on test results. Only in this way it is possible to acquire a proper understanding of the internal structure of the problem and of the interactions between the various phenomena occurring in the buffer. A better reproduction of the geometry of the test is achieved by means of a 2-D mesh representing and axisymmetric longitudinal section. Due to two-dimensional effects, the analyses carried out using this geometry exhibit some differences when compared with the results of the 1-D case, but the basic test behaviour is very similar. The test was started with an initial flooding stage with the purpose of closing the gaps between bentonite blocks. A limited number of compilations using recently developed joint elements have been performed to assess approximately the effect of this initial step on subsequent test behaviour. The analyses reported

  2. Validation of finite element code DELFIN by means of the zero power experiences at the nuclear power plant of Atucha I

    International Nuclear Information System (INIS)

    Code DELFIN, developed in CNEA, treats the spatial discretization using heterogeneous finite elements, allowing a correct treatment of the continuity of fluxes and currents among elements and a more realistic representation of the hexagonal lattice of the reactor. It can be used for fuel management calculation, Xenon oscillation and spatial kinetics. Using the HUEMUL code for cell calculation (which uses a generalized two dimensional collision probability theory and has the WIMS library incorporated in a data base), the zero power experiences performed in 1974 were calculated. (author). 8 refs., 9 figs., 3 tabs

  3. 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

  4. Utilization of the low power Musashi reactor

    International Nuclear Information System (INIS)

    Although the Musashi reactor is a low-power reactor of 100 kW, multi-purpose beam-experiments have been performed for the last ten years. Medical irradiation for boron neutron capture therapy (BNCT) is the most unique utilization of the reactor. Eighty-two patients had been treated in the reactor up to the end of August 1987. One of the horizontal beam ports has been used for a time-of-flight experiment by using a slow-chopper since 1977. The authors measured the total neutron cross sections of Mg, Al, Si, Zr, Nb and Mo in the energy range from 0.001 to 0.3 eV. A neutron radiography facility was designed and installed at another beam port in 1984. A real-time neutron TV system has been also installed for investigation of moving objects and for a neutron computed tomography study. A third beam port has been used for a filtered beam experiment and a capture γ-ray measurement. An Fe-filter for 24 keV neutrons and Si-filter for 54 and 144 keV neutrons are available for generating monochromatic neutrons. These beams have been used for the precise measurement of total neutron cross sections. The capture γ-ray measurements have been applied for the measurement of boron concentration in tissue in connection with the BNCT. The reactor has a Joint Use Program for university researchers in Japan under a grant-in-aid by the Ministry of Education, Science and Culture. (author)

  5. Advanced power reactors with improved safety characteristics

    International Nuclear Information System (INIS)

    The primary objective of nuclear safety is the protection of individuals, society and environment against radiological hazards from accidental releases of radioactive materials contained in nuclear reactors. Hereto, these materials are enclosed by several successive barriers and the barriers protected against mishaps and accidents by a multi-level system of safety precautions. The evolution of reactor technology continuously improves this concept and its implementation. At a world-wide scale, several advanced reactor concepts are currently being considered, some of them already at a design stage. Essential safety objectives include both further strengthening the prevention of accidents and improving the containment of fission products should an accident occur. The proposed solutions differ considerably with regard to technical principles, plant size and time scales considered for industrial application. Two typical approaches can be distinguished: The first approach basically aims at an evolution of power reactors currently in use, taking into account the findings from safety research and from operation of current plants. This approach makes maximum use of proven technology and operating experience but may nevertheless include new safety features. The corresponding designs are often termed 'large evolutionary'. The second approach consists in more fundamental changes compared to present designs, often with strong emphasis on specific passive features protecting the fuel and fuel cladding barriers. Owing to the nature and capability of those passive features such 'innovative designs' are mostly smaller in power output. The paper describes the basic objectives of such developments and illustrates important technical concepts focusing on next generation plants, i.e. designs to be available for industrial application until the end of this decade. 1 tab. (author)

  6. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    therapy machines. Today the majority of the cancer therapy cobalt-60 sources used in the world are manufactured using material from the NRU reactor in Chalk River. The same technology that was used for producing cobalt-60 in a research reactor was then adapted and transferred for use in a CANDU power reactor. In the early 1970s, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production was initiated in the four Pickering A CANDU reactors located east of Toronto. This was the first full scale production of millions of curies of cobalt-60 per year. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology in additional CANDUs. Over the years MDS Nordion has partnered with CANDU reactor owners to produce cobalt-60 at various sites. CANDU reactors that have, or are still producing cobalt-60, include Pickering A, Pickering B, Gentilly 2, Embalse in Argentina, and Bruce B. In conclusion, the technology for cobalt-60 production in CANDU reactors, designed and developed by MDS Nordion and Atomic Energy of Canada, has been safely, economically and successfully employed in CANDU reactors with over 195 reactor years of production. Today over forty percent of the world's disposable medical supplies are made safer through sterilization using cobalt-60 sources from MDS Nordion. Over the past 40 years, MDS Nordion with its CANDU reactor owner partners, has safely and reliably shipped more than 500 million curies of cobalt-60 sources to customers around the world. MDS Nordion is presently adding three more CANDU power reactors to its supply chain. These three additional cobalt producing CANDU's will help supplement the ability of the health care industry to provide safe, sterile, medical disposable products to people around the world. As new applications for cobalt-60 are identified, and the demand for bulk cobalt-60 increases, MDS Nordion and AECL

  7. Leaching of nuclear power reactor waste forms

    International Nuclear Information System (INIS)

    The leaching tests for immobilized power reactor wastes carried out at IPEN are described. These wastes forms consist mainly of spent resins and boric acid concentrates solidified in ordinary Portland cement. All tests were conducted according to the ISO and IAEA recommendations. Three years leaching results are reported. The cesium diffuvity coefficients determined out of these results are about 1 x 10-8 cm2/s for boric acid waste form and 9 x 10-9 cm2/s for ion-exchange resin waste. Strontium diffusivity coefficients found are about 3 x 10-11 cm2/s and 9 x 10-11 cm2/s respectively. (Author)

  8. Fuel assembly identification for nuclear power reactors

    International Nuclear Information System (INIS)

    The standard refers to fuel assemblies of light-water power reactors. It contains stipulations for uniform marking in order that the fuel assemblies may be identified. A figure consisting of 8 alpha-numerical characters is used for marking, the first three of which represent the operator who ordered the fuel assembly, while the four characters to follow symbolize a series number. The last character serves as a test mark to scrutinize reading mistakes. The alpha-numerical characters include the Arabic numerals 0-9 and, following them, the letters A-Y of the German alphabet, leaving out B, F, I, O, Q, Z (30 characters). (orig./HP)

  9. Fuel assembly identification for power reactors

    International Nuclear Information System (INIS)

    The standard refers to fuel assemblies of light-water power reactors. It contains stipulations for uniform marking in order that the fuel assemblies may be identified. A figure consisting of 8 alpha-numerical numbers is used for marking, the first three of which represent the operator who ordered the fuel assembly, while the four numbers to follow symbolize a series number. The last number serves as a test mark to scrutinize reading mistakes. The alpha-numerical numbers include the Arabic numerals 0-9 and, following them, the letters A-Y of the German alphabet, leaving out B, F, I, O, Q, Z (30 characters). (orig./HP)

  10. U.S. Nuclear Power Reactor Plant Status

    Data.gov (United States)

    Nuclear Regulatory Commission — Demographic data on U.S. commercial nuclear power reactors, including: plant name/unit number, docket number, location, licensee, reactor/containment type, nuclear...

  11. To question of NPP power reactor choice for Kazakhstan

    International Nuclear Information System (INIS)

    Full text: The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance

  12. To question of NPP power reactor choice for Kazakhstan

    International Nuclear Information System (INIS)

    The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance in the

  13. Nuclear Power Reactors in the World. 2010 Edition

    International Nuclear Information System (INIS)

    This is the thirtieth edition of Reference Data Series No. 2, Nuclear Power Reactors in the World, which is published once per year, to present the most recent reactor data available to the Agency. It contains the following summarized information: - General and technical information as of the end of 2009 on power reactors operating or under construction, and shut down; - Performance data on reactors operating in the Agency's Member States, as reported to the IAEA.

  14. 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

  15. 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)

  16. Cascade: a high-efficiency ICF power reactor

    International Nuclear Information System (INIS)

    Cascade attains a net power-plant efficiency of 49% and its cost is competitive with high-temperature gas-cooled reactor, pressurized-water reactor, and coal-fired power plants. The Cascade reactor and blanket are made of ceramic materials and activation is 6 times less than that of the MARS Tandem Mirror Reactor operating at comparable power. Hands-on maintenance of the heat exchangers is possible one day after shutdown. Essentially all tritium is recovered in the vacuum system, with the remainder recovered from the helium power conversion loop. Tritium leakage external to the vacuum system and power conversion loop is only 0.03 Ci/d

  17. 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

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

    International Nuclear Information System (INIS)

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

  19. Oregon State TRIGA reactor power calibration study

    International Nuclear Information System (INIS)

    As a result of a recent review of the Oregon State TRIGA Reactor (OSTR) power calibration procedure, an investigation was performed on the origin and correctness of the OSTR tank factor and the calibration method. It was determined that there was no clear basis for the tank factor which was being used (0.0525 deg. C/kwh) and therefore a new value was calculated (0.0493 deg. C/kwh). The calculational method and likely errors are presented in the paper. In addition, a series of experimental tests were conducted to decide if the power calibration was best performed with or without a mixer, at 100 KW or at 1 MW. The results of these tests along with the final recommendation are presented. (author)

  20. Neutron measurements at nuclear power reactors [55

    CERN Document Server

    Scherpelz, R I

    2002-01-01

    Staff from the Pacific Northwest National Laboratory (operated by Battelle Memorial Institute), have performed neutron measurements at a number of commercial nuclear power plants in the United States. Neutron radiation fields at light water reactor (LWR) power plants are typically characterized by low-energy distributions due to the presence of large amounts of scattering material such as water and concrete. These low-energy distributions make it difficult to accurately monitor personnel exposures, since most survey meters and dosimeters are calibrated to higher-energy fields such as those produced by bare or D sub 2 O-moderated sup 2 sup 5 sup 2 Cf sources. Commercial plants typically use thermoluminescent dosimeters in an albedo configuration for personnel dosimetry and survey meters based on a thermal-neutron detector inside a cylindrical or spherical moderator for dose rate assessment, so their methods of routine monitoring are highly dependent on the energy of the neutron fields. Battelle has participate...

  1. Scenario analysis of the nuclear power's role in future zero-carbon electricity system in Japan

    International Nuclear Information System (INIS)

    The realization of a zero-carbon electricity system is of vital importance to a future zero-carbon energy system and society, and nuclear power is expected to contribute to this much more than intermittent, complicated and costly renewald energies in the future in Japan. Therefore, in the present study, the role of nuclear power in Japan's future zero-carbon electricity system was studied using scenario analysis methods. Furthermore, technical feasibility analysis was conducted for electricity systems of the proposed scenarios in terms of reliability for the fluctuations of both daily and seasonal electrical demand and supply using an hour by hour simulation. The results show that nuclear power will contribute at least 60% of electricity production, and the whole systems were proven to be technically feasible with the help of EV batteries and hydrogen for daily and seasonal electricity storages respectively, operated based on smart gird control technologies. (author)

  2. UF6 breeder reactor power plants for electric power generation

    International Nuclear Information System (INIS)

    The reactor concept analyzed is a 233UF6 core surrounded by a molten salt (Li7F, BeF2, ThF4) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. A maximum breeding ratio of 1.22 was found. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. Optimization of a Rankine cycle for a gas core breeder reactor employing an intermediate heat exchanger gave a maximum efficiency of 37 percent. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. The advantages of the GCBR are as follows: (1) high efficiency, (2) simplified on-line reprocessing, (3) inherent safety considerations, (4) high breeding ratio, (5) possibility of burning all or most of the long-lived nuclear waste actinides, and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion

  3. UF6 breeder reactor power plants for electric power generation

    Science.gov (United States)

    Rust, J. H.; Clement, J. D.; Hohl, F.

