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

  1. Role of irradiation reactor mock-ups

    International Nuclear Information System (INIS)

    Casali, F.; Cerles, J.M.; Debrue, J.

    1977-01-01

    A survey is given of the utilization of low power facilities in support to irradiation reactor experiments. The BRO2, ISIS and RB3 facilities are described as neutronic mock-ups of the BR2, OSIRIS and ESSOR reactors respectively

  2. Determination of power density in VVER-1000 Mock-Up in LR-0 reactor

    Directory of Open Access Journals (Sweden)

    Košál Michal

    2017-01-01

    Full Text Available The pin power density is an important quantity which has to be monitored during the reactor operation, for two main reasons. Firstly, it is part of the limits and conditions of safe operation and, secondly, it is source term in neutron transport calculations used for the adequate assessing of the state of core structures and pressure vessel material. It is often calculated using deterministic codes which may have problems with an adequate definition of boundary conditions in subcritical regions. This may lead to overestimation of real situation, and therefore the validation of the utility codes contributes not only to better fuel utilization, but also to more precise description of radiation situation in structural components of core. Current paper presents methods developed at LR-0 reactor, as well as selected results for pin power density measurement in peripheral regions of VVER-1000 mock-up. The presented data show that the results of a utility diffusion code at core boundary overestimate the measurement. This situation, however satisfactory safe, may lead to unduly conservative approach in the determination of radiation damage of core structures.

  3. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2002-01-01

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

  4. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2001-01-01

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

  5. 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.

  6. 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.

  7. Reactor BR2

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2000-07-01

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

  8. Reactor BR2

    International Nuclear Information System (INIS)

    Gubel, P.

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    Vermeeren, L.; Dekeyser, J.; Gouat, P.; Kalcheva, S.; Koonen, E.; Kuzminov, V.; Verwimp, A.; Weber, M.

    2005-01-01

    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)

  10. BR2 Reactor: Introduction

    International Nuclear Information System (INIS)

    Moons, F.

    2007-01-01

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

  11. Reactor BR2: Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2000-01-01

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

  12. BR2 reactor neutron beams

    International Nuclear Information System (INIS)

    Neve de Mevergnies, M.

    1977-01-01

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

  13. Team training using full-scale reactor coolant pump seal mock-ups

    International Nuclear Information System (INIS)

    McDonald, T.J.; Hamill, R.W.

    1987-01-01

    The use of full-scale reactor coolant pump (RCP) seal mock-ups has greatly enhanced Northeast Utilities' ability to effectively utilize the team training approach to technical training. With the advent of the Institute of Nuclear Power Operations accreditation come a new emphasis and standards for the integrated training of plant engineering personnel, maintenance mechanics, quality control personnel, and health physics personnel. The results of purchasing full-scale RCP mock-ups to pilot the concept of team training have far exceeded expectations and cost-limiting factors. The initial training program analysis identified RCP seal maintenance as a task that required training for maintenance department personnel. Due to radiation exposure considerations and the unavailability of actual plant equipment for training purposes, the decision was made to procure a mock-up of an RCP seal assembly and housing. This mock-up was designed to facilitate seal cartridge removal, disassembly, assembly, and installation, duplicating all internal components of the seal cartridge and housing area in exact detail

  14. Dosimetry work and calculations in connection with the irradiation of large devices in the high flux materials testing reactor BR2

    International Nuclear Information System (INIS)

    De Raedt, C.; Leenders, L.; Tourwe, H.; Farrar, H. IV.

    1982-01-01

    For about fifteen years the high flux reactor BR2 has been involved in the testing of fast reactor fuel pins. In order to simulate the fast reactor neutron environment most devices are irradiated under cadmium screen, cutting off the thermal flux component. Extensive neutronic calculations are performed to help the optimization of the fuel bundle design. The actual experiments are preceded by irradiations of their mock-ups in BR02, the zero power model of BR2. The mock-up irradiations, supported by supplementary calculations, are performed for the determination of the main neutronic characteristics of the irradiation proper in BR2 and for the determination of the corresponding operation data. At the end of the BR2 irradiation, the experimental results, such as burn-ups, neutron fluences, helium production in the fuel pin claddings, etc. are correlated by neutronic calculations in order to examine the consistency of the post-irradiation results and to validate the routine calculation procedure and cross-section data employed. A comparison is made in this paper between neutronic calculation results and some post-irradiation data for MOL 7D, a cadmium screened sodium cooled loop containing a nineteen fuel pin bundle

  15. Refurbishing the BR2 materials testing reactor

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  16. Irradiation techniques at BR2 reactor

    International Nuclear Information System (INIS)

    Hebel, W.

    1978-01-01

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

  17. Ageing management of the BR2 research reactor

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  18. 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)

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

    International Nuclear Information System (INIS)

    Rabello, Emerson G.; Silva, Luiz L.; Gomes, Paulo T.V.

    2011-01-01

    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)

  20. BR2 Reactor: Irradiation of fuels

    International Nuclear Information System (INIS)

    Verwimp, A.

    2005-01-01

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

  1. Refurbishment programme for the BR2-reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-07-01

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

  2. Refurbishment programme for the BR2-reactor

    International Nuclear Information System (INIS)

    Koonen, E.

    1992-01-01

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

  3. Mock-up critical experiments for prototype fast breeder reactor 'Monju'

    International Nuclear Information System (INIS)

    Zukeran, Atsushi; Inoue, Teruji; Suzuki, Takeo; Kawashima, Kanau

    1976-01-01

    The mock-up criticality experiments for Monju are roughly divided into the full mock-up test using the ZEBRA of Winfrith Institute, UK AEA, and the partial mock-up experiment with FCA of JAERI. The former test has been carried out over 18 months from September 1971 as the Japan-UK cooperative research project MOZART. With the FCA, the experiment complementing the MOZART has been carried out, focusing on the nuclear characteristics of Monju which can be simulated with a relatively small core, and the experiment on highly enriched control rods and shielding is being continued now with the FCA 7 core. The experimental data of the MOZART and the ZPPR series in USA were exchanged at the international symposium in Tokyo, thus the prediction and the accuracy evaluation of the nuclear characteristics of Monju became possible, and the highly reliable core design was able to be accomplished. The simulated criticality experiment is necessary for directly grasping the reliability of calculated values in comparison with the experimental values, and also for the experimental prediction of the nuclear characteristics. The outline and the analysis of the simulated criticality experiment such as reactivity factor, control rod value, reaction rate distribution and sodium void reactivity are described, and the reflection of the results to the design of the core of Monju is explained. (Kako, I.)

  4. High RF power test of a lower hybrid module mock-up in carbon fiber composite

    International Nuclear Information System (INIS)

    Goniche, M.; Bibet, P.; Brossaud, J.; Cano, V.; Froissard, P.; Kazarian, F.; Rey, G.; Maebara, S.; Kiyono, K.; Seki, M.; Suganuma, K.; Ikeda, Y.; Imai, T.

    1999-02-01

    A mock-up module of a Lower Hybrid Current Drive antenna module of a Carbon Fiber Composite (CFC) was fabricated for the development of heat resistive front facing the plasma. This module is made from CFC plates and rods which are copper coated to reduce the RF losses. The withstand-voltage, the RF properties and outgassing rates for long pulses and high RF power were tested at the Lower Hybrid test bed facility of Cadarache. After the short pulse conditioning, long pulses with a power density ranging between 50 and 150 MW/m 2 were performed with no breakdowns. During these tests, the module temperature was increasing from 100-200 deg. C to 400-500 deg. C. It was also checked that high power density, up to 150 MW/m 2 , could be transmitted when the waveguides are filled with H 2 at a pressure of 5 x 10 -2 Pa. No significant change in the reflection coefficient is measured after the long pulse operation. During a long pulse, the power reflection increases during the pulse typically from 0.8% to 1.3%. It is concluded that the outgassing rate of Cu-plated CFC is about 6 times larger than of Dispersion Strengthened Copper (DSC) module at the module temperature of 300 deg. C. No significant increase of the global outgassing of the CFC module was measured after hydrogen pre-filling. After the test, visual inspection revealed that peeling of the copper coating occurred at one end of the module only on a very small area (0.2 cm 2 ). It is assessed that a CFC module is an attractive candidate for the hardening of the tip of the LHCD antenna. (authors)

  5. High RF power test of a lower hybrid module mock-up in Carbon Fiber Composite

    International Nuclear Information System (INIS)

    Maebara, Sunao; Kiyono, Kimihiro; Seki, Masami

    1997-11-01

    A mock-up module of a Lower Hybrid Current Drive antenna module of a Carbon Fiber Composite (CFC) was fabricated for the development of heat resistive front facing the plasma. This module is made from CFC plates and rods which are copper coated to reduce the RF losses. The withstand-voltage, the RF properties and outgassing rates for long pulses and high RF power were tested at the Lower Hybrid test bed facility of Cadarache. After the short pulse conditioning, long pulses with a power density ranging between 50 and 150 MW/m 2 were performed with no breakdowns. During these tests, the module temperature was increasing from 100-200degC to 400-500degC. It was also checked that high power density, up to 150 MW/m 2 , could be transmitted when the waveguides are filled with H 2 at a pressure of 5 x 10 -2 Pa. No significant change in the reflection coefficient is measured after the long pulse operation. During a long pulse, the power reflection increases during the pulse typically from 0.8 % to 1.3 %. It is concluded that the outgassing rate of Cu-plated CFC is about 6-7 times larger than of Dispersion Strengthened Copper (DSC) module at the module temperature of 300degC. No significant increase of the global outgassing of the CFC module was measured after hydrogen prefilling. After the test, visual inspection revealed that peeling of the copper coating occurred at one end of the module only on a very small area (0.2 cm 2 ). It is assessed that a CFC module is an attractive candidate for the hardening of the tip of the LHCD antenna. (author)

  6. BR2 reactor: medical and industrial applications

    International Nuclear Information System (INIS)

    Ponsard, B.

    2005-01-01

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

  7. 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.

  8. The Swedish Zero Power Reactor R0

    International Nuclear Information System (INIS)

    Landergaard, Olof; Cavallin, Kaj; Jonsson, Georg

    1961-05-01

    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

  9. Material test reactor fuel research at the BR2 reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-03-15

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

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

    Directory of Open Access Journals (Sweden)

    Košťál Michal

    2016-01-01

    Full Text Available 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.

  11. Behavior of underclad cracks in reactor pressure vessels - evaluation of mechanical analyses with tests on cladded mock-ups

    International Nuclear Information System (INIS)

    Moinereau, D.; Rousselier, G.; Bethmont, M.

    1993-01-01

    Innocuity of underclad flaws in the reactor pressure vessels must be demonstrated in the French safety analyses, particularly in the case of a severe transient at the end of the pressure vessel lifetime, because of the radiation embrittlement of the vessel material. Safety analyses are usually performed with elastic and elasto-plastic analyses taking into account the effect of the stainless steel cladding. EDF has started a program including experiments on large size cladded specimens and their interpretations. The purpose of this program is to evaluate the different methods of fracture analysis used in safety studies. Several specimens made of ferritic steel A508 C1 3 with stainless steel cladding, containing small artificial defects, are loaded in four-point bending. Experiments are performed at very low temperature to simulate radiation embrittlement and to obtain crack instability by cleavage fracture. Three tests have been performed on mock-ups containing a small underclad crack (with depth about 5 mn) and a fourth test has been performed on one mock-up with a larger crack (depth about 13 mn). In each case, crack instability occurred by cleavage fracture in the base metal, without crack arrest, at a temperature of about - 170 deg C. Each test is interpreted using linear elastic analysis and elastic-plastic analysis by two-dimensional finite element computations. The fracture are conservatively predicted: the stress intensity factors deduced from the computations (K cp or K j ) are always greater than the base metal toughness. The comparison between the elastic analyses (including two plasticity corrections) and the elastic-plastic analyses shows that the elastic analyses are often conservative. The beneficial effect of the cladding in the analyses is also shown : the analyses are too conservative if the cladding effects is not taken into account. (authors). 9 figs., 6 tabs., 10 refs

  12. Design characteristics of zero power fast reactor Lasta

    International Nuclear Information System (INIS)

    Milosevic, M.; Stefanovic, D.; Pesic, M.; Popovic, D.; Nikolic, D.; Antic, D.; Zavaljevski, N.

    1987-01-01

    The concept, purpose and preliminary design of a zero power fast reactor LASTA are described. The methods of computing the reactor core parameters and reactor kinetics are presented with the basic calculated results and analysis for one selected LASTA configuration. The nominal parameters are determined according to the selected reactor safety criteria and results of calculations. Important aspects related to the overall safety are examined in detail. (author)

  13. Uncertainty analysis and flow measurements in an experimental mock-up of a molten salt reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, Bogdan; Aszodi, Attila [Budapest University of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-09-15

    In the paper measurement results from the experimental modelling of a molten salt reactor concept will be presented along with detailed uncertainty analysis of the experimental system. Non-intrusive flow measurements are carried out on the scaled and segmented mock-up of a homogeneous, single region molten salt fast reactor concept. Uncertainty assessment of the particle image velocimetry (PIV) measurement system applied with the scaled and segmented model is presented in detail. The analysis covers the error sources of the measurement system (laser, recording camera, etc.) and the specific conditions (de-warping of measurement planes) originating in the geometry of the investigated domain. Effect of sample size in the ensemble averaged PIV measurements is discussed as well. An additional two-loop-operation mode is also presented and the analysis of the measurement results confirm that without enhancement nominal and other operation conditions will lead to strong unfavourable separation in the core flow. It implies that use of internal flow distribution structures will be necessary for the optimisation of the core coolant flow. Preliminary CFD calculations are presented to help the design of a perforated plate located above the inlet region. The purpose of the perforated plate is to reduce recirculation near the cylindrical wall and enhance the uniformity of the core flow distribution.

  14. Operation of the BR2 Reactor

    International Nuclear Information System (INIS)

    Gubel, P.

    2006-01-01

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

  15. Operation of the BR2 Reactor

    International Nuclear Information System (INIS)

    Gubel, P.

    2005-01-01

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

  16. Design characteristics of research zero power fast reactor Lasta

    International Nuclear Information System (INIS)

    Milosevic, M.; Stefanovic, D.; Pesic, M.; Nikolic, D.; Antic, D.; Zavaljevski, N.; Popovic, D.

    1990-01-01

    LASTA is a flexible zero power reactor with uranium and plutonium fuel designed for research in the neutron physics and in the fast reactor physics. Safety considerations and experimental flexibility led to the choice of a fixed vertical assembly with two safety blocks as the main safety elements, so that safety devices would be operated by gravity. The neutron and reactor physics, the control and safety philosophy adopted in our design, are described in this paper. Developed computer programs are presented. (author)

  17. Preliminary concept of a zero power nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D., E-mail: lamai@ipen.b, E-mail: ptsiquei@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  18. Preliminary concept of a zero power nuclear reactor core

    International Nuclear Information System (INIS)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D.

    2011-01-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  19. Low drift type N thermocouples in out-of-pile advanced gas reactor mock-up test: metallurgical analysis

    International Nuclear Information System (INIS)

    Scervini, M.; Palmer, J.; Haggard, D.C.; Swank, W.D.

    2015-01-01

    Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. They are crucial for the control of current nuclear reactors and for the development of GEN IV reactors. In nuclear applications thermocouples are strongly affected by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation for relatively short times compared to the life required for temperature sensors in nuclear applications. Previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of Nickel based thermocouples is limited to temperatures lower than 1000 deg. C due to drift related to phenomena other than nuclear irradiation. As part of a collaboration between Idaho National Laboratory (INL) and the University of Cambridge a variety of Type N thermocouples have been exposed at INL in an Advanced Gas Reactor mock-up test at 1150 deg. C for 2000 h, 1200 deg. C for 2000 h, 125 deg. C for 200 h and 1300 deg. C for 200 h, and later analysed metallurgically at the University of Cambridge. The use of electron microscopy allows to identify the metallurgical changes occurring in the thermocouples during high temperature exposure and correlate the time dependent thermocouple drift with the microscopic changes experienced by the thermoelements of different thermocouple designs. In this paper conventional Inconel 600 sheathed type N thermocouples and a type N using a customized sheath developed at the University of

  20. Low drift type N thermocouples in out-of-pile advanced gas reactor mock-up test: metallurgical analysis

    Energy Technology Data Exchange (ETDEWEB)

    Scervini, M. [University of Cambridge, Department of Materials Science and Metallurgy, 27 Charles Babbage Road, CB30FS Cambridge, (United Kingdom); Palmer, J.; Haggard, D.C.; Swank, W.D. [Idaho National Laboratory, Idaho Falls, ID 83415-3840, (United States)

    2015-07-01

    Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. They are crucial for the control of current nuclear reactors and for the development of GEN IV reactors. In nuclear applications thermocouples are strongly affected by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation for relatively short times compared to the life required for temperature sensors in nuclear applications. Previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of Nickel based thermocouples is limited to temperatures lower than 1000 deg. C due to drift related to phenomena other than nuclear irradiation. As part of a collaboration between Idaho National Laboratory (INL) and the University of Cambridge a variety of Type N thermocouples have been exposed at INL in an Advanced Gas Reactor mock-up test at 1150 deg. C for 2000 h, 1200 deg. C for 2000 h, 125 deg. C for 200 h and 1300 deg. C for 200 h, and later analysed metallurgically at the University of Cambridge. The use of electron microscopy allows to identify the metallurgical changes occurring in the thermocouples during high temperature exposure and correlate the time dependent thermocouple drift with the microscopic changes experienced by the thermoelements of different thermocouple designs. In this paper conventional Inconel 600 sheathed type N thermocouples and a type N using a customized sheath developed at the University of

  1. The state of art of the manufacturing technology of FW blanket and the development of mock-up for fusion reactor in Russia

    International Nuclear Information System (INIS)

    Baek, Jong Hyuk; Jeong, Y. H.; Park, J. Y.; Kim, J. H.; Kim, H. G.

    2004-08-01

    In early 1990s, Russia had carried out the performance tests to verify the optimization of Be tile geometry and the bonding integrity of small mock-up using a HHF (High Heat Flux) test and an in-pile test in a research reactor. They had obtained the reliability of the brazing technologies for the Be/Cu bonding. And they had manufactured the near real-size large mock-up (about 0.8 mm in length) to find the bonding integrity by a fast brazing technique. They had a satisfied results from the HHF test for the large mock-up. Additionally, an alternative FW mock-ups, which were manufactured by both casting and fast brazing techniques to reduce the joining parts, showed a good joining performance from the HHF test. Therefore, it was concluded that the fast brazing techniques could be strongly recommended as a one of the preferable joining techniques and be possible to apply to joining for the Be/Cu joining of FW blanket

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

    International Nuclear Information System (INIS)

    Girard, J.P.; Buravand, Y.

    1982-10-01

    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

  3. In-pile testing of ITER first wall mock-ups at relevant thermal loading conditions in the LVR-15 nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kysela, Jan [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Entler, Slavomir, E-mail: slavomir.entler@cvrez.cz [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Vsolak, Rudolf; Klabik, Tomas [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Zlamal, Ondrej [CEZ, Duhova 2/1444, 140 53 Praha 4 (Czech Republic); Bellin, Boris; Zacchia, Francesco [Fusion for Energy, Josep Pla, 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

    2015-10-15

    Highlights: • Irradiated thermal fatigue testing of the ITER primary first wall mock-ups. • Cyclic heat flux of 0.5 MW/m{sup 2} in the neutron field of the nuclear reactor core. • 17,040 thermal cycles. • Radiation damage in the range of 0.41–1.17 dpa depending on the material. - Abstract: The TW3 in-pile rig enabled the thermal fatigue testing of ITER primary first wall mock-ups in the core of the nuclear reactor. This experiment investigated the neutron irradiation influence on the design performance under high heat flux testing. A thermal flux of 0.5 MW/m{sup 2} in the neutron field of the core of the LVR-15 nuclear reactor was applied. Within the scope of the tests with simultaneous neutron irradiation, the TW3 rig reached a record of 17,040 thermal cycles with the radiation damage in the range of 0.41–1.17 dpa depending on the material. Even after a high number of thermal cycles, while being irradiated by neutrons, no damage of the tested mock-ups was visually observed. Further testing and analysis will follow in the Forschungszentrum Juelich.

  4. The zero-power basis of fast reactor dosimetry

    International Nuclear Information System (INIS)

    Sanders, J.E.

    1978-06-01

    Predictions of reaction rates, atomic displacements, and gamma-ray energy deposition in the Prototype Fast Reactor are based on cross-section data and calculation methods validated against the results of zero-power experiments. The paper reviews work in Zebra relevant to this dosimetry, including neutron spectrometry, power mapping, foil activations within core heterogeneities, and measurements with thermoluminescent detectors. Comparisons of experiment and calculation are discussed in relation to the accuracies required to meet materials testing objectives. (author)

  5. The zero-power basis of fast reactor dosimetry

    International Nuclear Information System (INIS)

    Sanders, J.E.

    1978-06-01

    Predictions of reaction rates, atomic displacements, and gamma-ray energy deposition in the Prototype Fast Reactor are based on cross-section data and calculation methods validated against the results of zero-power experiments. The paper reviews work in Zebra relevant to this dosimetry, including neutron spectrometry, power mapping, foil activations within core heterogeneities, and measurements with thermoluminescent detectors. Comparisons of experiment and calculation are discussed in relation to the accuracies required to meet material testing objectives. (author)

  6. Experimental measurement of zero power reactor transfer function

    International Nuclear Information System (INIS)

    Liang Shuhong

    2011-01-01

    In order to study the zero power reactor (ZPR) transfer function, the ZPR transfer function expression was deduced with the point reactor kinetics equation, which was disturbed by reactivity input response. Based on the Fourier analysis for the input of triangular wave, the relation between the transfer function and reactivity was got. Validating research experiment was made on the DF-VI fast ZPR. After the disturbed reactivity was measured, the experimental value of the transfer function was got. According to the experimental value and the calculated value, the expression of the ZPR transfer function is proved, whereas the disturbed reactivity is got from the transfer function. (authors)

  7. Optimized Control Rods of the BR2 Reactor

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, E.

    2007-01-01

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

  8. 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.

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

    International Nuclear Information System (INIS)

    Girardin, G.; Chawla, R.

    2011-01-01

    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.

  10. Bouyancy effects on sodium coolant temperature profiles measured in an electrically heated mock-up of a 61-rod breeder reactor blanket assembly

    International Nuclear Information System (INIS)

    Engel, F.C.; Markley, R.A.; Minushkin, B.

    1978-01-01

    The paper describes test results selected to demonstrate the effect of buoyancy on the temperature profiles in a 61-rod electrically heated mock-up of an LMFBR radial blanket assembly. In these assemblies, heat transfer occurs over a wide range of complex operating conditions. The range and complexity of conditions are the result of the steep flux and power gradients which are an inherent feature of the blanket region and the power generation level in an assembly which can vary from 20 to 1100 kW

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  12. Spent fuel strategy for the BR2 reactor

    International Nuclear Information System (INIS)

    Gubel, P.; Collard, G.

    1998-01-01

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

  13. Calculations of fission rate distribution in the core of WWER-1000 mock-up on the LR-0 reactor using alternative methods and comparison with results of measurements

    International Nuclear Information System (INIS)

    Zaritskiy, S.; Kovalishin, A.; Tsvetkov, T.; Rypar, V.; Svadlenkova, M.

    2011-01-01

    General review of experimental and calculation researches on WWER-440 and WWER-1000 mock-ups on the reactor LR-0 was introduced on the twentieth Symposium AER. The experimental core fission rate distribution was obtained by means of gamma-scanning of the fuel pins - 140La single peak (1596 keV) measurements and wide energy range (approximately 600-900 keV) measurements. Altogether from 260 to 500 fuel pins were scanned in different experiments. The measurements were arranged in the middle of the fuel (the active part of pin). Pin-to-pin calculations of the WWER-1000 mock-up core fission rate distribution were performed with several codes: Monte Carlo codes MCU-REA/2 and MCNPX with different nuclear data libraries, diffusion code RADAR (63 energy groups library) and code SVL based on Surface Harmonics Method (69 energy groups). Calculated data are compared with experimental ones. The obtained results allow developing the benchmark for core calculations methodologies, evaluating and validating source reliability for the out-of-core (inside and outside pressure vessel) neutron transport calculations. (Authors)

  14. Weld defects analysis of 60 mm thick SS316L mock-ups of TIG and EB welds by ultrasonic inspection for fusion reactor vacuum vessel applications

    International Nuclear Information System (INIS)

    Buddu, Ramesh Kumar; Shaikh, Shamsuddin; Raole, P.M.; Sarkar, B.

    2015-01-01

    The present paper reports the weld quality inspections carried with 60 mm thick AISI welds of SS316L. The high thickness steel plates requirement is due to the specific applications in case of advanced fusion reactor structural components like vacuum vessel and others. Different kind welds are proposed for the thick plate joints like Tungsten Inert Gas (TIG) welding, Electron beam welding as per stringent conditions (like very low distortions and residual stresses) for the vacuum vessel fabrication. Mock-ups of laboratory scale welds are fabricated by TIG (multi-pass) and EB (double pass) process techniques and different weld quality inspections are carried by different NDT tests. The welds are examined with Liquid penetrant examination to check sub surface cracks/discontinuities towards the defects observation

  15. BR2 reactor core steady state transient modeling

    International Nuclear Information System (INIS)

    Makarenko, A.; Petrova, T.

    2000-01-01

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

  16. Siloette, Siloe mock-up

    International Nuclear Information System (INIS)

    Delcroix, V.; Jeanne, G.; Mitault, G.; Schulhof, P.

    1964-01-01

    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) [fr

  17. Improved production of Br atoms near zero speed by photodissociating laser aligned Br2 molecules.

    Science.gov (United States)

    Deng, L Z; Yin, J P

    2014-10-28

    We theoretically investigated the improvement on the production rate of the decelerated bromine (Br) atoms near zero speed by photodissociating laser aligned Br2 precursors. Adiabatic alignment of Br2 precursors exposed to long laser pulses with duration on the order of nanoseconds was investigated by solving the time-dependent Schrödinger equation. The dynamical fragmentation of adiabatically aligned Br2 precursors was simulated and velocity distribution of the Br atoms produced was analyzed. Our study shows that the larger the degree of the precursor alignment, ⟨cos(2) θ⟩, the higher the production rate of the decelerated Br atoms near zero speed. For Br2 molecules with an initial rotational temperature of ~1 K, a ⟨cos(2) θ⟩ value of ~0.88 can result in an improvement factor of over ~20 on the production rate of the decelerated Br atoms near zero speed, requiring a laser intensity of only ~1 × 10(12) W/cm(2) for alignment.

  18. The zero power reactor SUR and its application

    International Nuclear Information System (INIS)

    Wesser, U.

    1986-01-01

    This low-power reactor, rated nominally at 100 milliwatts, has a cylindrical core of 26 cm in diameter and 24 cm high consisting of U 3 O 8 powder in a polyethylene matrix. The fuel is 20 percent enriched and the critical mass about 700 g. The excess reactivity is about 3 mk. The reactivity is controlled by two cadmium sheets in addition to a back-up system that drops the inner reflector. The reactor has no active cooling system. Personnel costs include a supervisor and an operator. The reactor is used for training in Reactor Theory (including use of a neutron chopper), reactor kinetics, nuclear technology, reactor operations and for doctoral thesis research. (author)

  19. Power Reactor Design at Zero Power; Etudes de Reacteurs de Puissance, au Moyen de Machines de Puissance Zero; Konstruktsiya ehnergeticheskogo reaktora nulevoj moshchnosti; Diseno de Reactores Generadores con Ayuda de Reactores de Potencia Nula

    Energy Technology Data Exchange (ETDEWEB)

    Redman, W. C.; Plumlee, K. E.; Baird, Q. L. [Argonne National Laboratory, Argonne, IL (United States)

    1964-02-15

    Numerous research, central station power, propulsion, isotope production, and test reactor designs have been investigated in Argonne's zero-power reactor facilities, and related exponential and clean critical assemblies have provided basic data. To present a representative account of recent experiments and to demonstrate the wide variety of reactor design information obtainable in low flux systems, the following experimental programmes are reviewed: 1. A study of the properties of thoria-urania fuel in heavy water, with particular attention to the requirements for design of a second core for Argonne's Experimental Boiling Water Reactor; 2. A mock-up of a proposed high flux research reactor to confirm the design calculations, optimize the geometry and estimate the effect of fuel burn-up; 3. A determination of the power distribution patterns and reactivity effect of fuel element flooding for a combined boiling-superheat reactor test; 4. The design of a sodium cooled. U{sup 235} fueled, plutonium producing fast breeder reactor core as a first loading for Argonne's Experimental Breeder Reactor II; and 5. An investigation of the characteristics of a reactor with interacting thermal and fast neutron zones. In the discussion of these programmes, the circumstances which influenced the choice among exponentials, clean criticals, zero-power mock-ups and in situ experiments for the acquisition of the required data are explained, as is the role played by supporting analytical effort. The extent to which reactor design data can be attained before actual operation at power is illustrated by specific examples. Such data include shutdown margin, excess reactivity for operational requirements, temperature coefficients, control and safety rods' effectiveness, reactor kinetics, power production patterns, requirements for start-up source and instrument sensitivity, shielding needs and neutron economy. This review of recent activities in zero-power experimentation reveals the strong

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

    International Nuclear Information System (INIS)

    Joppen, F.

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Ponsard, B.

    2000-01-01

    The BR2 reactor, put into operation in 1963 and refurbished from July 1995 till April 1997, is a 100 MW high-flux Materials Testing Reactor, using 93% 235 U enriched uranium as standard fuel, light water as coolant and beryllium as moderator. The present operating regime consists of five irradiation cycles per year at an operating power between 50 and 70 MW; each cycle is characterized by 21 days operation. In the framework of a 'qualification programme', six 72% 235 U fuel elements fabricated with uranium recovered from the reprocessing of BR2 spent fuel at UKAEA-Dounreay have been successfully irradiated in the period 1994-1995 reaching a maximum mean burnup of 48% without the release of fission products. Since 1998, this type of fuel element is irradiated routinely together with standard 93% 235 U fuel elements in order to optimize the utilization of the available HEU inventory. The purpose of this paper is to present the strategy developed in order to optimize the mixed core management of the BR2 reactor. (author)

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

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, Edgar

    2011-01-01

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

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

    International Nuclear Information System (INIS)

    Raisic, N.

    1962-09-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-05-01

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

  5. Design characteristics of zero power fast reactor Lasta; Osnovne karakteristike brzog reaktora nulte snage Lasta

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M; Stefanovic, D; Pesic, M; Popovic, D; Nikolic, D; Antic, D; Zavaljevski, N [Institut za nuklearne nauke Boris Kidric, Vinca, Beograd (Yugoslavia)

    1987-07-01

    The concept, purpose and preliminary design of a zero power fast reactor LASTA are described. The methods of computing the reactor core parameters and reactor kinetics are presented with the basic calculated results and analysis for one selected LASTA configuration. The nominal parameters are determined according to the selected reactor safety criteria and results of calculations. Important aspects related to the overall safety are examined in detail. (author)

  6. Control and instrumentation system of the Zero Power Reactor at IEA, Sao Paulo (Brazil)

    International Nuclear Information System (INIS)

    Peluso, M.A.V.; Matsuda, K.; Hukai, R.

    1974-01-01

    The control and instrumentation system of the Zero Power Reactor at the IEA (Institute of Atomic Energy - Sao Paulo, Brazil) is described. Technical specifications of the main items of equipment are presented in a general way. Information is also given on the connection between the system described and the electrical supply system of the IEA reactor physics laboratory [pt

  7. The state of the art report on the fabrication of FW blanket for the fusion reactor and mock-up development in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Whan; Baek, Jong Hyuk; Park, Jeong Yong; Kim, Hyun Gil; Jeong, Yong Hwan

    2004-08-01

    Blanket-shield system in ITER is the component where it directly is faced with high-heat plasma. Function of blanket is to sustain extremely high temperature environment as well as to remove heat flux generated its surface. It mainly consists of plasma facing part, heat sinking part and structural part. Plasma facing part is made of armour materials such as beryllium, tungsten and carbon fiber composite. Heat sinking part is made of copper alloy to maximize heat transfer into flowing coolant inside of blanket. Structural material is used in 316LN stainless steel. As joining such dissimilar materials emerged as an issue, many developed countries have spurred the development of joint technology. This technical report was focused on the activities of EU regarding joining beryllium, copper and stainless steel. EU have adopted to Hot Isostatic Pressing (HIP) to join beryllium, copper and stainless steel. Although brazing process is not actively investigated compared as HIP, it still investigated in some countries to support HIP. Fabrication of mock-up is accomplished by CEA in France to finish small scale mock-up in 1996, medium and large scale mock-up in 1997. In recent, FRAMATOME in EU has focused on manufacturing prototype used for ITER.

  8. The state of the art report on the fabrication of FW blanket for the fusion reactor and mock-up development in Europe

    International Nuclear Information System (INIS)

    Kim, Jun Whan; Baek, Jong Hyuk; Park, Jeong Yong; Kim, Hyun Gil; Jeong, Yong Hwan

    2004-08-01

    Blanket-shield system in ITER is the component where it directly is faced with high-heat plasma. Function of blanket is to sustain extremely high temperature environment as well as to remove heat flux generated its surface. It mainly consists of plasma facing part, heat sinking part and structural part. Plasma facing part is made of armour materials such as beryllium, tungsten and carbon fiber composite. Heat sinking part is made of copper alloy to maximize heat transfer into flowing coolant inside of blanket. Structural material is used in 316LN stainless steel. As joining such dissimilar materials emerged as an issue, many developed countries have spurred the development of joint technology. This technical report was focused on the activities of EU regarding joining beryllium, copper and stainless steel. EU have adopted to Hot Isostatic Pressing (HIP) to join beryllium, copper and stainless steel. Although brazing process is not actively investigated compared as HIP, it still investigated in some countries to support HIP. Fabrication of mock-up is accomplished by CEA in France to finish small scale mock-up in 1996, medium and large scale mock-up in 1997. In recent, FRAMATOME in EU has focused on manufacturing prototype used for ITER

  9. Assessment of the linear power level in fuel rods irradiated in the CALLISTO loop in the high flux materials testing reactor BR2

    International Nuclear Information System (INIS)

    Malambu, E.; Raedt, Ch. de; Weber, M.

    1999-01-01

    The pressurized light-water-cooled testing facility CALLISTO was designed to test the behaviour of advanced fuel rods (UO 2 or MOX, possibly with burnable poisons) under conditions representative of actual LWRs up to high burn-up rates. The accurate determination of the fission powers in each of the nine rods, and hence of the burn-up values, is carried out according to a rather elaborate procedure. (author)

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

    International Nuclear Information System (INIS)

    Intsiful, J.D.K.; Akaho, E.H.K.; Tetteh, G.K.

    1997-12-01

    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)

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

    Science.gov (United States)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    Lin Shenghuo; Luo Zhanglin; Su Zhuting

    1994-10-01

    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 U 3 O 8 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

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

    International Nuclear Information System (INIS)

    Peluso, M.A.V.

    1978-01-01

    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) [pt

  14. Characterization of LWR fuel rod irradiations with power transients in the BR2 reflector

    International Nuclear Information System (INIS)

    Ponsard, B.; Bodart, S.; Meer, K. van der; Raedt, C. de

    1996-01-01

    Fuel rod irradiations in reflector positions of the materials testing reactor BR2 are becoming increasingly important. A typical example is that of irradiation devices containing single LWR fuel rods, to be tested in the framework of a new international fuel investigation and development programme. Some of the irradiations will comprise power transients with central fuel melting (at 2800 deg. C), the power increase being obtained by decreasing the pressure in a He-3 neutron absorbing screen and/or by varying the BR2 reactor operating power. A total power variation by a factor of at least 2.5 in the fuel rod irradiated could thus be achieved. In some of the rods, central temperature measurements (up to 2000 deg. C) will be carried out. Both fresh and pre-irradiated fuel rods are concerned in the programme. For these irradiations, the accurate knowledge of the neutron-induced fission heating and of the gamma heating is required, as one of the purposes of the programme consists in establishing the correlation among the thermal conductivity, the burn-up and the irradiation temperature. Calibration work among various measuring methods and between measurements and one- and two-dimensional calculations is being pursued. (author). 10 refs, 15 figs, 3 tabs

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

    International Nuclear Information System (INIS)

    Ferreira, Antonio Carlos de Almeida

    1974-01-01

    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)

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

    International Nuclear Information System (INIS)

    Yao Zewu; Ji Huaxiang; Chen Zhicheng; Yao Zhiquan; Chen Chen; Li Yuwen

    1995-01-01

    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

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

    International Nuclear Information System (INIS)

    Martins, F.R.

    1992-01-01

    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)

  18. Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'

    International Nuclear Information System (INIS)

    Novelli, A.

    1982-01-01

    The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer. (author)

  19. Determination of spatially dependent transfer function of zero power reactor by using pseudo-random incentive

    International Nuclear Information System (INIS)

    Kostic, Lj.

    1973-01-01

    Specially constructed fast reactivity oscillator was stimulating the zero power reactor by a stimulus which caused pseudo-random reactivity changes. Measuring system included stochastic oscillator BCR-1 supplied by pseudo-random pulses from noise generator GBS-16, instrumental tape-recorder, system for data acquisition and digital computer ZUSE-Z-23. For measuring the spatially dependent transfer function, reactor response was measured at a number of different positions of stochastic oscillator and ionization chamber. In order to keep the reactor system linear, experiment was limited to small reactivity fluctuations. Experimental results were compared to theoretical ones

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    Koonen, E.; Gubel, P.

    1994-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Koonen, E; Gubel, P [BR2 Department, Belgian Nuclear Research Center, CEN/SCK, Mol (Belgium)

    1993-07-01

    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)

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

    International Nuclear Information System (INIS)

    Koonen, E.; Gubel, P.

    1993-01-01

    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)

  4. Core concept of fast power reactor with zero sodium void reactivity

    International Nuclear Information System (INIS)

    Matveev, V.I.; Chebeskov, A.N.; Krivitsky, I.Y.

    1991-01-01

    The paper presents a core concept of BN-800 - type fast power reactor with zero sodium void reactivity (SVR). Consideration is given to the layout-and some design features of such a core. Some considerations on the determination of the required SVR value as one of the fast reactor safety criteria in accidents with coolant boiling are presented. Some methodical considerations an the development of calculation models that give a correct description of the new core features are stated. The results of the integral SVR calculation studies are included. reactivity excursions under different scenarios of sodium boiling are estimated, some corrections into the calculated SVR value are discussed. (author)

  5. Accident at the zero power reactor which happened on October 15 1958

    International Nuclear Information System (INIS)

    Savic, P.

    1959-01-01

    During an experiment on the zero power heavy water reactor with natural uranium fuel in the Boris Kidric Institute of Nuclear Sciences, the reactor escaped control. Six staff members in the immediate surrounding of the bare assembly were exposed to high neutron and ionising irradiation. Other two employees who were at some bigger distance were exposed to doses higher than permitted. This paper deals with the circumstances that caused the accident, status of the dosimetry, control and alarm systems. Individual exposure doses were estimated according to the calculated neutron flux values obtained from measuring the activities of personal belongings made of gold and copper as well as radioactive phosphorous from urine

  6. Blanket Manufacturing Technologies : Thermomechanical Tests on HCLL Blanket Mocks Up

    International Nuclear Information System (INIS)

    Laffont, G.; Cachon, L.; Taraud, P.; Challet, F.; Rampal, G.; Salavy, J.F.

    2006-01-01

    In the Helium Cooled Lithium Lead (HCLL) Blanket concept, the lithium lead plays the double role of breeder and multiplier material, and the helium is used as coolant. The HCCL Blanket Module are made of steel boxes reinforced by stiffening plates. These stiffening plates form cells in which the breeder is slowly flowing. The power deposited in the breeder material is recovered by the breeder cooling units constituted by 5 parallel cooling plates. All the structures such as first wall, stiffening and cooling plates are cooled by helium. Due to the complex geometry of these parts and the high level of pressure and temperature loading, thermo-mechanical phenomena expected in the '' HCLL blanket concept '' have motivated the present study. The aim of this study, carried out in the frame of EFDA Work program, is to validate the manufacturing technologies of HCLL blanket module by testing small scale mock-up under breeder blanket representative operating conditions.The first step of this experimental program is the design and manufacturing of a relevant test section in the DIADEMO facility, which was recently upgraded with an He cooling loop (pressure of 80 bar, maximum temperature of 500 o C,flow rate of 30 g/s) taking the opportunity of synergies with the gas-cooled fission reactor R-and-D program. The second step will deal with the thermo-mechanical tests. This paper focuses on the program made to support the cooling plate mock up tests which will be carried out on the DIADEMO facility (CEA) by thermo-mechanical calculations in order to define the relevant test conditions and the experimental parameters to be monitored. (author)

  7. Zero energy reactor 'RB'

    Energy Technology Data Exchange (ETDEWEB)

    Popovic, D; Takac, S; Markovic, H; Raisic, N; Zdravkovic, Z; Radanovic, Lj [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1959-03-15

    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)

  8. 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....

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

    Energy Technology Data Exchange (ETDEWEB)

    Schuh, N J.H.

    1966-12-15

    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.

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

    International Nuclear Information System (INIS)

    Schuh, N.J.H.

    1966-12-01

    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

  11. The bent crystal diffraction spectrometer at the BR2 reactor in Mol

    Science.gov (United States)

    Kaerts, E.; Jacobs, L.; Vandenput, G.; Van Assche, P. H. M.

    1988-05-01

    The DuMond-type bent crystal diffraction spectrometer installed at the BR2 reactor in Mol is presented. The spectrometer is mainly designed to study nuclear γ-transitions following thermal neutron capture. It covers the energy interval 25 ≦ Eγ ≦ 1500 keV. Instead of the traditionally used quartz crystals, a highly perfect silicium crystal is chosen as analysing crystal. Diffraction occurs from the (220) plane. The "quasi-mosaic" width, introduced by bending the crystal, is as small as 0.2″. The integrated reflecting power R of the bent crystal stays constant up to 1.5 MeV in first, 680 keV in second and 300 keV in third diffraction order. For higher photon energies, only an E-1 energy dependence is observed in second and third diffraction order. Consequently, besides improving the energy resolution, the use of these silicium crystals substantially increases the spectrometer efficiency and extends the high energy limit of bent crystal diffraction spectrometers. The diffraction angles are measured with a symmetrical interferometer system which covers an angular range of -6° to +6° with a precision of about 0.01″. Minimum diffraction line widths of 0.9″ have been measured, corresponding to an energy resolution ΔE = 1.35 × 10 -6E2n-1 keV -1. The dominant contribution to the observed line widths arises from the finite extent of the source.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    Nicol, R.G.; Parrott, J.R.; Krichinsky, A.M.; Box, W.D.; Martin, C.W.; Whitson, W.R.

    1982-05-01

    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 233 U from rejected elements. This report describes the steps associated with conversion of purified uranyl nitrate (as solution) to U 3 O 8 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. Structural characteristics of a graphite moderated critical assembly for a Zero Power reactor at IEA (Brazil)

    International Nuclear Information System (INIS)

    Almeida Ferreira, A.C. de; Hukai, R.Y.

    1975-01-01

    The structural characteristics of a graphite moderated core of a critical assembly to be installed in the Zero Power Reactor of IEA have been defined. These characteristics are the graphite block dimensions, the number and dimensions of the holes in the graphite, the pitch, the dimensions of the sticks of fuel and graphite to be inserted in the holes, and the mechanical reproducibility of the system. The composition of the fuel and moderator sticks were also defined. The main boundary conditions were the range of the relation C/U and C/TH used in commercial HTGR and the neutronics homogeneity

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  17. Status of the Digital Mock-up System for the dismantling of the nuclear facilities

    International Nuclear Information System (INIS)

    Park, Hee Seoung; Kim, S. K.; Lee, K. W.; Oh, W. J.

    2004-12-01

    The database system have already developed is impossible to solve a quantitative evaluation about a various situation from the dismantle activities of the reactor had contaminated with radioactivity. To satisfy the requirements for safety and economical efficiency among a major decommissioning technologies, it need a system that can evaluate and estimate dismantling scheduling, amount of radioactive waste being dismantled, and decommissioning cost. We have review and analyzed status of the digital mock-up system to get a technical guide because we have no experience establishment of one relation to dismantling of research reactor and nuclear power plant

  18. Qualification of high density aluminide fuels for the BR2 reactor

    International Nuclear Information System (INIS)

    Beeckmans de West-Meerbeeck, Andre; Gubel, Pol; Ponsard, Bernard; Pin, Thomas; Falgoux, Jean Louis

    2005-01-01

    The BR2 operation still relies on the use of 90..93% enriched HEU aluminide fuel. The availability of a limited batch of 73% enriched HEU from reprocessed BR2 uranium in Dounreay justified 10 years ago the qualification and use of this material. After some preliminary test irradiations, various batches of fuel elements were fabricated by the UKAEA-Dounreay and successfully irradiated. Due to their lower 235 U content (0.050 g 235 U/cm 2 ), these elements were always irradiated together with standard 90...93% HEU fuel elements. A mixed-core strategy was developed at this occasion for an optimal utilization, and was reported during the 4th RRFM conference (March 19-21, 2000, Colmar, France). The availability of a new batch of fresh 73% HEU material was the occasion, a few years ago, to initiate the development, fabrication and qualification of a new high density fuel element. An order was placed with CERCA to assess the optimal fabrication methods and tooling required to meet as far as possible the existing BR2 standard specifications and 235 U content (0.060 g 235 U/cm 2 ). This development phase has been already reported during the 7th RRFM conference (March 9-12, 2003, Aix-en-Provence, France). Afterwards, six lead test fuel elements were ordered for qualification by irradiation. The neutronic properties of the fuel elements were adjusted and optimized. After a short summary of the main results of the development program, this paper describes the nuclear characteristics of the high density fuel elements and comments on the nuclear follow-up of the lead test fuel elements during their irradiation for five cycles in the BR2 reactor and the return of experience for CERCA. (author)

  19. Measurement and analysis of breeding index in the first mock-up core (XXII-1(65V)) for water-cooled breeder reactor at FCA. Contract research

    International Nuclear Information System (INIS)

    Fukushima, Masahiro; Okajima, Shigeaki; Andoh, Masaki; Yamane, Tsuyoshi; Kataoka, Masaharu

    2005-03-01

    Measurements and analyses of breeding index were performed in the FCA-XXII-1(65V) core simulating reduced-moderation lightwater reactor (RMWR) with void fraction 65% of moderator at Fast Critical Assembly (FCA). The measurement of the reaction rate ratio of 238 U capture to 235 U fission (C8/F5) was made by a foil activation technique using depleted and enriched uranium foils, and the reaction rate ratios of 239 Pu fission to 235 U fission (F9/F5) and 238 U fission to 235 U fission (F8/F5) were measured using absolutely calibrated fission chambers. Cell averaged reaction correction factors were derived by the Monte Carlo code (MVP) calculations with modeling the forms and positions of fission chambers and foils. Consequently, cell averaged reaction rate ratios were determined to be C8/F5 of 0.0916±1.4%, F9/F5 of 0.759±1.2% and F8/F5 of 0.0201±0.9%. Therefore, breeding index of C8/F9 was determined to be 0.121 ± 1.8%. The analyses were made by using the JFS-3-J3.2R group constant set which is generated from the JENDL-3.2 nuclear data library. The effective cross sections were calculated by the standard cell calculation code for fast reactor, SLAROM. Three-dimensional diffusion calculations by the CITATION code with 70-group structure have been performed to estimate the reaction rate ratios in the core center. Here, effective cross sections of fuel cells in the core center were obtained using the PEACO-X code with an ultra-fine group structure to consider self-shielding effect in the resonance energy range. Calculation to experiment ratios (C/E) of F9/F5 and F8/F5 were 1.02 and 1.03, respectively. These calculated values slightly overestimate the experimental one. The calculated value of C8/F5 overestimates the experimental one by about 6%. Consequently, the C/E value of C8/F9 was 1.03. The calculated value slightly overestimates the experimental one. The calculation code system developed for the thermal reactor, SRAC code system, was also used to analyze the

  20. In-pile testing of ITER first wall mock-ups at relevant thermal loading conditions

    Science.gov (United States)

    Litunovsky, N.; Gervash, A.; Lorenzetto, P.; Mazul, I.; Melder, R.

    2009-04-01

    The paper describes the experimental technique and preliminary results of thermal fatigue testing of ITER first wall (FW) water-cooled mock-ups inside the core of the RBT-6 experimental fission reactor (RIAR, Dimitrovgrad, Russia). This experiment has provided simultaneous effect of neutron fluence and thermal cycling damages on the mock-ups. A PC-controlled high-temperature graphite ohmic heater was applied to provide cyclic thermal load onto the mock-ups surface. This experiment lasted for 309 effective irradiation days with a final damage level (CuCrZr) of 1 dpa in the mock-ups. About 3700 thermal cycles with a heat flux of 0.4-0.5 MW/m 2 onto the mock-ups were realized before the heater fails. Then, irradiation was continued in a non-cycling mode.

  1. The making of a mock-up

    DEFF Research Database (Denmark)

    Rosenqvist, Tanja Schultz; Heimdal, Elisabeth Jacobsen

    2011-01-01

    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...

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

    International Nuclear Information System (INIS)

    Hashimoto, Toshiyuki; Takatsu, Hideyuki; Sato, Satoshi

    1994-07-01

    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)

  3. Use of highly enriched uranium in the material testing reactor BR2

    International Nuclear Information System (INIS)

    Beeckmans de West-Meerbeeck, A.

    1979-05-01

    In the material testing reactor BR2, the use of highly enriched uranium is determined by the consideration of the fast, epithermal and thermal neutron flux effectively available for the experimental devices. The choice of the core configuration is defined by combining the localisation of the experimental devices and of fuel elements of various burnup, such as to satisfy the irradiation conditions of the experimental load, compatible with an economic use of the fuel elements and safe operation of the reactor. Taking into account the present manufacturing technology for MTR fuels (37 Wt % uranium density in the fuel meat) the highly enriched uranium cannot be avoided; if higher concentration of uranium could be realised by some new manufacturing technology, the 235 U density of fuel elements at elimination should be kept at the required level and the enrichment could be reduced accordingly

  4. Use of highly enriched uranium in the material testing reactor BR2

    International Nuclear Information System (INIS)

    Beeckmans de West-Meerbeeck, A.

    1979-05-01

    In the material testing reactor BR2, the use of highly enriched uranium is determined by the consideration of the fast, epithermal and thermal neutron flux effectively available for the experimental devices. The choice of the core configuration is defined by combining the localisation of the experimental devices and of fuel elements of various burnup, such as to satisfy the irradiation conditions of the experimental load, compatible with an economic use of the fuel elements and safe operation of the reactor. Taking into account the present manufacturing technology for MTR fuels (37 Wt % uranium density in the fuel meat) the highly enriched uranium cannot be avoided: if higher concentration of uranium could be realised by some new manufacturing technology, the 235 U density of fuel elements at elimination should be kept at the required level and the enrichment could be reduced accordingly. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Bernander, G

    1967-07-15

    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.

  6. How tangible mock-ups support design collaboration

    DEFF Research Database (Denmark)

    Brandt, Eva

    2007-01-01

    This paper is a contribution to a more conscious use of tangible mock-ups in collaborative design processes. It describes a design team's use of mock-ups in a series of workshops involving potential customers and users. Focus is primarily on the use of three-dimensional design mock-ups and how...... 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...... things for different participants whereas the challenge for the design team was to find boundaries upon which everybody could agree. The level of details represented in a mock-up affected the communication so that a mock-up with few details evoked different issues whereas a very detailed mock-up evoked...

  7. Accident at the zero power reactor which happened on October 15 1958; Sur l'accident avec le reacteur de puissance zero du 15 octobre 1958

    Energy Technology Data Exchange (ETDEWEB)

    Savic, P [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1959-03-15

    During an experiment on the zero power heavy water reactor with natural uranium fuel in the Boris Kidric Institute of Nuclear Sciences, the reactor escaped control. Six staff members in the immediate surrounding of the bare assembly were exposed to high neutron and ionising irradiation. Other two employees who were at some bigger distance were exposed to doses higher than permitted. This paper deals with the circumstances that caused the accident, status of the dosimetry, control and alarm systems. Individual exposure doses were estimated according to the calculated neutron flux values obtained from measuring the activities of personal belongings made of gold and copper as well as radioactive phosphorous from urine.

  8. Current Mode Neutron Noise Measurements in the Zero Power Reactor CROCUS

    Science.gov (United States)

    Pakari, O.; Lamirand, V.; Perret, G.; Braun, L.; Frajtag, P.; Pautz, A.

    2018-01-01

    The present article is an overview of developments and results regarding neutron noise measurements in current mode at the CROCUS zero power facility. Neutron noise measurements offer a non-invasive method to determine kinetic reactor parameters such as the prompt decay constant at criticality α = βeff / λ, the effective delayed neutron fraction βeff, and the mean generation time λ for code validation efforts. At higher detection rates, i.e. above 2×104 cps in the used configuration at 0.1 W, the previously employed pulse charge amplification electronics with BF3 detectors yielded erroneous results due to dead time effects. Future experimental needs call for higher sensitivity in detectors, higher detection rates or higher reactor powers, and thus a generally more versatile measurement system. We, therefore, explored detectors operated with current mode acquisition electronics to accommodate the need. We approached the matter in two ways: 1) By using the two compensated 10B-coated ionization chambers available in CROCUS as operational monitors. The compensated current signal of these chambers was extracted from coremonitoring output channels. 2) By developing a new current mode amplification station to be used with other available detectors in core. Characteristics and first noise measurements of the new current system are presented. We implemented post-processing of the current signals from 1)and 2) with the APSD/CPSD method to determine α. At two critical states (0.5 and 1.5 W), using the 10B ionization chambers and their CPSD estimate, the prompt decay constant was measured after 1.5 hours to be α=(156.9 ± 4.3) s-1 (1σ). This result is within 1σ of statistical uncertainties of previous experiments and MCNPv5-1.6 predictions using the ENDF/B-7.1 library. The newsystem connected to a CFUL01 fission chamber using the APSDestimate at 100 mW after 33 min yielded α = (160.8 ± 6.3) s-1, also within 1σ agreement. The improvements to previous neutron noise

  9. Safety challenges encountered during the operating life of the almost 40 year old research reactor BR2

    International Nuclear Information System (INIS)

    Koonen, E.; Joppen, F.; Gubel, P.

    2001-01-01

    The BR2 reactor is one of the major MTR-type research reactors in the world. Its operation started in the early 1960's. Two major refurbishment operations have been carried out since then. Several safety reassessments were carried out over the years in order to keep the safety level in line with modern standards and to enhance operational safety. This paper gives an overview of the safety challenges encountered over the years and how those were met. (author)

  10. Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'

    Energy Technology Data Exchange (ETDEWEB)

    Novelli, A. (Politecnico di Milano (Italy). Centro Studi Nucleari E. Fermi)

    1982-01-01

    The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer.

  11. High heat flux test of tungsten brazed mock-ups developed for KSTAR divertor

    Energy Technology Data Exchange (ETDEWEB)

    Song, J.H. [National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, K.M., E-mail: kyungmin@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Hong, S.H.; Kim, H.T.; Park, S.H.; Park, H.K.; Ahn, H.J. [National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, S.K.; Lee, D.W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    The tungsten (W) brazed flat type mock-up which consists of W, OFHC-Cu (oxygen-free high conductive copper) and CuCrZr alloy has been designed for KSTAR divertor in preparation for KSTAR upgrade with 17 MW heating power. For verification of the W brazed mock-up, the high heat flux test is performed at KoHLT-EB (Korea High Heat Load Test Facility-Electron Beam) in KAERI (Korea Atomic Energy Research Institute). Three mock-ups are tested for several thousand thermal cycles with absorbed heat flux up to 5 MW/m{sup 2} for 20 s duration. There is no evidence of the failure at the bonding joints of all mock-ups after HHF test. Finite element analysis (FEA) is performed to interpret the result of the test. As a result, it is considered that the local area in the water is in the subcooled boiling regime.

  12. 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.

  13. 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)

    Martin, H.R.; Keil, W.M.; Pezzi, N.

    1991-01-01

    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) [es

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

    International Nuclear Information System (INIS)

    Vermeeren, L.

    2009-01-01

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

  15. Self-Powered Neutron Detector Qualification for Absolute On-Line In-Pile Neutron Flux Measurements in BR2

    Science.gov (United States)

    Vermeeren, L.; Wéber, M.

    2003-06-01

    A set of ten Self-Powered Neutron Detectors with Co, Rh and Ag emitters has been irradiated in several channels of the BR2 research reactor at SCK•CEN aiming at a comparison of their performance as thermal neutron flux detectors under various conditions. To allow for a correct interpretation of their signals, all detector sensitivity contributions (prompt and delayed) were calculated using a dedicated Monte Carlo model. The various contributions were also measured separately; the agreement between calculated and experimental data, including data from activation dosimetry, was excellent. Detailed neutron flux profiles were obtained from the SPND data, after correction for the finite detector lengths and for the slow response of delayed SPNDs.

  16. 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. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Kalcheva, S [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Sikik, E [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-09-01

    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. 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. A feasibility study for the conversion of the BR2 reactor from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel was previously performed to verify it can operate safely at the same maximum nominal steady-state heat flux. An assessment was also performed to quantify the heat fluxes at which the onset of flow instability and critical heat flux occur for each fuel type. This document updates and expands these results for the current representative core configuration (assuming a fresh beryllium matrix) by evaluating the onset of nucleate boiling (ONB), onset of fully developed nucleate boiling (FDNB), onset of flow instability (OFI) and critical heat flux (CHF).

  17. SoLid: Search for Oscillations with Lithium-6 Detector at the SCK-CEN BR2 reactor

    Science.gov (United States)

    Ban, G.; Beaumont, W.; Buhour, J. M.; Coupé, B.; Cucoanes, A. S.; D'Hondt, J.; Durand, D.; Fallot, M.; Fresneau, S.; Giot, L.; Guillon, B.; Guilloux, G.; Janssen, X.; Kalcheva, S.; Koonen, E.; Labare, M.; Moortgat, C.; Pronost, G.; Raes, L.; Ryckbosch, D.; Ryder, N.; Shitov, Y.; Vacheret, A.; Van Mulders, P.; Van Remortel, N.; Weber, A.; Yermia, F.

    2016-04-01

    Sterile neutrinos have been considered as a possible explanation for the recent reactor and Gallium anomalies arising from reanalysis of reactor flux and calibration data of previous neutrino experiments. A way to test this hypothesis is to look for distortions of the anti-neutrino energy caused by oscillation from active to sterile neutrino at close stand-off (˜ 6- 8m) of a compact reactor core. Due to the low rate of anti-neutrino interactions the main challenge in such measurement is to control the high level of gamma rays and neutron background. The SoLid experiment is a proposal to search for active-to-sterile anti-neutrino oscillation at very short baseline of the SCK•CEN BR2 research reactor. This experiment uses a novel approach to detect anti-neutrino with a highly segmented detector based on Lithium-6. With the combination of high granularity, high neutron-gamma discrimination using 6LiF:ZnS(Ag) and precise localization of the Inverse Beta Decay products, a better experimental sensitivity can be achieved compared to other state-of-the-art technology. This compact system requires minimum passive shielding allowing for very close stand off to the reactor. The experimental set up of the SoLid experiment and the BR2 reactor will be presented. The new principle of neutrino detection and the detector design with expected performance will be described. The expected sensitivity to new oscillations of the SoLid detector as well as the first measurements made with the 8 kg prototype detector deployed at the BR2 reactor in 2013-2014 will be reported.

  18. Adequacy of the analysis of mock-up control rod experiment with FCA

    International Nuclear Information System (INIS)

    Mizoo, Nobutatsu; Nakano, Masafumi

    1977-07-01

    A method of numerical analysis has been investigated for the mock-up control rod experiment of FCA VII-1 assembly constructed as the engineering mock-up of prototype fast breeder reactor MONJU. The results of criticality and B 4 C mock-up control rod worths analysis for the assembly are described in comparison with the experimental ones. The tendency of the C/E value with 10 B enrichment and the interaction effect of the multiple rods array was also examined. Reactivities and the mock-up rods worths were obtained with the X-Y geometry six groups diffusion theory. Twelve kinds of the mock-up rods with different 10 B contents and/or enrichments were used in the experiment; effective cross-sections are provided for each rod by calculation using the collision probability method. Criticality of VII-1 90Z assembly is underestimated for 3 reference critical configurations, ranging from -0.65%Δk/k to -0.77%Δk/k. The C/E values at core center for 12 kinds of B 4 C mock-up rods range from 1.03 to 1.09. The overestimate of the rod worth increases with macroscopic absorption cross-section of the rod region. The C/E values for 24 different arrays of the mock-up rods ranging from single rod to five rods lie between 1.04 and 1.08. The C/E value tends to decrease with increase in the number of rods inserted, the values for five rods arrays being about 4% lower than those for single rod arrays. The calculated interaction effects of the multiple rods arrays are slightly more negative than the experimental ones. (auth.)

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

    International Nuclear Information System (INIS)

    Koonen, E.; Beeckmans, A.; Gubel, P.

    1998-01-01

    The standard BR2 fuel element contains 400 g 235 U under the form of UAl x with burnable absorbers homogeneously mixed into the fuel meat. The uranium is highly enriched with a density of ∼1.30 g U/cm 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. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-01

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

  1. First results of the deployment of a SoLid detector module at the SCK•CEN BR2 reactor

    Science.gov (United States)

    Ryder, N.

    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 three years. In this talk the novel detector design is explained and initial detector performance results from the module level deployment are presented along with an estimation of the physics reach of the next phase.

  2. Measurement of zero power reactor dynamic response by cross correlation method; Merenje dinamickog odziva reaktora nulte snage kros korelacionom metodom

    Energy Technology Data Exchange (ETDEWEB)

    Kostic, Lj; Petrovic, M [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1969-07-15

    Pulse response is comprehensive description of linear system dynamics. In this paper, cross correlation method was used for measuring the response of zero power reactor. Reactor system was perturbed by pseudo-random signal, which was cross correlated with the reactor signal responding to this perturbation on the digital ZUSE Z-23 computer. Cross-correlation functions were measured for different positions of stochastic oscillator and ionization chamber in the critical system. From numerical processing of performed experimental data, it was concluded that a more powerful faster computer would be needed for processing statistical experiments. In that case it would be possible to obtain information about spatial effects in the reactor and propagation of neutron waves in the multiplication medium. Impulsni odziv je potpuni opis dinamike linearnog sistema. Za merenje impulsnog odziva nultog reaktora, u ovom radu, koriscena je kros korelaciona metoda. Reaktorski sistem je perturbovan pseudoslucajnim signalom, koji je u digitalnom racunaru ZUSE Z-23 kroskorelisan sa signalom odziva reaktora na ove perturbacije. Merene su kroskorelacione funkcije za razlicite polozaje stohastickog oscilatora i jonizacione komore u kriticnom sistemu. Iz numericki obradjivanih eksperimenta namece se kao zakljucak da bi za obradu statistickih eksperimenata kod nultih reaktora bio potreban racunar veceg kapaciteta i brzine. U tom slucaju bi se iz ovako postavljenog eksperimenta moglo doci i do informacija o prostornim efektima u reaktoru i prostiranju neutronskih talasa kroz multiplikativnu sredinu. (author)

  3. State of the art on nuclear heating measurement methods and expected improvements in zero power research reactors

    International Nuclear Information System (INIS)

    Le Guillou, M.; Gruel, A.; Destouches, C.; Blaise, P.

    2017-01-01

    The paper focuses on the recent methodological advances suitable for nuclear heating measurements in zero power research reactors. This bibliographical work is part of an experimental approach currently in progress at CEA Cadarache, aiming at optimizing photon heating measurements in low-power research reactors. It provides an overview of the application fields of the most widely used detectors, namely thermoluminescent dosimeters (TLDs) and optically stimulated luminescent dosimeters. Starting from the methodology currently implemented at CEA, the expected improvements relate to the experimental determination of the neutron component, which is a key point conditioning the accuracy of photon heating measurements in mixed n-γ field. A recently developed methodology based on the use of "7Li and "6Li-enriched TLDs, pre-calibrated both in photon and neutron fields, is a promising approach to de-convolute the 2 components of nuclear heating. We also investigate the different methods of optical fiber dosimetry, with a view to assess the feasibility of online photon heating measurements, whose primary benefit is to overcome constraints related to the withdrawal of dosimeters from the reactor immediately after irradiation. Moreover, a fiber-using setup could allow measuring the instantaneous dose rate during irradiation, as well as the delayed photon dose after reactor shutdown. Some insights from potential further developments are given. Obviously, any improvement of the technique has to lead to a measurement uncertainty at least equal to that of the currently used methodology (∼5% at 1 σ). (authors)

  4. State of the art on nuclear heating measurement methods and expected improvements in zero power research reactors

    Directory of Open Access Journals (Sweden)

    Le Guillou Mael

    2017-01-01

    Full Text Available The paper focuses on the recent methodological advances suitable for nuclear heating measurements in zero power research reactors. This bibliographical work is part of an experimental approach currently in progress at CEA Cadarache, aiming at optimizing photon heating measurements in low-power research reactors. It provides an overview of the application fields of the most widely used detectors, namely thermoluminescent dosimeters (TLDs and optically stimulated luminescent dosimeters. Starting from the methodology currently implemented at CEA, the expected improvements relate to the experimental determination of the neutron component, which is a key point conditioning the accuracy of photon heating measurements in mixed n–γ field. A recently developed methodology based on the use of 7Li and 6Li-enriched TLDs, precalibrated both in photon and neutron fields, is a promising approach to deconvolute the two components of nuclear heating. We also investigate the different methods of optical fiber dosimetry, with a view to assess the feasibility of online photon heating measurements, whose primary benefit is to overcome constraints related to the withdrawal of dosimeters from the reactor immediately after irradiation. Moreover, a fibered setup could allow measuring the instantaneous dose rate during irradiation, as well as the delayed photon dose after reactor shutdown. Some insights from potential further developments are given. Obviously, any improvement of the technique has to lead to a measurement uncertainty at least equal to that of the currently used methodology (∼5% at 1σ.

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

    International Nuclear Information System (INIS)

    W. C. Adams

    2007-01-01

    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

  6. Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor; Ispitivanje reaktorskih parametara na kriticnim sistemima, I faza: Izvestaj o sigurnosti reaktora nulte snage RB

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1962-09-15

    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.