    1976-01-01

    The reactor concept analyzed is a U-233F6 core surrounded by a molten salt (Li(7)F, BeF2, ThF4) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. Advantages of the gas core breeder reactor (GCBR) are as follows: (1) high efficiency; (2) simplified on-line reprocessing; (3) inherent safety considerations; (4) high breeding ratio; (5) possibility of burning all or most of the long-lived nuclear waste actinides; and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion.

  4. Power generation costs for alternate reactor fuel cycles

    International Nuclear Information System (INIS)

    The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U3O8 price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions

  5. Study on treatment of coking wastewater by biofilm reactors combined with zero-valent iron process

    International Nuclear Information System (INIS)

    Experiments were conducted to investigate the behavior of the integrated system with biofilm reactors and zero-valent iron (ZVI) process for coking wastewater treatment. Particular attention was paid to the performance of the integrated system for removal of organic and inorganic nitrogen compounds. Maximal removal efficiencies of chemical oxygen demand (COD), ammonia nitrogen (NH3-N) and total inorganic nitrogen (TIN) were up to 96.1, 99.2 and 92.3%, respectively. Moreover, it was found that some phenolic compounds were effectively removed. The refractory organic compounds were primarily removed in ZVI process of the integrated system. These compounds, with molecular weights either ranged 10,000-30,000 Da or 0-2000 Da, were mainly the humic acid (HA) and hydrophilic (HyI) compounds. Oxidation-reduction and coagulation were the main removal mechanisms in ZVI process, which could enhance the biodegradability of the system effluent. Furthermore, the integrated system showed a rapid recovery performance against the sudden loading shock and remained high efficiencies for pollutants removal. Overall, the integrated system was proved feasible for coking wastewater treatment in practical applications

  6. Experience in using a research reactor for the training of power reactor operators

    International Nuclear Information System (INIS)

    A research reactor facility such as the one at the Omaha Veterans Administration Hospital would have much to offer in the way of training reactor operators. Although most of the candidates for the course had either received previous training in the Westinghouse Reactor Operator Training Program, had operated nuclear submarine reactors or had operated power reactors, they were not offered the opportunity to perform the extensive manipulations of a reactor that a small research facility will allow. In addition the AEC recommends 10 research reactor startups per student as a prerequisite for a cold operator?s license and these can easily be obtained during the training period

  7. Instrumentation and control for reactor power setback in PFBR

    International Nuclear Information System (INIS)

    In Prototype Fast Breeder Reactor (PFBR), a 500 MWe plant, Reactor Power Setback is a special operation envisaged for bulk power reduction on occurrence of certain events in Balance of Plant. The bulk power reduction requires a large negative reactivity perturbation if reactor is operating on nominal power. This necessitates a reliable monitoring system with fault tolerant I and C architecture in order to inhibit reactor SCRAM on negative reactivity trip signal. The impact of above events on the process is described. Design of a functional prototype module to carry out RPSB logic operation and its interface with other instruments has been discussed. (author)

  8. Management of radioactive wastes at power reactor sites in India

    International Nuclear Information System (INIS)

    Indian nuclear power programme, at the present stage, is based on natural uranium fuelled heavy water moderated CANDU type reactors except for the first nuclear power station consisting of two units of enriched uranium fuelled, light water moderated, BWR type of reactors. Some of the salient aspects of radioactive waste management at power reactor sites in India are discussed. Brief reviews are presented on treatment of wastes, their disposal and environmental aspects. Indian experience in power reactor waste management is also summarised identifying some of the areas needing further work. (auth.)

  9. Pellet bed reactor for multi-modal space power

    International Nuclear Information System (INIS)

    A review of forthcoming space power needs for both civil and military missions indicates that power requirements will be in tens of megawatts. It is envisioned that the electrical power requirements will be two-fold; long-duration low power will be needed for station keeping, communications and/or surveillance, while short-duration high power will be required for pulsed power devices. These power characteristics led to authors to propose a multi-modal space power reactor using a pellet bed design. Characteristics desired for such a multi-megawatt reactor power source are the following: standby, alert and pulsed power modes; high thermal output heat source (around 1000 MWt peak power); long lifetime standby power (10-30 yrs); high temperature output (1500-1750 K); rapid burst power transition; high reliability (>95%); and meets stringent safety requirements. The proposed pellet bed reactor concept is designed to satisfy these characteristics

  10. Nuclear problems of power reactors safety

    International Nuclear Information System (INIS)

    The objective of this presentation is to emphasize the conditions that would be of high importance in safety analyses concerned first of all with reactor core. It describes reactor kinetics processes in the core and build up of fission products, classification of reactor accidents related to the core, risk estimation and includes a list of importance reactor accidents

  11. Project WAGR: The UK demonstration project for power reactor decommissioning - removing the core and looking to completion

    Energy Technology Data Exchange (ETDEWEB)

    Benest, T. G. [United Kingdom Atomic Energy Authority, Headquarters, London (United Kingdom)

    2003-07-01

    The United Kingdom Atomic Energy Authority (UKAEA) has built and operated a wide range of nuclear facilities since the late 1940's. UKAEA's present mission is to restore the environment of these facilities in a safe and environmentally responsible manner. This restoration includes the decommissioning of a number of redundant research and power reactors, one of which is the Windscale Advanced Gas-cooled Reactor (WAGR). Following shut down, UKAEA decided to continue the prototype function of the reactor into the decommissioning phase to develop dismantling techniques and establish waste routes. The reactor core and pressure vessel are now being dismantled in a programme of 10 campaigns, seven of which have been completed since 1998. It is anticipated that the current programme will be completed by summer 2005. This paper outlines the history of the reactor, the operation of the waste-processing route, the installed dismantling equipment and the successful completion of the first seven campaigns. This earlier work has been described in a number of publications and conferences, so this paper concentrates on recent work to select and develop cutting equipment to dismantle the core support structures and the pressure vessel. The decommissioning of the Windscale Advance Gas-cooled reactor is being undertaken to demonstrate that a power reactor can be decommissioned shortly after shutdown. The removal of the core and pressure vessel has been broken down into a series of 10 campaigns associated with particular core components. The first 7 campaigns have been successfully completed and the 8., is expected to commence in September 2003 17 months earlier than planned. Dismantling methodologies and tools have been developed specifically for each of these campaigns. Full-scale mock-ups have been used to test the tools, train the operators and assess the duration of operations. However, despite successful trials, operational experience has shown that some of these tools

  12. Intercomparison of liquid metal fast reactor seismic analysis codes. V.1: Validation of seismic analysis codes using reactor core experiments. Proceedings of a research co-ordination meeting held in Vienna, 16-17 November 1993

    International Nuclear Information System (INIS)

    The Research Co-ordination Meeting held in Vienna, 16-17 November 1993, was attended by participants from France, India, Italy, Japan and the Russian Federation. The meeting was held to discuss and compare the results obtained by various organizations for the analysis of Italian tests on PEC mock-up. The background paper by A. Martelli, et al., Italy, entitled Fluid-Structure Interaction Experiments of PEC Core Mock-ups and Numerical Analysis Performed by ENEA presented details on the Italian PEC (Prova Elementi di Combustibile, i.e. Fuel Element Test Facility) test data for the benchmark. Several papers were presented on the analytical investigations of the PEC reactor core experiments. The paper by M. Morishita, Japan, entitled Seismic Response Analysis of PEC Reactor Core Mock-up, gives a brief review of the Japanese data on the Monju mock-up core experiment which had been distributed to the participating countries through the IAEA. Refs, figs and tabs

  13. 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)

  14. Full-core pin-power calculations using Monte Carlo codes

    International Nuclear Information System (INIS)

    Pin wise calculations of core power distribution have been performed for a criticality mock up installation that models a WWER-1000 reactor. Two Monte Carlo codes have been applied for solving of this problem: the MCNP4B code and the KENO-VI code from the SCALE 4.4 system. The codes use different kinds of neutron cross section data: pointwise continuous-energy ENDF/B-VI data and multigroup ENDF/B-V data. Comparisons of calculated results show that the MCNP4B and KENO-VI results are in good agreement. (authors)

  15. In-core dosimetry in CAGR - measurements on power reactors and laboratory facilities

    International Nuclear Information System (INIS)

    The problem of radiolytic corrosion of the graphite moderator in CAGR has led to a need for more accurate information on the radiation dose to the coolant gas in the pores of the graphite. An experimental in-core dosimetry programme is in progress to acquire this data. The problems of in-core dosimetry, particularly that of measuring gamma dose in the presence of high thermal neutron fluences, are described with reference to calorimetry, ionisation chambers and thermoluminescence dosimeters. Progress made in the refinement of these techniques for reactor dosimetry is described. An experiment is described in which dosimetry measurements in components of a Heysham Power Station reactor were made during its commissioning. The major facility of this dosimetry programme is a zero-energy research reactor constructed from CAGR components; this reactor and its experimental facilities are described, together with the results of some of the first experiments. (author)

  16. An overview of future sustainable nuclear power reactors

    OpenAIRE

    Andreas Poullikkas

    2013-01-01

    In this paper an overview of the current and future nuclear power reactor technologies is carried out. In particular, the nuclear technology is described and the classification of the current and future nuclear reactors according to their generation is provided. The analysis has shown that generation II reactors currently in operation all around the world lack significantly in safety precautions and are prone to loss of coolant accident (LOCA). In contrast, generation III reactors, which are ...

  17. Reactor Physics Tests for the Full Power Operation of HANARO

    International Nuclear Information System (INIS)

    The initial criticality of HANARO was achieved on the Feb. 8th of 1995. As HANARO is a unique reactor, there were difficulties to get a license to its full power operation, in which the design power of HANARO is 30 MW. There were two operation license conditions that limited the operation power to 80% of the design power. They were resolved in 2003 and the power ascension tests were conducted for the full power operation. This paper presents the several reactor physics tests for the power ascension to the full power of HANARO

  18. Reactor/Brayton power systems for nuclear electric spacecraft

    International Nuclear Information System (INIS)

    Studies are currently underway to assess the technological feasibility of a nuclear-reactor-powered spacecraft propelled by electric thrusters. The purpose of this study was to provide comparative information on a closed cycle gas turbine power conversion system

  19. Service hall in Number 1 Fukushima Nuclear Power Station, Tokyo Electric Power Company, Inc

    International Nuclear Information System (INIS)

    There are six BWR type nuclear power plants in the Number 1 Fukushima Nuclear Power Station, Tokyo Electric Power Company, Inc. The service hall of the station is located near the entrance of the station. In the center of this service hall, there is the model of a nuclear reactor of full scale. This mock-up shows the core region in the reactor pressure vessel for the number one plant. The diameter and the thickness of the pressure vessel are about 5 m and 16 cm, respectively. The fuel assemblies and control rods are set just like the actual reactor, and the start-up operation of the reactor is shown colorfully and dynamically by pushing a button. When the control rods are pulled out, the boiling of water is demonstrated. The 1/50 scale model of the sixth plant with the power generating capacity of 1100 MWe is set, and this model is linked to the mock-up of reactor written above. The operations of a recirculating loop, a turbine and a condenser are shown by switching on and off lamps. The other exhibitions are shielding concrete wall, ECCS model, and many kinds of panels and models. This service hall is incorporated in the course of study and observation of civics. The good environmental effects to fishes and shells are explained in this service hall. Official buildings and schools are built near the service hall utilizing the tax and grant concerning power generation. This service hall contributes to give much freedom from anxiety to the public by the tour. (Nakai, Y.)

  20. 多领域功能样机可交换模型规范实现研究%Research on Functional Mock-up Interface for Multi-domain Model Exchange

    Institute of Scientific and Technical Information of China (English)

    吴紫俊; 赵建军

    2012-01-01

    针对不同领域仿真工具模型的重用互换需求,欧洲仿真界提出了功能样机可交换接口规范Functional Mock-up Interfacefor Model Exchange(FMI)。基于FMI接口规范,详细分析了功能样机可交换模型的结构与功能,研究了规范对仿真模型变量、方程和仿真算法的描述,并在多领域物理系统建模仿真平台MWorks中设计7FMl接口,实现7MWorks模型的重用,扩展7MWorks模型的使用范围。目前,MWorks是国内唯一支持该规范的建模仿真工具。%According to the needs of model reuse and exchange between different modeling tools, the standard of Functional Mock-up Interface for model exchange was proposed. Based on the FMI standard, the structure and the function of Functional Mock-up Unit were studied, and the description of variables, equations and algorithms of model were investigated. The FMI interface was designed for the multi-domain modeling and simulation platform MWorks. The reuse of the model which was built in MWorks was achieved. The application of MWorks models was expanded. Recently, MWorks is the only simulation tool which supports the FMI standard.

  1. Nuclear modular power stations with lead-based coolant reactors

    International Nuclear Information System (INIS)

    In the present report, the projects of reactors with the lead-based coolant are considered. This class of reactors has the advantages and limitations. The main of advantages is enhanced safety and the main of restrictions is the limitation on power, both originate from natural properties of lead-based coolants. This limitation uniquely determines lead-technology reactors as medium - and small-power systems. (author)

  2. Comparison of Pickering NGS performance with world power reactors, 1977

    International Nuclear Information System (INIS)

    Pickering NGS performance is compared, in highly graphic form, with the perfomance of other nuclear power plants around the world. The four Pickering reactors score in the top six, rated by gross capacity factor. Major system suppliers for world power reactors above 500 MW are cataloged. (E.C.B.)

  3. The calculation of the YALINA BOOSTER zero power sub critical assembly driven by external neutron sources: Brazillian contribution

    Science.gov (United States)

    Carluccio, Thiago; Rossi, Pedro Carlos Russo; Maiorino, José Rubens

    2011-08-01

    The YALINA-Booster is an experimental zero power Accelerator Driven Reactor (ADS), which consists of a sub-critical assemby driven by external neutron sources. It has a fast spectrum booster zone in the center, surrounded by a thermal one. The sub-critical core is driven by external neutron sources. Several experiments have been proposed in the framework of IAEA Coordinated Reserch Project (CRP) on ADS. This work shows results obtained by IPEN modelling and simulating experiments proposed at CRP, using the MCNP code. The comparison among our results, the experimental one and the results obtained by other participants is being done by CRP coordinators. This coolaborative work has an important role in the qualification and improvement of calculational methodologies.

  4. Static and dynamic performance tests of nuclear powered ship Mutsu reactor (report on nuclear ship Mutsu power-up tests)

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, Toshihisa; Kusunoki, Tsuyoshi; Ochiai, Masa-aki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Tanaka, Yoshimi; Inoue, Kimio; Yao, Toshiaki; Kamai, Satoshi; Kitamura, Toshikatsu

    1992-08-01

    The power-up tests of the Mutsu reactor were performed from March 29th 1990 to December 14th. The tests were divided into six phases: The tests Phase 0 and Phase 1 were done in the state that the ship was moored at the quay of Sekinehama port in March and April; The tests Phase 2, Phase 3, Phase 4, and Phase 5 were done on the Pacific Ocean from July to December. Present report describes the test results on the static and dynamic plant performance. On static plant performance tests, there are 13 test items including measurements of primary system heat balance at low and high power levels, a virgin run of feed water pump with SG steam, a change-over test of steam supply of auxiliary boiler to SG. On the dynamic plant performance, there are 11 test items including a test of reactor power auto-control system, a test of main feed water auto-control system, a test of small load variation, a load increasing test, a turbine trip test, tests of ahead and astern maneuvering, a test of single loop operation, and a reactor scram test. The reactor power for each item`s test was increased step by step from zero power to the goal of rated power of 100 %, 36 MWt. In order to confirm proper reactor system performance, criteria were laid down for the static and dynamic tests: for example, (1) reactor scram shall not occur, (2) pressurizer relief valve and steam generator safety valve shall not work, and (3) after the transients reactor systems shall become the steady state without manual adjustment of the reactor control system. The test results satisfied these criteria and some of test data showed that reactor had much more margin in any performance for design. It is verified, therefore, that the Mutsu reactor systems have adequate performances as a marine reactor and that one is capable to respond smoothly and safely to the load of ship`s demand. (author).

  5. Static and dynamic performance tests of nuclear powered ship Mutsu reactor (report on nuclear ship Mutsu power-up tests)

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, Toshihisa; Kusunoki, Tsuyoshi; Ochiai, Masa-aki (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment); Tanaka, Yoshimi; Inoue, Kimio; Yao, Toshiaki; Kamai, Satoshi; Kitamura, Toshikatsu.

    1992-08-01

    The power-up tests of the Mutsu reactor were performed from March 29th 1990 to December 14th. The tests were divided into six phases: The tests Phase 0 and Phase 1 were done in the state that the ship was moored at the quay of Sekinehama port in March and April; The tests Phase 2, Phase 3, Phase 4, and Phase 5 were done on the Pacific Ocean from July to December. Present report describes the test results on the static and dynamic plant performance. On static plant performance tests, there are 13 test items including measurements of primary system heat balance at low and high power levels, a virgin run of feed water pump with SG steam, a change-over test of steam supply of auxiliary boiler to SG. On the dynamic plant performance, there are 11 test items including a test of reactor power auto-control system, a test of main feed water auto-control system, a test of small load variation, a load increasing test, a turbine trip test, tests of ahead and astern maneuvering, a test of single loop operation, and a reactor scram test. The reactor power for each item's test was increased step by step from zero power to the goal of rated power of 100 %, 36 MWt. In order to confirm proper reactor system performance, criteria were laid down for the static and dynamic tests: for example, (1) reactor scram shall not occur, (2) pressurizer relief valve and steam generator safety valve shall not work, and (3) after the transients reactor systems shall become the steady state without manual adjustment of the reactor control system. The test results satisfied these criteria and some of test data showed that reactor had much more margin in any performance for design. It is verified, therefore, that the Mutsu reactor systems have adequate performances as a marine reactor and that one is capable to respond smoothly and safely to the load of ship's demand. (author).

  6. Axial power monitoring uncertainty in the Savannah River Reactors

    International Nuclear Information System (INIS)

    The results of this analysis quantified the uncertainty associated with monitoring the Axial Power Shape (APS) in the Savannah River Reactors. Thermocouples at each assembly flow exit map the radial power distribution and are the primary means of monitoring power in these reactors. The remaining uncertainty in power monitoring is associated with the relative axial power distribution. The APS is monitored by seven sensors that respond to power on each of nine vertical Axial Power Monitor (APM) rods. Computation of the APS uncertainty, for the reactor power limits analysis, started with a large database of APM rod measurements spanning several years of reactor operation. A computer algorithm was used to randomly select a sample of APSs which were input to a code. This code modeled the thermal-hydraulic performance of a single fuel assembly during a design basis Loss-of Coolant Accident. The assembly power limit at Onset of Significant Voiding was computed for each APS. The output was a distribution of expected assembly power limits that was adjusted to account for the biases caused by instrumentation error and by measuring 7 points rather than a continuous APS. Statistical analysis of the final assembly power limit distribution showed that reducing reactor power by approximately 3% was sufficient to account for APS variation. This data confirmed expectations that the assembly exit thermocouples provide all information needed for monitoring core power. The computational analysis results also quantified the contribution to power limits of the various uncertainties such as instrumentation error

  7. Steady performance of a zero valent iron packed anaerobic reactor for azo dye wastewater treatment under variable influent quality

    Institute of Scientific and Technical Information of China (English)

    Yaobin Zhang; Yiwen Liu; Yanwen Jing; Zhiqiang Zhao; Xie Quan

    2012-01-01

    Zero valent iron (ZVI) is expected to help create an enhanced anaerobic environment that might improve the performance of anaerobic treatment.Based on this idea,a novel ZVI packed upflow anaerobic sludge blanket (ZVI-UASB) reactor was developed to treat azo dye wastewater with variable influent quality.The results showed that the reactor was less influenced by increases of Reactive Brilliant Red X-3B concentration from 50 to 1000 mg/L and chemical oxygen demand (COD) from 1000 to 7000 mg/L in the feed than a reference UASB reactor without the ZVI.The ZVI decreased oxidation-reduction potential in the reactor by about 80 mV.Iron ion dissolution from the ZVI could buffer acidity in the reactor,the amount of which was related to the COD concentration.Fluorescence in situ hybridization test showed the abundance of methanogens in the sludge of the ZVI-UASB reactor was significantly greater than that of the reference one.Denaturing gradient gel electrophoresis showed that the ZVI increased the diversity of microbial strains responsible for high efficiency.

  8. Quality surveillance for PWR power plant reactor internals manufacturing

    International Nuclear Information System (INIS)

    The structure and the function of the reactor internals of the improved generation Ⅱ PWR power plant is instructed briefly, the critical factors and difficulties in the manufacture process for reactor internals are analyzed, the quality control and surveillance of reactor internals manufacturing is discussed, especially the critical factors and difficulties of the quality control in the manufacture process for the main parts of reactor internals and in the reactor internals assembling process are represented in detail, the key points of the resident manufacture supervision is presented, and other key points of quality control in the manufacture process are also given, such as the documents control and personnel control. (author)

  9. Regulation concerning installation and operation of reactors for power generation

    International Nuclear Information System (INIS)

    The regulations applying to reactors for power generation mentioned in the law for the regulations of nuclear source materials, nuclear fuel materials and reactors. Covered are the following: definitions of terms, application for the permission to install reactors, application for the permission to alter installations, reactor operation plans, keeping of various records, limitations on ingress and ingress in radiation controlled areas, measures concerning radiation exposure doses, operation of reactors, on-site transport, storage of nuclear fuel materials and radioactive waste, security regulations, steps taken during times of danger, making of various reports, and so on. (Mori, K.)

  10. Nuclear Power Reactors in the World. 2012 Ed

    International Nuclear Information System (INIS)

    This is the 32nd edition of Reference Data Series No. 2, which presents the most recent reactor data available to the IAEA. It contains summarized information as of the end of 2011 on power reactors that are in operation, under construction and shut down, and performance data on reactors operating in IAEA Member States, as reported to the IAEA. The information is collected through designated national correspondents in the Member States and the data area used to maintain the IAEA's Power Reactor Information System.