  7. Investigating heavy water zero power reactors with a new core configuration based on experiment and calculation results

    Energy Technology Data Exchange (ETDEWEB)

    Nasrazadani, Zahra; Salimi, Raana; Askari, Afrooz; Khorsandi, Jamshid; Mirvakili, Mohammad; Mashayekh, Mohammad [Reactor Research School, Nuclear Science and Technology Research Institute, Atomic Energy Organization of Iran, Esfahan (Iran, Islamic Republic of)

    2017-02-15

    The heavy water zero power reactor (HWZPR), which is a critical assembly with a maximum power of 100 W, can be used in different lattice pitches. The last change of core configuration was from a lattice pitch of 18-20 cm. Based on regulations, prior to the first operation of the reactor, a new core was simulated with MCNP (Monte Carlo N-Particle)-4C and WIMS (Winfrith Improved Multigroup Scheme)-CITATON codes. To investigate the criticality of this core, the effective multiplication factor (Keff) versus heavy water level, and the critical water level were calculated. Then, for safety considerations, the reactivity worth of D2O, the reactivity worth of safety and control rods, and temperature reactivity coefficients for the fuel and the moderator, were calculated. The results show that the relevant criteria in the safety analysis report were satisfied in the new core. Therefore, with the permission of the reactor safety committee, the first criticality operation was conducted, and important physical parameters were measured experimentally. The results were compared with the corresponding values in the original core.

  8. SoLid: An innovative anti-neutrino detector for searching oscillations at the SCK•CEN BR2 reactor

    Science.gov (United States)

    Abreu, Yamiel; SoLid Collaboration

    2017-02-01

    The SoLid experiment intends to search for active-to-sterile anti-neutrino oscillations at a very short baseline from the SCK•CEN BR2 research reactor (Mol, Belgium). A novel detector approach to measure reactor anti-neutrinos was developed based on an innovative sandwich of composite polyvinyl-toluene and 6LiF:ZnS(Ag) scintillators. The system is highly segmented and read out by a network of wavelength shifting fibers and SiPM. High experimental sensitivity can be achieved compared to other standard technologies thanks to the combination of high granularity, good neutron-gamma discrimination using 6LiF:ZnS(Ag) scintillator and precise localisation of the Inverse Beta Decay products. This technology can be considered as a new generation of an anti-neutrino detector. This compact system requires limited passive shielding and relies on spatial topology to determine the different classes of backgrounds. We will describe the principle of detection and the detector design. Particular focus on the neutron discrimination will be made, as well as on the capability to use cosmic muons for channel equalisation and energy calibration. The performance of the first 288 kg SoLid module (SM1), based on the data taken at BR2 from February to September 2015, will be presented. We will conclude with the next phase, which will start in 2016, and the future plans of the experiment.

  9. SoLid: An innovative anti-neutrino detector for searching oscillations at the SCK• CEN BR2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abreu, Yamiel, E-mail: yamiel.abreu@uantwerpen.be

    2017-02-11

    The SoLid experiment intends to search for active-to-sterile anti-neutrino oscillations at a very short baseline from the SCK• CEN BR2 research reactor (Mol, Belgium). A novel detector approach to measure reactor anti-neutrinos was developed based on an innovative sandwich of composite polyvinyl-toluene and {sup 6}LiF:ZnS(Ag) scintillators. The system is highly segmented and read out by a network of wavelength shifting fibers and SiPM. High experimental sensitivity can be achieved compared to other standard technologies thanks to the combination of high granularity, good neutron–gamma discrimination using {sup 6}LiF:ZnS(Ag) scintillator and precise localisation of the Inverse Beta Decay products. This technology can be considered as a new generation of an anti-neutrino detector. This compact system requires limited passive shielding and relies on spatial topology to determine the different classes of backgrounds. We will describe the principle of detection and the detector design. Particular focus on the neutron discrimination will be made, as well as on the capability to use cosmic muons for channel equalisation and energy calibration. The performance of the first 288 kg SoLid module (SM1), based on the data taken at BR2 from February to September 2015, will be presented. We will conclude with the next phase, which will start in 2016, and the future plans of the experiment.

  10. Experimental investigation of the IFMIF target mock-up

    International Nuclear Information System (INIS)

    Loginov, N.; Mikheyev, A.; Morozov, V.; Aksenov, Yu.; Arnol'dov, M.; Berensky, L.; Fedotovsky, V.; Chernov, V.; Nakamura, H.

    2009-01-01

    The international fusion materials irradiation facility (IFMIF) lithium neutron target mock-ups have been constructed and tested at water and lithium test facilities in the IPPE of Russia. Jet velocity in both mock-ups was up to 20 m/s. Calculations and experiments showed lithium flow instability at conjunction point of straight and concave sections of the mock-up back wall. Water velocity profile across the mock-up width, jet thickness, and wave height were measured. The significant increase of thickness of both water and lithium jets near the mock-up sidewalls was observed. The influence of shape of the nozzle outlet part on jet stability was investigated. Lithium evaporation from the jet free surface was investigated as well as lithium deposition on vacuum pipe walls of the target mock-up. It was shown that these phenomena are not very critical for the target efficiency. The possibility of lithium denitration down to 2 ppm (at 10 ppm requested) by means of aluminium getter was shown. Two types of cold traps and plug indicators of impurities were tested. The results are presented in the paper.

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

    Energy Technology Data Exchange (ETDEWEB)

    Perret, G.; Pattupara, R. M. [Paul Scherrer Inst., 5232 Villigen (Switzerland); Girardin, G. [Ecole Polytechnique Federale de Lausanne, 1015 Lausanne (Switzerland); Chawla, R. [Paul Scherrer Inst., 5232 Villigen (Switzerland); Ecole Polytechnique Federale de Lausanne, 1015 Lausanne (Switzerland)

    2012-07-01

    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 PuO{sub 2}/UO{sub 2} 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 {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

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

    International Nuclear Information System (INIS)

    Paola Batistoni, P.; Angelone, M.; Bettinali, L.

    2006-01-01

    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 Li 2 CO 3 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 Li 2 CO 3 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

  13. Use of zero power plutonium reactor measurements as a support of criticality prediction for the SNR-300

    International Nuclear Information System (INIS)

    Pilate, S.; de Wouters, R.; Wehmann, U.; Helm, F.; Scholtyssek, W.

    1978-01-01

    Evaluations of criticality measurements performed in various SNEAK and Zero Power Plutonium Reactor (ZPPR) cores are compared. The best available methods of calculations (including transport theory) are used. The ZPPR results support well the trend indicated by the SNEAK evaluations for clean cores and for cores with followers; for cores with absorbers partially inserted, the agreement is only rough. Evaluations of control rod worth measurements are therefore also compared, using the routine method of calculation for SNR-300 (diffusion theory). The control rod worths are largely underestimated in SNEAK (C/E = 0.89), but only slightly underestimated in the ZPPR (C/E = 0.97). The difference in the nature of core fuel (uranium in SNEAK, plutonium in the ZPPR) could be at the origin of this discrepancy

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

    International Nuclear Information System (INIS)

    Zdarek, J.; Brabec, P.; Frybort, O.; Lahodova, Z.; Vit, J.; Stemberk, P.

    2015-01-01

    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

  15. MSR - SPHINX concept program Eros (Experimental zero power Salt reactor SR-0) - The proposed experimental program as a basis for validation of reactor physics methods

    Energy Technology Data Exchange (ETDEWEB)

    Hron, M.; Juricek, V.; Kyncl, J.; Mikisek, M.; Rypar, V. [Nuclear Research Institute Rez plc, Rez (Czech Republic)

    2007-07-01

    The Molten Salt Reactor (MSR) - SPHINX (SPent Hot fuel Incinerator by Neutron fluX) concept solves this principal problem of spent fuel treatment by means of so-called nuclear incineration. It means the burning of fissionable part of its inventory and transmutation of other problematic radionuclides by use of nuclear reactions with neutrons in a MSR-SPHINX system. This reactor system is an actinide burner (most in resonance neutron spectrum) and a radionuclide transmuter in a well-thermalized neutron spectrum. In the frame of the physical part, there are computational analyses and experimental activities. The experimental program has been focused, in its first stage, on a short-term irradiation of small size samples of molten-salt systems as well as structural materials proposed for the MSR blanket in the field of high neutron flux of research reactors. The proposed next stage of the program will focus on a large-scale experimental verification of design inputs by use of MSR-type inserting zones into the existing light water moderated experimental reactor LR-0, which may allow us to modify it into the experimental zero power salt reactor SR-0. There will be a detail description of the proposed program given in the paper together with the so far performed experiments and their first results. These realized experiments help us also to verify computational codes used, and to recognize some anomalies related to molten fluorides utilization. (authors)

  16. Neutron diffraction instruments at the BR2 reactor and their use in the study of crystal and magnetic structures

    International Nuclear Information System (INIS)

    Legrand, E.

    1977-01-01

    The study of structural properties of condensed matted is frequently performed by means of methods based on X-ray, electron or neutron scattering. In many particular cases, the latter technique offers definite advantages which are mainly based on the following characteristics of neutron-matter interaction: The large penetration depth for neutrons in matter (with the exception of a few elements), which allow the study of large samples and the use of wavelengths from 0.5 to 5A, and even to 10 A; the absence of an atomic form factor; the irregular dependence of the scattering length for the different elements as a function of their atomic number; the large cross section for magnetic scattering; the energy of thermal neutrons which allows the direct measuremtment of the enrgy exchange between the neutrons and the scatterer. Three of the four neutron diffraction instruments for solid state research, installed at the BR2 reactor, are described in detail. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Haeggblom, H

    1965-08-15

    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.

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

    International Nuclear Information System (INIS)

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

    2007-01-01

    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.

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

    International Nuclear Information System (INIS)

    Bess, John D.; Fujimoto, Nozomu

    2014-01-01

    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

  20. 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)

    1961-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1961-06-22

    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.

  2. Electron beam irradiation experiments of monoblock divertor mock-up

    International Nuclear Information System (INIS)

    Satoh, Kazuyoshi; Akiba, Masato; Araki, Masanori; Suzuki, Satoshi; Yokoyama, Kenji; Smid, I.; Cardella, A.; Duwe, R.; Di Pietro, E.

    1993-03-01

    It is one of the key issues for ITER to develop the divertor plate. Electron beam irradiation tests were carried out on a NET divertor mock-up using JEBIS at JAERI under a collaboration between The NET team, JAERI and KFA Juelich. Screening tests (maximum heat flux of 23 MW/m 2 ) and thermal cycling tests (18 MW/m 2 , 30s, 1000cycle) were carried out. As a result of the screening tests, the erosion caused by sublimation of C/C was observed on the surface of armor tile. No serious damage such as cracks or detachments, however, were found. As a result of the thermal cycling tests, no major damage was detected on the C/C surface. However cooling time constant of the divertor mock-up increased over 600cycle. Therefore it implies that some defects would occur at the brazing interface of the divertor mock-up. (author)

  3. ABOUT DIGITAL MOCK-UP FOR MECHANICAL PRODUCTS

    Directory of Open Access Journals (Sweden)

    GHERGHINA George

    2015-06-01

    The digital mock-up of the product is built at a design stage, and is applicable to the whole life-cycle of the product, including design, manufacture, marketing and aftermarket. The digital mock-up could achieve interference check, motion analysis, simulation of performance and manufacturing, technical training, advertising and maintenance, planning etc. The DMU of mechanical products, as important engineering data in a company, is supposed to be able to support all the activities in the whole life-cycle of the product including design, manufacture, marketing and aftermarket

  4. Two-detector cross-correlation noise technique and its application in measuring reactor kinetic parameters

    International Nuclear Information System (INIS)

    Lu Guiping; Peng Feng; Yi Jieyi

    1988-01-01

    The two-detector cross-correlation noise technique is a new method of measuring reactor kinetic parameters developed in the sixties. It has the advantages of non-perturbation in core, high signal to noise ratio, low space dependent effect, and simple and reliable in measurement. A special set of cross-correlation analyzer has been prepared for measuring kinetic parameters of several reactor assemblies, such as the High Flux Engineering Test Reactor, its zero power mock up facility and a low enriched uranium light water lattice zero power facility

  5. Analysis of the Temporal Response of Coupled Asymmetrical Zero-Power Subcritical Bare Metal Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Klain, Kimberly L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-21

    The behavior of symmetrical coupled-core systems has been extensively studied, yet there is a dearth of research on asymmetrical systems due to the increased complexity of the analysis of such systems. In this research, the multipoint kinetics method is applied to asymmetrical zeropower, subcritical, bare metal reactor systems. Existing research on asymmetrical reactor systems assumes symmetry in the neutronic coupling; however, it will be shown that this cannot always be assumed. Deep subcriticality adds another layer of complexity and requires modification of the multipoint kinetics equations to account for the effect of the external neutron source. A modified set of multipoint kinetics equations is derived with this in mind. Subsequently, the Rossi-alpha equations are derived for a two-region asymmetrical reactor system. The predictive capabilities of the radiation transport code MCNP6 for neutron noise experiments are shown in a comparison to the results of a series of Rossi-alpha measurements performed by J. Mihalczo utilizing a coupled set of symmetrical bare highly-enriched uranium (HEU) cylinders. The ptrac option within MCNP6 can generate time-tagged counts in a cell (list-mode data). The list-mode data can then be processed similarly to measured data to obtain values for system parameters such as the dual prompt neutron decay constants observable in a coupled system. The results from the ptrac simulations agree well with the historical measured values. A series of case studies are conducted to study the effects of geometrical asymmetry in the coupling between two bare metal HEU cylinders. While the coupling behavior of symmetrical systems has been reported on extensively, that of asymmetrical systems remains sparse. In particular, it appears that there has been no previous research in obtaining the coupling time constants for asymmetrically-coupled systems. The difficulty in observing such systems is due in part to the inability to determine the

  6. 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)

  7. 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.

  8. FBR core mock-up RAPSODIE I - experimental analysis

    International Nuclear Information System (INIS)

    Brochard, D.; Buland, P.; Gantenbein, F.

    1990-01-01

    The main phenomena which influence the LMFBR core response to a seismic excitation are the fluid structure interaction and the impacts between subassemblies. To study the core behaviour, seismic tests have been performed on the core mock-up RAPSODIE with and without fluid and restraint ring and for different levels of excitation. This paper summarizes the results of these tests. (author)

  9. Performance test results of mock-up test facility of HTTR hydrogen production system

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Inaba, Yoshitomo; Nishihara, Tetsuo

    2004-01-01

    For the purpose to demonstrate effectiveness of high-temperature nuclear heat utilization, Japan Atomic Energy Research Institute has been developing a hydrogen production system and has planned to connect the hydrogen production system to High Temperature Engineering Test Reactor (HTTR). Prior to construction of a HTTR hydrogen production system, a mock-up test facility was constructed to investigate transient behavior of the hydrogen production system and to establish system controllability. The Mock-up test facility with a full-scale reaction tube is an approximately 1/30-scale model of the HTTR hydrogen production system and an electric heater is used as a heat source instead of a reactor. After its construction, a performance test of the test facility was carried out in the same pressure and temperature conditions as those of the HTTR hydrogen production system to investigate its performance such as hydrogen production ability, controllability and so on. It was confirmed that hydrogen was stably produced with a hot helium gas about 120m 3 /h, which satisfy the design value, and thermal disturbance of helium gas during the start-up could be mitigated within the design value by using a steam generator. The mock-up test of the HTTR hydrogen production system using this facility will continue until 2004. (author)

  10. 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

    International Nuclear Information System (INIS)

    Dionne, B.; Tzanos, C.P.

    2011-01-01

    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.

  11. Development of An Embedded FPGA-Based Data Acquisition System Dedicated to Zero Power Reactor Noise Experiments

    Directory of Open Access Journals (Sweden)

    Arkani Mohammad

    2014-08-01

    Full Text Available An embedded time interval data acquisition system (DAS is developed for zero power reactor (ZPR noise experiments. The system is capable of measuring the correlation or probability distribution of a random process. The design is totally implemented on a single Field Programmable Gate Array (FPGA. The architecture is tested on different FPGA platforms with different speed grades and hardware resources. Generic experimental values for time resolution and inter-event dead time of the system are 2.22 ns and 6.67 ns respectively. The DAS can record around 48-bit x 790 kS/s utilizing its built-in fast memory. The system can measure very long time intervals due to its 48-bit timing structure design. As the architecture can work on a typical FPGA, this is a low cost experimental tool and needs little time to be established. In addition, revisions are easily possible through its reprogramming capability. The performance of the system is checked and verified experimentally.

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

    International Nuclear Information System (INIS)

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

    1979-09-01

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

  13. Numerical studies on helium cooled divertor finger mock up with sectorial extended surfaces

    International Nuclear Information System (INIS)

    Rimza, Sandeep; Satpathy, Kamalakanta; Khirwadkar, Samir; Velusamy, Karupanna

    2014-01-01

    Highlights: • Studies on heat transfer enhancement for divertor finger mock-up. • Heat transfer characteristics of jet impingement with extended surfaces have been investigated. • Effect of critical parameters that influence the thermal performance of the finger mock-up by CFD approach. • Effect of extended surface in enhancing heat removal potential with pumping power assessed. • Practicability of the chosen design is verified by structural analysis. - Abstract: Jet impinging technique is an advance divertor concept for the design of future fusion power plants. This technique is extensively used due to its high heat removal capability with reasonable pumping power and for safe operation. In this design, plasma-facing components are fabricated with numerous fingers cooled by helium jets to reduce the thermal stresses. The present study is focused towards finding an optimum performance of one such finger mock-up through systematic computational fluid dynamics (CFD) studies. Heat transfer characteristics of jet impingement have been numerically investigated with sectorial extended surfaces (SES). The result shows that addition of SES enhances heat removal potential with minimum pumping power. Detailed parametric studies on critical parameters that influence thermal performance of the finger mock-up have been analyzed. Thermo-mechanical analysis has been carried out through finite element based approach to know the state of stress in the assembly as a result of large temperature gradients. It is seen that the stresses are within the permissible limits for the present design. The whole numerical simulation has been carried out using general-purpose CFD software (ANSYS FLUENT, Release 14.0, User Guide, Ansys, Inc., 2011). Benchmark validation studies have been performed against high-heat flux experiments (B. Končar, P. Norajitra, K. Oblak, Appl. Therm. Eng., 30, 697–705, 2010) and a good agreement is noticed between the present simulation and the reported

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

    International Nuclear Information System (INIS)

    Ishitsuka, Etsuo; Kawamura, Hiroshi; Sagawa, Hisashi; Nagakura, Masaaki; Kanzawa, Toru.

    1993-03-01

    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)

  15. Monte Carlo simulation of irradiation of MTR fuel plates in the BR2 reactor using a full-scale 3-d model with inclined channels

    International Nuclear Information System (INIS)

    Kuzminov, V. V; Koonen, E.; Ponsard, B.

    2002-01-01

    A three-dimensional full-scale Monte Carlo model of the BR2 reactor has been developed for simulation of irradiation conditions of materials and fuel loaded in various irradiation devices. This new reactor model includes a detailed geometrical description of the inclined reactor channels, the irradiation devices loaded in these channels including the materials to be tested/loaded in these devices, the burn-up of the BR2 fuel elements and the poisoning of the beryllium matrix. Recently a benchmark irradiation of new irradiation device for testing and qualification of MTR fuel plates has been performed. For this purpose the detailed irradiation conditions of fuel plates had to be predetermined. Monte Carlo calculations of neutron fluxes and heat load distributions in irradiated MTR fuel plates were performed taking into account the contents of all loaded experimental devices in the reactor channels. A comparison of the calculated and measured values of neutron fluxes and of heat loads in the BR2 reactor is presented in this paper. The comparison is part of the validation process of the new reactor model. It also serves to establish the capability to conduct a fuel plate irradiation program under requested and well- known irradiation conditions. (author)

  16. 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.  

  17. Analysis of high heat flux testing of mock-ups

    International Nuclear Information System (INIS)

    Salavy, J.-F.; Giancarli, L.; Merola, M.; Picard, F.; Roedig, M.

    2003-01-01

    ITER EU Home Team is performing a large R and D effort in support of the development of high heat flux components for ITER. In this framework, this paper describes the thermal analyses, the fatigue lifetime evaluation and the transient VDE with material melting related to the high heat flux thermo-mechanical tests performed in the JUDITH facility. It reports on several mock-ups representative of different proposed component designs based on Be, W and CFC as armour materials

  18. 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)

    1983-09-01

    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

  19. Blanket Cooling Plates Mock-ups Manufactured in different Diffusion Weld Setup

    International Nuclear Information System (INIS)

    Von Der Weth, A.; Aktaa, J.

    2007-01-01

    Full text of publication follows: The breeding blanket box is considered as one of the most important components of a future fusion power plant. It will be assembled by so called cooling plates (CP) with a system of internal cooling channels. Such a CP is produced by two symmetric half pieces with half milled-in channels. Both pieces will be joined by a diffusion weld (DW) process. Within recent years a two step DW process for different EUROFER batches has been developed. It has been first applied to small laboratory scaled samples with dimensions of 25 mm x 30 mm x 40 mm. Then the DW process had then been successfully transferred to so called compact mock ups which are small CPs with dimensions of 67 mm x 70 mm x 50 mm. As third step this process has been used to manufacture a CP (465 mm x 205 mm x 50 mm) of a breeder unit in an industrial uniaxial diffusion weld setup. This paper treats the manufacturing sequence of a cooling plate and a first wall mock up in an industrial hot isostatic pressing (HIP) setup. The firstly laboratory specimens scaled diffusion weld process has been adjusted to different cooling channel dimensions and a different DW setup. The weld quality is investigated by tensile and Charpy impact testing. This allows comparison of the weld quality of mock ups welded in different DW setups. (authors)

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  1. Experimental Investigation of the IFMIF Target Mock-up

    International Nuclear Information System (INIS)

    Loginov, N.; Mikheyev, A.; Morozov, V.; Aksenov, Y.; Arnoldov, M.; Berensky, L.; Fedotovsky, V.; Chernov, V.M.; Nakamura, H.

    2007-01-01

    Full text of publication follows: The IFMIF lithium neutron target mock-ups have been constructed and tested at the water and lithium test facilities. Description of the mock-ups and test facilities is presented in the paper, as well as the main results obtained. Reference geometry was used but the mockup flow cross-section was decreased. Velocity of water and lithium was up to reference value of 20 m/s. Features of lithium and water hydrodynamics were observed. The calculations and experiments showed that conjunction point of back wall straight and concave sections generated instability of lithium flow because of centrifugal force sudden change at this place. Therefore, it was proposed to use parabolic shape of the target back wall. Generation of wakes at the corners of cross-section of the Shima nozzle outlet was observed, and, as a result, surface waves appeared on the lithium jet. Observations of lithium and water jets and measurements of water jet thickness showed significant increasing the thickness near sidewalls of the mock-up concave section. It is because of absence of the centrifugal force at these places. Very large instability of the water jet surface was observed when outlet part of the Shima nozzle was divergent slightly (about 1 deg.), and vice versa very smooth jet surface occurred in confusing case (of about 0.5 deg.). So, nozzle outlet shape is very critical. Evaporation of lithium from the jet surface was investigated as well as deposition of vapor on vacuum pipe wall. It turned out to be not so critical. Significant part of the work concerned purification of lithium and monitoring impurities. The possibility of denitration of lithium down to 2 ppm by means of aluminum soluble getter was showed. Two types of both cold traps and plug indicators of impurities were tested. The results are presented in the paper. (authors)

  2. Digital mock-up for the spent fuel disassembly processes

    International Nuclear Information System (INIS)

    Lee, J. Y.; Kim, S. H.; Song, T. G.; Kim, Y. H.; Hong, D. H.; Yoon, J. S.

    2000-12-01

    In this study, the graphical design system is developed and the digital mock-up is implemented for designing the spent fuel handling and disassembly processes. The system consists of a 3D graphical modeling system, a devices assembling system, and a motion simulation system. This system is used throughout the design stages from the conceptual design to the motion analysis. By using this system, all the process involved in the spent fuel handling and disassembly processes are analyzed and optimized. Also, this system is used in developing the on-line graphic simulator which synchronously simulates the motion of the equipment in a real time basis by connecting the device controllers with the graphic server through the TCP/IP network. This simulator can be effectively used for detecting the malfunctions of the process equipment which is remotely operated. Thus, the simulator enhances the reliability and safety of the spent fuel handling process by providing the remote monitoring function of the process. The graphical design system and the digital mock-up system can be effectively used for designing the process equipment, as well as the optimized process and maintenance process. And the on-line graphic simulator can be an alternative of the conventional process monitoring system which is a hardware based system

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

    International Nuclear Information System (INIS)

    Diniz, Ricardo

    2005-01-01

    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 β i and λ i 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 β eff 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 β eff determination is obtained with JENDL3.3. In contrast, ENDF/B-VI.8 and its revised version performed at LANL overestimate β eff by as much as 4%. The β eff results of this work support totally the proposal of reducing the thermal delayed neutron number for 235 U 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)

  4. Critical experiments supporting close proximity water storage of power reactor fuel. Technical progress report, July 1, 1978-September 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Baldwin, M.N.; Hoovler, G.S.; Eng, R.L.; Welfare, F.G.

    1978-11-01

    Experimental measurements are being taken on critical configurations of clusters of fuel rods mocking up LWR-type fuel elements in close proximity water storage. The results will serve to benchmark the computer codes used in designing nuclear power reactor fuel storage racks. KENO calculations of Cores I to VI are within two standard deviations of the measured k/sub eff/ values.

  5. BR2 mixed core management

    International Nuclear Information System (INIS)

    Ponsard, B.; Beeckmans, A.

    1997-01-01

    The BR2 fuel cycle management can be optimized by the fabrication and the irradiation of fuel elements with uranium recovered from the reprocessing of BR2 spent fuel. The VIn E fuel performances could be upgraded by increasing the amount of burnable poisons, the fuel mass, the fuel density, ... in order to obtain a higher reactivity effect at a burnup of about β=12% and a longer cycle duration. The preliminary results of the calculations need however to be confirmed by measurements on effective reactor loads. (author)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  7. Project W-314 performance mock-up test procedure

    International Nuclear Information System (INIS)

    CARRATT, R.T.

    1999-01-01

    The purpose of this Procedure is to assist construction in the pre-operational fabrication and testing of the pit leak detection system and the low point drain assembly by: (1) Control system testing of the pit leak detection system will be accomplished by actuating control switches and verifying that the control signal is initiated, liquid testing and overall operational requirements stated in HNF-SD-W314-PDS-003, ''Project Development Specification for Pit Leak Detection''. (2) Testing of the low point floor drain assembly by opening and closing the drain to and from the ''retracted'' and ''sealed'' positions. Successful operation of this drain will be to verify that the seal does not leak on the ''sealed'' position, the assembly holds liquid until the leak detector actuates and the assembly will operate from on top of the mock-up cover block

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

    CERN Document Server

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

    2018-05-03

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

  9. Manufacturing of a HCLL cooling plate mock up

    International Nuclear Information System (INIS)

    Rigal, E.; Dinechin, G. de; Rampal, G.; Laffont, G.; Cachon, L.

    2007-01-01

    The European DEMO blankets and associated Test Blanket Modules (TBM) are made of a set of components cooled by flowing helium at 80bar pressure. Hot Isostatic Pressing (HIP) is one of the very few processes that allow manufacturing such components exhibiting complex cooling channels. In HIP technology, the parts used to manufacture components with embedded channels are usually machined plates, blocks and tubes. Achievable geometries are limited in shape because it is not always possible to figure the channels by bent tubes. This occurs for example when channels present sharp turns, when the cross section of the channels is rectangular or when the rib between channels is so small that very thin tubes would be required. In these cases, bending is unpractical. The breeder unit cooling plates of the Helium Cooled Lithium Lead (HCLL) blanket have eight 4 x 4.5 mm parallel channels that run following a double U scheme. Turns are sharp and the wall thickness is small (1mm), so the manufacturing process described above cannot be used. An alternative process has been developed which has many advantages. It consists in machining grooves in a base plate, then closing the top of the grooves using thin welded strips, and finally adding a plate by HIP. There is then no need for the use of tubes with associated bending and deformation issues. The final component contains welds, but it must be stressed out that these potentially brittle zones do not connect the channels to the external surface because they are covered by the HIPed plate. Furthermore, the welds are homogenised during the HIP operation and further heat treatments. This paper describes the design of a simplified cooling plate mock up and its fabrication using this so-called weld+HIP process. The thermal fatigue testing of this mock up is presented somewhere else in this conference. (orig.)

  10. 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)

    Martin, H.R; Carballido, C.; Oliveras, T.

    1991-01-01

    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) [es

  11. Zero-Power Radio Device.

    Energy Technology Data Exchange (ETDEWEB)

    Brocato, Robert W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2018-02-01

    This report describes an unpowered radio receiver capable of detecting and responding to weak signals transmit ted from comparatively long distances . This radio receiver offers key advantages over a short range zero - power radio receiver previously described in SAND2004 - 4610, A Zero - Power Radio Receiver . The device described here can be fabricated as an integrated circuit for use in portable wireless devices, as a wake - up circuit, or a s a stand - alone receiver operating in conjunction with identification decoders or other electroni cs. It builds on key sub - components developed at Sandia National Laboratories over many years. It uses surface acoustic wave (SAW) filter technology. It uses custom component design to enable the efficient use of small aperture antennas. This device uses a key component, the pyroelectric demodulator , covered by Sandia owned U.S. Patent 7397301, Pyroelectric Demodulating Detector [1] . This device is also described in Sandia owned U.S. Patent 97266446, Zero Power Receiver [2].

  12. High heat flux tests of mock-ups for ITER divertor application

    International Nuclear Information System (INIS)

    Giniatulin, R.; Gervash, A.; Komarov, V.L.; Makhankov, A.; Mazul, I.; Litunovsky, N.; Yablokov, N.

    1998-01-01

    One of the most difficult tasks in fusion reactor development is the designing, fabrication and high heat flux testing of actively cooled plasma facing components (PFCs). At present, for the ITER divertor project it is necessary to design and test components by using mock-ups which reflect the real design and fabrication technology. The cause of failure of the PFCs is likely to be through thermo-cycling of the surface with heat loads in the range 1-15 MW m -2 . Beryllium, tungsten and graphite are considered as the most suitable armour materials for the ITER divertor application. This work presents the results of the tests carried out with divertor mock-ups clad with beryllium and tungsten armour materials. The tests were carried out in an electron beam facility. The results of high heat flux screening tests and thermo-cycling tests in the heat load range 1-9 MW m -2 are presented along with the results of metallographic analysis carried out after the tests. (orig.)

  13. Gamma-ray spectrometric measurements of fission rate ratios between fresh and burnt fuel following irradiation in a zero-power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kröhnert, H., E-mail: hanna.kroehnert@ensi.ch [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); École Polytechnique Fédérale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland); Perret, G.; Murphy, M.F. [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); Chawla, R. [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); École Polytechnique Fédérale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland)

    2013-01-11

    The gamma-ray activity from short-lived fission products has been measured in fresh and burnt UO{sub 2} fuel samples after irradiation in a zero-power reactor. For the first time, short-lived gamma-ray activity from fresh and burnt fuel has been compared and fresh-to-burnt fuel fission rate ratios have been derived. For the measurements, well characterized fresh and burnt fuel samples, with burn-ups up to 46 GWd/t, were irradiated in the zero-power research reactor PROTEUS. Fission rate ratios were derived based on the counting of high-energy gamma-rays above 2200 keV, in order to discriminate against the high intrinsic activity of the burnt fuel. This paper presents the measured fresh-to-burnt fuel fission rate ratios based on the {sup 142}La (2542 keV), {sup 89}Rb (2570 keV), {sup 138}Cs (2640 keV) and {sup 95}Y (3576 keV) high-energy gamma-ray lines. Comparisons are made with the results of Monte Carlo modeling of the experimental configuration, carried out using the MCNPX code. The measured fission rate ratios have 1σ uncertainties of 1.7–3.4%. The comparisons with calculated predictions show an agreement within 1–3σ, although there appears to be a slight bias (∼3%).

  14. Some particular aspects of control in nuclear power reactors

    International Nuclear Information System (INIS)

    Vathaire, F. de; Vernier, Ph.; Pascouet, A.

    1964-01-01

    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) [fr

  15. New design procedure development of future reactor critical power estimation. (1) Practical design-by-analysis method for BWR critical power design correlation

    International Nuclear Information System (INIS)

    Yamamoto, Yasushi; Mitsutake, Toru

    2007-01-01

    For present BWR fuels, the full mock-up thermal-hydraulic test, such as the critical power measurement test, pressure drop measurement test and so on, has been needed. However, the full mock-up test required the high costs and large-scale test facility. At present, there are only a few test facilities to perform the full mock-up thermal-hydraulic test in the world. Moreover, for future BWR, the bundle size tends to be larger, because of reducing the plant construction costs and minimizing the routine check period. For instance, AB1600, improved ABWR, was proposed from Toshiba, whose bundle size was 1.2 times larger than the conventional BWR fuel size. It is too expensive and far from realistic to perform the full mock-up thermal-hydraulic test for such a large size fuel bundle. The new design procedure is required to realize the large scale bundle design development, especially for the future reactor. Therefore, the new design procedure, Practical Design-by-Analysis (PDBA) method, has been developed. This new procedure consists of the partial mock-up test and numerical analysis. At present, the subchannel analysis method based on three-fluid two-phase flow model only is a realistic choice. Firstly, the partial mock-up test is performed, for instance, the 1/4 partial mock-up bundle. Then, the first-step critical power correlation coefficients are evaluated with the measured data. The input data, such as the spacer effect model coefficient, on the subchannel analysis are also estimated with the data. Next, the radial power effect on the critical power of the full-bundle size was estimated with the subchannel analysis. Finally, the critical power correlation is modified by the subchannel analysis results. In the present study, the critical power correlation of the conventional 8x8 BWR fuel was developed with the PDBA method by 4x4 partial mock-up tests and the subchannel analysis code. The accuracy of the estimated critical power was 3.8%. The several themes remain to

  16. Results of water chemistry control in the in-pile ''Callisto'' loop (an experimental PWR rig installed in the BR2 reactor)

    International Nuclear Information System (INIS)

    Weber, M.; Benoit, P.; Dekeyser, J.; Verwimp, A.

    1994-01-01

    Since June 1992, a new experimental facility, called CALLISTO, is being irradiated in the BR2 materials testing reactor at Mol, Belgium. The main objective of the present test campaign is to study the behaviour of advanced fuel to high burn-up rates in a realistic PWR environment. Three in-pile sections, containing each 9 fuel rods, are loaded inside the reactor vessel and are connected to a common out-of-pile pressurized water circulation loop (ref.1). The later is branched-off into a purification circuit (feed-bleed concept) and further equipped with safety and auxiliary systems. To cope with the test programme, the equipments are designed so that the guidelines of a PWR primary water chemistry can be followed (ref.2). Real steady-state conditions cannot be observed because the typical BR2 cycle (3 weeks running/3 weeks shut-down) is much shorter and because the rig is cooled down during each reactor shut-down. The purpose of this poster is to provide results of chemical parameters recorded during the cycling behaviour of the CALLISTO primary water. (authors). 4 figs., 1 tab., 2 refs

  17. Direct tritium measurement in lithium titanate for breeding blanket mock-up experiments with D-T neutrons

    International Nuclear Information System (INIS)

    Klix, A.; Ochiai, K.; Nishitani, T.; Takahashi, A.

    2004-01-01

    At Fusion Neutronics Source (FNS) of JAERI, tritium breeding experiments with blanket mock-ups consisting of advanced fusion reactor materials are in progress. The breeding zones are thin layers of lithium titanate which is one of the candidate tritium breeder materials for the DEMO fusion power reactor. It is anticipated that the application of small pellet-shaped lithium titanate detectors manufactured from the same material as the breeding layer will reduce experimental uncertainties arising from necessary corrections due to different isotopic lithium volume concentrations in breeding material and detector. Therefore, a method was developed to measure the local tritium production by means of lithium titanate pellet detectors and a liquid scintillation counting technique. The lithium titanate pellets were dissolved in concentrated hydrochloric acid solution and the resulting acidic solution was neutralized. Two ways of further processing were followed: direct incorporation into a liquid scintillation cocktail and distillation of the solution followed by mixing with liquid scintillator. Two types of lithium titanate pellets were investigated with different 6 Li enrichment and manufacturing procedure. It was found that lithium titanate is suitable for tritium production measurements. However some discrepancies in the measurement accuracy remained with one of the investigated pellet detectors when compared with a well-established lithium carbonate measurement technique and this issue needs further investigation

  18. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

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

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

    International Nuclear Information System (INIS)

    Dreier, J.; Winkler, H.

    1985-01-01

    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)

  20. Interactive virtual mock-ups for Remote Handling compatibility assessment of heavy components

    Energy Technology Data Exchange (ETDEWEB)

    Oosterhout, J. van, E-mail: j.vanoosterhout@differ.nl [FOM Institute DIFFER (Dutch Institute for Fundamental Energy Research), Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Heemskerk, C.J.M.; Koning, J.F. [Heemskerk Innovative Technology B.V., Jonckerweg 12, 2201 DZ Noordwijk 6 (Netherlands); Ronden, D.M.S.; Baar, M. de [FOM Institute DIFFER (Dutch Institute for Fundamental Energy Research), Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster, PO Box 1207, 3430 BE Nieuwegein (Netherlands)

    2014-10-15

    Highlights: •Specific ITER components require RHCA on hardware mock-ups. •Hardware mock-ups are expensive and have a long lead time. •Interactive Virtual Reality mock-ups are readily available and easily adapted. •This paper analysis and proposes improvements to simulator capabilities. -- Abstract: ITER standards Tesini (2009) require hardware mock-ups to validate the Remote Handling (RH) compatibility of RH class 1- and critical class 2-components. Full-scale mock-ups of large ITER components are expensive, have a long lead time and lose their relevance in case of design changes. Interactive Virtual Reality simulations with real time rigid body dynamics and contact interaction allow for RH Compatibility Assessment during the design iterations. This paper explores the use of interactive virtual mock-ups to analyze the RH compatibility of heavy component handling and maintenance. It infers generic maintenance operations from the analysis and proposes improvements to the simulator capabilities.

  1. Interactive virtual mock-ups for Remote Handling compatibility assessment of heavy components

    International Nuclear Information System (INIS)

    Oosterhout, J. van; Heemskerk, C.J.M.; Koning, J.F.; Ronden, D.M.S.; Baar, M. de

    2014-01-01

    Highlights: •Specific ITER components require RHCA on hardware mock-ups. •Hardware mock-ups are expensive and have a long lead time. •Interactive Virtual Reality mock-ups are readily available and easily adapted. •This paper analysis and proposes improvements to simulator capabilities. -- Abstract: ITER standards Tesini (2009) require hardware mock-ups to validate the Remote Handling (RH) compatibility of RH class 1- and critical class 2-components. Full-scale mock-ups of large ITER components are expensive, have a long lead time and lose their relevance in case of design changes. Interactive Virtual Reality simulations with real time rigid body dynamics and contact interaction allow for RH Compatibility Assessment during the design iterations. This paper explores the use of interactive virtual mock-ups to analyze the RH compatibility of heavy component handling and maintenance. It infers generic maintenance operations from the analysis and proposes improvements to the simulator capabilities

  2. Experimental test campaign on an ITER divertor mock-up

    International Nuclear Information System (INIS)

    Dell'Orco, G.; Malavasi, A.; Merola, M.; Polazzi, G.; Simoncini, M.; Zito, D.

    2002-01-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body (CB), designed with some mechanical and hydraulic simplifications with respect to the reference body and its actively cooled Dummy Armour Prototype (DAP). This DAP consists of a Vertical Target (VT), a Wing (WI) and a Dump Target (DT), manufactured by European industries, which are integrated to the Gas Box Liner (GBL) supplied by the Russian Federation ITER Home Team. In 1999, in parallel with the manufacturing activity, the ITER European Home Team decided to assign to ENEA a Task for checking the component integration and performing the thermal-hydraulic and thermal mechanical testing of the DAP and CB. In 1999-2000, ENEA performed the experimental campaign at Brasimone Labs. The present work presents the experimental results of the component integration and the thermal-hydraulic and thermo-mechanical fatigue tests

  3. Experimental test campaign on an ITER divertor mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Dell' Orco, G. E-mail: giovanni.dellorco@brasimone.enea.it; Malavasi, A.; Merola, M.; Polazzi, G.; Simoncini, M.; Zito, D

    2002-11-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body (CB), designed with some mechanical and hydraulic simplifications with respect to the reference body and its actively cooled Dummy Armour Prototype (DAP). This DAP consists of a Vertical Target (VT), a Wing (WI) and a Dump Target (DT), manufactured by European industries, which are integrated to the Gas Box Liner (GBL) supplied by the Russian Federation ITER Home Team. In 1999, in parallel with the manufacturing activity, the ITER European Home Team decided to assign to ENEA a Task for checking the component integration and performing the thermal-hydraulic and thermal mechanical testing of the DAP and CB. In 1999-2000, ENEA performed the experimental campaign at Brasimone Labs. The present work presents the experimental results of the component integration and the thermal-hydraulic and thermo-mechanical fatigue tests.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, Etsuo; Kawamura, Hiroshi; Sagawa, Hisashi (Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment); Nagakura, Masaaki; Kanzawa, Toru.

    1993-03-01

    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 [approx] -60 %, and that the pressure drop was not affected by moisture concentration (< 100 ppm). (author).

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

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, Etsuo; Kawamura, Hiroshi; Sagawa, Hisashi [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Nagakura, Masaaki; Kanzawa, Toru

    1993-03-01

    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 {approx} -60 %, and that the pressure drop was not affected by moisture concentration (< 100 ppm). (author).

  6. Fabrication of a 1/6-scale mock-up and manifolds for the Korea first wall in the ITER

    International Nuclear Information System (INIS)

    Yoon, Jae Sung; Kim, Suk Kwon; Lee, Eo Hwak; Lee, Dong Won

    2012-01-01

    Korea has developed and participated in the Test Blanket Module (TBM) program of the International Thermo-nuclear Experimental Reactor (ITER). The first wall (FW) of the TBM is an important component that faces the plasma directly and therefore it is subjected to high heat and neutron loads. To fabricate the TBM FW, the Hot Isostatic Pressing (HIP) bonding method has been investigated. In the present study, the manufacturing method of the TBM FW is introduced through the fabrication and testing of a 1/6-scale mockup. To distribute fluid uniformly in the mock-up, a manifold was designed and fabricated using the ANSYS-CFX analysis. After the mock-up was fabricated and its fluid distribution tests performed, we compared the results of tests with the simulated results

  7. Progress on pebble bed experimental activity for the HE-FUS3 mock-ups

    International Nuclear Information System (INIS)

    Dell'Orco, G.; Sansone, L.; Simoncini, M.; Zito, D.

    2002-01-01

    The EU Long Term for DEMO Programme foresees the qualification of the reference design of the helium cooled pebble bed (HCPB) - test blanket module (TBM) to be tested in ITER Reactor. In this frame, FZK and ENEA have launched many experimental activities for the evaluation of the interactions between the Tritium breeder and neutron multiplier pebble beds and the steel containment walls. Main aim of these activities is the measuring the pebble bed effective thermal conductivity, the wall heat transfer coefficient as well as their dependency from the mechanical constraints. The paper presents the progress of the testing activity and results of the tests on two mock-up, called Tazza and Helichetta, carried out on the HE-FUS3 facility at ENEA Brasimone. (orig.)

  8. Mock-up experiment and analysis for the primary shield of the N.S. MUTSU

    International Nuclear Information System (INIS)

    Miyasaka, S.; Asaoka, T.; Taji, Y.; Ise, T.; Koyama, K.; Tsutsui, T.; Takeuchi, M.; Fuse, T.; Miura, T.; Yamaji, Y.

    1977-01-01

    A series of shielding mock-up experiments was performed at JRR-4, a swimming pool type reactor, of Japan Atomic Energy Research Institute (JAERI) to obtain the necessary experimental data and the sufficiently accurate method of calculation adopted for the modification of the MUTSU primary shield. Analyses for the experiments were carried out by using of the Ssub(n) codes, ANISN and TWOTRAN. The two dimensional calculations were performed with the P 1 -S 8 approximation. The neutron streaming through the annular gap between the pressure vessel and the primary shield has been confirmed to be estimated from the present method of calculation. The agreement between the calculated and the measured values is generally in about a factor of 2 to 4. (orig.) [de

  9. Validation results of satellite mock-up capturing experiment using nets

    Science.gov (United States)

    Medina, Alberto; Cercós, Lorenzo; Stefanescu, Raluca M.; Benvenuto, Riccardo; Pesce, Vincenzo; Marcon, Marco; Lavagna, Michèle; González, Iván; Rodríguez López, Nuria; Wormnes, Kjetil

    2017-05-01

    The PATENDER activity (Net parametric characterization and parabolic flight), funded by the European Space Agency (ESA) via its Clean Space initiative, was aiming to validate a simulation tool for designing nets for capturing space debris. This validation has been performed through a set of different experiments under microgravity conditions where a net was launched capturing and wrapping a satellite mock-up. This paper presents the architecture of the thrown-net dynamics simulator together with the set-up of the deployment experiment and its trajectory reconstruction results on a parabolic flight (Novespace A-310, June 2015). The simulator has been implemented within the Blender framework in order to provide a highly configurable tool, able to reproduce different scenarios for Active Debris Removal missions. The experiment has been performed over thirty parabolas offering around 22 s of zero-g conditions. Flexible meshed fabric structure (the net) ejected from a container and propelled by corner masses (the bullets) arranged around its circumference have been launched at different initial velocities and launching angles using a pneumatic-based dedicated mechanism (representing the chaser satellite) against a target mock-up (the target satellite). High-speed motion cameras were recording the experiment allowing 3D reconstruction of the net motion. The net knots have been coloured to allow the images post-process using colour segmentation, stereo matching and iterative closest point (ICP) for knots tracking. The final objective of the activity was the validation of the net deployment and wrapping simulator using images recorded during the parabolic flight. The high-resolution images acquired have been post-processed to determine accurately the initial conditions and generate the reference data (position and velocity of all knots of the net along its deployment and wrapping of the target mock-up) for the simulator validation. The simulator has been properly

  10. Mock-up-CZ: dismantling of the experiment - Geotechnical results

    International Nuclear Information System (INIS)

    Svoboda, J.; Vasicek, R.

    2010-01-01

    Document available in extended abstract form only. The issue of the disposal of radioactive waste is one of the most pressing challenges of our age, for which, in most countries, the deep repository concept is generally considered to be the most suitable final solution. In order to make such a repository both safe and reliable, intensive research is underway worldwide. The construction of physical models is one approach to the study of the engineered barriers for deep geological repositories; one such experiment, Mock-Up-CZ, has been performed at the Centre of Experimental Geotechnics, CTU in Prague. The Mock-Up-CZ experiment simulated the vertical placement of a container with radioactive waste, an approach that is in line with the Swedish KBS-3 system. The physical model consisted of a barrier made up of bentonite blocks, powdered bentonite backfill, a heater and hydration and monitoring systems. The whole experiment was enclosed in a cylindrical box, whose construction was able to withstand high pressure due to bentonite swelling. A number of sensors (monitoring changes in temperature, pressure and moisture) were placed inside the bentonite barrier. The basic material used in the experiment consisted of a mixture of Czech bentonite from the Rokle deposit (85%), quartz sand (10%) and graphite (5%). The first phase of the experiment commenced on 7 May 2002, during which the heater was switched on, with no water input. After 6 months the second phase commenced in which water was introduced through the hydration system. This phase ended on 2nd January 2006 when the heater was switched off. After allowing time for cooling, the dismantling phase commenced (30 January 2006). After a further one and a half months (17 March 2006) the dismantling of the experimental vessel was completed. Post-decommissioning analysis continued until the end of 2007. Dismantling and post-decommissioning analysis were carried out according to a very detailed plan which included not only

  11. Manufacturing of In-Pile Test Section(IPS) Mock-up for the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. M.; Park, K. N.; Chi, D. Y. (and others)

    2005-10-15

    Manufacturing process of IPS Mock-up was initiated in late of 2003 with DAEWOO Precision industries Company. Manufacturing drawings due to detail drawings are composed of Outer assembly and Inner assembly. Welding of IPS Mock-up was performed by the GMAW(Gas Metal Arc Welding) process. After the welding process, non-destructive examination was conducted. Leak test was performed to the Main cooling water part and Neon gas inter-space gap part by the He gas injection with the pressure of 6.0 kg{sub f}/cm{sup 2} and 30 minutes holding time. the result was shown that there was no leak at the Neon gas inter-space gap part but leak was occurred at Main cooling water part according to imperfect screw of purge plug. so, it was re-finished and test was performed to certify the leak tightness. To satisfy the HANARO Limiting Operation Condition, IPS should be tested ahead of installation at the HANARO reactor by the use of test facilities. IPS Mock-up and its test facilities will be designed and used for the test of 'HANARO flow tube pressure drop', 'IPS inner pressure drop' and 'IPS inner vibration'.

  12. Manufacturing, testing and post-test examination of ITER divertor vertical target W small scale mock-ups

    International Nuclear Information System (INIS)

    Visca, Eliseo; Cacciotti, Emanuele; Komarov, Anton; Libera, Stefano; Litunovsky, Nikolay; Makhankov, Alexey; Mancini, Andrea; Merola, Mario; Pizzuto, Aldo; Riccardi, Bruno; Roccella, Selanna

    2011-01-01

    ENEA is involved in the International Thermonuclear Experimental Reactor (ITER) R and D activities. During the last years ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP), suitable for the construction of high heat flux plasma-facing components, such as the divertor targets. In the frame of the EFDA contract six mock-ups were manufactured by HRP in the ENEA labs using W monoblocks supplied by the Efremov Institute in St. Petersburg, Russian Federation and IG CuCrZr tubes. According to the technical specifications the mock-ups were examined by ultrasonic technique and after their acceptance they were delivered to the Efremov Institute TSEFEY-M e-beam facility for the thermal fatigue testing. The test consisted in 3000 cycles of 15 s heating and 15 s cooling at 10 MW/m 2 and finally 1000 cycles at 20 MW/m 2 . After the testing the ultrasonic non-destructive examination was repeated and the results compared with the investigation performed before the testing. A microstructure modification of the W monoblock material due to the overheating of the surfaces and the copper interlayer structure modification were observed in the high heat flux area. The leakage points of the mock-ups that did not conclude the testing were localized in the middle of the monoblock while they were expected between two monoblocks. This paper reports the manufacturing route, the thermal fatigue testing, the pre and post non destructive examination and finally the results of the destructive examination performed on the monoblock small scale mock-ups.

  13. Simulation in full-scale mock-ups: an ergonomics evaluation method?

    DEFF Research Database (Denmark)

    Andersen, Simone Nyholm; Broberg, Ole

    2014-01-01

    This paper presents and exploratory study of four simulation sessions in full-scale mock-ups of future hospital facilities.......This paper presents and exploratory study of four simulation sessions in full-scale mock-ups of future hospital facilities....

  14. Development of control technology for HTTR hydrogen production system with mock-up test facility

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Inaba, Yoshitomo; Nishihara, Tetsuo; Takeda, Tetsuaki; Hayashi, Koji; Takada, Shoji; Inagaki, Yoshiyuki

    2006-01-01

    The Japan Atomic Energy Agency has been planning the demonstration test of hydrogen production with the High Temperature Engineering Test Reactor (HTTR). In a HTTR hydrogen production system (HTTR-H2), it is required to control a primary helium temperature within an allowable value at a reactor inlet to prevent a reactor scram. A cooling system for a secondary helium with a steam generator (SG) and a radiator is installed at the downstream of a chemical rector in a secondary helium loop in order to mitigate the thermal disturbance caused by the hydrogen production system. Prior to HTTR-H2, the simulation test with a mock-up test facility has been carried out to establish the controllability on the helium temperature using the cooling system against the loss of chemical reaction. It was confirmed that the fluctuations of the helium temperature at chemical reactor outlet, more than 200 K, at the loss of chemical reaction could be successfully mitigated within the target of ±10 K at SG outlet. A dynamic simulation code of the cooling system for HTTR-H2 was verified with the obtained test data

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

    International Nuclear Information System (INIS)

    Kupperman, D.S.; Bakhtiari, S.

    1997-01-01

    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

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

    International Nuclear Information System (INIS)

    Kupperman, D.S.; Bakhtiari, S.

    1996-10-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Campos, Wagner R.C.; Rabello, Emerson G.; Silva, Luiz L.; Mansur, Tanius R., E-mail: wrcc@cdtn.br, E-mail: egr@cdtn.br, E-mail: silvall@cdtn.br, E-mail: tanius@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte (Brazil). Servico de Integridade Estrutural; Martins, Ketsia S., E-mail: ketshinoda@hotmail.com [Universidade Federal de Minas Gerais (UFMG), Nelo Horizonte (Brazil). Departamento de Engenharia Metalurgica

    2015-07-01

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

    International Nuclear Information System (INIS)

    Campos, Wagner R.C.; Rabello, Emerson G.; Silva, Luiz L.; Mansur, Tanius R.; Martins, Ketsia S.

    2015-01-01

    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)

  19. Fabrication and testing of small scale mock-ups of ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Hatano, Toshihisa; Sato, Satoshi; Suzuki, Satoshi; Yokoyama, Kenji; Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Takatsu, Hideyuki; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1998-12-01

    Small scale mock-ups of the primary first wall, the baffle first wall, the shield block and a partial model for the edge of the primary first wall module were designed and fabricated incorporating most of the key design features of the ITER shielding blanket. All mock-ups featured the DSCu heat sink, the built-in SS coolant tubes within the heat sink and the SS shield block. CFC tiles was used as the protection armor for the baffle first wall mock-up. The small scale shield block mock-up, integrated with the first wall, was designed to have a poloidal curvature specified in the ITER design. Fabrication routes of mock-ups were decided based on the single step solid HIP of DSCu/DSCu, DSCu/SS and SS/SS reflecting the results of previous joining techniques development and testing. For attaching the CFC tiles onto DSCu heat sink in the fabrication of the baffle first wall mock-up, a two-step brazing was tried. All mock-ups and the partial model were successfully fabricated with a satisfactory dimensional accuracy. The small scale primary first wall mock-up was thermo-mechanically tested under high heat fluxes of 5-7 MW/m{sup 2} for 2500 cycles in total. Satisfactory heat removal performance and integrity of the mock-up against cyclic high heat flux loads were confirmed by measurement during the tests and destructive examination after the tests. Similar high heat flux tests were also performed with the small scale baffle first wall mock-up under 5-10 MW/m{sup 2} for 4500 cycles in total resulting in sufficient heat removal capability and integrity confirmed by measurements during the tests. (author)

  20. Fabrication and testing of small scale mock-ups of ITER shielding blanket

    International Nuclear Information System (INIS)

    Hatano, Toshihisa; Sato, Satoshi; Suzuki, Satoshi; Yokoyama, Kenji; Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Takatsu, Hideyuki; Ohara, Yoshihiro

    1998-12-01

    Small scale mock-ups of the primary first wall, the baffle first wall, the shield block and a partial model for the edge of the primary first wall module were designed and fabricated incorporating most of the key design features of the ITER shielding blanket. All mock-ups featured the DSCu heat sink, the built-in SS coolant tubes within the heat sink and the SS shield block. CFC tiles was used as the protection armor for the baffle first wall mock-up. The small scale shield block mock-up, integrated with the first wall, was designed to have a poloidal curvature specified in the ITER design. Fabrication routes of mock-ups were decided based on the single step solid HIP of DSCu/DSCu, DSCu/SS and SS/SS reflecting the results of previous joining techniques development and testing. For attaching the CFC tiles onto DSCu heat sink in the fabrication of the baffle first wall mock-up, a two-step brazing was tried. All mock-ups and the partial model were successfully fabricated with a satisfactory dimensional accuracy. The small scale primary first wall mock-up was thermo-mechanically tested under high heat fluxes of 5-7 MW/m 2 for 2500 cycles in total. Satisfactory heat removal performance and integrity of the mock-up against cyclic high heat flux loads were confirmed by measurement during the tests and destructive examination after the tests. Similar high heat flux tests were also performed with the small scale baffle first wall mock-up under 5-10 MW/m 2 for 4500 cycles in total resulting in sufficient heat removal capability and integrity confirmed by measurements during the tests. (author)

  1. 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.

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

    Science.gov (United States)

    Leenaers, A.; Van den Berghe, S.; Koonen, E.; Kuzminov, V.; Detavernier, C.

    2015-03-01

    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.

  3. 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.

  4. The text of the Agreement of 7 October 1983 between Cuba and the Agency for the application of safeguards in connection with the supply of a zero-power nuclear reactor from the Hungarian People's Republic

    International Nuclear Information System (INIS)

    1984-01-01

    The full text of the agreement of 7 October 1983 between Cuba and the Agency for the application of safeguards in connection with the supply of a zero-power nuclear reactor from the Hungarian People's Republic and to the nuclear material to be used therein to be supplied by the Union of Soviet Socialist Republics is presented

  5. Mock-up test on key components of ITER blanket remote handling system

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Matsumoto, Yasuhiro; Taguchi, Koh; Kozaka, Hiroshi; Shibanuma, Kiyoshi; Tesini, Alessandro

    2009-01-01

    The maintenance operation of the ITER in-vessel component, such as a blanket and divertor, must be executed by the remote equipment because of the high gamma-ray environment. During the Engineering Design Activity (EDA), the Japan Atomic Energy Agency (then called as Japan Atomic Energy Research Institute) had been fabricated the prototype of the vehicle manipulator system for the blanket remote handling and confirmed feasibility of this system including automatic positioning of the blanket and rail deployment procedure of the articulated rail. The ITER agreement, which entered into force in the last year, formally decided that Japan will procure the blanket remote handling system and the JAEA, as the Japanese Domestic Agency, is continuing several R and Ds so that the system can be procured smoothly. The residual key issues after the EDA are rail connection and cable handling. The mock-ups of the rail connection mechanism and the cable handling system were fabricated from the last year and installed at the JAEA Naka Site in this March. The former was composed of the rail connecting mechanism, two rail segments and their handling systems. The latter one utilized a slip ring, which implemented 80 lines for power and 208 lines for signal, because there is an electrical contact between the rotating spool and the fixed base. The basic function of these systems was confirmed through the mock-up test. The rail connection mechanism, for example, could accept misalignment of 1.5-2 mm at least. The future test plan is also mentioned in the paper.

  6. Fabrication data package for HEDL dosimetry in the ORNL Poolside Facility: LWR Pressure Vessel Mock-up irradiation

    International Nuclear Information System (INIS)

    Lippincott, E.P.; McElroy, W.N.; Kellogg, L.S.; Gold, R.; Guthrie, G.L.; Ruddy, F.H.; Ulseth, J.A.

    1981-09-01

    This document provides a complete description of the HEDL dosimetry inserted in the metallurgical specimen irradiation in the LWR Pressure Vessel Mock-up at the Oak Ridge Reactor Poolside Facility (PSF). This experiment is being conducted under the Nuclear Regulatory Commission sponsored program on Surveillance Dosimetry Improvement. The irradiation started April 1980 with recovery of the 2 x 10 19 (nominal fluence with E > 1 MeV) capsule in September 1980, the 4 x 10 19 surveillance capsule in November 1981 and the pressure vessel and void box capaules about August 1982

  7. Fabrication of mock-up with Be armour tiles diffusion bonded to the CuCrZr heat sink

    International Nuclear Information System (INIS)

    Moreschi, L.F.; Pizzuto, A.; Alessandrini, I.; Agostini, M.; Visca, E.; Merola, M.

    2001-01-01

    The aim of this work is the manufacture of high heat flux mock-ups with Be armour tiles on a CuCrZr heat sink for fabricating the beryllium section of the divertor vertical target (DVT) in the ITER reactor. Diffusion bonding between the CuCrZr bar and the beryllium tiles was obtained by inserting an aluminium interlayer to accommodate surface irregularities as well as to provide a compliant layer for accommodating thermal mismatches during both manufacturing and operation and cycles

  8. HTTR hydrogen production system. Structure and main specifications of mock-up test facility (Contract research)

    International Nuclear Information System (INIS)

    Kato, Michio; Aita, Hideki; Inagaki, Yoshiyuki; Hayashi, Koji; Ohashi, Hirofumi; Sato, Hiroyuki; Iwatsuki, Jin; Takada, Shoji; Inaba, Yoshitomo

    2007-03-01

    The mock-up test facility was fabricated to investigate performance of the steam generator for mitigation of the temperature fluctuation of helium gas and transient behavior of the hydrogen production system for HTTR and to obtain experimental data for verification of a dynamic analysis code. The test facility has an approximate hydrogen production capacity of 120Nm 3 /h and the steam reforming process of methane; CH 4 +H 2 O=3H 2 +CO, was used for hydrogen production of the test facility. An electric heater was used as a heat source instead of the reactor in order to heat helium gas up to 880degC (4MPa) at the chemical reactor inlet which is the same temperature as the HTTR hydrogen production system. Fabrication of the test facility was completed in February in 2002, and seven cycle operations were carried out from March in 2002 to December in 2004. This report describes the structure and main specifications of the test facility. (author)

  9. Study on control characteristics for HTTR hydrogen production system with mock-up test facility

    International Nuclear Information System (INIS)

    Inaba, Yoshitomo; Ohashi, Hirofumi; Nishihara, Tetsuo; Sato, Hiroyuki; Inagaki, Yoshiyuki; Takeda, Tetsuaki; Hayashi, Koji; Takada, Shoji

    2005-01-01

    The Japan Atomic Energy Research Institute has a demonstration test plan of a hydrogen production system by steam reforming of methane coupling with the High-Temperature Engineering Test Reactor (HTTR). Prior to the coupling of a hydrogen production plant with the HTTR, simulation tests with a mock-up test facility of the HTTR hydrogen production system (HTTR-H2) is underway. The test facility is a 1/30-scale of the HTTR-H2 and simulates key components downstream from an intermediate heat exchanger of the HTTR. The main objective of the simulation tests is the establishment and demonstration of control technology, focusing on the mitigation of a thermal disturbance to the reactor by a steam generator (SG) and on the controllability of the pressure difference between the helium and process gases at the reaction tube in a steam reformer (SR). It was confirmed that the fluctuation of the outlet helium gas temperature at the SG and the pressure difference in the SR can be controlled within the allowable range for the HTTR-H2 in the case of the system controllability test for the fluctuation of chemical reaction. In addition, a dynamic simulation code for the HTTR-H2 was verified with the obtained test data

  10. Reactor power control device

    International Nuclear Information System (INIS)

    Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

    1998-01-01

    The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

  11. Increased SRP reactor power

    International Nuclear Information System (INIS)

    MacAfee, I.M.

    1983-01-01

    Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

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

    Energy Technology Data Exchange (ETDEWEB)

    Hernández, F., E-mail: francisco.hernandez@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (INR) (Germany); Kolb, M. [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-WPT) (Germany); Ilić, M.; Kunze, A. [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (INR) (Germany); Németh, J. [KFKI Research Institute for Particle and Nuclear Physics (Hungary); Weth, A. von der [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (INR) (Germany)

    2013-10-15

    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 Li{sub 4}SiO{sub 4} 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

  13. Guide to power reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-07-15

    The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

  14. Dismantling Experiment of Mock-up Tube Bundle of Steam Generator

    International Nuclear Information System (INIS)

    Kim, Sung Kyun; Lee, Kune Woo

    2010-01-01

    A SG (steam generator) is one of the biggest decommissioning components in nuclear power plants and one has been replaced 2∼6 times during the whole operation of a nuclear power plant. The old SG should be decommissioned for the purpose of the volume reduction of radioactive waste. Among the components of SG, the tube bundle is one of the most difficult items to be dismantled due to the fact that it is very hard to cut since it is made of Inconel 600 which has high resistance of corrosion and abrasion. Moreover, All cutting process should be performed by remotely since radioactive contamination of the internal surface of SG tubes is very high (about 150,000∼300,000 Bq/cm 2 ). Therefore, it is necessary to choose the appropriate cutting methods by the pros and cons analysis for candidate dismantling technologies and to do experiment study for the validation. In this study, the results of cutting experiment for a mock-up bundle by using band saw cutting method are described herein

  15. Water Mock-up for the Sodium Waste Treatment Process

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Ho Yun; Kim, Jong Man; Kim, Byung Ho; Lee, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    It is important to safely treat the waste sodium which was produced from the sodium cooled fast reactors and the sodium facilities. About 1.3 tons of sodium waste has accumulated at KAERI from the sodium experiments which have been carried out since 1990. Also, large scaled sodium experiments are scheduled to verify the design of the sodium cooled fast reactor. As a treatment method for the waste sodium produced at the sodium facility, an investigation of the reaction procedure of the waste sodium with the sodium hydroxide aqueous has been developed. The NOAH process was developed in France for the treatment of waste sodium produced from sodium facilities and reactors. In the NOAH process, a small amount of sodium waste is continuously injected into the upper space which is formed on the free surface of the aqueous and slowly reacted with sodium hydroxide aqueous. Since the density of the sodium is lower than that of the aqueous, the injected sodium waste sometimes accumulates above the free surface of the sodium hydroxide aqueous, and its reaction rate becomes slow or suddenly increases. In the improved process, the sodium was injected into a reaction vessel filled with a sodium hydroxide aqueous through an atomizing nozzle installed on a lower level than that of the aqueous to maintain the reaction uniformly. Fig.1 shows the sodium waste process which was proposed in KAERI. The aqueous is composed of 60% sodium hydroxide, and its temperature is about 60 .deg. C. The process is an exothermic reaction. The hydrogen gas is generated, and the concentration of the sodium hydroxide increases in this process. It needs several systems for the process, i.e. a waste sodium injection, a cooling of the aqueous, hydrogen ventilation, and neutralization with nitric acid. The atomizing nozzle was designed to inject the sodium with the nitrogen gas which supplies a heat to the sodium to prevent its solidification and to uniformly mix the sodium with the aqueous. There are

  16. 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)

    1995-10-01

    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

  17. 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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-10-01

    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.