  11. Power start up of the Dalat nuclear research reactor

    International Nuclear Information System (INIS)

    After accomplishing the physical start-up of the reactor, the power start-up was carried out in February 1984. The power of the reactor has reached: 10 KW on 6/2/1984, 100 KW on 7/2/1984, 200 KW and 300 KW on 8/2/1984; 400 KW and nominal power 500 KW on 9/2/1984. The reactivity temperature coefficient and the xenon poisoning were determined. 3 figs., 12 tabs

  12. Burnup analysis of the power reactor, 1

    International Nuclear Information System (INIS)

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

  13. Performance analysis of a near zero CO2 emission solar hybrid power generation system

    International Nuclear Information System (INIS)

    Highlights: • A novel solar hybrid power system with near zero CO2 emission has been proposed. • The system integrates fuel reforming, solar-driven steam generation and CO2 capture. • Solar heat upgrading and high-efficiency heat-to-power conversion are achieved. • The system accomplishes near zero CO2 emission with oxy-fuel combustion. • The system thermodynamic performances have been investigated and compared. - Abstract: A novel solar hybrid power generation system with near zero CO2 emission (ZE-SOLRGT) has been proposed in the previous work, which is based on a GRAZ-like cycle integrating methane–steam reforming, solar-driven steam generation and CO2 capture. Solar heat assistance increases power output and reduces fossil fuel consumption. Besides near zero CO2 emission with oxy-fuel combustion and cascade recuperation of turbine exhaust heat, the system is featured with indirect upgrading of low-mid temperature solar heat and its high efficiency heat-to-power conversion. A performance analysis of ZE-SOLRGT cycle has been carried out using ASPEN PLUS code to explore the effects of key parameters on system performances. It is concluded that ∼54% exergy efficiency can be attained with ∼100% CO2 capture. The net solar-to-electricity efficiency can reach up to 34.7% in the base case. Steam-to-methane molar ratio of 2–3 is suitable for system performance improvement. High system efficiency can be obtained as the HPT pressure ratio is in the range of 15–18. The system integration achieves the complementary utilization of fossil fuel and solar heat, as well as their high-efficiency conversion into electricity

  14. Power plant systems for fusion reactors

    International Nuclear Information System (INIS)

    To investigate and compare the applicability of thermal cycles for power generation to nuclear fusion, four plant systems, i.e. direct steam turbine, in-direct steam turbine, direct gas turbine and in-direct steam turbine with gas-cooled blanket, have been designed with the estimates of their thermal efficiencies. An plant designs here are based on the same power core: fusion power 2300 MW, external heating power 58 MW, thermal power of blanket 2420 MW and divertor 490 MW. In addition, it is assumed that the construction of the power plant is near future so that the structural material would be a ferritic/martensitic steel such as F82H to be used at temperatures lower than 500-550degC. Also the divertor is always cooled by water at pressure of 10 MPa, and inlet/outlet temperature of 150-200degC/200-250degC. The removal heat from the divertor is utilized to heat the coolant fed to the blanket inlet in all above plants. The direct steam turbine cycle employs supercritical pressure water at 25 MPa and blanket inlet/outlet temperatures of 280degC/500degC. The steam out from the blanket directly flows into a high pressure turbine. The steam intermediately extracted from the high pressure turbine and/or a part of main steam from the blanket outlet is utilized to reheat the steam coming out of the high pressure turbine. Also the regenerative cycle is applied by using steams extracted from high, medium and low pressure turbines. Eventually the obtained thermal efficiency is 41.4%. The in-direct steam turbine cycle consists of the primary loop which removes the heat from the blanket and the secondary (power generation) loop by using a steam generator at their interface. The primary coolant is supercritical pressure water similar to that of above direct steam cycle, i.e. 25 MPa, 290degC/510degC. The secondary coolant is also water but with the condition of a fast breeder fission reactor, i.e. 16.3 MPa, 210degC/480degC. With reheat and regenerative cycle, the thermal

  15. Consumption of the electric power inside silent discharge reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yehia, Ashraf, E-mail: yehia30161@yahoo.com [Department of Physics, Faculty of Science, Assiut University, Assiut 71516, Arab Republic of Egypt and Department of Physics, College of Science and Humanitarian Studies at Alkharj, Salman bin Abdulaziz University, P.O. Box 83, Alkharj 11942 (Saudi Arabia)

    2015-01-15

    An experimental study was made in this paper to investigate the relation between the places of the dielectric barriers, which cover the surfaces of the electrodes in the coaxial cylindrical reactors, and the rate of change of the electric power that is consumed in forming silent discharges. Therefore, silent discharges have been formed inside three coaxial cylindrical reactors. The dielectric barriers in these reactors were pasted on both the internal surface of the outer electrode in the first reactor and the external surface of the inner electrode in the second reactor as well as the surfaces of the two electrodes in the third reactor. The reactor under study has been fed by atmospheric air that flowed inside it with a constant rate at normal temperature and pressure, in parallel with the application of a sinusoidal ac voltage between the electrodes of the reactor. The electric power consumed in forming the silent discharges inside the three reactors was measured as a function of the ac peak voltage. The validity of the experimental results was investigated by applying Manley's equation on the same discharge conditions. The results have shown that the rate of consumption of the electric power relative to the ac peak voltage per unit width of the discharge gap improves by a ratio of either 26.8% or 80% or 128% depending on the places of the dielectric barriers that cover the surfaces of the electrodes inside the three reactors.

  16. Improvements in the management of safety in research reactor operation through appropriate application of selected power reactor good practices

    International Nuclear Information System (INIS)

    Research reactor managers are increasingly implementing improvements in their management of safety through the application of good practices originally developed as power reactor programs. This paper considers ways to select practices to emulate, effectively incorporate them into a research reactor program and evaluate their contribution to safety. Relative to research reactors, power reactor programs look relatively homogeneous when considering source terms, stored energy, core power density, operating cycles, plant systems and staff sizes. They have potential hazard consequences that require effective safety management programs. Finally, power reactors generate a stream of revenue to fund these programs. The power reactor community has combined their resources with the homogeneity of their challenge to create impressive safety management tools, many of which can be effectively implemented in the research reactor community. However, not all programs can be effectively implemented in all research reactors. number of power reactor programs are analyzed in the paper with consideration of their effective implementation and potential contribution to research reactor. (author)

  17. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    The technology for cobalt-60 production in CANDU reactors, designed and developed by MDS Nordion and AECL, has been safely,economically and successfully employed in CANDU reactors with over 195 reactor years of production. Today over forty percent of the world's disposable medical supplies are made safer through sterilization using cobalt-60 sources from MDS Nordion. Over the past 40 years, MDS Nordion with its CANDU reactor owner partners, has safely and reliably shipped more than 500 million curies of cobalt-60 sources to customers around the world

  18. Characterization of the TRIGA Mark II reactor full-power steady state

    CERN Document Server

    Cammi, Antonio; Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica; Magrotti, Giovanni; Prata, Michele; Salvini, Andrea

    2015-01-01

    In this work, the characterization of the full-power steady state of the TRIGA Mark II nuclear reactor of the University of Pavia is performed by coupling Monte Carlo (MC) simulation for neutronics with "Multiphysics" model for thermal-hydraulics. Neutronic analyses have been performed starting from a MC model of the entire reactor system, based on the MCNP5 code, that was already validated in fresh fuel and zero-power configuration (in which thermal effects are negligible) using the available experimental data as benchmark. In order to describe the full-power reactor configuration, the temperature distribution in the core is necessary. To evaluate it, a thermal-hydraulic model has been developed, using the power distribution results from MC simulation as input. The thermal-hydraulic model is focused on the core active region and takes into account sub-cooled boiling effects present at full reactor power. The obtained temperature distribution is then introduced in the MC model and a benchmark analysis is carr...

  19. Experimental study of the IPR-R1 TRIGA reactor power channels responses

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Henrique F.A.; Ferreira, Andrea V., E-mail: hfam@cdtn.br, E-mail: avf@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The IPR-R1 nuclear reactor installed at Centro de Desenvolvimento da Tecnologia Nuclear CDTN/CNEN, Belo Horizonte, Brazil, is a Mark I TRIGA reactor (Training, Research, Isotopes, General Atomics) and became operational on November of 1960. The reactor has four irradiation devices: a rotary specimen rack with 40 irradiation channels, the central tube, and two pneumatic transfer tubes. The nuclear reactor is operated in a power range between zero and 100 kW. The instrumentation for IPR-R1 operation is mainly composed of four neutronic channels for power measurements. The aim of this work is to investigate the responses of neutronic channels of IPR-R1, Linear, Log N and Percent Power channels, and to check their linearity. Gold foils were activated at low powers (0.125-1.000 kW), and cobalt foils were activated at high powers (10-100kW). For each sample irradiated at rotary specimen rack, another one was irradiated at the same time at the pneumatic transfer tube-2. The obtained results allowed evaluating the linearity of the neutronic channels responses. (author)

  20. 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.

  1. Influence of radiation on maintenance in a nuclear power station

    International Nuclear Information System (INIS)

    Maintenance in nuclear power plant differs from that in fossil fuel power plant in many aspects because the maintenance in the former has to be carried out in radiation area. These aspects are : (1) manpower planning to minimise time of repair in order to reduce the radiation dose received by the maintenance crew, (2) difficulties in isolating components to be repaired from reactor which is normally filled with water, (3) shielding and decontamination to reduce radiation fields around equipment and (4) need to write the detailed procedures and use special tools, brief and train personnel before-hand on similar equipment or mock-ups. These aspects are discussed. Two of the major repair jobs carried out at the Tarapur Atomic Power Station are described in brief. The jobs were : (1) tube plugging of secondary steam generators and (2) repair to the guide brackets of dryer -separator assembly in the reactor vessel. (M.G.B.)

  2. KLT-20 reactor for a floating power unit. Annex VI

    International Nuclear Information System (INIS)

    The KLT-20 reactor installation is being designed by the Experimental Design Bureau of Machine Building (OKBM, Nizhny Novgorod) as a power source for floating nuclear power plants (NPPs). At present, the activities are most advanced for the project of a pilot floating heat and power plant with the KLT-40S reactor installations, advanced analogues of the commercial KLT-40 reactors of the Russian icebreaker fleet. For the KLT-40S, detailed design of the reactor unit and floating power unit has been developed and approved; the Rostechnadzor of Russia license for plant siting and floating power unit construction in Severodvinsk (Russian Federation) has been obtained. The KLT-20, based on a pressurized light-water reactor of 20 MW(e), is a two-loop modification of the KLT-40S reactor with several improvements in the main equipment and a long-refuelling interval, achieved with the enrichment of less than 20%. The reactor design with a long refuelling interval was developed based on the engineering solutions of the pilot KLT-40S reactor installation; different from it, the KLT-20 provides for no on-site refuelling. The refuelling, radioactive waste management and repairs of a floating NPP with the KLT-20 would be performed at special maintenance centres. The infrastructure of nuclear ship maintenance centres in Russia could be used for these purposes

  3. Role of advanced reactors in further nuclear power development

    International Nuclear Information System (INIS)