  18. Pre-brazed casting and hot radial pressing: A reliable process for the manufacturing of CFC and W monoblock mock-ups

    Energy Technology Data Exchange (ETDEWEB)

    Visca, Eliseo [Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, Via E. Fermi, 45, IT-00044 Frascati, RM (Italy)], E-mail: visca@frascati.enea.it; Libera, S.; Mancini, A.; Mazzone, G.; Pizzuto, A. [Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, Via E. Fermi, 45, IT-00044 Frascati, RM (Italy); Testani, C. [CSM S.p.A., IT-00128 Castel Romano, RM (Italy)

    2007-10-15

    ENEA is involved in the European International Thermonuclear Experimental Reactor (ITER) R and D activities and, in particular, for the manufacturing of high heat flux plasma-facing components (HHFC), such as the divertor targets, the baffles and the limiters. During last years, ENEA has manufactured actively cooled mock-ups by using different technologies, namely brazing, diffusion bonding and hot isostatic pressing (HIPping). A new manufacturing process has been set up and tested. It was successfully applied for the manufacturing of W armoured monoblock mock-ups. This technique is the HRP (hot radial pressing) based on performing a radial diffusion bonding between the cooling tube and the armour tile by pressurizing only internal tube and by keeping the joining zone in vacuum at the required bonding temperature. The heating is obtained by a standard air furnace. The HRP technique is now used for the manufacturing of CFC armoured monoblock components. For this purpose, some issues have to be faced, like the low CFC tensile strength, the pure copper interlayer between the heat sink and the armour necessary to mitigate the stress at the joint interface, and the low wettability of the pure copper on the CFC matrix. This paper reports the research path followed to manufacture a medium scale vertical target CFC and W armoured mock-up by HRP. A casting of a soft copper interlayer between the tube and the tile was obtained by a new technique: the pre-brazed casting (PBC, ENEA patent). Some preliminary mock-ups with three NB31 CFC tiles were successfully manufactured and tested to thermal fatigue using electron beam facilities. They all reached at least 1000 cycles at 20 MW/m{sup 2} without suffering any damage. The manufactured medium scale vertical target mock-up is now under testing at the FE2000 (France) facility. These activities were performed in the frame of ITER-EFDA contracts.

  19. Pre-brazed casting and hot radial pressing: A reliable process for the manufacturing of CFC and W monoblock mock-ups

    International Nuclear Information System (INIS)

    Visca, Eliseo; Libera, S.; Mancini, A.; Mazzone, G.; Pizzuto, A.; Testani, C.

    2007-01-01

    ENEA is involved in the European International Thermonuclear Experimental Reactor (ITER) R and D activities and, in particular, for the manufacturing of high heat flux plasma-facing components (HHFC), such as the divertor targets, the baffles and the limiters. During last years, ENEA has manufactured actively cooled mock-ups by using different technologies, namely brazing, diffusion bonding and hot isostatic pressing (HIPping). A new manufacturing process has been set up and tested. It was successfully applied for the manufacturing of W armoured monoblock mock-ups. This technique is the HRP (hot radial pressing) based on performing a radial diffusion bonding between the cooling tube and the armour tile by pressurizing only internal tube and by keeping the joining zone in vacuum at the required bonding temperature. The heating is obtained by a standard air furnace. The HRP technique is now used for the manufacturing of CFC armoured monoblock components. For this purpose, some issues have to be faced, like the low CFC tensile strength, the pure copper interlayer between the heat sink and the armour necessary to mitigate the stress at the joint interface, and the low wettability of the pure copper on the CFC matrix. This paper reports the research path followed to manufacture a medium scale vertical target CFC and W armoured mock-up by HRP. A casting of a soft copper interlayer between the tube and the tile was obtained by a new technique: the pre-brazed casting (PBC, ENEA patent). Some preliminary mock-ups with three NB31 CFC tiles were successfully manufactured and tested to thermal fatigue using electron beam facilities. They all reached at least 1000 cycles at 20 MW/m 2 without suffering any damage. The manufactured medium scale vertical target mock-up is now under testing at the FE2000 (France) facility. These activities were performed in the frame of ITER-EFDA contracts

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

    International Nuclear Information System (INIS)

    Lapena, J.; Sol Garcia-Redondo, M. del; Perosanz, F.J.; Saez, A.; Gomez-Briceno, D.; Castelao, C.

    2015-01-01

    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)

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

    International Nuclear Information System (INIS)

    Dumortier, P.; Lamalle, P.; Messiaen, A.; Vervier, M.

    2006-01-01

    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

  2. Experiment and analysis of hypervapotron mock-ups for preparing the 2nd qualification of the ITER blanket first wall

    International Nuclear Information System (INIS)

    Lee, Dong Won; Bae, Young Dug; Kim, Suk Kwon; Bang, In Cheol

    2010-01-01

    According to the increased heat flux condition up to 5 MW/m 2 in the International Thermonuclear Experimental Reactor (ITER), new design of the blanket first wall (FW) has been considered and the analysis was performed with ANSYS-CFX for checking its temperature with the ITER operation conditions. And a semi-prototype of the FW was proposed to be tested with the similar heat flux conditions under the second qualification for the FW procurement. In order to investigate the fabrication procedure and analysis capability of the code, two types of mock-up were fabricated according to the current semi-prototype design except for bending shape; one with hypervapotron and another without it. They were tested with KoHLT-2 (Korea Heat Load Test) facility and the results were compared with the ones by CFX code. The mass flow rate of inlet coolant was the same as the ITER condition and heat flux was loaded up to 0.48 MW/m 2 heat flux. The results show that the temperature of the mock-up can be predicted using the CFX code even with the complex geometry and the hypervapotron shows its function to increase the cooling.

  3. Reactor power measuring device

    International Nuclear Information System (INIS)

    Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

    1993-01-01

    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.)

  4. Mock-up test of remote controlled dismantling apparatus for large-sized vessels (contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Myodo, Masato; Miyajima, Kazutoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Okane, Shogo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2001-03-01

    The Remote dismantling apparatus, which is equipped with multi-units for functioning of washing, cutting, collection of cut pieces and so on, has been constructed to dismantle the large-sized vessels in the JAERI's Reprocessing Test Facility (JRTF). The apparatus has five-axis movement capability and its operation is performed remotely. The mock-up tests were performed to evaluate the applicability of the apparatus to actual dismantling activities by using the mock-ups of LV-3 and LV-5 in the facility. It was confirmed that each unit was satisfactory functioned by remote operation. Efficient procedures for dismantling the large-sized vessel was studied and various date was obtained in the mock-up tests. This apparatus was found to be applicable for the actual dismantling activity in JRTF. (author)

  5. Mock-up qualification and prototype manufacture for ITER current leads

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Tingzhi, E-mail: tingszhou@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Lu, Kun; Ran, Qingxiang; Ding, Kaizhong; Feng, Hansheng; Wu, Huan; Liu, Chenglian; Song, Yuntao [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Niu, Erwu [CNDA, Ministry of Science and Technology, Beijing (China); Bauer, Pierre; Devred, Arnaud [Magnet Division, ITER Organization, Cadarache (France)

    2015-10-15

    Highlights: • Vacuum brazing and electron beam welding qualification. • Machine and assembly strategy of fin type heat exchanger. • Soldering and joint resistance test of superconducting joint. • Pre-preg technology with vacuum bag on insulation. - Abstract: Three types of high temperature superconducting current leads (HTSCL) are designed to carry 68 kA, 55 kA or 10 kA to the ITER magnets. Before the supply of the HTS current lead series, the design and manufacturing process is qualified through mock-ups and prototypes. Seven mock-ups, representing the critical technologies of the current leads, were built and tested successfully in the Institute of Plasma Physics of the Chinese Academy of Sciences (ASIPP) in 2013. After the qualification some design features of the HTS leads were updated. This paper summarizes the qualification through mock-ups. In 2014 ASIPP started the manufacture of the prototypes. The preparation and manufacturing process are also described.

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

    International Nuclear Information System (INIS)

    Liu, C.; Nielsen, R.W.; Kruy, T.L.; Shu, D.; Kuzay, T.M.

    1994-01-01

    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

  7. Thermo-mechanical tests of a CFC divertor mock-up

    International Nuclear Information System (INIS)

    Cardella, A.; Akiba, M.; Duwe, R.; Di Pietro, E.; Suzuki, S.; Satoh, K.; Reheis, N.

    1994-01-01

    Thermo-mechanical tests have been performed on a divertor mock-up consisting of a metallic tube armoured with five carbon fibre composite tiles. The tube is inserted the tiles and brazed with TiCuSil braze (monoblock concept). The tube material is TZM, a molybdenum alloy, and the armour material is SEP CARB N112, a high conductivity carbon-carbon composite. Using special surface preparation consisting of laser drilling, small (≅ 500 μm) holes in the composite have been made to increase the surface wetted by the braze and the resistance. The mock-up has been tested at the JAERI 400 kW electron beam test facility JEBIS. The aim of the test was to assess the performance of the mock-up in screening and thermal fatigue tests with particular attention to the behaviour of the armour to heat sink joint. (orig.)

  8. Mock-up test of remote controlled dismantling apparatus for large-sized vessels (contract research)

    International Nuclear Information System (INIS)

    Myodo, Masato; Miyajima, Kazutoshi; Okane, Shogo

    2001-03-01

    The Remote dismantling apparatus, which is equipped with multi-units for functioning of washing, cutting, collection of cut pieces and so on, has been constructed to dismantle the large-sized vessels in the JAERI's Reprocessing Test Facility (JRTF). The apparatus has five-axis movement capability and its operation is performed remotely. The mock-up tests were performed to evaluate the applicability of the apparatus to actual dismantling activities by using the mock-ups of LV-3 and LV-5 in the facility. It was confirmed that each unit was satisfactory functioned by remote operation. Efficient procedures for dismantling the large-sized vessel was studied and various date was obtained in the mock-up tests. This apparatus was found to be applicable for the actual dismantling activity in JRTF. (author)

  9. ITER baffle module small-scale mock-ups: first wall thermo-mechanical testing results

    International Nuclear Information System (INIS)

    Severi, Y.; Giancarli, L.; Poitevin, Y.; Salavy, J.F.; Le Marois, G.; Roedig, M.; Vieider, G.

    1998-01-01

    The EU-home team is in charge of the R and D related to the ITER baffle first wall. Five small-scale mock-ups, using Be, CFC and W tiles and different armour/heat-sink material joints under development, have been fabricated and thermomechanically tested in FE200 (Le Creusot) and JUDITH (Juelich) electron beam facilities. The small-scale mock-ups have been submitted to thermo-mechanical fatigue tests (up to failure using accelerating techniques). The objective was to determine the performances of the armour material joints under high heat flux cycles. (orig.)

  10. Fabrication of ITER first wall mock-ups with beryllium armour

    International Nuclear Information System (INIS)

    Mohri, K.; Nomoto, Y.; Uda, M.; Enoeda, M.; Akiba, M.

    2004-01-01

    This paper presents the fabric ability development for the ITER first wall through the fabrication of a real size first wall panel mock-up without beryllium armor and a partial mock-up of the first wall panel with beryllium armor. Microscopic observation and mechanical test of the hot isostatic pressed Be/Cu-alloy joints were also performed of which results showed good bond ability of the joints. Finally the fabrication procedure of the ITER first wall panel has been established. (author)

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

    International Nuclear Information System (INIS)

    Tsuruga, Keisuke

    2010-01-01

    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)

  12. Status of the BR2 refurbishment programme

    International Nuclear Information System (INIS)

    Koonen, E.

    1995-01-01

    The operation of the BR2 reactor with its second beryllium matrix is foreseen up to mid-1995. A refurbishment programme has been established in order to allow for future operation during at least ten years. Recently a positive decision to effectively carry out this programme has been taken. The refurbishment action plan follows from a general assessment of the different systems of BR2, with respect to their actual status, the operational experience and the evolution of safety standards and criteria. Ageing considerations were of uppermost importance in those assessments, not only to assure safety of future operation, but also to guarantee future availability and reliability. (orig.)

  13. Experimental Investigation Into Thermal Siphon Used as an Intermediate Circuit of an Integrated Cooling System Reactor

    International Nuclear Information System (INIS)

    Adamovich, L.A.; Gabaraev, B.A.; Solovjev, S.L.; Shpansky, S.B.

    2002-01-01

    In the paper the results of study in heat transfer capacity of the thermosyphon mock-up which is considered as an intermediate circuit of the reactor under design, are presented. The mock-up design, the test rig and the experimental results are described. It is shown that the simplest mathematical model describes the processes of power transfer by the thermosyphon under certain conditions. (authors)

  14. The review of the reactor physics experiments carried out on the LR-0 research reactor NRI Rez plc for reactors of the VVER type

    International Nuclear Information System (INIS)

    Hudec, Frantisek; Jansky, Bohumil; Juricek, Vlastimil; Mikus, Jan; Novak, Evzen; Osmera, Bohumil; Posta, Severin; Rypar, Vojtech; Svadlenkova, Marie

    2010-01-01

    LR-0 is an experimental zero power reactor mainly used for the determination of the neutron-physical characteristics of WWER and PWR type reactor lattices and shielding with UO2 or MOX fuel. Its major assets include capability to design and operate multizone cores, i.e. substituted cores, with an inner inserted part in hexagonal or square geometry (driven by LR-0 standard assemblies); Standard and special supporting plates for mock-up experiments; special supporting plates, which enables the triangular symmetrical assembly arrangement with an arbitrary pitch; Modeling neutron field parameters of power reactors; Wide range benchmarking possibilities, with high reproducibility of the benchmark design parameters; Wide range of measurement techniques including equipment and experienced personal; Flexible rearrangements of the core. The main experiments included: Pin wise flux distribution measurements; VVER-440 and VVER-1000 mock-ups; compact spent fuel storage; space kinetics experiment; core parameters experimental determination; experiment with new design fuel assembly; WWER-440 control assembly influence; and burnable absorber study. International research projects are also described. (P.A.)

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

    International Nuclear Information System (INIS)

    Koonen, E.

    2006-01-01

    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. Weld distortion prediction and control of the ITER vacuum vessel manufacturing mock-ups

    International Nuclear Information System (INIS)

    Ottolini, Marco; Barbensi, Andrea

    2014-01-01

    The fabrication of the ITER Vacuum Vessel Sectors is an unprecedented challenge, due to their dimensions, the close tolerances, the complex 'D' shape. The technological issues were faced by the production of full scale mock ups to confirm the manufacturing feasibility to achieve very tight tolerances and qualify the main manufacturing processes, by a step by step welding distortion control, by the qualification of not conventional NDT inspection techniques and by innovative 3D dimensional inspections. The Supplier is required to fabricate at least two mock ups, inboard and outboard, related to the manufacturing method of the VV Sectors, to demonstrate the control of the welding distortions to achieve tolerances, optimizing welding sequences and calibrating of welding distortions computer simulations. The stages of this preparatory activity are: prediction of welding distortion for fabrication mock ups representative of selected segments; demonstration that distortion predictions are consistent with experimental results from 3D dimensional inspection; understanding of reasons of possible deviations between numerical and experimental results and definition of action to solve these issues; demonstration that possible calculation simplifications, adopted to speed up the analysis process, do not affect significantly the welding distortion prediction. This paper describes the weld distortion prediction and control on the manufacturing mock-ups of ITER Vacuum Vessel Sectors, with particular emphasis to the lessons learned. (authors)

  17. F.B.R. Core mock-up RAPSODIE- I: Experimental analysis

    International Nuclear Information System (INIS)

    Brochard, D.; Buland, P.; Gantenbein, F.

    1990-01-01

    The main phenomena which influence the LMFBR core response to a seismic excitation are the fluid structure interaction and the impacts between subassemblies. To study the core behaviour, seismic tests have been performed on the core mock-up RAPSODIE with or without fluid and restraint ring and for different levels of excitation. This paper summarizes the results of these tests

  18. In-Pile thermal fatigue of First Wall mock-ups under ITER relevant conditions

    International Nuclear Information System (INIS)

    Blom, F.; Schmalz, F.; Kamer, S.; Ketema, D.J.

    2006-01-01

    The objective of this study is to perform in-pile thermal fatigue testing of three actively cooled First Wall (FW) mock-ups to check the effect of neutron irradiation on the Be/CuCrZr joints under representative FW operation conditions. Three FW mock-ups with Beryllium armor tiles will be neutron irradiated at 1 dpa (in Be) with parallel thermal fatigue testing for 30,000 cycles. The temperatures, stress distributions and stress amplitudes at the Be/CuCrZr interface of the mock-ups will be as close as possible to the values calculated for ITER FW panels. For this objective the PWM mocks-up subjected to thermal fatigue will be integrated with high density (W) plates on the Be-side to provide heat flux by nuclear heating. The assembly will be placed in the pool-side facility of the HFR and thermal cycling is then arranged by mechanical movement towards and from the core box. As the thermal design of the irradiation rig is very critical a pilot-irradiation will be performed to cross check the models used in the thermal design of the rig. The project is currently in the design phase of both the pilot and actual irradiation rig. The irradiation of the actual rig is planned to start at mid 2007 and last for two years. (author)

  19. F.B.R. Core mock-up RAPSODIE - II - numerical models

    International Nuclear Information System (INIS)

    Brochard, D.; Hammami, L.; Gantenbein, F.

    1990-01-01

    To study the behaviour of LMFBR cores excited by a seism, tests have been performed on the RAPSODIE core mock-up. The aim of this paper is to present the numerical models used to interprete these tests and the comparisons between calculations and experimental results

  20. Reactor power monitoring device

    International Nuclear Information System (INIS)

    Dogen, Ayumi; Ozawa, Michihiro.

    1983-01-01

    Purpose: To significantly improve the working efficiency of a nuclear reactor by reflecting the control rod history effect on thermal variants required for the monitoring of the reactor operation. Constitution: An incore power distribution calculation section reads the incore neutron fluxes detected by neutron detectors disposed in the reactor to calculate the incore power distribution. A burnup degree distribution calculation section calculates the burnup degree distribution in the reactor based on the thus calculated incore power distribution. A control rod history date store device supplied with the burnup degree distribution renews the stored control rod history data based on the present control rod pattern and the burnup degree distribution. Then, thermal variants of the nuclear reactor are calculated based on the thus renewed control rod history data. Since the control rod history effect is reflected on the thermal variants required for the monitoring of the reactor operation, the working efficiency of the nuclear reactor can be improved significantly. (Seki, T.)

  1. Fabrication of divertor mock-up with ODS-Cu and W by the improved brazing technique

    Science.gov (United States)

    Tokitani, M.; Hamaji, Y.; Hiraoka, Y.; Masuzaki, S.; Tamura, H.; Noto, H.; Tanaka, T.; Muroga, T.; Sagara, A.; FFHR Design Group

    2017-07-01

    Copper alloy has been considered as a divertor cooling tube or heat sink not only in the helical reactor FFHR-d1 but also in the tokamak DEMO reactor, because it has a high thermal conductivity. This work focused on applying an oxide dispersion strengthened copper alloy (ODS-Cu), GlidCop® (Cu-0.3 wt%Al2O3) as the divertor heat sink material of FFHR-d1. This alloy has superior high temperature yield strength exceeding 300 MPa at room temperature even after annealing up to ~1000 °C. The change in material properties of Pure-Cu, GlidCop® and CuCrZr by neutron irradiation are summarized in this paper. A primary dose limit is the radiation-induced hardening/softening (~0.2 dpa/1-2 dpa) which has a temperature dependence. According to such an evaluation, the GlidCop® can be selected as the current best candidate material in the commercial base of the divertor heat sink, and its temperature should be maintained as close as possible to 300 °C during operation. Bonding between the W armour and the GlidCop® heat sink was successfully performed by using an improved brazing technique with BNi-6 (Ni-11%P) filler material. The bonding strength was measured by a three-point bending test and reached up to approximately 200 MPa. Surprisingly, several specimens showed an obvious yield point. This means that the BNi-6 brazing (bonding) layer caused relaxation of the applied stress. The small-scale divertor mock-up of the W/BNi-6/GlidCop® was successfully fabricated by using the improved brazing technique. The heat loading test was carried out by the electron beam device ACT2 in NIFS. The mock-up showed an excellent heat removal capability for use in the FFHR-d1 divertor.

  2. High heat flux testing of EU tungsten monoblock mock-ups for the ITER divertor

    International Nuclear Information System (INIS)

    Gavila, P.; Riccardi, B.; Pintsuk, G.; Ritz, G.; Kuznetsov, V.; Durocher, A.

    2015-01-01

    Highlights: • All the tested items sustained the ITER Full W divertor qualification program requirements. This confirms that the technology for the manufacturing of the first set of the ITER Divertor is available in Europe. • The surface roughening and local melting of the W surface under high heat flux was proven to be significantly reduced for an armour thickness lower or equal to 6 mm. • However, this campaign highlighted some specific areas of improvement to be implemented ideally before the upcoming ITER Divertor IVT serial production. • The issue of the self-castellation of the W monoblocks, which typically appears after a few tenths of cycles at 20 MW/m"2, is critical because it generates some uncontrolled defects at the amour to heat sink joints. Besides, they create a gap which exposure is almost perpendicular to the magnetic field lines and which might lead to local W melting in the strike point region. • This campaign also evidenced that the minimum IO requirements on the CuCrZr ductility could be revised to avoid the occurrence of rather early fatigue failures. Although the W material characterization program has been set up by the IO, the strategy on the CuCrZr still needs to be defined. - Abstract: With the aim to assess the option to start the ITER operation with a full tungsten divertor, an R&D program was launched in order to evaluate the performances of tungsten (W) armoured plasma facing components (PFCs) under high heat flux. The F4E program consisted in the manufacturing and high heat flux (HHF) testing of W monoblock mock-ups and medium scale prototypes up to 20 MW/m"2. During the test campaign, 26 W mock-ups and two medium scale prototypes manufactured by Plansee SE (Austria) and by Ansaldo Nucleare (Italy) have been tested at the FE200 (AREVA, Le Creusot, France) and ITER Divertor Test Facility (IDTF) (Efremov Institute Saint Petersburg, Russian Federation) electron beam test facilities. The high heat flux (HHF) testing program

  3. High heat flux testing of EU tungsten monoblock mock-ups for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Gavila, P., E-mail: pierre.gavila@f4e.europa.eu [Fusion for Energy, 08019 Barcelona (Spain); Riccardi, B. [Fusion for Energy, 08019 Barcelona (Spain); Pintsuk, G. [Forschungszentrum Juelich, 52425 Juelich (Germany); Ritz, G. [AREVA NP, Centre Technique France, 71205 Le Creusot (France); Kuznetsov, V. [JCS “Efremov Institute”, Doroga na Metallostroy 3, Metallostroy, Saint-Petersburg 196641 (Russian Federation); Durocher, A. [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 Saint Paul-lez-Durance (France)

    2015-10-15

    Highlights: • All the tested items sustained the ITER Full W divertor qualification program requirements. This confirms that the technology for the manufacturing of the first set of the ITER Divertor is available in Europe. • The surface roughening and local melting of the W surface under high heat flux was proven to be significantly reduced for an armour thickness lower or equal to 6 mm. • However, this campaign highlighted some specific areas of improvement to be implemented ideally before the upcoming ITER Divertor IVT serial production. • The issue of the self-castellation of the W monoblocks, which typically appears after a few tenths of cycles at 20 MW/m{sup 2}, is critical because it generates some uncontrolled defects at the amour to heat sink joints. Besides, they create a gap which exposure is almost perpendicular to the magnetic field lines and which might lead to local W melting in the strike point region. • This campaign also evidenced that the minimum IO requirements on the CuCrZr ductility could be revised to avoid the occurrence of rather early fatigue failures. Although the W material characterization program has been set up by the IO, the strategy on the CuCrZr still needs to be defined. - Abstract: With the aim to assess the option to start the ITER operation with a full tungsten divertor, an R&D program was launched in order to evaluate the performances of tungsten (W) armoured plasma facing components (PFCs) under high heat flux. The F4E program consisted in the manufacturing and high heat flux (HHF) testing of W monoblock mock-ups and medium scale prototypes up to 20 MW/m{sup 2}. During the test campaign, 26 W mock-ups and two medium scale prototypes manufactured by Plansee SE (Austria) and by Ansaldo Nucleare (Italy) have been tested at the FE200 (AREVA, Le Creusot, France) and ITER Divertor Test Facility (IDTF) (Efremov Institute Saint Petersburg, Russian Federation) electron beam test facilities. The high heat flux (HHF) testing

  4. Reactor power control device

    International Nuclear Information System (INIS)

    Doi, Kazuyori.

    1981-01-01

    Purpose: To automatically control the BWR type reactor power by simple and short-time searching the load pattern nearest to the required pattern at a nuclear power plant side. Constitution: The reactor power is automatically regulated by periodical modifying of coefficients fitting to a reactor core model, according as a required load pattern. When a load requirement pattern is given, a simulator estimates the total power change and the axial power distribution change from a xenon density change output calculated by a xenon dynamic characteristic estimating device, and a load pattern capable of being realized is searched. The amount to be recirculated is controlled on the basis of the load patteren thus searched, and the operation of the BWR type reactor is automatically controlled at the side of the nuclear power plant. (Kamimura, M.)

  5. Thermal transient and the temperature profile in a HELICA mock-up simulated by a new finite element homogenous model

    International Nuclear Information System (INIS)

    Zaccari, Nicola; Aquaro, Donato

    2013-01-01

    Highlights: • We have developed a numerical model of the pebble beds is based on the results of a theoretical and experimental research activity performed. • The model has been used to simulate the experimental tests performed on HELICA mock-up (ENEA Italy). • Moreover the numerical results are compared with the experimental ones. Finally, a discussion on results obtained by other authors involved in the benchmark is reported. -- Abstract: This paper deals with a numerical approach for simulating the thermal and mechanical behaviour of pebble beds used as breeder and neutron multiplier in breeding blanket of nuclear fusion reactor. The model of the pebble beds is based on the results of a theoretical and experimental research activity performed by the Authors on ceramic pebble beds (lithium ortosilicate and lithium metatitanate). The results of this activity permitted to determine the effective thermal conductivity of the beds, versus the temperature and the axial pressure and to implement a homogenous model of pebble bed in a FEM code. This paper illustrates an application of the implemented model, considering pebble beds under several cycles of heating and cooling. The examined geometry corresponds to the HELICA mock-up tested by ENEA in the research centre Brasimone. The experimental tests performed on HELICA have been used as a benchmark problem in order to assess the different approaches for simulating pebble beds. In this paper, the simulations performed with two-dimensional models are illustrated. Moreover the numerical results are compared with the experimental ones. Finally, a discussion on results obtained by other authors involved in the benchmark is reported

  6. Reactor power distribution monitor

    International Nuclear Information System (INIS)

    Hoizumi, Atsushi.

    1986-01-01

    Purpose: To grasp the margin for the limit value of the power distribution peaking factor inside the reactor under operation by using the reactor power distribution monitor. Constitution: The monitor is composed of the 'constant' file, (to store in-reactor power distributions obtained from analysis), TIP and thermocouple, lateral output distribution calibrating apparatus, axial output distribution synthesizer and peaking factor synthesizer. The lateral output distribution calibrating apparatus is used to make calibration by comparing the power distribution obtained from the thermocouples to the power distribution obtained from the TIP, and then to provide the power distribution lateral peaking factors. The axial output distribution synthesizer provides the power distribution axial peaking factors in accordance with the signals from the out-pile neutron flux detector. These axial and lateral power peaking factors are synthesized with high precision in the three-dimensional format and can be monitored at any time. (Kamimura, M.)

  7. Production of radioisotopes with BR2 facilities

    International Nuclear Information System (INIS)

    Fallais, C.J.; Morel de Westfaver, A.; Heeren, L.; Baugnet, J.M.; Gandolfo, J.M.; Boeykens, W.

    1978-01-01

    After a brief account on the isotopes production evolution in the industrialized countries the irradiation devices and the types of standardized capsules used in the BR2 reactor are described as well as the thermal neutron flux. Production of most important radioisotopes like 131 Iodine, 60 Cobalt, 192 Iridium and 99 Molybdenum and their main utilizations (uses)are described. The mean specific activities and the limit of use for different radioisotopes are reported. (A.F.)

  8. Reactor power control system

    International Nuclear Information System (INIS)

    Tomisawa, Teruaki.

    1981-01-01

    Purpose: To restore reactor-power condition in a minimum time after a termination of turbine bypass by reducing the throttling of the reactor power at the time of load-failure as low as possible. Constitution: The transient change of the internal pressure of condenser is continuously monitored. When a turbine is bypassed, a speed-control-command signal for a coolant recirculating pump is generated according as the internal pressure of the condenser. When the signal relating to the internal pressure of the condenser indicates insufficient power, a reactor-control-rod-drive signal is generated. (J.P.N.)

  9. Measurement and control system for the ITER remote handling mock-up test

    International Nuclear Information System (INIS)

    Oka, K.; Kakudate, S.; Takiguchi, Y.; Ako, K.; Taguchi, K.; Tada, E.; Ozaki, F.; Shibanuma, K.

    1998-01-01

    The mock-up test platforms composed of full-scale remote handling (RH) equipment were developed for demonstrating remote replacement of the ITER blanket and divertor. In parallel, the measurement and control system for operating these RH equipment were constructed on the basis of open architecture with object oriented feature, aiming at realization of fully-remoted automatic operation required for ITER. This paper describes the design concept of the measurement and control system for the remote handling equipment of ITER, and outlines the measured performances of the fabricated measurement system for the remote handling mock-up tests, which includes Data Acquisition System (DAS), Visual Monitoring System (VMS) and Virtual Reality System (VRS). (authors)

  10. 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.

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

    International Nuclear Information System (INIS)

    Lottin, J.; Brunetti, L.; Formosa, F.; Adloff, C.; Bastian, Y.; Bolzon, B.; Cadoux, F.; Geffroy, N.; Girard, C.; Jeremie, A.; Karyotakis, Y.; Peltier, F.

    2006-01-01

    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

  12. Fabrication of small mock-ups for the KO HCCR TBM

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Jae Sung; Kim, Suk Kwon; Lee, Eo Hwak; Jin, Hyung Gon; Lee, Dong Won [KAERI, Daejeon (Korea, Republic of); Cho, Seung Yon [NFRI, Daejeon (Korea, Republic of)

    2016-05-15

    A fabrication procedure for the manufacturing of the HCCR TBM sub-module was performed and small mock-ups were fabricated using an E-beam and laser beam weld to verify the manufacturing procedure and method of the HCCR TBM sub-module. To establish and optimize the welding procedure in an E-beam weld from ARAA material, the distortion and radiographic tests were carried out from the E-beam weld results. It could be noted that a small amount of distortion occurred, but the values are small enough to neglect for the fabrication. In addition, a helium leak test and water pressure test will be performed for verification of the fabricated small mock-ups.

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

    Energy Technology Data Exchange (ETDEWEB)

    Lottin, J.; Brunetti, L.; Formosa, F. [Universite de Savoie, ESIA, 74 - Annecy (France); Adloff, C.; Bastian, Y.; Bolzon, B.; Cadoux, F.; Geffroy, N.; Girard, C.; Jeremie, A.; Karyotakis, Y.; Peltier, F. [LAPP-IN2P3-CNRS, 74 - Annecy-le-Vieux (France)

    2006-07-01

    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. Overview of the EU small scale mock-up tests for ITER high heat flux components

    International Nuclear Information System (INIS)

    Vieider, G.; Barabash, V.; Cardella, A.

    1998-01-01

    This task within the EU R and D for ITER was aimed at the development of basic manufacturing solutions for the high heat flux plasma facing components such as the divertor targets, the baffles and limiters. More than 50 representative small-scale mock-ups have been manufactured with beryllium, carbon and tungsten armour using various joining technologies. High heat flux testing of 20 of these mock-ups showed the carbon mono-blocks to be the most robust solution, surviving 2000 cycles at absorbed heat fluxes of up to 24 MW m -2 . With flat armour tiles rapid joint failures occurred at 5-16 MW m -2 depending on joining technology and armour material. These test results serve as a basis for the selection of manufacturing options and materials for the prototypes now being ordered. (orig.)

  15. Power reactors operational diagnosis

    International Nuclear Information System (INIS)

    Dach, K.; Pecinka, L.

    1976-01-01

    The definition of reactor operational diagnostics is presented and the fundamental trends of research are determined. The possible sources of power reactor malfunctions, the methods of defect detection, the data evaluation and the analysis of the results are discussed in detail. In view of scarcity of a theoretical basis and of insufficient in-core instrumentation, operational diagnostics cannot be as yet incorporated in a computer-aided reactor control system. (author)

  16. Irradiation of novel MTR fuel plates in BR2

    International Nuclear Information System (INIS)

    Verboomen, B.; Aoust, Th.; Beeckmans De Westmeerbeeck, A.; De Raedt, Ch.

    2000-01-01

    Since the end of 1999, novel MTR fuel plates with very high-density meat are being irradiated in BR2. The purpose of the irradiation is to investigate the behaviour of these fuel plates under very severe reactor operation conditions. The novel fuel plates are inserted in two standard six-tube BR2 fuel elements in the locations normally occupied by the standard outer fuel plates. The irradiation in BR2 was prepared by carrying out detailed neutron Monte Carlo calculations of the whole BR2 core containing the two experimental fuel elements for various positions in the reactor and for various azimuthal orientations of the fuel elements. Comparing the thus determined fission density levels and azimuthal profiles in the new MTR fuel plates irradiated in the various channels allowed the experimenters to choose the most appropriate BR2 channel and the most appropriate fuel element orientation. (author)

  17. Tests on a mock-up of the feedback controlled matching options of the ITER ICRH system

    International Nuclear Information System (INIS)

    Grine, D.; Vervier, M.; Messiaen, A.; Dumortier, P.

    2009-01-01

    Automatic control of the matching of the ITER ICRH antenna array on a reference load is presently developed and tested for optimization on a low-powered scaled (1:5) mock-up. Resilience to fast load variations is obtained either by 4 Conjugate-T (CT) or 4 quadrature hybrid circuits, the latter being the reference option. The main results are (i) for the CT option: successful implementation of the simultaneous feedback control of 11 actuators for the matching of the 4 CT and for the control of the array toroidal phasing; (ii) for the hybrid option: the matching and the array current control via feedback control of the decouplers and double stub tuners. This system is being progressively implemented and the simultaneous control of matching and antenna current has already been successfully tested on half of the array for heating and current drive phasings.

  18. Full scale mock-up tests for rod bundle thermal-hydraulics in Japan

    International Nuclear Information System (INIS)

    Sugawara, S.

    1995-01-01

    This poster describes tests aimed at development and validation of principal design methodology of rod bundle thermal-hydraulics correlations. The works are based on domestic data base using the full-scale mock-up test facilities. The scope of the tests comprises DNB heat flux, transient DNB heat flux, post DNB heat transfer, pressure drop and void distribution. The works have been performed under collaboration among electric facilities, NPP vendors, universities, governmental corporations. 1 tab., 14 figs

  19. ASTP crewmen in Soyuz orbital module mock-up during training session at JSC

    Science.gov (United States)

    1975-01-01

    An interior view of the Soyuz orbital module mock-up in bldg 35 during Apollo Soyuz Test Project (ASTP) joint crew training at JSC. The ASTP crewmen are Astronaut Vance D. Brand (on left), command module pilot of the American ASTP prime crew; and Cosmonaut Valeriy N. Kubasov, engineer on the Soviet ASTP first (prime) crew. The training session simulated activities on the second day in Earth orbit.

  20. Mock-up test results of monoblock-type CFC divertor armor for JT-60SA

    Energy Technology Data Exchange (ETDEWEB)

    Higashijima, S. [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)], E-mail: higashijima.satoru@jaea.go.jp; Sakurai, S.; Suzuki, S.; Yokoyama, K.; Kashiwa, Y.; Masaki, K.; Shibama, Y.K.; Takechi, M.; Shibanuma, K.; Sakasai, A.; Matsukawa, M.; Kikuchi, M. [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

    2009-06-15

    The JT-60 Super Advanced (JT-60SA) tokamak project starts under both the Japanese domestic program and the international program 'Broader Approach'. The maximum heat flux to JT-60SA divertor is estimated to {approx}15 MW/m{sup 2} for 100 s. Japan Atomic Energy Agency (JAEA) has developed a divertor armor facing high heat flux in the engineering R and D for ITER, and it is concluded that monoblock-type CFC divertor armor is promising for JT-60SA. The JT-60SA armor consists of CFC monoblocks, a cooling CuCrZr screw-tube, and a thin oxygen-free high conductivity copper (OFHC-Cu) buffer layer between the CFC monoblock and the screw-tube. CFC/OFHC-Cu and OFHC-Cu/CuCrZr joints are essential for the armor, and these interfaces are brazed. Needed improvements from ITER engineering R and D are good CFC/OFHC-Cu and OFHC-Cu/CuCrZr interfaces and suppression of CFC cracking. For these purposes, metalization inside CFC monoblock is applied, and we confirmed again that the mock-up has heat removal capability in excess of ITER requirement. For optimization of the fabrication method and understanding of the production yield, the mock-ups corresponding to quantity produced in one furnace at the same time is also produced, and the half of the mock-ups could remove 15 MW/m{sup 2} as required. This paper summarizes the recent progress of design and mock-up test results for JT-60SA divertor armor.

  1. Major accident analyses for experimental zero-power fast reactor assemblies; Analyse des accidents graves pouvant survenir dans les reacteurs experimentaux a neutrons rapides de puissance zero; Analiz krupnoj avarii dlya ehksperimental'ny kh reaktornykh ustanovok nulevoj moshchnosti na bystrykh nejtronakh; Analisis de los accidentes graves que pueden producirse en los reactores experimentales rapidos de potencia cero

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, G.; Barts, E. W.; Kapil, S.; Tomabechi, K. [Argonne National Laboratory, Argonne, IL (United States)

    1962-03-15

    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 U{sup 235}-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 U{sup 235}-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 U{sup 235} 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

  2. Thermal fatigue tests with actively cooled divertor mock-ups for ITER

    International Nuclear Information System (INIS)

    Roedig, M.; Duwe, R.; Linke, J.; Schuster, A.; Wiechers, B.; Ibbott, C.; Jacobson, D.; Le Marois, G.; Lind, A.; Lorenzetto, P.; Vieider, G.; Peacock, A.; Ploechl, L.; Severi, Y.; Visca, E.

    1998-01-01

    Mock-ups for high heat flux components with beryllium and CFC armour materials have been tested by means of the electron beam facility JUDITH. The experiments concerned screening tests to evaluate heat removal efficiency and thermal fatigue tests. CFC monoblocks attached to DS-Cu (Glidcop Al25) and CuCrZr tubes by active metal casting and Ti brazing showed the best thermal fatigue behaviour. They survived more than 1000 cycles at heat loads up to 25 MW m -2 without any indication of failure. Operational limits are given only by the surface temperature on the CFC tiles. Most of the beryllium mock-ups were of the flat tile type. Joining techniques were brazing, hot isostatic pressing (HIP) and diffusion bonding. HIPed and diffusion bonded Be/Cu modules have not yet reached the standards for application in high heat flux components. The limit of this production method is reached for heat loads of approximately 5 MW m -2 . Brazing with and without silver seems to be a more robust solution. A flat tile mock-up with CuMnSnCe braze was loaded at 5.4 MW m -2 for 1000 cycles without damage The first test with a beryllium monoblock joined to a CuCrZr tube by means of Incusil brazing shows promising results; it survived 1000 cycles at 4.5 MW m -2 without failure. (orig.)

  3. Destructive analysis on the ITER FW small scale mock-ups

    International Nuclear Information System (INIS)

    Wang, Pinghuai; Chen, Jiming; Liu, Danhua; Jin, Fanya; Yang, Bo

    2015-01-01

    As one of the core components of ITER, the first wall (FW) panel of shield blanket defines a physical boundary for the plasma transients and exhausts the majority of the plasma heat flux. China will undertake 12.64% of FW manufacturing tasks, and all of them are enhanced heat flux (EHF) components which will suffer surface heat flux of 4 - 5 MW/m 2 . The FW will be manufactured by a combination technology of explosion bonding CuCrZr alloy/316L (N) stainless steel plate and hot iso-static pressing (HIP) joining of beryllium tiles/CuCrZr alloy. The Be/Cu joint qualities is the key issue for the manufacturing of the FW panels. Several small scale mock-ups were manufactured for the qualification of the HIP technology for the FW. To avoid the brittle Be-Cu phase formed during the HIPing process, different thick Ti and pure Cu were coated on the beryllium tiles before HIPing to CuCrZr alloy. Ultrasonic testing was conducted on the mock-ups and destructive analysis was carried out on the mock-ups. For the failed ones, the results show that in the UT indication area brittle fracture occurs at the Be/Ti interface and then Ti/Cu interface in other areas. Based on these results, the manufacturing technology was improved mainly on the beryllium tiles quality, coating process and canister design. (author)

  4. The supply of small scale mock-ups of the primary wall module concepts for ITER

    International Nuclear Information System (INIS)

    Walsh, G.; Cheyne, K.; Lorenzetto, P.

    1998-01-01

    The present design of Blanket Shield and Primary Wall for ITER envisages construction of the wall with a water cooled, stainless steel outer layer and a water cooled, copper liner on the inside plasma facing surface. Protection of the inner copper surface with an armour layer is necessary to cope with plasma to wall interaction. There are a number of armour materials under consideration, for this project beryllium was used. The scope of work was to produce a series of mock-ups, each consisting of a different combination of materials, which included Dispersion Strengthened Copper, Copper-Chrome-Zirconium alloy, Beryllium and Stainless Steel. Hot Isostatic Pressing (HIP) was the method used to ensure that a fully diffused bonded joint was achieved giving the necessary strength and thermal conductivity. The first five of the mock ups have been successfully completed and are being tested at the various laboratories in Europe. The remaining mock ups are awaiting the results of this test work prior to being completed. (authors)

  5. Beryllium mock-ups development and ultrasonic testing for ITER divertor conditions

    International Nuclear Information System (INIS)

    Barabash, V.R.; Bykov, V.A.; Giniyatulin, R.N.; Gervash, A.A.; Gurieva, T.M.; Egorov, K.E.; Komarov, V.L.; Korolkov, M.D.; Mazul, I.V.; Gitarsky, L.S.; Strulia, I.L.; Sizenev, V.S.; Pronyakin, V.T.

    1995-01-01

    At the present time beryllium is considered as the most suitable armour material for the ITER divertor application. Different types of Be-divertor mock-up construction are compared in the report. Two different technologies of beryllium tiles joining to a heat sink body are analysed: high temperature brazing and thermodiffusion bonding. The comparative analysis of different constructions has been performed on the basis of 2-D finite element calculation for temperatures and stresses. The main parameters and diagnostic capabilities of electron beam facility for HHF testing of beryllium mock-ups are described. The first results of HHF tests of ''beryllium-copper saddle-MAGT tube'' and ''beryllium-copper plate-SS body'' mock-ups are presented. The reasons of the damages during the HHF are analysed. The technique of ultrasonic testing of the thermodifussion bonding and brazing quality for beryllium-copper joints is presented. The recorded results are prepared in the form of ultrasound grams. The testing results are compared with the metallographic analysis. (orig.)

  6. Thermodynamic analysis of the advanced zero emission power plant

    Directory of Open Access Journals (Sweden)

    Kotowicz Janusz

    2016-03-01

    Full Text Available The paper presents the structure and parameters of advanced zero emission power plant (AZEP. This concept is based on the replacement of the combustion chamber in a gas turbine by the membrane reactor. The reactor has three basic functions: (i oxygen separation from the air through the membrane, (ii combustion of the fuel, and (iii heat transfer to heat the oxygen-depleted air. In the discussed unit hot depleted air is expanded in a turbine and further feeds a bottoming steam cycle (BSC through the main heat recovery steam generator (HRSG. Flue gas leaving the membrane reactor feeds the second HRSG. The flue gas consist mainly of CO2 and water vapor, thus, CO2 separation involves only the flue gas drying. Results of the thermodynamic analysis of described power plant are presented.

  7. HOMOGENEOUS NUCLEAR POWER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  8. Reactor power control device

    International Nuclear Information System (INIS)

    Kobayashi, Akira.

    1980-01-01

    Purpose: To prevent misoperation in a control system for the adjustment of core coolant flow rate, and the increase in the neutron flux density caused from the misoperation in BWR type reactors. Constitution: In a reactor power control system adapted to control the reactor power by the adjustment of core flow rate, average neutron flux signals of a reactor core, entire core flow rate signals and operation state signals for coolant recycling system are inputted to a microcomputer. The outputs from the computer are sent to a recycling MG set speed controller to control the reactor core flow rate. The computer calculates the change ratio with time in the average neutron flux signals, correlation between the average neutron flux signals and the entire core flow rate signals, change ratio with time in the operation state signals for the coolant recycling system and the like and judges the abnormality in the coolant recycling system based on the calculated results. (Ikeda, J.)

  9. 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)

  10. Operating US power reactors

    International Nuclear Information System (INIS)

    Silver, E.G.

    1988-01-01

    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 September 30, 1987, 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 (July, August, and September 1987) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. In addition to the tabular data, this article discusses other significant occurrences and developments that affected licensed US power reactors during this reporting period. Status changes at Braidwood Unit 1, Nine Mile Point 2, and Beaver Valley 2 are discussed. Other occurrences discussed are: retraining of control-room operators at Peach Bottom; a request for 25% power for Shoreham, problems at Fermi 2 which delayed the request to go to 75% power; the results of a safety study of the N Reactor at Hanford; a proposed merger of Pacific Gas and Electric with Sacramento Municipal Utility District which would result in the decommissioning of Rancho Seco; the ordered shutdown of Oyster Creek; a minor radioactivity release caused by a steam generator tube rupture at North Anna 1; and 13 fines levied by the NRC on reactor licensees

  11. Remote handling of the blanket segments: Testing of 1/3 scale mock-ups on the ROBERTINO facility

    International Nuclear Information System (INIS)

    Maisonnier, D.; Amelotti, F.; Chiasera, A.

    1994-01-01

    The remotized replacement of the blanket segments inside the Vacuum Vessel of a fusion reactor is one of the critical tasks for reactor components design, operational procedures, and safety. This open-quotes hostile environmentclose quotes task must be accomplished by a specific Blanket Handling Device, with a grasping device acting as open-quotes end-effectorclose quotes, because of intervention complexity, of components dimensions and weights, and of consequences of possible accidents during the blanket segments handling operations. Therefore, specific support experimental studies in this field appear to be necessary in order to: select appropriate blanket handling devices and procedures; assess the design of all components involved in the handling operations; perform checks in all field related to the robotized handling control (kinematics and dynamics of the grasping device trajectory planning and motion control, sensing and intelligence of the blanket handling devices, etc.); improve reliability and safety for the replacement sequences; give a realistic estimation of the time duration of the replacement duration. During the test phase, handling operations were carried out on the blanket mock-ups by means of different gripping devices. The operations were driven in the control room by means of the Motion command computer and the real time sensing data display allowed operations' control. The results were analyzed by charting the sensors' data

  12. Compact power reactor

    International Nuclear Information System (INIS)

    Wetch, J.R.; Dieckamp, H.M.; Wilson, L.A.

    1978-01-01

    There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector

  13. Reactor power control device

    International Nuclear Information System (INIS)

    Imaruoka, Hiromitsu.

    1994-01-01

    A high pressure water injection recycling system comprising injection pipelines of a high pressure water injection system and a flow rate control means in communication with a pool of a pressure control chamber is disposed to a feedwater system of a BWR type reactor. In addition, the flow rate control means is controlled by a power control device comprising a scram impossible transient event judging section, a required injection flow rate calculation section for high pressure water injection system and a control signal calculation section. Feed water flow rate to be supplied to the reactor is controlled upon occurrence of a scram impossible transient event of the reactor. The scram impossible transient event is judged based on reactor output signals and scram operation demand signals and injection flow rate is calculated based on a predetermined reactor water level, and condensate storage tank water or pressure control chamber pool water is injected to the reactor. With such procedures, water level can be ensured and power can be suppressed. Further, condensate storage tank water of low enthalpy is introduced to the pressure suppression chamber pool to directly control elevation of water temperature and ensure integrity of the pressure vessel and the reactor container. (N.H.)

  14. Power reactor design trends

    International Nuclear Information System (INIS)

    Hogan, W.J.

    1985-01-01

    Cascade and Pulse Star represent new trends in ICF power reactor design that have emerged in the last few years. The most recent embodiments of these two concepts, and that of the HYLIFE design with which they will compare them, are shown. All three reactors depend upon protecting structural elements from neutrons, x rays and debris by injecting massive amounts of shielding material inside the reaction chamber. However, Cascade and Pulse Star introduce new ideas to improve the economics, safety, and environmental impact of ICF reactors. They also pose different development issues and thus represent technological alternatives to HYLIFE

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

    International Nuclear Information System (INIS)

    Jeskanen, H.; Lahdenperae, K.; Kauppinen, P.; Taehtinen, S.

    1998-06-01

    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.)

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

    International Nuclear Information System (INIS)

    Taehtinen, S.; Kauppinen, P.; Jeskanen, H.; Lahdenperae, K.; Ehrnsten, U.

    1997-04-01

    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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Rivard, J B; Thome, F V [Sandia Laboratories (United States)

    1974-07-01

    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 {approx}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)

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

    International Nuclear Information System (INIS)

    Rivard, J.B.; Thome, F.V.

    1974-01-01

    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)

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

    International Nuclear Information System (INIS)

    Belolipetskiy, A.A.; Lalinina, E.A.; Panina, L.V.

    2010-01-01

    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

  20. High heat load tests on W/Cu mock-ups and evaluation of their application to EAST device

    Energy Technology Data Exchange (ETDEWEB)

    Li, H. [Institute of Plasma Physics, Chinese Academy of Science, Hefei, Anhui 230031 (China); Hefei Electronic Engineering Institute, Hefei, Anhui 230037 (China)], E-mail: lih72@hotmail.com; Chen, J.L.; Li, J.G. [Institute of Plasma Physics, Chinese Academy of Science, Hefei, Anhui 230031 (China); Sun, X.J. [Hefei Electronic Engineering Institute, Hefei, Anhui 230037 (China)

    2009-01-15

    Tungsten has been considered as the primary candidate plasma-facing materials (PFM) for the EAST device. Three actively cooled W/Cu mock-ups with an interlayer made of tungsten-copper alloy (1.5 mm) were designed and manufactured. The tungsten armors, pure sintered tungsten plate (1 mm) and plasma-sprayed tungsten coatings (0.3 and 0.9 mm), were bonded to the interlayer by brazing and depositing respectively. All mock-ups can withstand high heat flux up to 5 MW/m{sup 2} and no obvious failure was found after tests. The thermal performance experiments and microstructure analyses indicated the structure of mock-ups possess good thermal contact and high heat transfer capability. WCu alloy as an interlayer can largely reduce the stress due to the mismatch and improve the reliability. The mock-up with 0.9 mm coating had the highest surface temperature than the other two mock-ups, delaminations of this mock-up were found in the near surface by SEM. The primary results show that pure sintered tungsten brazed to WCu alloy is a possible way, and thick plasma-sprayed coating technique still need to be improved.

  1. Sensitivity and uncertainty analyses of the HCLL mock-up experiment

    International Nuclear Information System (INIS)

    Leichtle, D.; Fischer, U.; Kodeli, I.; Perel, R.L.; Klix, A.; Batistoni, P.; Villari, R.

    2010-01-01

    Within the European Fusion Technology Programme dedicated computational methods, tools and data have been developed and validated for sensitivity and uncertainty analyses of fusion neutronics experiments. The present paper is devoted to this kind of analyses on the recent neutronics experiment on a mock-up of the Helium-Cooled Lithium Lead Test Blanket Module for ITER at the Frascati neutron generator. They comprise both probabilistic and deterministic methodologies for the assessment of uncertainties of nuclear responses due to nuclear data uncertainties and their sensitivities to the involved reaction cross-section data. We have used MCNP and MCSEN codes in the Monte Carlo approach and DORT and SUSD3D in the deterministic approach for transport and sensitivity calculations, respectively. In both cases JEFF-3.1 and FENDL-2.1 libraries for the transport data and mainly ENDF/B-VI.8 and SCALE6.0 libraries for the relevant covariance data have been used. With a few exceptions, the two different methodological approaches were shown to provide consistent results. A total nuclear data related uncertainty in the range of 1-2% (1σ confidence level) was assessed for the tritium production in the HCLL mock-up experiment.

  2. Experimental investigation of MHD pressure losses in a mock-up of a liquid metal blanket

    Science.gov (United States)

    Mistrangelo, C.; Bühler, L.; Brinkmann, H.-J.

    2018-03-01

    Experiments have been performed to investigate the influence of a magnetic field on liquid metal flows in a scaled mock-up of a helium cooled lead lithium (HCLL) blanket. During the experiments pressure differences between points on the mock-up have been recorded for various values of flow rate and magnitude of the imposed magnetic field. The main contributions to the total pressure drop in the test-section have been identified as a function of characteristic flow parameters. For sufficiently strong magnetic fields the non-dimensional pressure losses are practically independent on the flow rate, namely inertia forces become negligible. Previous experiments on MHD flows in a simplified test-section for a HCLL blanket showed that the main contributions to the total pressure drop in a blanket module originate from the flow in the distributing and collecting manifolds. The new experiments confirm that the largest pressure drops occur along manifolds and near the first wall of the blanket module, where the liquid metal passes through small openings in the stiffening plates separating two breeder units. Moreover, the experimental data shows that with the present manifold design the flow does not distribute homogeneously among the 8 stacked boxes that form the breeding zone.

  3. 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.

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

    International Nuclear Information System (INIS)

    Brossa, M.; Ciric, D.; Deksnis, E.; Falter, H.; Guerreschi, U.; Peacock, A.; Pick, M.; Rossi, M.; Shen, Y.; Zacchia, F.

    1995-01-01

    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/m 2 . (orig.)

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

    International Nuclear Information System (INIS)

    Singh, Kongkham Premjit; Khirwadkar, Samir S.; Bhope, Kedar; Patel, Nikunj; Mokaria, Prakash K.; Mehta, Mayur

    2015-01-01

    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)

  6. Analysis of free and forced excitation tests of 394 KN isolated structure mock-up

    International Nuclear Information System (INIS)

    Serino, G.; Martelli, A.; Bonacina, G.

    1993-01-01

    At the 1991 ASME-PVP Conference, some first experimental results obtained from static and dynamic tests on high damping steel laminated rubber bearings (Martelli et al., 1991) and from free and forced excitation tests on a 394 kN isolated structure mock-up were presented (Forni et al., 1991). In this paper, the most significant test data are reorganized and discussed in order to assess the suitability of single bearing test results to predict the dynamic response of an isolated structure. Three mathematical models of the single isolator having different levels of approximation are proposed, and their capability to estimate the experimental response of the mock-up is evaluated. It is shown that a non-linear hysteretic model, defined by three rubber parameters only, allows a very good complete simulation of the dynamic behavior of the isolated structure in both free and forced vibration tests. A simpler equivalent linear viscous model permits a good prediction of the peak absolute acceleration and relative displacement values if bearing stiffness and damping parameters are properly selected, and can be used in a response spectrum analysis, but reproduces less exactly the experimental behavior. An equivalent linear hysteretic model represents more correctly the actual rubber damping behavior, but gives results very similar to those obtained through the equivalent linear viscous model because of the practically mono-frequencial response of the isolated structure

  7. Plasma cleaning of ITER edge Thomson scattering mock-up mirror in the EAST tokamak

    Science.gov (United States)

    Yan, Rong; Moser, Lucas; Wang, Baoguo; Peng, Jiao; Vorpahl, Christian; Leipold, Frank; Reichle, Roger; Ding, Rui; Chen, Junling; Mu, Lei; Steiner, Roland; Meyer, Ernst; Zhao, Mingzhong; Wu, Jinhua; Marot, Laurent

    2018-02-01

    First mirrors are the key element of all optical and laser diagnostics in ITER. Facing the plasma directly, the surface of the first mirrors could be sputtered by energetic particles or deposited with contaminants eroded from the first wall (tungsten and beryllium), which would result in the degradation of the reflectivity. The impurity deposits emphasize the necessity of the first mirror in situ cleaning for ITER. The mock-up first mirror system for ITER edge Thomson scattering diagnostics has been cleaned in EAST for the first time in a tokamak using radio frequency capacitively coupled plasma. The cleaning properties, namely the removal of contaminants and homogeneity of cleaning were investigated with molybdenum mirror insets (25 mm diameter) located at five positions over the mock-up plate (center to edge) on which 10 nm of aluminum oxide, used as beryllium proxy, were deposited. The cleaning efficiency was evaluated using energy dispersive x-ray spectroscopy, reflectivity measurements and x-ray photoelectron spectroscopy. Using argon or neon plasma without magnetic field in the laboratory and with a 1.7 T magnetic field in the EAST tokamak, the aluminum oxide films were homogeneously removed. The full recovery of the mirrors’ reflectivity was attained after cleaning in EAST with the magnetic field, and the cleaning efficiency was about 40 times higher than that without the magnetic field. All these results are promising for the plasma cleaning baseline scenario of ITER.

  8. Development of a digital mock-up system for selecting a decommissioning scenario

    International Nuclear Information System (INIS)

    Kim, Sung-Kyun; Park, Hee-Sung; Lee, Kune-Woo; Jung, Chong-Hun

    2006-01-01

    The evaluation of decommissioning scenarios is critical to the successful development and execution of a decommissioning project. In the past, many experts have used a physical mock-up system to find the exact work processes and the working positions. Nowadays, these jobs are being done by a Digital Mock-Up (DMU) system. The DMU, which is a technology to realize an effective work process by using virtual environments through representing the physical and logical schema and the behavior of a real decommissioning work, can save on the cost and time, reduce the risk of making later changes, and develop various decommissioning scenarios. In this research, a decommissioning DMU system was developed for simulating the relevant dismantling processes. Decommissioning data-computing modules which can calculate a dismantling schedule, quantify a radioactive waste, visualize a radioactive inventory, estimate a decommissioning cost, and estimate a worker's exposure were also developed to qualitatively assess the decommissioning information. And an analytic hierarchy process (AHP) model was developed to evaluate the decommissioning scenarios which reflected the quantitative and qualitative considerations. To establish the proper scenario for the thermal column in KRR-1, the developed decommissioning DMU system was applied to evaluate the two candidate scenarios of it

  9. Tests and measurements with a thermal VXD mock-up for BELLE II

    International Nuclear Information System (INIS)

    Huebner, Lars

    2015-03-01

    As part of the Belle detector upgrade, located at the KEK in Tsukuba, Japan, a CO 2 cooling system will be added. Using new detector components, which are easily damageable or influenced by heat, make this step necessary. Particularly the next to the beam pipe located PXD is strained by high thermal load and therefore requires cooling. The CDC needs a constant temperature for precise measurements, but it could be influenced by heat from the SVD. Knowledge about the heat generation and distribution is needed before assembling the full detector. A mock-up of the innermost parts of the detector and a CO 2 cooling system is under construction at DESY in Hamburg, Germany, to gather such knowledge. The mock-up should be able to emulate the thermal properties of the final detector. Within the scope of this bachelor's thesis, the outermost VXD Layer 6 was studied in a flat arrangement. Focus lay on the heat dissipation at the sensors and on pressure drop measurements of the cooling pipe. It was investigated whether the applied heat load can be sufficiently lead away and how large the pressure drop is along the experiment line. Despite cooling was applied, a remarkable rise in temperature was observed. However, the unfavorable position of the thermistors make reliable quantitative statements of the sensor dummies' temperatures impossible. The pressure drop was determined, but is of limited accuracy due to large uncertainties. Further investigations have to be made with a better set-up.

  10. Neutrino Mass Models: impact of non-zero reactor angle

    International Nuclear Information System (INIS)

    King, Stephen F.

    2011-01-01

    In this talk neutrino mass models are reviewed and the impact of a non-zero reactor angle and other deviations from tri-bi maximal mixing are discussed. We propose some benchmark models, where the only way to discriminate between them is by high precision neutrino oscillation experiments.

  11. UV-Induced Anisotropy In CdBr2-CdBr2: Cu Nanostructures

    Directory of Open Access Journals (Sweden)

    El-Naggar A. M.

    2015-09-01

    Full Text Available We have found an occurrence of anisotropy in the nanostructure CdBr2-CdBr2: Cu nanocrystalline films. The film thickness was varied from 4 nm up to 80 nm. The films were prepared by successive deposition of the novel layers onto the basic nanocrystals. The detection of anisotropy was performed by occurrence of anisotropy in the polarized light at 633 nm He-Ne laser wavelength. The occurrence of anisotropy was substantially dependent on the film thickness and the photoinduced power density. Possible mechanisms of the observed phenomena are discussed.

  12. Reactor power distribution monitor

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1980-01-01

    Purpose: To improve the performance and secure the safety of a nuclear reactor by rapidly computing and display the power density in the nuclear reactor by using a plurality of processors. Constitution: Plant data for a nuclear reactor containing the measured values from a local power monitor LPRM are sent and recorded in a magnetic disc. They are also sent to a core performance computer in which burn-up degree distribution and the like are computed, and the results are sent and recorded in the magnetic disc. A central processors loads programs to each of the processors and applies data recorded in the magnetic disc to each of the processors. Each of the processors computes the corresponding power distribution in four fuel assemblies surrounding the LPRM string by the above information. The central processor compiles the computation results and displays them on a display. In this way, power distribution in the fuel assemblies can rapidly be computed to thereby secure the improvement of the performance and safety of the reactor. (Seki, T.)

  13. Nuclear power reactor safety

    International Nuclear Information System (INIS)

    Pon, G.A.

    1976-10-01

    This report is based on the Atomic Energy of Canada Limited submission to the Royal Commission on Electric Power Planning on the safety of CANDU reactors. It discusses normal operating conditions, postulated accident conditions, and safety systems. The release of radioactivity under normal and accident conditions is compared to the limits set by the Atomic Energy Control Regulations. (author)

  14. 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

  15. A constitutive model for the thermo-mechanical behaviour of fusion-relevant pebble beds and its application to the simulation of HELICA mock-up experimental results

    International Nuclear Information System (INIS)

    Vella, G.; Maio, P.A. Di; Giammusso, R.; Tincani, A.; Orco, G. Dell

    2006-01-01

    Within the framework of the activities promoted by European Fusion Development Agreement on the technology of the Helium Cooled Pebble Bed Test Blanket Module to be irradiated in one of the ITER equatorial ports, attention has been focused on the theoretical modelling of the thermo-mechanical constitutive behaviour of both beryllium and lithiated ceramics pebble beds, that are envisaged to act respectively as neutron multiplier and tritium breeder. The thermo-mechanical behaviour of the pebble beds and their nuclear performances in terms of tritium production depend on the reactor relevant conditions (heat flux and neutron wall load), the pebble sizes and the breeder cell geometries (bed thickness, pebble packing factor, bed overall thermal conductivity). ENEA-Brasimone and the Department of Nuclear Engineering (DIN) of the Palermo University have performed intense research activities intended to investigate fusion-relevant pebble bed thermo-mechanical behaviour by adopting both experimental and theoretical approaches. In particular, ENEA has carried out several experimental campaigns on small scale mock-ups tested in out-of-pile conditions, while DIN has developed a proper constitutive model that has been implemented on commercial FEM code, for the prediction of the thermal and mechanical performances of fusion-relevant pebble beds and for the comparison with the experimental results of the ENEA tests. In that framework, HELICA mock-up has been set-up and tested to investigate the behaviour of pebble bed in reactor-relevant geometries, providing useful data sets to be numerically reproduced by means of the DIN constitutive model, contributing to its assessment. The paper presents the constitutive model developed and the main experimental results of two test campaigns on HELICA mock-up carried out at HE-FUS 3 facility of ENEA Brasimone, the geometry of the mock-up, the adopted thermal and mechanical boundary conditions and the test operating conditions. The most

  16. Life extension of the BR2 aluminium vessel

    International Nuclear Information System (INIS)

    Koonen, E.; Fabry, A.; Chaouadi, R.; Verwerft, M.; Raedt, C. de; Winckel, S. van; Wacquier, W.; Dadoumont, J.; Verwimp, A.

    2000-01-01

    The BR2 reactor has recently undergone a major refurbishment comprising the replacement of all vessel internals. The vessel itself however was not replaced. An important requalification programme has been executed to prove that the vessel would remain fit during the contemplated life extension period of BR2. Representative material samples could be obtained from the shroud surrounding the vessel. A comprehensive in-service inspection was carried out and a vessel surveillance programme has been established. (author)

  17. Power reactor noise

    International Nuclear Information System (INIS)

    Thie, J.A.

    1981-01-01

    This book concentrates on the different types of noise present in power reactors and how the analysis of this noise can be used as a tool for reactor monitoring and diagnostics. Noise analysis is a growing field that offers advantages such as simplicity, low cost, and natural multivariable interactions. A major advantage, continuous and undisturbed monitoring, supplies a means of obtaining early warnings of possible reactor malfunctions thus preventing further complications by alerting operators to a problem - and aiding in the diagnosis of that problem - before it demands major repairs. Following an introductory chapter, the theoretical basis for the various methods of noise analysis is explained, and full chapters are devoted to the fundamentals of statistics for time-domain analysis and Fourier series and related topics for frequency-domain analysis. General experimental techniques and associated theoretical considerations are reviewed, leading to discussion of practical applications in the latter half of the book. Besides chapters giving examples of neutron noise and acoustical noise, chapters are also devoted to extensive examples from pressurized water reactor and boiling water reactor power plants

  18. Zero energy reactor RB technical characteristics and experimental possibilities

    Energy Technology Data Exchange (ETDEWEB)

    Jovanovic, S; Takac, S; Raisic, N; Lolic, B; Markovic, H [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1963-04-15

    The zero energy reactor RB was constructed in 1958 in accordance with the nuclear reactor development programme of the Boris Kidric Institute of Nuclear Sciences. The reactor was in operation until the middle of 1959 when the heavy water, serving as the moderator, was transported to the high flux reactor RA, built at the same time at the Boris Kidric Institute. Owing to the fact that the purchase of new quantities of heavy water was planned for 1961 it was decided to reconstruct the RB reactor in order to improve the safety of the system and to obtain better flexibility in performing the experiments. New control, safety and radiation monitoring systems were constructed. Some changes were also made on the reactor tank, water circulation system and the water level monitoring equipment. The reconstruction was completed in 1961. and the heavy water was delivered early in 1962. The reconstructed reactor was critical for the first time in summer 1962, and from that time was in continuous operation. This report presents an outline of the design and construction characteristics of the reactor. The main intention is to inform potential users of the reactor about experimental possibilities, advantages and disadvantages of such a critical facility.

  19. Zero energy reactor RB technical characteristics and experimental possibilities

    International Nuclear Information System (INIS)

    Jovanovic, S.; Takac, S.; Raisic, N.; Lolic, B.; Markovic, H.