    As a part of the national long-term nuclear R and D program launched in 1992, an endeavor has been made in Korea to develop advanced nuclear reactor systems with significantly enhanced safety and economics from those of the current generation nuclear power plants. The advanced PWR nuclear reactor systems under development in Korea include 1300 MWe Korean Next Generation Reactor (KNGR), 330 MWt Integral Type System Integrated Modular Advanced Reactor (SMART) for nuclear cogeneration, and 330 MWe Korea Advanced Liquid Metal Reactor (KALIMER) in addition to the evolutionary enhancement of the 1000 MWt KSNPP (Korea Standard Nuclear Power Plant). Three point design philosophy has been adopted for the development of the advanced reactors in Korea : enhancements on safety, economics and public acceptance of nuclear power. To enhance the safety of the advanced reactor systems, a strategy has been adopted to employ advanced design features as well as the passive safety design features. Economically viable design concepts also have been implemented in the evolutionary KSNPP, KNGR, and the SMART development. Economic competitiveness against the fossil plants also has been set as a major objective of the ALWR development program in Korea. These safer and more economical advanced reactors will better promote the public acceptance of the commercial use of the nuclear power and thus could be utilized to meet the forecasted national energy need in the early 21st century. International cooperation in the areas of ALWR development as well as improving public acceptance of the nuclear power is required. (author)

  4. Data management for spent fuel from power reactors in Argentina

    International Nuclear Information System (INIS)

    Updated data for the spent fuel management from the operating power reactors - as of December 31st, 2004 - as well as research reactors - as of March 1st, 2003 - in Argentina are presented. Data for the power reactor spent fuel are received from the nuclear power plant operator (Nucleoelectrica Argentina S.A.) twice a year for the cumulative spent fuel arising up to June 30th and December 31st. Data for the research reactor spent fuel are collected once a year from CNEA operators. At the time being, such data are not managed in a database management system but some of them are handled with a spreadsheet program in order to get total, average, lower and higher values. These values are being used to built the input of codes for calculating the composition, activity and thermal power for the spent fuel as a whole as well as the mass, activity and thermal power for spent fuel elements or nuclides. (author)

  5. Tokamak power systems studies: A second stability power reactor

    International Nuclear Information System (INIS)

    A number of innovative physics and engineering features have been studied which promise to greatly improve the reactor prospects of tokamaks relative to STARFIRE. A reference design point has been developed with the following features: large aspect ratio (A = 6); high beta (β ≅ 0.20), with only mild shaping and no indentation, which brings the maximum toroidal field down to 7 T; low toroidal current (I ≅ 5MA), which reduces the cost of the current drive and EF coil system; and steady-state operation with combined fast wave and lower hybrid wave current drive. The key to high beta operation with low toroidal current lies in utilizing second stability regime equilibria with control of the current achieved by the appropriate choice of wave frequencies and spectra. By selecting an axial safety factor q(o) = 2.0, MHD stability has been found above β ≅ 0.20. Additional features include: impurity control with self-pumped limiters which bury helium on continuously deposited metal surfaces; liquid Li-cooled blanket which provides good performance with low pressure operation; vanadium alloy blanket structure for higher thermal efficiency (eta = 0.42), longer lifetime and reduced activation; and reduced reactor mass (higher power density) due to smaller TF coil, less shielding, fewer blanket penetrations, and higher wall loading. At low neutron wall loads this device represents a minimum capital cost unit. However, economies of scale are strong, and eventually higher wall loads (W ≅ 8 MW/m2, P/sub net/ = 1400 MW) may prove most attractive. Preliminary investigations show inherently safe operation is likely at W ≥ 5 MW/m2. 15 refs., 3 figs., 1 tab

  6. Radiation embrittlement in pressure vessels of power reactors

    International Nuclear Information System (INIS)

    It is presented the project to study the effect of lead factors on the mechanical behavior of Reactor Pressure Vessel steels. It is described the facility designed to irradiate Charpy specimens with V notch of SA-508 type 3 steel at power reactor temperature, installed in the RA-1 reactor. The objective is to obtain the fracture behavior of irradiated specimens with different lead factors and to know their dependence with the diffusion of alloy elements. (author)

  7. The next generation of power reactors - safety characteristics

    International Nuclear Information System (INIS)

    The next generation of commercial nuclear power reactors is characterized by a new approach to achieving reliability of their safety systems. In contrast to current generation reactors, these designs apply passive safety features that rely on gravity-driven transfer processes or stored energy, such as gas-pressurized accumulators or electric batteries. This paper discusses the passive safety system of the AP600 and Simplified Boiling Water Reactor (SBWR) designs

  8. EBT reactor characteristics consistent with stability and power balance requirements

    International Nuclear Information System (INIS)

    This paper summarizes the results of a recent EBT reactor study that includes both ring and core plasma properties and 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. A number of concept improvements are also proposed that are found to offer the potential for further improvement of the reactor size and parameters

  9. Nuclear power reactors in the world. April 2003 ed

    International Nuclear Information System (INIS)

    This is the twenty third edition of Reference Data Series No. 2, Nuclear Power Reactors in the World, which is published once per year, to present the most recent reactor data available to the Agency. It contains the following summarized information: General information as of the end of 2002 on power reactors operating or under construction, and shut down; Performance data on reactors operating in the Agency's Member States, as reported to the IAEA. The information is collected by the Agency by circulating questionnaires to Member States through the designated national correspondents. The replies are used to maintain computerized files on general and design data of, and operating experience with, power reactors. The Agency's Power Reactor Information System (PRIS) comprising the above files provides all the information and data previously published in the Agency's Power Reactors in Member States and currently published in the Agency's Operating Experience with Nuclear Power Stations in Member States and available at the Internet address http://www.iaea.or.at/programmes/a2

  10. Nuclear power reactors in the world. April 2001 ed

    International Nuclear Information System (INIS)

    This is the twenty-first edition of Reference Data Series No.2, Nuclear Power Reactors in the World, which is published once per year, to present the most recent reactor data available to the Agency. It contains the following summarized information: General information as of the end of 2000 on power reactors operating or under construction, and shut down; Performance data on reactors operating in the Agency's Member States, as reported to the IAEA. The information is collected by the Agency by circulating questionnaires to Member States through the designated national correspondents. The replies are used to maintain computerized files on general and design data of, and operating experience with, power reactors. The Agency's Power Reactor Information System (PRIS) comprising the above files provides all the information and data previously published in the Agency's Power Reactors in Member States and currently published in the Agency's Operating Experience with Nuclear Power Stations in Member States and is available at the Internet address http://www.iaea.or.at/programmes/a2

  11. Nuclear power reactors in the world. April 2002 ed

    International Nuclear Information System (INIS)

    This is the twenty-second edition of Reference Data Series No. 2, Nuclear Power Reactors in the World, which is published once per year, to present the most recent reactor data available to the Agency. It contains the following summarized information: General information as of the end of 2001 on power reactors operating or under construction, and shut down; Performance data on reactors operating in the Agency's Member States, as reported to the IAEA. The information is collected by the Agency by circulating questionnaires to Member States through the designated national correspondents. The replies are used to maintain computerized files on general and design data of, and operating experience with, power reactors. The Agency's Power Reactor Information System (PRIS) comprising the above files provides all the information and data previously published in the Agency's Power Reactors in Member States and currently published in the Agency's Operating Experience with Nuclear Power Stations in Member States and available at the Internet address http://www. iaea.or.at/programmes/a2

  12. Future zero-carbon energy systems in Japan with different nuclear power development scenarios

    International Nuclear Information System (INIS)

    An integrated scenario analysis has been conducted toward zero-carbon energy system from 2010 to 2100 in Japan, wherein the effect of Fukushima nuclear accident happened in March, 2011 is more or less taken into account. In the study, various service demands are firstly estimated based on social-economic data and then best technology and energy mixes are obtained using the optimization model to meet the service demand. On the conductance of integrated scenario analysis towards the year 2100 when zero-carbon energy system will be attained, three different scenarios of nuclear power development are taken, i.e., (1) no further introduction of nuclear, (2) fixed portion and (3) no limit of nuclear. The results show that, in the end user side, zero-carbon energy scenario can be attained at 2100 with electricity supplies 75% of total energy utilization. And for the electricity supply, three different power generation scenarios are proposed: (Scenario 1) 30% renewable and 70% gas-CCS (Carbon Capture and Storage), (Scenario 2) every one third by nuclear, by renewable and by gas-CCS, and (Scenario 3) 60% nuclear power, 20% renewable and 10% gas-CCS. Lastly by the inter-comparison of the three scenarios from the four aspects of cost, CO2 emission, risk and diversity, Scenario 2 is rated as the most balanced scenario among the three by putting emphasis on the availability of diversified electric source of nuclear, renewable and gas-CCS. (author)

  13. Power coefficient of reactivity in CANDU 6 Reactors

    International Nuclear Information System (INIS)

    The Power Coefficient of Reactivity (PCR) measures the change in reactor core reactivity per unit change in reactor power and is an integral quantity which captures the contributions of the fuel temperature, coolant void and coolant temperature reactivity feedbacks. All nuclear reactor designs provide a balance between the inherent nuclear characteristics and the engineered reactivity control features, to ensure that changes in reactivity in all operating conditions are maintained within a safe range. The CANDU reactor design takes advantage of the inherent nuclear characteristics of small reactivity coefficient, minimal excess reactivity and very long prompt neutron lifetime to mitigate the magnitude of the demand on the engineered systems for controlling reactivity. In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with its design characteristics, such that the overall design of the reactor does not depend on the sign of the PCR. This is a contrast to other reactor design concepts which are dependent on a PCR which is both large and negative in the design of their engineered systems for controlling reactivity. It will be demonstrated that during a Loss of Regulation Control (LORC) event, the impact of having a positive power coefficient, or of hypothesizing a PCR larger than that estimated for CANDU, has no significant impact on the reactor safety. Since the CANDU 6 PCR is small, its role in the operation or safety of the reactor is not significant

  14. Nuclear Power Reactors in the World. 2016 Ed

    International Nuclear Information System (INIS)

    Nuclear Power Reactors in the World is an annual publication that presents the most recent data pertaining to reactor units in IAEA Member States. This thirty-sixth edition of Reference Data Series No. 2 provides a detailed comparison of various statistics up to and including 31 December 2015. The tables and figures contain the following information: — General statistics on nuclear reactors in IAEA Member States; — Technical data on specific reactors that are either planned, under construction or operational, or that have been shut down or decommissioned; — Performance data on reactors operating in IAEA Member States, as reported to the IAEA. The data compiled in this publication is a product of the IAEA’s Power Reactor Information System (PRIS). The PRIS database is a comprehensive source of data on all nuclear power reactors in the world. It includes specification and performance history data on operational reactors as well as on reactors under construction or in the decommissioning process. Data is collected by the IAEA via designated national correspondents in Member States

  15. Design for reactor core safety in nuclear power plants

    International Nuclear Information System (INIS)

    This Guide covers the neutronic, thermal, hydraulic, mechanical, chemical and irradiation considerations important to the safe design of a nuclear reactor core. The Guide applies to the types of thermal neutron reactor power plants that are now in common use and fuelled with oxide fuels: advanced gas cooled reactor (AGR), boiling water reactor (BWR), pressurized heavy water reactor (PHWR) (pressure tube and pressure vessel type) and pressurized water reactor (PWR). It deals with the individual components and systems that make up the core and associated equipment and with design provisions for the safe operation of the core and safe handling of the fuel and other core components. The Guide discusses the reactor vessel internals and the reactivity control and shutdown devices mounted on the vessel. Possible effects on requirements for the reactor coolant, the reactor coolant system and its pressure boundary (including the pressure vessel) are considered only as far as necessary to clarify the interface with the Safety Guide on Reactor Coolant and Associated Systems in Nuclear Power Plants (IAEA Safety Series No. 50-SG-D13) and other Guides. In relation to instrumentation and control systems the guidance is mainly limited to functional requirements

  16. Small size modular fast reactors in large scale nuclear power

    International Nuclear Information System (INIS)