    1963-04-01

    The zero energy reactor RB was constructed in 1958 in accordance with the nuclear reactor development programme of the Boris Kidric Institute of Nuclear Sciences. The reactor was in operation until the middle of 1959 when the heavy water, serving as the moderator, was transported to the high flux reactor RA, built at the same time at the Boris Kidric Institute. Owing to the fact that the purchase of new quantities of heavy water was planned for 1961 it was decided to reconstruct the RB reactor in order to improve the safety of the system and to obtain better flexibility in performing the experiments. New control, safety and radiation monitoring systems were constructed. Some changes were also made on the reactor tank, water circulation system and the water level monitoring equipment. The reconstruction was completed in 1961. and the heavy water was delivered early in 1962. The reconstructed reactor was critical for the first time in summer 1962, and from that time was in continuous operation. This report presents an outline of the design and construction characteristics of the reactor. The main intention is to inform potential users of the reactor about experimental possibilities, advantages and disadvantages of such a critical facility

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

    Energy Technology Data Exchange (ETDEWEB)

    Harris, S.P. [Westinghouse Savannah River Company, AIKEN, SC (United States)

    1997-09-18

    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.Either new DWPF analytical method could result in a two to three fold improvement in sample analysis time.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.The insert sample is named after the initial trials which placed the container inside the sample (peanut) vials. 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

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

    International Nuclear Information System (INIS)

    Harris, S.P.

    1997-01-01

    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

  2. Reactor power monitoring device

    International Nuclear Information System (INIS)

    Kono, Shigehiro.

    1990-01-01

    Among a plurality of power monitoring programs in a reactor power monitoring device, rapid response is required for a scram judging program for the power judging processing of scram signals. Therefore, the scram judging program is stored independently from other power monitoring programs, applied with a priority order, and executed in parallel with other programs, to output scram signals when the detected data exceeds a predetermined value. As a result, the capacity required for the scram judging program is reduced and the processing can be conducted in a short period of time. In addition, since high priority is applied to the scram judging program which is divided into a small capacity, it is executed at higher frequency than other programs when they are executed in parallel. That is, since the entire processings for the power monitoring program are repeated in a short cycle, the response speed of the scram signals required for high responsivity can be increased. (N.H.)

  3. Design and Analysis of the Korean Small Semi-prototype Mock-up for the 2nd Qualification of the ITER Blanket First Wall

    International Nuclear Information System (INIS)

    Lee, Dong Won; Kim, Suk Kwon; Yoon, Jae Sung; Lee, Eo Hwak; Lee, Seung Jae; Choi, Bo Guen; Park, Jeong Yong; Jung, Yang Il; Choi, Byung Kwon; Kim, Byoung Yoon

    2011-01-01

    Since the blanket First Wall (FW) of the International Thermonuclear Experimental Reactor (ITER) is subjected to a high heat and high neutron loads, it is one of the most important components. It composed of a beryllium (Be) layer as a plasma facing material, a copper alloy (CuCrZr) layer as a heat sink and type 316L authentic stainless steel (SS316L) as a structure material. The joining of the three different metals is the key issue to be solved. And more, the peak heat load was assumed to be 0.5 MW/m 2 in the initial design of the FW, but it was changed to be up to 5 MW/m 2 In Korea, the joining method has developed and it was proved through the several mock-up fabrication and high heat flux tests for confirming the joining integrity. Some of them were tested in the foreign facilities such as JEBIS at JAEA in Japan, TSEFEY at Efremov in Russia, and JUDITH at FZJ in Germany, and others were tested in our own facilities such as KoHLT-1 and -2. And finally, the 1 st , recently. Therefore, the FW panel design has been changed for enhancing the cooling and ITER Organization will provide the proposed design. Qualification was passed, in which two 80x80x3 Be/Cu/SS mock-ups were tested under 0.625 and 0.875 MW/m 2 heat fluxes for 12,000 cycles and then tested under 1.75 and 1.40 MW/m 2 Currently, the 2 heat fluxes for 1,000 cycles at FZJ and SNL, respectively. Currently, the 2 nd qualification program was started and the semi-prototype should be fabricated by the end of 2011 for testing under 5.0 MW/m 2 heat flux for certain number of cycles. In order to prepare the semi-prototype, several fabrication methods should be developed through the fabrication and test with the several mock-ups. In the present study, small Be mock-up was fabricated as the first step for the preparation. It was fabricated according to the designs considering the currently modified design of the FW. In the present paper, the fabrication objectives, methods, results and related tests were

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

    International Nuclear Information System (INIS)

    Song, In Ho; Chung, Sung Chong

    2007-01-01

    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

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

    International Nuclear Information System (INIS)

    Queval, J.C.; Gantenbein, F.; Lebelle, M.

    1995-01-01

    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

  6. A generalized approach for historical mock-up acquisition and data modelling: Towards historically enriched 3D city models

    Science.gov (United States)

    Hervy, B.; Billen, R.; Laroche, F.; Carré, C.; Servières, M.; Van Ruymbeke, M.; Tourre, V.; Delfosse, V.; Kerouanton, J.-L.

    2012-10-01

    Museums are filled with hidden secrets. One of those secrets lies behind historical mock-ups whose signification goes far behind a simple representation of a city. We face the challenge of designing, storing and showing knowledge related to these mock-ups in order to explain their historical value. Over the last few years, several mock-up digitalisation projects have been realised. Two of them, Nantes 1900 and Virtual Leodium, propose innovative approaches that present a lot of similarities. This paper presents a framework to go one step further by analysing their data modelling processes and extracting what could be a generalized approach to build a numerical mock-up and the knowledge database associated. Geometry modelling and knowledge modelling influence each other and are conducted in a parallel process. Our generalized approach describes a global overview of what can be a data modelling process. Our next goal is obviously to apply this global approach on other historical mock-up, but we also think about applying it to other 3D objects that need to embed semantic data, and approaching historically enriched 3D city models.

  7. A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

    Energy Technology Data Exchange (ETDEWEB)

    Nakahira, Masataka; Takeda, Nobukazu; Onozuka, Masanori [Japan Atomic Energy Agency (Japan); Kakudate, Satoshi [Mitsubishi Heavy Industries, Ltd. (Japan)

    2007-07-01

    The structure and fabrication methods of the ITER vacuum vessel have been investigated and defined by the ITER international team. However, some of the current specifications are very difficult to be achieved from the manufacturing point of view and will lead to cost increase. In the mock-up fabrication, it is planned to conduct the following items: 1. Feasibility of the Japanese proposed VV structure and fabrication methods and the applicability to the ITER are to be confirmed; 2. Assembly procedure and inspection procedure are to be confirmed; 3. Manufacturing tolerances are to be assessed; 4. Manufacturing schedule is to be assessed. This report summarizes the Japanese proposed specification of the VV mock-up describing differences between the ITER supplied design. General scope of the mock-up fabrication and the detailed dimensions are also shown. In the VV fabrication, several types of weld joint configuration will be used. This report shows the joint configurations proposed by Japan to be used for the inner shell connection, the rib-to-shell connection and outer shell connection, and the housing-to-shell connection, respectively. Non-destructive testing considered to be applied to each joint configuration is also presented. A series of the fabrication and assembly procedures for the mock-up are presented in this report, together with candidates of welding configurations. Finally, the report summarizes the results of mock-up fabrication, including results of nondestructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. (orig.)

  8. Validation of CLIC Re-Adjustment System Based on Eccentric Cam Movers One Degree of Freedom Mock-Up

    CERN Document Server

    Kemppinen, J; Lackner, F

    2011-01-01

    Compact Linear Collider (CLIC) is a 48 km long linear accelerator currently studied at CERN. It is a high luminosity electron-positron collider with an energy range of 0.5-3 TeV. CLIC is based on a two-beam technology in which a high current drive beam transfers RF power to the main beam accelerating structures. The main beam is steered with quadrupole magnets. To reach CLIC target luminosity, the main beam quadrupoles have to be actively pre-aligned within 17 µm in 5 degrees of freedom and actively stabilised at 1 nm in vertical above 1 Hz. To reach the pre-alignment requirement as well as the rigidity required by nano-stabilisation, a system based on eccentric cam movers is proposed for the re-adjustment of the main beam quadrupoles. Validation of the technique to the stringent CLIC requirements was started with tests in one degree of freedom on an eccentric cam mover. This paper describes the dedicated mock-up as well as the tests and measurements carried out with it. Finally, the test results are present...

  9. Nuclear power reactor physics

    International Nuclear Information System (INIS)

    Barjon, Robert

    1975-01-01

    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 [fr

  10. Reactor power control device

    International Nuclear Information System (INIS)

    Watanabe, Mitsutaka

    1997-01-01

    Hardware of an analog nuclear instrumentation system is reformed, a function generator is added to a setting calculation circuit of the nuclear instrumentation system, and each of setting lines of the nuclear instrumentation system is set in parallel with an upper limit curve in an operation region defined by a second order or third order equation. Upon transient change of abnormal power elevation during operation, scram signals are generated by power change in the same state as 100% rated operation due to elevation of reactor thermal power. Since the operation limit value relative to transient change due to power elevation can be made substantially equal with the same as that upon rated operation, the operation limit value for partial power operation state can be kept substantially the same level as that upon rated operation. When transition change caused by abnormal control rod withdrawal occurs during operation, a control rod withdrawal inhibition signal can ensure the power elevation width equal with that upon rated power operation, and since the withdrawal inhibition signal is generated in substantially the same withdrawing state, the operation limit value relative to a partial power operation state can be kept at the same level as that during rated operation. (N.H.)

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

    International Nuclear Information System (INIS)

    Hansen, G.E.; Audas, J.H.; Martin, E.R.; Pederson, R.A.; Spriggs, G.D.; White, R.H.

    1988-01-01

    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% 235 U) 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

  12. Analysis of the impacts of the J-TEXT TBM mock-up on the equilibrium magnetic field

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Zhengqing [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, and College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Rao, Bo, E-mail: borao@hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, and College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Zhang, Ming; Zhang, Jun [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, and College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Wang, Weihua [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); New Star Institute of Applied Technology, Hefei 230031 (China); Liu, Sumei [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); School of Engineering,Anhui Agricultural University, Hefei 230036 (China)

    2016-11-01

    Highlights: • J-TEXT TBM mock-up was designed and fabricated to test and study the distribution of eddy current, electromagnetic and thermal load on the TBM during plasma disruption. • This paper focuses on evaluating the influence of the TBM structural material (RAMF steel) to tokamak discharge and security. The simulation data presents a relatively complete assessment of impacts of the J-TEXT TBM mock-up on the equilibrium magnetic field. • The conclusion of the simulation will offer the guidance for installation interface design of the TBM mock-up. - Abstract: The Test Blanket Module (TBM) will be used in the test port of ITER to demonstrate tritium self-sufficiency and the extraction of high grade heat for electricity production. J-TEXT TBM mock-up using reduced activation ferritic/martensitic (RAFM) steel as structural material was designed and fabricated to perform and validate relevant electromagnetic and thermal technologies of the China Helium-Cooled Ceramic Breeder Test Blanket Module (CN HCCB-TBM) on the J-TEXT. Its size is one third of the CN HCCB-TBM. By using the finite element analysis technology, this paper analyzed the impacts on the equilibrium magnetic field over the plasma region after introducing the structure material RAFM steel. The distribution of toroidal field (TF) ripple and the magnitude of the error field with the mock-up at different positions were given. Simulation shows the distribution of the null field region formed by poloidal field (PF). The influence to tokamak discharge has been evaluated by drawing the magnetic field lines. Based on the results above, we have optimized and finished the installation of the mock-up to J-TEXT which meets the needs of the experiments and to ensure the normal discharge.

  13. Refurbishment of BR2 (Phases 4 and 5)

    International Nuclear Information System (INIS)

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

    1998-01-01

    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

  14. 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.

  15. Chemically deposited tungsten fibre-reinforced tungsten – The way to a mock-up for divertor applications

    Directory of Open Access Journals (Sweden)

    J. Riesch

    2016-12-01

    Full Text Available The development of advanced materials is essential for sophisticated energy systems like a future fusion reactor. Tungsten fibre-reinforced tungsten composites (Wf/W utilize extrinsic toughening mechanisms and therefore overcome the intrinsic brittleness of tungsten at low temperature and its sensitivity to operational embrittlement. This material has been successfully produced and tested during the last years and the focus is now put on the technological realisation for the use in plasma facing components of fusion devices. In this contribution, we present a way to utilize Wf/W composites for divertor applications by a fabrication route based on the chemical vapour deposition (CVD of tungsten. Mock-ups based on the ITER typical design can be realized by the implementation of Wf/W tiles. A concept based on a layered deposition approach allows the production of such tiles in the required geometry. One fibre layer after the other is positioned and ingrown into the W-matrix until the final sample size is reached. Charpy impact tests on these samples showed an increased fracture energy mainly due to the ductile deformation of the tungsten fibres. The use of Wf/W could broaden the operation temperature window of tungsten significantly and mitigate problems of deep cracking occurring typically in cyclic high heat flux loading. Textile techniques are utilized to optimise the tungsten wire positioning and process speed of preform production. A new device dedicated to the chemical deposition of W enhances significantly, the available machine time for processing and optimisation. Modelling shows that good deposition results are achievable by the use of a convectional flow and a directed temperature profile in an infiltration process.

  16. Power reactor noise

    International Nuclear Information System (INIS)

    Thie, J.A.

    1981-01-01

    Noise analysis is a growing field that offers advantages such as simplicity, low cost, and natural multivariable interactions. A major advantage, continuous and undisturbed monitoring, supplies a means of obtaining early warnings of possible reactor malfunctions, thus preventing further complications by alerting opeators to a problem - and aiding in the diagnosis of that problem - before it demands major repairs. Dr. Thie hopes to further, through detailed explanations and over 70 illustrations, the acceptance of the use of noise analysis by the nuclear utility industry. Following an introductory chapter, the theoretical basis for the various methods of noise analysis is explained, and full chapters are devoted to the fundamentals of statistics for time-domain analysis and Fourier series and related topics for frequency-domain analysis. General experimental techniques and associated theoretical considerations are reviewed, leading to discussions of practical applications in the latter half of the book. Besides chapters giving examples of neutron noise and acoustical noise, chapters are also devoted to extensive examples from pressurized water reactor and boiling water reactor power plants

  17. Nuclear reactor power supply

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

    The redundant signals from the sensor assemblies measuring the process parameters of a nuclear reactor power supply are transmitted each in its turn to a protection system which operates to actuate the protection apparatus for signals indicating off-process conditions. Each sensor assembly includes a number of like sensors measuring the same parameters. The sets of process signals derived from the sensor assemblies are each in its turn transmitted from the protection system to the control system which impresses control signals on the reactor or its components to counteract the tendency for conditions to drift off-normal status requiring operation of the protection system. A parameter signal selector is interposed between the protection system and the control system. This selector prevents a parameter signal of a set of signals, which differs from the other parameters signals of the set by more than twice the allowable variation of the sensors which produce the set, from passing to the control system. The selectors include a pair of signal selection units, one unit sending selected process signals to primary control channels and the other sending selected process signals to back-up control channels. Test signals are periodically impressed by a test unit on a selected pair of a selected unit and control channels. When test signals are so impressed the selected control channel is disabled from transmitting control signals to the reactor and/or its associated components. This arrangement eliminates the possibility that a single component failure which may be spurious will cause an inadvertent trip of the reactor during test

  18. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

    1978-01-01

    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. MTR fuel testing in BR2

    International Nuclear Information System (INIS)

    Jacquet, P.; Verwimp, A.; Wirix, S.

    2000-01-01

    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)

  20. Implementation of a Digital Mock-up for Remote Hot cell Operations

    International Nuclear Information System (INIS)

    Park, Hee Seong; Park, Byung Suk; Kim, Sung Hyun; Kim, Ki Ho; Kim, Ho Dong

    2010-01-01

    A remote manipulation environment that a human operator has to observe is the inner side of a hotcell through a lead grass window which has many obstacles due to many existing 'blind-spots' where are several cameras installed. The lack of visual information when operating in a cluttered environment makes manoeuvering a manipulator very difficult and when this situation is exacerbated by strict time limits for a task completion, then a manipulator and environmental collisions and resultant damage can occur. To cope with these problems, there has been efforts to develop a virtual simulator to validate control programs visually and to establish maintainability-engineering tools that automate generation assembly/disassembly procedures by using Computer Aided Design(CAD) visualization systems with human figure models to virtual reality systems where engineers can interact with the system using virtual input devices. This article introduces a system that can simulate a deployment analysis on a digital mock-up effectively and proposes a scheme to enable an operator to improve a remote manipulation by using a haptic device

  1. Development of the ITER IOIS assembly tool and mock-up

    International Nuclear Information System (INIS)

    Nam, Kyoungo; Kim, Dongjin; Park, Hyunki; Ahn, Heejae; Kim, Kyoungkyu; Yoo, Yongsoo; Watson, Emma; Shaw, Robert

    2014-01-01

    The ITER toroidal field coils (TFCs) are connected by 3 different connecting structures as follows; Outer Intercoil Structure (OIS), Inner Intercoil Structure (IIS), Intermediate Outer Intercoil Structure (IOIS). In assessing the assembly, requirements and environmental conditions of each Intercoil structure, the IOIS and IIS assembly were thought to be the most challenging compared to the OIS assembly due to the very limited assembly space available and the strict requirements requested by IO, especially the IOIS assembly, which has particularly difficult installation requirements including complicated shear pin assemblies. A conceptual and preliminary design has been developed by the Korean domestic agency (KODA) for the sub assembly and final assembly phase; the tool includes the ability to control both IOIS plates simultaneously. For design verification of the IOIS assembly tool mentioned above, structural analysis has been carried out considering seismic event. Also, a half sized mock-up has been fabricated and tested according to assembly procedures. In this paper, a description of tool design and the results of analysis and mock-test will be introduced

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

    International Nuclear Information System (INIS)

    Facy, L.; Perrin De Brichambaut, C.; Doury, A.; Le Quinio, R.

    1962-01-01

    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) [fr

  3. Fabrication of small mock-ups reflecting the design features of the ITER semi-prototype

    International Nuclear Information System (INIS)

    Jung, Yang-Il; Choi, Byoung-Kwon; Park, Jeong-Yong; Kim, Suk-Kwon; Lee, Dong Won; Kim, Byoung Yoon

    2012-01-01

    The ITER semi-prototype was designed to qualify the manufacturing technology for the ITER blanket first wall. However, its fabrication is expected to face great difficulty due to a design complexity. Even though joining technology for different materials such as beryllium, CuCrZr, and stainless steel (SS) was developed during the first stage of qualification, the joining is still a key issue for the fabrication of the semi-prototype. In this study, small mock-ups (SMU) were fabricated to realize and verify the manufacturing of the semi-prototype reflecting the described design features. The joining of multiple beryllium tiles on the angled CuCrZr surface was confirmed with SMU no. 1. Six beryllium tiles were joined using hot isostatic pressing (HIP), and slitting was then performed to form multiple tiles. In SMU no. 2, HIP was performed two times in order to facilitate the cooling channels at the CuCrZr/SS interface, and to join the beryllium tiles on CuCrZr/SS. The method used to form a pressure boundary for the complex cooling channels was also developed by fabricating the SMU no. 3. The SMUs confirmed the applicability of the HIP for the manufacturing of the semi-prototype.

  4. 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.

  5. The HIE-ISOLDE alignment and monitoring system software and test mock up

    CERN Document Server

    Kautzmann, G; Kadi, Y; Leclercq, Y; Waniorek, S; Williams, L

    2012-01-01

    For the HIE Isolde project a superconducting linac will be built at CERN in the Isolde facility area. The linac will be based on the creation and installation of 2 high- β and 4 low- β cryomodules containing respectively 5 high-β superconducting cavities and 1 superconducting solenoid for the two first ones, 6 low-β superconducting cavities and 2 superconducting solenoids for the four other ones. An alignment and monitoring system of the RF cavities and solenoids placed inside the cryomodules is needed to reach the optimum linac working conditions. The alignment system is based on opto-electronics, optics and precise mechanical instrumentation. The geometrical frame configuration, the data acquisition and the 3D adjustment will be managed using a dedicated software application. In parallel to the software development, an alignment system test mock-up has been built for software validation and dimensional tests. This paper will present the software concept and the development status, and then will describe...

  6. Characterization of ITER tungsten qualification mock-ups exposed to high cyclic thermal loads

    Energy Technology Data Exchange (ETDEWEB)

    Pintsuk, Gerald, E-mail: g.pintsuk@fz-juelich.de [Forschungszentrum Jülich GmbH, D-52425 Jülich (Germany); Bednarek, Maja; Gavila, Pierre [Fusion for Energy, E-08019 Barcelona (Spain); Gerzoskovitz, Stefan [Plansee SE, Innovation Services, 6600 Reutte (Austria); Linke, Jochen [Forschungszentrum Jülich GmbH, D-52425 Jülich (Germany); Lorenzetto, Patrick; Riccardi, Bruno [Fusion for Energy, E-08019 Barcelona (Spain); Escourbiac, Frederic [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 Saint Paul lez Durance (France)

    2015-10-15

    Highlights: • Mechanical deformation of CuCrZr in case a thermal barrier layer has been formed due to impurity content in the cooling water. • Crack formation at the W/Cu interface starting at the block edge. • Porosity formation in the pure Cu interlayer. • Microstructural changes in tungsten down to the W/Cu interface, which indicates also high temperatures for the pure Cu interlayer. • Macrocrack formation in tungsten which is assumed to be ductile at the initiation point and brittle when proceeding toward the cooling tube. - Abstract: High heat flux tested small-scale tungsten monoblock mock-ups (5000 cycles at 10 MW/m{sup 2} and up to 1000 cycles at 20 MW/m{sup 2}) manufactured by Plansee and Ansaldo were characterized by metallographic means. Therein, the macrocrack formation and propagation in tungsten, its recrystallization behavior and the surface response to different heat load facilities were investigated. Furthermore, debonding at the W/Cu interface, void formation in the soft copper interlayer and microcrack formation at the inner surface of the CuCrZr cooling tube were found.

  7. Mock-up development of new warship protective armor structure and feasibility analysis of ship installation

    Directory of Open Access Journals (Sweden)

    ZHENG Pan

    2017-05-01

    Full Text Available To ensure the installation of the new design of protective armor structure on larger warships,a study into the installation process of the structure of this armor is carried out to improve installation efficiency and ensure the protective effect. This paper proposes a typical composite armor structure design which is composed of ‘silicate aerogel/ballistic ceramic/high-strength polyethylene/silicate aerogel’. The study analyzes the modeling design,down-selection of materials and equipment,and real ship mock-up technical development. The reliability and application of high strength polyethylene in response to high temperatures in the real ship installation process is discussed. The results show that high-temperatures during welding have no negative impact on the high strength polyethylene of the armored structure. The design demonstrates that this installation process is feasible and can be provided as an alternative solution by virtues of its good maneuverability,controllable precision,checkable quality and high reliability.

  8. Fabrication of a full-size mock-up for inboard 10o section of ITER vacuum vessel thermal shield

    International Nuclear Information System (INIS)

    Chung, W.; Nam, K.; Noh, C.H.; Kang, D.K.; Kang, S.M.; Oh, Y.G.; Choi, S.W.; Kang, S.H.; Utin, Y.; Ioki, K.; Her, N.; Yu, J.

    2011-01-01

    A full-scale mock-up of VVTS inboard section was made in order to validate its manufacturing processes before manufacturing the vacuum vessel thermal shield (VVTS) for ITER tokamak. VVTS inboard 10 o section consists of 20 mm shells on which cooling tubes are welded and flange joints that connect adjacent thermal shield sectors. The whole VVTS inboard is divided into two by bisectional flange joint located at the center. All the manufacturing processes except silver coating were tested and verified in the fabrication of mock-up. For the forming and the welding, pre-qualification tests were conducted to find proper process conditions. Shell thickness change was measured after bending, forming and buffing processes. Shell distortion was adjusted after the welding. Welding was validated by non-destructive examination. Bisectional flange joint was successfully assembled by inserting pins and tightening with bolt/nut. Bolt hole margin of 2 mm for sector flange was revealed to be sufficient by successful sector assembly of upper and lower parts of mock-up. Handling jig was found to be essential because the inboard section was flexible. Dimensional inspection of the fabricated mock-up was performed with a 3D laser scanner.

  9. Using an integrative mock-up simulation approach for evidence-based evaluation of operating room design prototypes.

    Science.gov (United States)

    Bayramzadeh, Sara; Joseph, Anjali; Allison, David; Shultz, Jonas; Abernathy, James

    2018-07-01

    This paper describes the process and tools developed as part of a multidisciplinary collaborative simulation-based approach for iterative design and evaluation of operating room (OR) prototypes. Full-scale physical mock-ups of healthcare spaces offer an opportunity to actively communicate with and to engage multidisciplinary stakeholders in the design process. While mock-ups are increasingly being used in healthcare facility design projects, they are rarely evaluated in a manner to support active user feedback and engagement. Researchers and architecture students worked closely with clinicians and architects to develop OR design prototypes and engaged clinical end-users in simulated scenarios. An evaluation toolkit was developed to compare design prototypes. The mock-up evaluation helped the team make key decisions about room size, location of OR table, intra-room zoning, and doors location. Structured simulation based mock-up evaluations conducted in the design process can help stakeholders visualize their future workspace and provide active feedback. Copyright © 2018 Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

    Hundal, T.S.

    1995-01-01

    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

  11. High heat flux testing of ITER ICH&CD antenna beryllium faraday screen bars mock-ups

    International Nuclear Information System (INIS)

    Courtois, X.; Meunier, L.; Kuznetsov, V.; Beaumont, B.; Lamalle, P.; Conchon, D.; Languille, P.

    2016-01-01

    Highlights: • ITER ICH&CD antenna beryllium faraday screen bars mock-ups were manufactured. • The mock-ups are submitted to high heat loads to test their heat exhaust capabilities. • The mock-ups withstand without damage the design limit load. • Lifetime is gradually reduced when the heat load is augmented beyond the design limit. • Thermal and mechanical behavior are reproducible, and coherent with the calculation. - Abstract: The Faraday Screen (FS) is the plasma facing component of ITER ion cyclotron heating antennas shielding. The requirement for the high heat exhaust, and the limitation of the temperatures to minimize strain and thus offer sufficient resistance to fatigue, imply the need for high conductivity materials and a high cooling flow rate. The FS bars are constructed by a hipping process involving beryllium tiles, a pure copper layer, a copper chrome zirconium alloy for the cooling channel and a stainless steel backing strip. Two FS bars small scale mock-ups were manufactured and tested under high heat flux. They endured 15,000 heating cycles without degradation under nominal heat flux, and revealed growing flaws when the heat flux was progressively augmented beyond. In this case, the ultrasonic test confirms a strong delamination of the Be tiles.

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

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    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

  13. Improvement works report on mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system (Contract research)

    International Nuclear Information System (INIS)

    Sakaki, Akihiro; Kato, Michio; Hayashi, Koji; Fujisaki, Katsuo; Aita, Hideki; Ohashi, Hirofumi; Takada, Shoji; Shimizu, Akira; Morisaki, Norihiro; Maeda, Yukimasa; Sato, Hiroyuki; Hanawa, Hiromi; Yonekawa, Hideo; Inagaki, Yoshiyuki

    2005-04-01

    In order to establish the system integration technology to connect a hydrogen production system to a high temperature gas cooled reactor; the mock-up test facility with a full-scale reaction tube for the steam reforming HTTR hydrogen production system was constructed in fiscal year 2001 and its functional test operation was performed in the year. Seven experimental test operations were performed from fiscal year 2001 to 2004. On a period of each test operation, there happened some troubles. For each trouble, the cause was investigated and the countermeasures and the improvement works were performed to succeed the experiments. The tests were successfully achieved according to plan. This report describes the improvement works on the test facility performed from fiscal year 2001 to 2004. (author)

  14. Power reactors in member states

    International Nuclear Information System (INIS)

    1975-01-01

    This is the first issue of a periodical computer-based listing of civilian nuclear power reactors in the Member States of the IAEA, presenting the situation as of 1 April 1975. It is intended as a replacement for the Agency's previous annual publication of ''Power and Research Reactors in Member States''. In the new format, the listing contains more information about power reactors in operation, under construction, planned and shut down. As far as possible all the basic design data relating to reactors in operation have been included. In future these data will be included also for other power reactors, so that the publication will serve to give a clear picture of the technical progress achieved. Test and research reactors and critical facilities are no longer listed. Of interest to nuclear power planners, nuclear system designers, nuclear plant operators and interested professional engineers and scientists

  15. Reactivity effects due to beryllium poisoning of BR2

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  16. Fast reactors in nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Kazachkovskii, O

    1981-02-01

    The possible applications are discussed of fast reactor nuclear power plants. Basic differences are explained in fast and thermal reactors, mainly with a view to nuclear fuel utilization. Discussed in more detail are the problems of nuclear fuel reproduction and the nost important technical problems of fast reactors. Flow charts are shown of heat transfer for fast reactors BN-350 (loop design) and BN-600 (integral coolant circuit design). Main specifications are given for demonstration and power fast reactors in operation, under construction and in project-stage.

  17. Tokamak reactor startup power

    International Nuclear Information System (INIS)

    Weldon, D.M.; Murray, J.G.

    1983-01-01

    Tokamak startup with ohmic heating (OH)-induced voltages requires rather large voltages and power supplies. On present machines, with no radiofrequency (rf)-assist provisions, hundreds of volts have been specified for their designs. With the addition of electron cyclotron resonant heating (ECRH) assist, the design requirements have been lowered. To obtain information on the cost and complexity associated with this ECRH-assisted, OH-pulsed startup voltage for ignition-type machines, a trade-off study was completed. The Fusion Engineering Device (FED) configuration was selected as a model because information was available on the structure. The data obtained are applicable to all tokamaks of this general size and complexity, such as the Engineering Test Reactor

  18. Fractional power operation of tokamak reactors

    International Nuclear Information System (INIS)

    Mau, T.K.; Vold, E.L.; Conn, R.W.

    1986-01-01

    Methods to operate a tokamak fusion reactor at fractions of its rated power, identify the more effective control knobs and assess the impact of the requirements of fractional power operation on full power reactor design are explored. In particular, the role of burn control in maintaining the plasma at thermal equilibrium throughout these operations is studied. As a prerequisite to this task, the critical physics issues relevant to reactor performance predictions are examined and some insight into their impact on fractional power operation is offered. The basic tool of analysis consists of a zero-dimensional (0-D) time-dependent plasma power balance code which incorporates the most advanced data base and models in transport and burn plasma physics relevant to tokamaks. Because the plasma power balance is dominated by the transport loss and given the large uncertainty in the confinement model, the authors have studied the problem for a wide range of energy confinement scalings. The results of this analysis form the basis for studying the temporal behavior of the plasma under various thermal control mechanisms. Scenarios of thermally stable full and fractional power operations have been determined for a variety of transport models, with either passive or active feedback burn control. Important power control parameters, such as gas fueling rate, auxiliary power and other plasma quantities that affect transport losses, have also been identified. The results of these studies vary with the individual transport scaling used and, in particular, with respect to the effect of alpha heating power on confinement

  19. New control system for BR2. Preventive approach to process control

    International Nuclear Information System (INIS)

    Van den Branden, G.; Koonen, E.

    2011-01-01

    In 1961, the BR2 reactor became critical for the first time. Yet the multi-functional research reactor at SCK-CEN is not out of date, quite the contrary. Regular upgrades and innovations keep the reactor in step with the latest advancements in technology. In 2010, the control system of BR2, a vital part of the reactor, was replaced as a preventive measure.

  20. Calibration of RB reactor power

    International Nuclear Information System (INIS)

    Sotic, O.; Markovic, H.; Ninkovic, M.; Strugar, P.; Dimitrijevic, Z.; Takac, S.; Stefanovic, D.; Kocic, A.; Vranic, S.

    1976-09-01

    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

  1. Tendencies in operating power reactors

    International Nuclear Information System (INIS)

    Brinckmann, H.F.

    1987-01-01

    A survey is given about new tendencies in operating power reactors. In order to meet the high demands for control and monitoring of power reactors modern procedures are applicated such as the incore-neutron flux detection by means of electron emission detectors and multi-component activation probes, the noise diagnostics as well as high-efficient automation systems

  2. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Bertoncini, P.J.

    1976-01-01

    A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 yr. The EPR operates in a pulsed mode at a frequency of approximately 1/min, with approximately 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2 cm thick stainless steel, and has 2 cm thick detachable, beryllium-coated coolant panels mounted on the interior. A 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H 2 O. Sixteen niobium-titanium superconducting toroidal field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic heating and equilibrium field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam injectors which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-convertors

  3. Preparation of mandatory documentation before the start up of the RA-0 `zero power` nuclear reactor at Cordoba National University; Preparacion de la documentacion mandatoria para la puesta en marcha del reactor nuclear RA-0 en la Universidad Nacional de Cordoba

    Energy Technology Data Exchange (ETDEWEB)

    Martin, H R; Keil, W M; Pezzi, N

    1992-12-31

    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). [Espanol] Con motivo de la nueva puesta en servicio del REACTOR NUCLEAR RA-0 fue necesario elaborar la documentacion mandatoria requerida por la Autoridad Regulatoria Nacional. Siguiendo los lineamientos de las normas y recomendaciones vigentes e incluyendo criterios propios en lo que debia ser el contenido final de dicha documentacion, fue preparado lo que se ha denominado el INFORME DE SEGURIDAD DEL REACTOR NUCLEAR RA-0. Este documento que se describe en este trabajo, si bien contiene las habituales descripciones de todos los Informes de Seguridad, incluye otros aspectos que no siendo requeridos expresamente en el mismo, han dado una mayor coherencia a la conformacion de todos los aspectos que interrelacionan las areas de seguridad fisica, radiologica, nuclear y de control de materiales nucleares bajo salvaguardias. (Autor).