    The report presents an innovative nuclear power technology (NPT) based on usage of modular type fast reactors (FR) (SVBR-75/100) with heavy liquid metal coolant (HLMC) i. e. eutectic lead-bismuth alloy mastered for Russian nuclear submarines' (NS) reactors. Use of this NPT makes it possible to eliminate a conflict between safety and economic requirements peculiar to the traditional reactors. Physical features of FRs, an integral design of the reactor and its small power (100 MWe), as well as natural properties of lead-bismuth coolant assured realization of the inherent safety properties. This made it possible to eliminate a lot of safety systems necessary for the reactor installations (RI) of operating NPPs and to design the modular NPP which technical and economical parameters are competitive not only with those of the NPP based on light water reactors (LWR) but with those of the steam-gas electric power plant. Multipurpose usage of transportable reactor modules SVBR-75/100 of entirely factory manufacture assures their production in large quantities that reduces their fabrication costs. The proposed NPT provides economically expedient change over to the closed nuclear fuel cycle (NFC). When the uranium-plutonium fuel is used, the breeding ratio is over one. Use of proposed NPT makes it possible to considerably increase the investment attractiveness of nuclear power (NP) with fast neutron reactors even today at low costs of natural uranium. (authors)

  17. Ultrasonic level and temperature sensor for power reactor applications

    International Nuclear Information System (INIS)

    An ultrasonic waveguide employing torsional and extensional acoustic waves has been developed for use as a level and temperature sensor in pressurized and boiling water nuclear power reactors. Features of the device include continuous measurement of level, density, and temperature producing a real-time profile of these parameters along a chosen path through the reactor vessel

  18. MEASUREMENT ERROR EFFECT ON THE POWER OF CONTROL CHART FOR ZERO-TRUNCATED POISSON DISTRIBUTION

    Directory of Open Access Journals (Sweden)

    Ashit Chakraborty

    2013-09-01

    Full Text Available Measurement error is the difference between the true value and the measured value of a quantity that exists in practice and may considerably affect the performance of control charts in some cases. Measurement error variability has uncertainty which can be from several sources. In this paper, we have studied the effect of these sources of variability on the power characteristics of control chart and obtained the values of average run length (ARL for zero-truncated Poisson distribution (ZTPD. Expression of the power of control chart for variable sample size under standardized normal variate for ZTPD is also derived.

  19. Gas core reactor power plants designed for low proliferation potential

    International Nuclear Information System (INIS)

    The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF6 and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on 233U born from thorium. Fission product removal was continuous. Newly born 233U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of 233U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors

  20. Noise Reduction of an Ultrasonic System for the IRR2 Research Reactor Tank

    International Nuclear Information System (INIS)

    As part of the IRR2 reactor preventive maintenance, an ultrasonic scanning system was developed for inspection of the reactor tank wall. The design validation was carried out in a full-sized mock-up. An important issue tested in the mock-up experiments was the ultrasonic scanning system sensitivity to electromagnetic noise. Indeed, the system was approved only after the noise levels were found to be tolerable. However, when the Ultrasonic Inspection System was first operated in the IRR2 reactor, unacceptable noise levels were measured, i. e., system immunity to electromagnetic noise demonstrated in the mock-up testing proved to be irrelevant for the IRR2 reactor itself Various methods of noise reduction were attempted; consequently, noise interference was reduced to an acceptable level

  1. Radionuclides in United States commercial nuclear power reactors

    International Nuclear Information System (INIS)

    In the next ten to twenty years, many of the commercial nuclear power reactors in the United States will be reaching their projected lifetime of forty years. As these power plants are decommissioned, it seems prudent to consider the recycling of structural materials such as stainless steel. Some of these materials and components have become radioactive through either nuclear activation of the elements within the components or surface contamination with radioactivity form the operational activities. In order to understand the problems associated with recycling stainless steel from decommissioned nuclear power reactors, it is necessary to have information on the radionuclides expected on or in the contaminated materials. A study has been conducted of radionuclide contamination information that is available for commercial nuclear power reactors in the United States. There are two types of nuclear power reactors in commercial use in the United States, pressurized water reactors (PWRs) and boiling water reactors (BWRs). Before presenting radionuclide activities information, a brief discussion is given on the major components and operational differences for the PWRs and BWRs. Radionuclide contamination information is presented from 11 PWRs and over 8 BWRs. These data include both the radionuclides within the circulating reactor coolant water as well as radionuclide contamination on and within component parts

  2. Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

    International Nuclear Information System (INIS)

    In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO2 fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO2 LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

  3. Measurement of the power and temperature reactivity coefficients of the RTP TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rabir, Mohamad Hairie, E-mail: m_hairie@nuclearmalaysia.gov.my

    2013-12-15

    This paper presents the experimental results of the power and temperature coefficients of reactivity of the RTP TRIGA reactor at the Malaysian Nuclear Agency. The power coefficient of reactivity obtained was approximately −0.26 ¢ kW{sup −1} (−1.81 × 10{sup −5} kW{sup –1}), and the measured temperature reactivity coefficient of the reactor was −0.82 ¢ °C{sup −1} (−5.77 × 10{sup −5} °C{sup −1}) and −1.15 ¢ °C{sup −1} (−8.08 × 10{sup −5} °C{sup −1}) in IFE C12 and IFE F16, respectively. The power defect, which is the change in reactivity taking place between zero power and the power of 850 kW was ∼2.19 $. Because of the negative temperature coefficient, a significant amount of reactivity is needed to compensate for the temperature change and allows the reactor to operate at the higher power levels in steady state. Throughout this experiment, it is the temperature of the fuel that was measured, not the isothermal temperature coefficient (ITC), which comprises both moderator and fuel.

  4. Cermet-fueled reactors for multimegawatt space power applications

    International Nuclear Information System (INIS)

    The cermet-fueled reactor has evolved as a potential power source for a broad range of multimegawatt space applications. In particular, the fast spectrum reactor concept can be used to deliver 10s of megawatts of electric power for continuous, long term, unattended operation, and 100s of megawatts of electric power for times exceeding several hundred seconds. The system can also be utilized with either a gas coolant in a Brayton power conversion cycle, or a liquid metal coolant in a Rankine power conversion cycle. Extensive testing of the cermet fuel element has demonstrated that the fuel is capable of operating at very high temperatures under repeated thermal cycling conditions, including transient conditions which approach the multimegawatt burst power requirements. The cermet fuel test performance is reviewed and an advanced cermet-fueled multimegawatt nuclear reactor is described in this paper

  5. Natural uranium fueled light water moderated breeding hybrid power reactors

    International Nuclear Information System (INIS)

    The feasibility of fission-fusion hybrid reactors based on breeding light water thermal fission systems is investigated. The emphasis is on fuel-self-sufficient (FSS) hybrid power reactors that are fueled with natural uranium. Other LWHRs considered include FSS-LWHRs that are fueled with spent fuel from LWRs, and LWHRs which are to supplement LWRs to provide a tandem LWR-LWHR power economy that is fuel-self-sufficient

  6. Assessment of tritium breeding requirements for fusion power reactors

    International Nuclear Information System (INIS)

    This report presents an assessment of tritium-breeding requirements for fusion power reactors. The analysis is based on an evaluation of time-dependent tritium inventories in the reactor system. The method presented can be applied to any fusion systems in operation on a steady-state mode as well as on a pulsed mode. As an example, the UWMAK-I design was analyzed and it has been found that the startup inventory requirement calculated by the present method significantly differs from those previously calculated. The effect of reactor-parameter changes on the required tritium breeding ratio is also analyzed for a variety of reactor operation scenarios

  7. Power distribution control of CANDU reactors based on modal representation of reactor kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Xia, Lingzhi, E-mail: lxia4@uwo.ca [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Jiang, Jin, E-mail: jjiang@eng.uwo.ca [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Luxat, John C., E-mail: luxatj@mcmaster.ca [Department of Engineering Physics, McMaster University, Hamilton, Ontario L8S 4L7 (Canada)

    2014-10-15

    Highlights: • Linearization of the modal synthesis model of neutronic kinetic equations for CANDU reactors. • Validation of the linearized dynamic model through closed-loop simulations by using the reactor regulating system. • Design of a LQR state feedback controller for CANDU core power distribution control. • Comparison of the results of this new controller against those of the conventional reactor regulation system. - Abstract: Modal synthesis representation of a neutronic kinetic model for a CANDU reactor core has been utilized in the analysis and synthesis for reactor control systems. Among all the mode shapes, the fundamental mode of the power distribution, which also coincides with the desired reactor power distribution during operation, is used in the control system design. The nonlinear modal models are linearized around desired operating points. Based on the linearized model, linear quadratic regulator (LQR) control approach is used to synthesize a state feedback controller. The performance of this controller has been evaluated by using the original nonlinear models under load-following conditions. It has been demonstrated that the proposed reactor control system can produce more uniform power distribution than the traditional reactor regulation systems (RRS); in particular, it is more effective in compensating the Xenon induced transients.

  8. Studies on transferring the safety features of the module reactor to a large power reactor

    International Nuclear Information System (INIS)

    The German industries and research institutions have developed the HTR module reactor, which is strongly characterized by inherent safety features. The power output is limited to about 200 MWth because of its core configuration. It has been investigated in this work, whether the safety features of the module reactor can be transferred to larger power reactors. For this purpose the conceptual design of a ring core pebble bed reactor has been made with a thermal power output of 3000 MW. By means of computer calculations, the principal physical, thermohydraulical and safety features of the ring reactor have been studied. It has been shown that the 3000-MWth ring reactor basically possesses the same safety characteristics as the small module reactor. At reactivity disturbances, the reactor is shut down passively by the strongly negative temperature coefficient. The decay heat removal is also realized based on the passive priniciple. In the case of a total loss of coolant, the maximum fuel element temperature remains below 1600deg C; and consequently the retention of fission products in the fuel elements is fully attained. The control of xenon oscillations takes place inherently due to the mutual coupling between the local power production and the fuel temperature. (orig.)

  9. Preliminary nuclear power reactor technology qualitative assessment for Malaysia

    International Nuclear Information System (INIS)

    Since the worlds first nuclear reactor major breakthrough in December 02, 1942, the nuclear power industry has undergone tremendous development and evolution for more than half a century. After surpassing moratorium of nuclear power plant construction caused by catastrophic accidents at Three-mile island (1979) and Chernobyl (1986), today, nuclear energy is back on the policy agendas of many countries, both developed and developing, signaling nuclear revival or nuclear renaissance. Selection of suitable nuclear power technology has thus been subjected to primary attention. This short paper attempts to draw preliminary technology assessment for the first nuclear power reactor technology for Malaysia. Methodology employed is qualitative analysis collating recent finding of tnb-kepco preliminary feasibility study for nuclear power program in peninsular malaysia and other published presentations and/or papers by multiple experts. The results suggested that pressurized water reactor (PWR) is the prevailing technology in terms of numbers and plant performances, and while the commercialization of generation IV reactors is remote (e.g. Not until 2030), generation III/ III+ NPP models are commercially available on the market today. Five (5) major steps involved in reactor technology selection were introduced with a focus on introducing important aspects of selection criteria. Three (3) categories for the of reactor technology selection were used for the cursory evaluation. The outcome of these analyses shall lead to deeper and full analyses of the recommended reactor technologies for a comprehensive feasibility study in the near future. Recommendations for reactor technology option were also provided for both strategic and technical recommendations. The paper shall also implore the best way to select systematically the first civilian nuclear power reactor. (Author)

  10. Research and development into power reactor fuel performance

    International Nuclear Information System (INIS)

    The nuclear fuel in a power reactor must perform reliably during normal operation, and the consequences of abnormal events must be researched and assessed. The present highly reliable operation of the natural UO2 in the CANDU power reactors has reduced the need for further work in this area; however a core of expertise must be retained for purposes such as training of new staff, retaining the capability of reacting to unforeseen circumstances, and participating in the commercial development of new ideas. The assessment of fuel performance during accidents requires research into many aspects of materials, fuel and fission product behaviour, and the consolidation of that knowledge into computer codes used to evaluate the consequences of any particular accident. This work is growing in scope, much is known from out-reactor work at temperatures up to about 1500 degreesC, but the need for in-reactor verification and investigation of higher-temperature accidents has necessitated the construction of a major new in-reactor test loop and the initiation of the associated out-reactor support programs. Since many of the programs on normal and accident-related performance are generic in nature, they will be applicable to advanced fuel cycles. Work will therefore be gradually transferred from the present, committed power reactor system to support the next generation of thorium-based reactor cycles

  11. Self-operation type power control device for nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru.

    1993-07-23

    The device of the present invention operates by sensing the temperature change of a reactor core in all of LMFBR type reactors irrespective of the scale of the reactor core power. That is, a region where liquid poison is filled is disposed at the upper portion and a region where sealed gases are filled is disposed at the lower portion of a pipe having both ends thereof being closed. When the pipe is inserted into the reactor core, the inner diameter of the pipe is determined smaller than a predetermined value so that the boundary between the liquid poison and the sealed gases in the pipe is maintained relative to an assumed maximum acceleration. The sealed gas region is disposed at the reactor core region. If the liquid poison is expanded by the elevation of the reactor core exit temperature, it is moved to the lower gas region, to control the reactor power. Since high reliability can be maintained over a long period of time by this method, it is suitable to FBR reactors disposed in such environments that maintenance can not easily be conducted, such as desserts, isolated islands and undeveloped countries. Further, it is also suitable to ultra small sized nuclear reactors disposed at environments that the direction and the magnitude of gravity are different from those on the ground. (I.S.).

  12. Analysis of TRIGA reactor thermal power calibration method

    International Nuclear Information System (INIS)

    Analysis of thermal power method of the nuclear instrumentation of the TRIGA reactor in Ljubljana is described. Thermal power calibration was performed at different power levels and at different conditions. Different heat loss processes from the reactor pool to the surrounding are considered. It is shown that the use of proper calorimetric calibration procedure and the use of heat loss corrections improve the accuracy of the measurement. To correct the position of the control rods, perturbation factors are introduced. It is shown that the use of the perturbation factors enables power readings from nuclear instrumentation with accuracy better than without corrections.(author)

  13. Power instability and stochastic dynamics of periodic pulsed reactors

    International Nuclear Information System (INIS)

    This paper reports that physicists dealing with conventional reactor dynamics recognize two types of instability and reactor behavior beyond the stability region: asymptotic excursions and nonlinear periodic oscillations. A periodically pulsed reactor (PPR) has another peculiar instability: Under certain conditions, its power tends to oscillate at a frequency just twice less than the reactor pulsation frequency. The PPR dynamics far beyond the stability region are analyzed by using a discrete nonlinear model. A PPR with a negative temperature reactivity effect inevitably shows the chaotic power pulse energy behavior known as deterministic chaos. The way by which a reactor goes to chaos is defined by the time dependence of the feedback and by the kind of dynamics model used

  14. Management of Spent Nuclear Fuel from Nuclear Power Plant Reactor

    International Nuclear Information System (INIS)

    Management of spent nuclear fuel from Nuclear Power Plant (NPP) reactor had been studied to anticipate program of NPP operation in Indonesia. In this paper the quantity of generated spent nuclear fuel (SNF) is predicted based on the national electrical demand, power grade and type of reactor. Data was estimated using Pressurized Water Reactor (PWR) NPP type 1.000 MWe and the SNF management overview base on the experiences of some countries that have NPP. There are four strategy nuclear fuel cycle which can be developed i.e: direct disposal, reprocessing, DUPlC (Direct Use of Spent PWR Fuel In Candu) and wait and see. There are four alternative for SNF management i.e : storage at the reactor building (AR), away from reactor (AFR) using wet centralized storage, dry centralized storage AFR and prepare for reprocessing facility. For the Indonesian case, centralized facility of the wet type is recommended for PWR or BWR spent fuel. (author)

  15. Technology and use of low power research reactors

    International Nuclear Information System (INIS)

    The report contains a summary of discussions and 10 papers presented at the Consultants' Meeting on the Technology and Use of Low Power Research Reactors organized by the IAEA and held in Beijing (China) during 30 April - 3 May 1985. The following topics have been covered: reactor utilization in medicine and biology, in universities, for training, as a neutron source for radiography and some remarks on the safety of low power research reactors. A separate abstract was prepared for each paper presented at the meeting

  16. Extension of incompressible algorithms to compressible flows: validation on a governing valve mock up; Extension des algorithmes pour ecoulements incompressibles aux ecoulements compressibles: validation sur une maquette de vanne regulatrice

    Energy Technology Data Exchange (ETDEWEB)

    Baron, F.; Caruso, A.; Duplex, J.; Lefevre, L.

    1993-12-01

    The capacity of turbogenerators in PWR is regulated with governing valves located at the admission of the high-pressure turbine. In this paper we present a comparison between measurements and a numerical simulation of the flow in a 2D mock up of this governing valve. To predict and simulate transonic flow at low Mach numbers, we present a new extension of two codes initially devoted to incompressible and unsteady flows (pressure based method). The codes use either FInite Difference Method or, for complex geometry, Finite Element Method. Predicting those kinds of flows is difficult due to strong coupling between physical phenomena like turbulence on one hand, and the complexity of industrial geometry on the other hand. The comparison of numerical results with pressure measurements and also with Schlieren photographs confirms the validation of this approach. The results show clearly how the method correctly captures the structure of the jet. (authors). 10 figs., 11 refs.

  17. An overview of future sustainable nuclear power reactors

    Directory of Open Access Journals (Sweden)

    Andreas Poullikkas

    2013-01-01

    Full Text Available In this paper an overview of the current and future nuclear power reactor technologies is carried out. In particular, the nuclear technology is described and the classification of the current and future nuclear reactors according to their generation is provided. The analysis has shown that generation II reactors currently in operation all around the world lack significantly in safety precautions and are prone to loss of coolant accident (LOCA. In contrast, generation III reactors, which are an evolution of generation II reactors, incorporate passive or inherent safety features that require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. Today, partly due to the high capital cost of large power reactors generating electricity and partly due to the consideration of public perception, there is a shift towards the development of smaller units. These may be built independently or as modules in a larger complex, with capacity added incrementally as required. Small reactors most importantly benefit from reduced capital costs, simpler units and the ability to produce power away from main grid systems. These factors combined with the ability of a nuclear power plant to use process heat for co-generation, make the small reactors an attractive option. Generally, modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production and reduced installation costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction. Generation III+ designs are generally extensions of the generation III concept, which include advanced passive safety features. These designs can maintain the safe state without the use of any active control components. Generation IV reactors, which are future designs that are currently under research and development, will

  18. Regulation concerning installation and operation of reactors for power generation

    International Nuclear Information System (INIS)

    The regulation is defined under the law for the regulations of nuclear source materials, nuclear fuel materials and reactors and provisions concerning installation and operation of reactors for power generation in the order for execution of the law. Basic concepts and terms are explained, such as: radioactive waste; fuel assembly; exposure dose; accumulative dose; controlled area; safeguarded area; inspected surrounding area and employee. The application for permission of installation of reactors shall include location and general structure of reactor facilities, structure and equipment of reactors, handling and storing facilities of nuclear fuel materials and facilities for measurement and control, etc. Operation program of reactors shall be prepared for each reactor according to the form attached and filed every fiscal year from that one when the operation is expected to begin. Records shall be made for each reactor and kept for particular periods on inspection of reactor facilities, operation, fuel assembly, control of radiation and maintenance, etc. Entrance to the controlled area shall be limited through specified measures. Exposure dose, inspection, check up, independent examination and operation of reactors, transport and disposal in the works or the enterprise and others are in detail stipulated. Reports shall be submitted to the Minister of International Trade and Industry on concentration of radioactive materials, exposure dose of employees and other designated matters. (Okada, K.)

  19. Conceptual design of nuclear fusion power reactor DREAM. Reactor structures and remote maintenance

    International Nuclear Information System (INIS)

    Nuclear fusion reactors are required to be able to compete another energy sources in economy, reliability, safety and environmental integrity for commercial use. In the DREAM (DRastically EAsy Maintenance) reactor, a very low activated material of SiC/SiC composite has been introduced for the structural material, a reactor configuration for very easy maintenance and the helium gas of a high temperature for the cooling system, and hence DREAM has been proven to be very attractively as the commercial power reactor due to the high availability and efficiency of the plant and minimization of radioactive wastes. (author)

  20. Fast-breeder-power reactor records in the INIS database

    International Nuclear Information System (INIS)

    This report presents a statistical analysis of more than 19,700 records of publications concerned with research and technology in the field of fast breeder power fission reactors which are included in the INIS Bibliographic Database for the period from 1970. to 1999. The main objectives of this bibliometric study were: to make an inventory of the fast breeder power reactor related records in the INIS Database; to provide statistics and scientific indicators for the INIS users, namely science managers, researchers, engineers, operators, scientific editors and publishers, decision-makers in the field of fast breeder power reactors related subjects; to extract other useful information from the INIS Bibliographic Database about articles published in fast breeder reactors research and technology. The quantitative data in this report are obtained for various properties of relevant INIS records such as year of publication, secondary subject categories, countries of publication, language, publication types, literary types, etc. (author)

  1. Training simulator for nuclear power plant reactor monitoring

    International Nuclear Information System (INIS)

    A description is given of a method and apparatus for the real-time dynamic simulation of a nuclear power plant that includes a control and nuclear instrumentation console for operating the reactor and monitoring three-dimensional physical values in the reactor core. A digital computer is connected to the console to calculate physical values such as nuclear flux, power, and temperature including the distribution thereof throughout the core with such calculations including the effect of full length, part length, and malfunctioned reactor control rods, as well as xenon, decay heat and boron, for example, on the output and distribution of power within the core. The simulation also includes instrumentation that responds to the calculated physical values by recording a continuous trace of the flux value in the reactor core from the top to the bottom

  2. Small reactor power systems for manned planetary surface bases

    International Nuclear Information System (INIS)

    A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options

  3. Design of megawatt power level heat pipe reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mcclure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dasari, Venkateswara Rao [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reid, Robert Stowers [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-12

    An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors. The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.

  4. Sealing device for nuclear power reactor

    International Nuclear Information System (INIS)

    The sealing device is to stop a leak on a reactor pressure vessel where control of the output of reactor is arranged by control rods which are handled by drives connected to control rods and bars in tubes which penetrate the reactor wall. Each tube has a supporting case on the inside of the wall opened to the hole and welded to the tube. The weld may crack and leak. Then an inner sealing tube made of soft metallic material whose outer surface is conical is drawn on to the tube over which an outer sealing tube made of hard metallic material and conical inner surface is placed. On both sides of the crack special adhering planes are formed between the inner sealing tube and the tubes or the supporting case. When the outer sealing tube is pressed over the inner sealing tube, the conical surfaces tighten it against the tube and the supporting case

  5. Application of reactor-pumped lasers to power beaming

    Science.gov (United States)

    Repetti, T. E.

    1991-10-01

    Power beaming is the concept of centralized power generation and distribution to remote users via energy beams such as microwaves or laser beams. The power beaming community is presently performing technical evaluations of available lasers as part of the design process for developing terrestrial and space-based power beaming systems. This report describes the suitability of employing a nuclear reactor-pumped laser in a power beaming system. Although there are several technical issues to be resolved, the power beaming community currently believes that the AlGaAs solid-state laser is the primary candidate for power beaming because that laser meets the many design criteria for such a system and integrates well with the GaAs photodiode receiver array. After reviewing the history and physics of reactor-pumped lasers, the advantages of these lasers for power beaming are discussed, along with several technical issues which are currently facing reactor-pumped laser research. The overriding conclusion is that reactor-pumped laser technology is not presently developed to the point of being technically or economically competitive with more mature solid-state technologies for application to power beaming.

  6. Hiberarchy of requirement analysis of reactor protection system for advanced pressurized water reactor nuclear power plant

    International Nuclear Information System (INIS)

    In order to improve the security and the margin of safety of nuclear power plant, the research on requirement analysis of digital reactor protection system for advanced pressurized water reactor nuclear power plant was developed. Based on the known technology, a requirement analysis report was performed. A kind of three-levels pyramidal hierarchy was adopted in the requirement analysis, and the design characteristics of the requirement analysis were described in the analysis report. This hiberarchy can directly illuminate the design characters and logical achievement of the requirement analysis for advanced pressurized water reactor digital protection system. (authors)

  7. Automatic power control for a pressurized water reactor

    International Nuclear Information System (INIS)

    During a normal operation of a pressurized water reactor (PWR), the reactivity is controlled by control rods, boron, and the average temperature of the primary coolant. Especially in load follow operation, the reactivity change is induced by changes in power level and effects of xenon concentration. The control of the core power distribution is concerned, mainly, with the axial power distribution which depends on insertion and withdrawal of the control rods resulting in additional reactivity compensation. The utilization of part strength control element assemblies (PSCEAs) is quite appropriate for a control of the power distribution in the case of Yonggwang Nuclear Unit 3 (YGN Unit 3). However, control of the PSCEAs is not automatic, and changes in the boron concentration by dilution/boration are done manually. Thus, manual control of the PSCEAs and the boron concentration require the operator's experience and knowledge for a successful load follow operation. In this thesis, the new concepts have been proposed to adapt for an automatic power control in a PWR. One of the new concepts is the mode K control, another is a fuzzy power control. The system in mode K control implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial power shape change, which allows automatic control of the axial power distribution. And the mode K enables precise regulation, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1000-MWe PWR which is a reference plant for the YGN Unit 3. The simulation results illustrate that the mode K would be

  8. Preliminary Development of Thermal Power Calculation Code H-Power for a Supercritical Water Reactor

    OpenAIRE

    2014-01-01

    SCWR (Supercritical Water Reactor) is one of the promising Generation IV nuclear systems, which has higher thermal power efficiency than current pressurized water reactor. It is necessary to perform the thermal equilibrium and thermal power calculation for the conceptual design and further monitoring and calibration of the SCWR. One visual software named H-Power was developed to calculate thermal power and its uncertainty of SCWR, in which the advanced IAPWS-IF97 industrial formulation was us...

  9. Fixing the 'zero state' at the main equipment of the primary coolant circuit at the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    To determine the reference, so-called 'zero' state of the main reactor components and to prepare the original fault-map, a four-fold inspection system was applied to the primary circuit equipment of the Paks NPP based on the inspection of the reactor components and weldings before and after assembly works and after cold and hot washings. Investigations are carried out either manually or using special manipulators able to inspect the welded joints, and the structural materials of the reactor vessels by means of ultrasonic measurements and to monitor visually the vessel mantles by means of TV cameras. (V.N.)

  10. Gas-cooled reactor for space power systems

    International Nuclear Information System (INIS)

    Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors

  11. Automatic Meter Reading using Power Line Signaling and Voltage Zero-crossing Detection

    Directory of Open Access Journals (Sweden)

    C.L. Vasu

    2015-06-01

    Full Text Available In India, the electric power transmission and distribution loss is very high, about 7% in transmission and 26% in distribution. Though deployment of automated meter reading system will reduce losses, particularly in distribution, penetration of automated meter reading is low due to high costs involved. World over, the Two-Way Automatic Communications System (TWACS is the most widely used power line communications technology offering two-way communication between substation and end users. The TWACS introduces disturbance on the power system voltage for short durations near zero-crossing to generate the outbound (from substation to end user signal. To generate the inbound (from end user to substation signal, short duration current pulses are introduced, near voltage zero-crossings. Information is generated as a sequential combination of voltage disturbances for the outbound signal and current pulses for the inbound signal. The current study proposes a low-cost modification of the TWACS to reduce voltage and current harmonics. The proposed system has been modelled and simulated using SIMULINK/SIMPOWER Systems. The simulation results show that there is a reduction in voltage harmonics from 0.84 to 0.17% and in current harmonics from 2.07 to 1.10%.

  12. Laser fusion power reactor system (LFPRS)

    Energy Technology Data Exchange (ETDEWEB)

    Kovacik, W. P.

    1977-12-19

    This report gives detailed information for each of the following areas: (1) reference concept description, (2) nuclear design, (3) structural design, (4) thermal and fluid systems design, (5) materials design and analysis, (6) reactor support systems and balance of plant, (7) instrumentation and control, (8) environment and safety, (9) economics assessment, and (10) development requirements. (MOW)

  13. Laser fusion power reactor system (LFPRS)

    International Nuclear Information System (INIS)

    This report gives detailed information for each of the following areas: (1) reference concept description, (2) nuclear design, (3) structural design, (4) thermal and fluid systems design, (5) materials design and analysis, (6) reactor support systems and balance of plant, (7) instrumentation and control, (8) environment and safety, (9) economics assessment, and (10) development requirements

  14. Nuclear Power from Fission Reactors. An Introduction.

    Science.gov (United States)

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

  15. Special Nuclear Material Control by the Power Reactor Operator

    International Nuclear Information System (INIS)

    A relatively new and extremely valuable fuel for electric power production, uranium, requires very careful inventory control from the time the reactor operator assumes financial responsibility for this material until, as partially expended fuel, it is transferred to another facility and the remaining part of its initial value is recovered. Most power reactor operators were operating fossil-fuelled power plants before the advent of nuclear power and have long since established rather complete and adequate controls for these fossil fuels. The reactor operator must have no less adequate controls for the special nuclear material used in his nuclear plant. Power reactor, operation is not an ancient science and during its relatively short history our engineers and scientists have been constantly improving plant designs and methods of operation to reduce costs and make our nuclear plants competitive with fossil-fuelled conventional plants. Nuclear material management must be as modern and efficient as is humanly possible to ensure that technological advances leading to reduced costs are not lost by poor handling of nuclear fuel and the records pertaining to fuel inventory. Nuclear material management requires the maintaining of complete and informative records by the power reactor operator. These records need not be complex to satisfy the criteria of completeness and adequacy. In fact, simplicity is extremely desirable. Despite the fact that nuclear fuel is new and completely different to our conventional fuels no mystery should be attached thereto. Nuclear material control as part of nuclear material management is not limited to simple inventory work but it is the basis for a great deal of other activity that is an inherent part of any power reactor operations such as irradiated fuel shipments, reprocessing of spent fuel, with its associated accounting for reclaimed fuel and material produced during reactor operation, and the establishing and maintaining of an adequate

  16. The core design of the advanced power reactor plus (APR+)

    International Nuclear Information System (INIS)

    Advance Power Reactor Plus (APR+), a pressurized water reactor and an improved nuclear power reactor based on the Advanced Power Reactor 1400 MWe (APR1400) in Korea, has been developed with 18-month cycle operation strategy from its initial core. The APR+ core power is 4290 MWth which corresponds to a 1500 MWe class nuclear power plant. The reactor core consists of 257 fuel assemblies. Comparing with APR1400 core design, 16 fuel assemblies are added. Its cycle length is expected about 450 EFPD directly from initial core, although most of previous other plants had been started according to their annual or 15-month cycle operation schedule at their initial core and gone to 18-month after third - fourth cycle. In order to reduce the peaking power, fuel pin configurations of the assembly, are optimized by using some low enriched fuel pins and gadolinia bearings. APR+ core has been met the requirements as well as the above cycle length requirement; 1) peaking factor, 2) Negative MTC(Moderator Temperature Coefficient), 3) sufficient shutdown margin, 4) convergent Xenon stability Index. The maximum rod burnup and the discharge fuel assembly burnup are also satisfied those of the limit. It is expected to acquire the standard design approval by the end of 2012 by the Korean nuclear regulatory. (authors)

  17. Single-switch three-phase zero-current-transition rectifier with power factor correction

    OpenAIRE

    Gatari?, Slobodan

    1994-01-01

    A novel, zero-current-transition (ZCT) topology of the single-switch three-phase boost PFC rectifier is proposed. The soft transition is achieved with a low-power auxiliary circuit employing an additional switch. The circuit can be used with an IGBT at switching frequencies up to 50. Its operation is analyzed in detail, and design guidelines are provided. The small signal model of the circuit is developed, and voltage mode control is designed. The results are verified on a 4 kW...

  18. New Whole-House Solutions Case Study: New Town Builders' Power of Zero Energy Center - Denver, Colorado

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-10-01

    New Town Builders, a builder of energy efficient homes in Denver, Colorado, offers a zero energy option for all the homes it builds. To attract a wide range of potential homebuyers to its energy efficient homes, New Town Builders created a "Power of Zero Energy Center" linked to its model home in the Stapleton community. This case study presents New Town Builders' marketing approach, which is targeted to appeal to homebuyers' emotions rather than overwhelming homebuyers with scientific details about the technology. The exhibits in the Power of Zero Energy Center focus on reduced energy expenses for the homeowner, improved occupant comfort, the reputation of the builder, and the lack of sacrificing the homebuyers' desired design features to achieve zero net energy in the home. This case study also contains customer and realtor testimonials related to the effectiveness of the Center in influencing homebuyers to purchase a zero energy home.

  19. Building America Case Study: New Town Builders' Power of Zero Energy Center, Denver, Colorado (Brochure)

    Energy Technology Data Exchange (ETDEWEB)

    2014-10-01

    New Town Builders, a builder of energy efficient homes in Denver, Colorado, offers a zero energy option for all the homes it builds. To attract a wide range of potential homebuyers to its energy efficient homes, New Town Builders created a 'Power of Zero Energy Center' linked to its model home in the Stapleton community of Denver. This case study presents New Town Builders' marketing approach, which is targeted to appeal to homebuyers' emotions rather than overwhelming homebuyers with scientific details about the technology. The exhibits in the Power of Zero Energy Center focus on reduced energy expenses for the homeowner, improved occupant comfort, the reputation of the builder, and the lack of sacrificing the homebuyers' desired design features to achieve zero net energy in the home. The case study also contains customer and realtor testimonials related to the effectiveness of the Center in influencing homebuyers to purchase a zero energy home.

  20. PC version of PRIS (Power Reactor Information System)

    International Nuclear Information System (INIS)

    The IAEA has been collecting operating experience data on nuclear power plants in the Member States since 1970. In 1980 a computerized database was established, the IAEA Power Reactor Information System (PRIS). To make PRIS data available to the Member States in a more convenient format, the development of a PC version of PRIS started in 1989