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

    International Nuclear Information System (INIS)

    Grimstone, M.J.; Stevenson, J.M.

    1979-01-01

    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)

  5. Oxidation mechanisms of CF2Br2 and CH2Br2 induced by air nonthermal plasma.

    Science.gov (United States)

    Schiorlin, Milko; Marotta, Ester; Dal Molin, Marta; Paradisi, Cristina

    2013-01-02

    Oxidation mechanisms in air nonthermal plasma (NTP) at room temperature and atmospheric pressure were investigated in a corona reactor energized by +dc, -dc, or +pulsed high voltage.. The two bromomethanes CF(2)Br(2) and CH(2)Br(2) were chosen as model organic pollutants because of their very different reactivities with OH radicals. Thus, they served as useful mechanistic probes: they respond differently to the presence of humidity in the air and give different products. By FT-IR analysis of the postdischarge gas the following products were detected and quantified: CO(2) and CO in the case of CH(2)Br(2), CO(2) and F(2)C ═ O in the case of CF(2)Br(2). F(2)C ═ O is a long-lived oxidation intermediate due to its low reactivity with atmospheric radicals. It is however removed from the NTP processed gas by passage through a water scrubber resulting in hydrolysis to CO(2) and HF. Other noncarbon containing products of the discharge were also monitored by FT-IR analysis, including HNO(3) and N(2)O. Ozone, an important product of air NTP, was never detected in experiments with CF(2)Br(2) and CH(2)Br(2) because of the highly efficient ozone depleting cycles catalyzed by BrOx species formed from the bromomethanes. It is concluded that, regardless of the type of corona applied, CF(2)Br(2) reacts in air NTP via a common intermediate, the CF(2)Br radical. The possible reactions leading to this radical are discussed, including, for -dc activation, charge exchange with O(2)(-), a species detected by APCI mass spectrometry.

  6. Development of refined MCNPX-PARET multi-channel model for transient analysis in research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kalcheva, S.; Koonen, E. [SCK-CEN, BR2 Reactor Dept., Boeretang 200, 2400 Mol (Belgium); Olson, A. P. [RERTR Program, Nuclear Engineering Div., Argonne National Laboratory, Cass Avenue, Argonne, IL 60439 (United States)

    2012-07-01

    Reactivity insertion transients are often analyzed (RELAP, PARET) using a two-channel model, representing the hot assembly with specified power distribution and an average assembly representing the remainder of the core. For the analysis of protected by the reactor safety system transients and zero reactivity feedback coefficients this approximation proves to give adequate results. However, a more refined multi-channel model representing the various assemblies, coupled through the reactivity feedback effects to the whole reactor core is needed for the analysis of unprotected transients with excluded over power and period trips. In the present paper a detailed multi-channel PARET model has been developed which describes the reactor core in different clusters representing typical BR2 fuel assemblies. The distribution of power and reactivity feedback in each cluster of the reactor core is obtained from a best-estimate MCNPX calculation using the whole core geometry model of the BR2 reactor. The sensitivity of the reactor response to power, temperature and energy distributions is studied for protected and unprotected reactivity insertion transients, with zero and non-zero reactivity feedback coefficients. The detailed multi-channel model is compared vs. simplified fewer-channel models. The sensitivities of transient characteristics derived from the different models are tested on a few reactivity insertion transients with reactivity feedback from coolant temperature and density change. (authors)

  7. Realisation of a test facility for the ITER ICRH antenna plug-in by means of a mock-up with salted water load

    International Nuclear Information System (INIS)

    Messiaen, A.; Dumortier, P.; Koch, R.; Lamalle, P.; Louche, F.; Martini, J.L.; Vervier, M.

    2005-01-01

    By the use of a mock-up operated at higher frequency it is possible to measure with good accuracy the rf characteristics of an ICRH antenna, the plasma loading being simulated by a water tank in front of it. This concept has motivated the construction of the mock-up of the antenna array foreseen for ITER

  8. Visualization test using piping group mock up specimen for evaluation of wastage phenomena in steam generator for FBR

    International Nuclear Information System (INIS)

    Kato, Keisuke; Yoshida, Atsuro; Arae, Kunihiko; Narabayashi, Tadashi; Ohshima, Hiroyuki; Kurihara, Akikazu

    2012-01-01

    There is a need for quantitative evaluation of wastage phenomena in steam generator for FBR. We focused attention on liquid droplet impingement erosion (LDIE) in wastage phenomena and performed basic study with piping group mock up specimen for quantitative evaluation of LDIE. First, we did visualization test of high pressure and high speed jet into the water. Test section mock up the crack of heat exchanger tube and neighboring heat exchanger tubes. We did the test under the following test conditions. Upstream pressure is 0.3MPa, vapor temperature is 300K, crack width is 0.1mm, and crack length is 40mm. (crack diameter is 0.2mm) Second, we did pressure and temperature measurement test in the same test conditions as before. We evaluated jet behavior at test section by those two tests. In addition, we did two phase flow analysis of the jet with TRAC code. (author)

  9. Mock-up experiment at Birmingham University for BNCT project of Osaka University – Neutron flux measurement with gold foil

    International Nuclear Information System (INIS)

    Tamaki, S.; Sakai, M.; Yoshihashi, S.; Manabe, M.; Zushi, N.; Murata, I.; Hoashi, E.; Kato, I.; Kuri, S.; Oshiro, S.; Nagasaki, M.; Horiike, H.

    2015-01-01

    Mock-up experiment for development of accelerator based neutron source for Osaka University BNCT project was carried out at Birmingham University, UK. In this paper, spatial distribution of neutron flux intensity was evaluated by foil activation method. Validity of the design code system was confirmed by comparing measured gold foil activities with calculations. As a result, it was found that the epi-thermal neutron beam was well collimated by our neutron moderator assembly. Also, the design accuracy was evaluated to have less than 20% error. - Highlights: • Accelerator based neutron source for BNCT is being developed in Osaka University. • Mock-up experiment was carried out at Birmingham University, UK. • Neutronics performance of our assembly was evaluated from gold foil activation. • Gold foil activation was determined by using HPGe detectors. • Validity of the neutronics design code system was confirmed.

  10. A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

    International Nuclear Information System (INIS)

    Nakahira, M.; Takeda, N.; Kakudate, S.; Onozuka, M.

    2008-01-01

    The structure and fabrication methods of the ITER vacuum vessel (VV) have been investigated and defined by the ITER International Team (IT). However, some of the current technical specifications are difficult to be achieved from the manufacturing point of view. To solve such an issue, this paper proposes an alternative specification of the VV to the IT's design. A series of the fabrication and assembly procedures for the mock-up are presented, together with candidates of welding configurations. Finally, the paper summarizes the results of mock-up fabrication, such as non-destructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. Based on the results, it is suggested that several issues such as clarification of conditions of repair welding, demonstration of welding distortion control and detectability/localization of internal defects should be solved before manufacturing the ITER VV

  11. A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

    Energy Technology Data Exchange (ETDEWEB)

    Nakahira, M. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan)], E-mail: nakahira.masataka@jaea.go.jp; Takeda, N.; Kakudate, S. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Onozuka, M. [Mitsubishi Nuclear Energy Systems, Inc., 1700K Street NW, Suite 440, Washington, DC 20006 (United States)

    2008-12-15

    The structure and fabrication methods of the ITER vacuum vessel (VV) have been investigated and defined by the ITER International Team (IT). However, some of the current technical specifications are difficult to be achieved from the manufacturing point of view. To solve such an issue, this paper proposes an alternative specification of the VV to the IT's design. A series of the fabrication and assembly procedures for the mock-up are presented, together with candidates of welding configurations. Finally, the paper summarizes the results of mock-up fabrication, such as non-destructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. Based on the results, it is suggested that several issues such as clarification of conditions of repair welding, demonstration of welding distortion control and detectability/localization of internal defects should be solved before manufacturing the ITER VV.

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

    International Nuclear Information System (INIS)

    Hamada, Shozo; Fukaya, Kiyoshi; Kato, Chiaki; Yanagihara, Takao; Doi, Masamitu; Kiuchi, Kiyoshi

    2001-10-01

    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)

  13. Original Research. Statistical Study Regarding Differences Between the Wax-Up, Mock-Up, and Final Restoration

    Directory of Open Access Journals (Sweden)

    Jánosi Kinga

    2017-03-01

    Full Text Available The aesthetic rehabilitation of patients remains a challenge for practicians. To facilitate the clinicians’ and technicians’ task, several innovative methods were developed, like the diagnostic wax-up and mock-up. The width-to-length ratio of the maxillary frontal teeth can be used to evaluate dentofacial aesthetics. Our study presents the variations between the teeth size measured on casts obtained during the prosthodontic treatment.

  14. Safety of nuclear power reactors

    International Nuclear Information System (INIS)

    MacPherson, H.G.

    1982-01-01

    Safety is the major public issue to be resolved or accommodated if nuclear power is to have a future. Probabilistic Risk Analysis (PRA) of accidental releases of low-level radiation, the spread and activity of radiation in populated areas, and the impacts on public health from exposure evolved from the earlier Rasmussen Reactor Safety Study. Applications of the PRA technique have identified design peculiarities in specific reactors, thus increasing reactor safety and establishing a quide for evaluating reactor regulations. The Nuclear Regulatory Commission and reactor vendors must share with utilities the responsibility for reactor safety in the US and for providing reasonable assurance to the public. This entails persuasive public education and information that with safety a top priority, changes now being made in light water reactor hardware and operations will be adequate. 17 references, 2 figures, 2 tables

  15. A nuclear power reactor

    International Nuclear Information System (INIS)

    Borrman, B.E.; Broden, P.; Lundin, N.

    1979-12-01

    The invention consists of shock absorbing support beams fastened to the underside of the reactor tank lid of a BWR type reactor, whose purpose is to provide support to the steam separator and dryer unit against accelerations due to earthquakes, without causing undue thermal stresses in the unit due to differential expansion. (J.I.W.)

  16. Measurement and analysis of neutron flux spectra in a neutronics mock-up of the HCLL test blanket module

    International Nuclear Information System (INIS)

    Klix, A.; Batistoni, P.; Boettger, R.; Lebrun-Grandie, D.; Fischer, U.; Henniger, J.; Leichtle, D.; Villari, R.

    2010-01-01

    Fast neutron and gamma-ray flux spectra and time-of-arrival spectra of slow neutrons have been measured in a neutronics mock-up of the European Helium-Cooled Lithium-Lead Test Blanket Module with the aim to validate nuclear cross-section data. The mock-up was irradiated with fusion peak neutrons from the DT neutron generator of the Technical University of Dresden. A well characterized cylindrical NE-213 scintillator was inserted into two positions in the LiPb/EUROFER assembly. Pulse height spectra from neutrons and gamma-rays were recorded from the NE-213 output. The spectra were then unfolded with experimentally obtained response matrices of the NE-213 detector. Time-of-arrival spectra of slow neutrons were measured with a 3 He counter placed in the mock-up, and the neutron generator was operated in pulsed mode. Monte Carlo calculations using the MCNP code and nuclear cross-section data from the JEFF-3.1.1 and FENDL-2.1 libraries were performed and the results are compared with the experimental results. A good agreement of measurement and calculation was found with some deviations in certain energy intervals.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

    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

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

    International Nuclear Information System (INIS)

    Uda, M.; Iwadachi, T.; Uchida, M.; Yamada, H.; Nakamichi, M.; Kawamura, H.

    2004-01-01

    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/m 2 , 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)

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

    International Nuclear Information System (INIS)

    Young-Joon Shin; Ik-Soo Kim; Seung-Chul Oh; Soo-Haeng Cho; Yo-Taik Song; Hyun-Soo Park

    2000-01-01

    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)

  20. Status report about the works for the start up of the RA-0 `zero power` nuclear reactor at the Cordoba National University; Estado actual de avance de las tareas para la nueva puesta en marcha del reactor nuclear RA-0 en la Universidad Nacional de Cordoba

    Energy Technology Data Exchange (ETDEWEB)

    Martin, H R; Carballido, C; Oliveras, T

    1992-12-31

    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). [Espanol] Luego de aproximadamente dos anos de trabajo para la nueva puesta en marcha del REACTOR NUCLEAR RA-0, se han alcanzado los resultados presentados en este trabajo. Partiendo de una infraestructura practicamente inexistente en cuanto a recursos humanos y estado de las instalaciones, los avances logrados son significativos. Comenzando por la capacitacion y el entrenamiento del futuro personal de operacion y pasando por la adecuacion de los equipos y componentes, hasta la confeccion de la documentacion mandatoria, se muestran los aspectos mas destacables de los trabajos realizados. Una atencion especial se dedica a la insercion de una instalacion de este tipo en el ambito universitario, el cual por sus particulares caracteristicas, ha debido ser tenido en cuenta permanentemente para la futura operacion de las instalaciones. (Autor).

  1. Program of RA reactor start-up to nominal power

    International Nuclear Information System (INIS)

    1959-01-01

    The zero start-up program is followed by the program of RA reactor start-up to nominal power. This program is described in detail and includes the following measurements: radiation characteristics at the exit of the channels; gamma and fast neutron dose distribution in the reactor; influence of absorbers on the reactivity; temperature effect; absolute flux and calibration of ionization chambers; xenon effect; thermal and hydraulics; dosimetry around the reactor; neutron flux in the reactor core and in the reactor hall; heavy water level; thermal characteristics after shutdown. A list of measuring devices and instrumentation is included with the detailed action plan and list of responsible staff members

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

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

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

  3. Fractals in Power Reactor Noise

    International Nuclear Information System (INIS)

    Aguilar Martinez, O.

    1994-01-01

    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

  4. Qualification and post-mortem characterization of tungsten mock-ups exposed to cyclic high heat flux loading

    Energy Technology Data Exchange (ETDEWEB)

    Pintsuk, G., E-mail: g.pintsuk@fz-juelich.de [Forschungszentrum Jülich GmbH, Euratom Association, D-52425 Jülich (Germany); Bobin-Vastra, I.; Constans, S. [AREVA NP PTCMI-F, Centre Technique, Fusion, F-71200 Le Creusot (France); Gavila, P. [Fusion for Energy, E-08019 Barcelona (Spain); Rödig, M. [Forschungszentrum Jülich GmbH, Euratom Association, D-52425 Jülich (Germany); Riccardi, B. [Fusion for Energy, E-08019 Barcelona (Spain)

    2013-10-15

    Highlights: • We characterize tungsten mono-block components after exposure to ITER relevant heat loads. • We qualify the manufacturing technology, i.e., hot isostatic pressing and hot radial pressing, and repair technologies. • We determine the microstructural influences, i.e., rod vs. plate material, on the damage evolution. • Needle like microstructures increase the risk of deep crack formation due to a limited fracture strength. -- Abstract: In order to evaluate the option to start the ITER operation with a full tungsten (W) divertor, high heat flux tests were performed in the electron beam facility FE200, Le Creusot, France. Thereby, in total eight small-scale and three medium-scale monoblock mock-ups produced with different manufacturing technologies and different tungsten grades were exposed to cyclic steady state heat loads. The applied power density ranges from 10 to 20 MW/m{sup 2} with a maximum of 1000 cycles at each particular loading step. Finally, on a reduced number of tiles, critical heat flux tests in the range of 30 MW/m{sup 2} were performed. Besides macroscopic and microscopic images of the loaded surface areas, detailed metallographic analyses were performed in order to characterize the occurring damages, i.e., crack formation, recrystallization, and melting. Thereby, the different joining technologies, i.e., hot radial pressing (HRP) vs. hot isostatic pressing (HIP) of tungsten to the Cu-based cooling tube, were qualified showing a higher stability and reproducibility of the HIP technology also as repair technology. Finally, the material response at the loaded top surface was found to be depending on the material grade, microstructural orientation, and recrystallization state of the material. These damages might be triggered by the application of thermal shock loads during electron beam surface scanning and not by the steady state heat load only. However, the superposition of thermal fatigue loads and thermal shocks as also expected

  5. Power reactors in Member States. 1978 edition

    International Nuclear Information System (INIS)

    1978-01-01

    The computer-based reactor listing gives information on reactor core characteristics and plant systems for all power reactors in operation under construction and planned. The following two tables are included to give a general picture of the overall situation: Reactor types and net electrical power; Reactor units and net electrical power by country and cumulated by year

  6. 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.

  7. The probability safety assessment impact on the BR2 refurbishment

    International Nuclear Information System (INIS)

    Pouleur, Yvan

    1995-01-01

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

  8. Power reactor information system (PRIS)

    International Nuclear Information System (INIS)

    1989-06-01

    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

  9. Conceptual design of a First Wall mock-up experiment in preparation for the qualification of breeding blanket technologies in the Helium Loop Karlsruhe (HELOKA) facility

    Energy Technology Data Exchange (ETDEWEB)

    Zeile, C., E-mail: christian.zeile@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Abou-Sena, A.; Boccaccini, L.V.; Ghidersa, B.E.; Kang, Q.; Kunze, A. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Lamberti, L. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Dipartimento Energia, Politecnico di Torino (Italy); Maione, I.A.; Rey, J.; Weth, A. von der [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2016-11-01

    Highlights: • Experiment in preparation for the qualification of Breeding Blanket technologies in HELOKA facility is proposed. • Experimental capabilities, instrumentation of the mock-up and experimental program are presented. • Design and manufacturing of the mock-up is described. • Design of modular attachment system to obtain different stress levels and distributions on the mock-up is discussed. - Abstract: An experimental program based on a First Wall mock-up is presented as preparation for the qualification of breeding blanket mock-ups at high heat flux in the Helium Loop Karlsruhe (HELOKA) facility. Two objectives of the experimental program have been defined: testing of the experimental setup and a first validation of FE models. The design and manufacturing of mock-up representing about 1/3 of the heated zone of an ITER Test Blanket Module (TBM) First Wall is discussed. A modular attachment system concept has been developed for the fixation of the mock-up in order to be able to generate different stress distributions and levels on the plate, which is confirmed by thermo-mechanical analyses. The HELOKA facility is able to provide a TBM relevant helium cooling system and to generate the required surface heat flux by an electron beam gun. An installed IR camera can be used to measure the temperature distribution on the surface.

  10. Power oscillations in BWR reactors

    International Nuclear Information System (INIS)

    Espinosa P, G.

    2002-01-01

    One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

  11. Refurbishment of BR2 (Phase 4 and 5)

    International Nuclear Information System (INIS)

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

    1998-01-01

    The extensive refurbishment of the BR-2 materials testing reactor should allow another 10 to 15 years 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

  12. Power Reactor Information System (PRIS)

    International Nuclear Information System (INIS)

    Spiegelberg, R.

    1992-01-01

    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.)

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

    International Nuclear Information System (INIS)

    Conte, F.; Dambrine, C.; Gaussot, D.

    1963-01-01

    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) [fr

  14. Development of Digital Mock-Up for the Assessment of Dismantling Scenarios

    International Nuclear Information System (INIS)

    Kim, Sung-Kyun; Park, Hee-Sung; Lee, Kune-Woo; Jung, Chong-Hun

    2008-01-01

    As the number of superannuated research reactors and nuclear power plants increase, dismantling nuclear power facilities has become a big issue. However, decommissioning a nuclear facility is still a costly and possibly hazardous task. So prior to an actual decommission, what should be done foremost is to establish a proper procedure. Due to the fact that a significant difference in cost, exposure to a radiation, and safety might occur, a proper procedure is imperative for the entire engineering process. The purpose of this paper is to develop a system for evaluating the decommissioning scenarios logically and systematically. So a digital mockup system with functions such as a dismantling schedule, decommissioning costs, wastes, worker's exposure dose, and a radiation distribution was developed. Also on the basis of the quantitative information calculated from a DMU system and the data evaluated by decommissioning experts about qualitatively evaluating the items, the best decommissioning scenarios were established by using the analytic hierarchy process (AHP) method. Finally, the DMU was implemented in the thermal column of KRR-1 and adequate scenarios were provided after comparing and analyzing the two scenarios. In this paper, we developed the virtual environment of KRR-1 by using computer graphic technology and simulating the dismantling processes. The data-computing modules were also developed for quantitatively comparing the decommissioning scenarios. The decommissioning DMU system was integrated with both the VE system and the data-computing modules. In addition, we presented a decision-making method for selecting the best decommissioning scenario through the AHP. So the scenarios can be evaluated logically and quantitatively through the decommissioning DMU. As an implementation of the AHP, the plasma cutting scenario and the nibbler cutting scenario of the thermal column were prioritized. The fact that the plasma cutting scenario ranked the better than the

  15. Full-scale model development of the WWER-440 reactor fuel rod bundle for core temperature regime study under reflooding conditions

    International Nuclear Information System (INIS)

    Bezrukov, Yu.A.; Logvinov, S.A.; Levchuk, S.V.; Nakladnov, V.D.; Onshin, V.P.; Sokolov, A.S.

    1982-01-01

    Consideration is given to the issues of a full scale WWER-440 fuel rod bundle imitation. An imitator contains a molybdenum heating rod inclosed in stainless steel shell. The shell diameter is 9 mm, the heated length is 2500 mm, the total len.o.th is 2855 mm. 125 fuel rod imitators are set in the bundle mock-up. The experiments were run on a test facility imitating the WWER-440 reactor primary loop, providing the conditions of the loop breaking. The mock-up thermal hydraulics has been studied during the refloodino. stage. The mock-up was heated up to predetermined initial temperature at a low power level with saturated steam cooling. Then the steam input was stopped, the power level rarapidly rised up to a given value and the cooling water injected. Simultaneously with water injection all the measured parameters monitoring was started. Both at the top spraying and combined cooling temperature oscillations in the upper and middle parts of the mock-up were observed. At the bottom reflooding the mock-up cooling down took more time, thereat temperature inthe upper part first slowly rised during reflooding then decreased and then dropped abruptly at thefront coming up [ru

  16. Nuclear power reactor technology

    International Nuclear Information System (INIS)

    1978-09-01

    Risoe National Laboratory was established more than twenty years ago with research and development of nuclear reactor technology as its main objective. The Laboratory has by now accumulated many years of experience in a number of areas vital to nuclear reactor technology. The work and experience of, and services offered by the Laboratory within the following fields are described: Health physics site supervision; Treatment of low and medium level radioactive waste; Core performance evaluation; Transient analysis; Accident analysis; Fuel management; Fuel element design, fabrication and performance evaluation; Non-destructive testing of nuclear fuel; Theoretical and experimental structural analysis; Reliability analysis; Site evaluation. Environmental risk and hazard calculation; Review and analysis of safety documentation. Risoe has already given much assistance to the authorities, utilities and industries in such fields, carrying out work on both light and heavy water reactors. The Laboratory now offers its services to others as a consultant, in education and training of staff, in planning, in qualitative and quantitative analysis, and for the development and specification of fabrication techniques. (author)

  17. The Mock-up of the "Ratto Delle Sabine" by Giambologna: Making and Utilization of a 3D Model

    Directory of Open Access Journals (Sweden)

    Grazia Tucci

    2015-02-01

    Full Text Available Within a project for the knowledge and preservation of the mock-up of Giambologna's Ratto delle Sabine housed in the Galleria dell'Accademia in Florence, the GeCO laboratory has made laser scanner acquisitions to create surface models at different resolutions for structural analysis, on which to check the coverage of the photographic campaign and to create a three-dimensional thematic mapping of data relating to investigations and restoration works. The PDF3D file format has been used to easily manage data on a platform immediately available to all operators.

  18. The BR2 refurbishment: from concept to achievements

    International Nuclear Information System (INIS)

    Gubel, P.

    2002-01-01

    The BR2 reactor is one of the major research reactors in the world. It's operation started in the early 1960's. Two major refurbishments operation have been carried out since then. The report gives an overview of the methodology and inspections, which resulted in a refurbishment action plan. The main realizations and complementary actions required by the Licensing Authorities are summarized. Finally the operation experience feedback, four years now after start-up, is briefly discussed as well as the main aspects of the present safety reassessment [ru

  19. Cascade ICF power reactor

    International Nuclear Information System (INIS)

    Hogan, W.J.; Pitts, J.H.

    1986-01-01

    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 LiAlO 2 . 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

  20. Mock-up tests of rail-mounted vehicle type in-vessel transporter/manipulator

    International Nuclear Information System (INIS)

    Oka, K.; Kakaudate, S.; Fukatsu, S.

    1995-01-01

    A rail-mounted vehicle system has been developed for remote maintenance of in-vessel components for fusion experimental reactor. In this system, a rail deploying/storing system is installed at outside of the reactor core and used to deploy a rail transporter and vehicle/manipulator for the in-vessel maintenance. A prototype of the rail deploying/storing system has been fabricated for mockup tests. This paper describes structural design of the prototypical rail deploying/storing system and results of the performance tests such as payload capacity, position control and rail deployment/storage performance

  1. 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.

  2. Nuclear reactor power control device

    International Nuclear Information System (INIS)

    Koshi, Yuji; Sakata, Akira; Karatsu, Hiroyuki.

    1987-01-01

    Purpose: To control abrupt changes in neutron fluxes by feeding back a correction signal obtained from a deviation between neutron fluxes and heat fluxes for changing the reactor core flow rate to a recycling flow rate control system upon abrupt power change of a nuclear reactor. Constitution: In addition to important systems, that is, a reactor pressure control system and a recycling control system in the power control device of a BWR type power plant, a control circuit for feeding back a deviation between neutron fluxes and heat fluxes to a recycling flow rate control system is disposed. In the suppression circuit, a deviation signal is prepared in an adder from neutron flux and heat flux signals obtained through a primary delay filter. The deviation signal is passed through a dead band and an advance/delay filter into a correction signal, which is adapted to be fed back to the recycling flow rate control system. As a result, the reactor power control can be conducted smoothly and it is possible to effectively suppress the abrupt change or over shoot of the neutron fluxes and abrupt power change. (Kamimura, M.)

  3. 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.

  4. A comparative ab initio study of Br2*- and Br2 water clusters.

    Science.gov (United States)

    Pathak, A K; Mukherjee, T; Maity, D K

    2006-01-14

    The work presents ab initio results on structure and electronic properties of Br2*-.nH2O(n=1-10) and Br2.nH2O(n=1-8) hydrated clusters to study the effects of an excess electron on the microhydration of the halide dimer. A nonlocal density functional, namely, Becke's half-and-half hybrid exchange-correlation functional is found to perform well on the present systems with a split valence 6-31++G(d,p) basis function. Geometry optimizations for all the clusters are carried out with several initial guess structures and without imposing any symmetry restriction. Br2*-.nH2O clusters prefer to have symmetrical double hydrogen-bonding structures. Results on Br2.nH2O(n>or=2) cluster show that the O atom of one H2O is oriented towards one Br atom and the H atom of another H2O is directed to other Br atom making Br2 to exist as Br+-Br- entity in the cluster. The binding and solvation energies are calculated for the Br2*-.nH2O and Br2.nH2O clusters. Calculations of the vibrational frequencies show that the formation of Br2*- and Br2 water clusters induces significant shifts from the normal modes of isolated water. Excited-state calculations are carried out on Br2*-.nH2O clusters following configuration interaction with single electron excitation procedure and UV-VIS absorption profiles are simulated. There is an excellent agreement between the present theoretical UV-VIS spectra of Br2*-.10H2O cluster and the reported transient optical spectra for Br2*- in aqueous solution.

  5. Damage and fatigue crack growth of Eurofer steel first wall mock-up under cyclic heat flux loads. Part 2: Finite element analysis of damage evolution

    International Nuclear Information System (INIS)

    You, Jeong-Ha

    2014-01-01

    Highlights: • The surface heat flux load of 3.5 MW/m 2 produced substantial stresses and inelastic strains in the heat-loaded surface region, especially at the notch root. • The notch root exhibited a typical notch effect such as stress concentration and localized inelastic yield leading to a preferred damage development. • The predicted damage evolution feature agrees well with the experimental observation. • The smooth surface also experiences considerable stresses and inelastic strains. However, the stress intensity and the amount of inelastic deformation are not high enough to cause any serious damage. • The level of maximum inelastic strain is higher at the notch root than at the smooth surface. On the other hand, the amplitude of inelastic strain variation is comparable at both positions. • The amount of inelastic deformation is significantly affected by the length of pulse duration time indicating the important role of creep. - Abstract: In the preceding companion article (part 1), the experimental results of the high-heat-flux (3.5 MW/m 2 ) fatigue tests of a Eurofer bare steel first wall mock-up was presented. The aim was to investigate the damage evolution and crack initiation feature. The mock-up used there was a simplified model having only basic and generic structural feature of an actively cooled steel FW component for DEMO reactor. In that study, it was found that microscopic damage was formed at the notch root already in the early stage of the fatigue loading. On the contrary, the heat-loaded smooth surface exhibited no damage up to 800 load cycles. In this paper, the high-heat-flux fatigue behavior is investigated with a finite element analysis to provide a theoretical interpretation. The thermal fatigue test was simulated using the coupled damage-viscoplastic constitutive model developed by Aktaa. The stresses, inelastic deformation and damage evolution at the notch groove and at the smooth surface are compared. The different damage

  6. Power control system in BWR type reactors

    International Nuclear Information System (INIS)

    Nishizawa, Yasuo.

    1980-01-01

    Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

  7. Low power unattended defense reactor

    International Nuclear Information System (INIS)

    Kirchner, W.L.; Meier, K.L.

    1984-01-01

    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 (OandM) costs as well as no fuel logistics, it is particularly attractive for remote, unattended applications such as the North Warning System

  8. Low power unattended defense reactor

    International Nuclear Information System (INIS)

    Kirchner, W.L.; Meier, K.L.

    1984-01-01

    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

  9. Power Reactor Embrittlement Data Base

    International Nuclear Information System (INIS)

    Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

    1990-01-01

    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

  10. Power reactor core safety research

    International Nuclear Information System (INIS)

    Rim, C.S.; Kim, W.C.; Shon, D.S.; Kim, J.

    1981-01-01

    As a part of nuclear safety research program, a project was launched to develop a model to predict fuel failure, to produce the data required for the localizaton of fuel design and fabrication technology, to establish safety limits for regulation of nuclear power plants and to develop reactor operation method to minimize fuel failure through the study of fuel failure mechanisms. During 1980, the first year of this project, various fuel failure mechanisms were analyzed, an experimental method for out-of-pile tests to study the stress corrosion cracking (SCC) behaviour of Zircaloy cladding underiodine environment was established, and characteristics of PWR and CANDU Zircaloy specimens were examined. Also developed during 1980 were the methods and correlations to evaluate fuel failures in the reactor core based on operating data from power reactors

  11. Experimental power reactor

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    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

  12. 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)

    Shimizu, Akira; Ohashi, Hirofumi; Kato, Michio; Hayashi, Koji; Aita, Hideki; Nishihara, Tetsuo; Inaba, Yoshitomo; Takada, Shoji; Morisaki, Norihiro; Sakaki, Akihiro; Maeda, Yukimasa; Sato, Hiroyuki; Inagaki, Yoshiyuki; Hanawa, Hiromi; Fujisaki, Katsuo; Yonekawa, Hideo

    2005-06-01

    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 Nm 3 /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)

  13. Dimensioning the EVITA semi-open loop at BR2 for qualification of full size JHR fuel elements

    International Nuclear Information System (INIS)

    Gouat, Philippe

    2011-01-01

    Research highlights: → Research reactor fuel (LEU) qualification as part of the licensing process of the JHR reactor. → Thermal-hydraulic dimensioning process of fuel irradiation installation. → We compare the predicted pressure profile in the installation with in situ measured values. - Abstract: The Jules Horowitz Reactor (JHR) is the next generation research reactor from CEA and which commissioning is foreseen in 2014. Prior to acquiring the exploitation license, the fuel elements have to be qualified for their intended functioning power. The only facility capable to perform this task is the Belgian research reactor BR2, due to its similar thermal-hydraulic parameters. At the moment, one has already tested the fuel plates separately. The preparation of the JHR safety report still needs the test of full size elements. This JHR fuel element is broader and more powerful than a standard BR2 fuel element, and one cannot perform an irradiation by simply interchanging them. However, BR2 has 200 mm channels at its disposal, which can be adapted to give the correct hydraulic diameter. One also needs an additional pump to deliver the necessary cooling flow rate for the higher power. This paper describes the dimensioning of the EVITA semi-open loop, which has been built at BR2 to irradiate full size JHR fuel elements and qualify them for the foreseen exploitation parameters. One explains here the followed methodology to quantify the required additional head for the booster pump and to determine the pressure profile along the circuit and the safety margin on the fuel. This methodology relies only on a priori calculations without any measurement on full size installation subpart as usual before the assembly in controlled zone. The article also explains how the original JHR thermal hydraulic safety calculation scheme was adapted to the BR2 environment. One also compares the measurement results on the fully built installation with our previsions. Our models compare well

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

    International Nuclear Information System (INIS)

    Vold, E.L.; Mau, T.K.; Conn, R.W.

    1986-06-01

    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

  15. Ultrasonic testing results of fatigue cracks in PWR mock-up

    International Nuclear Information System (INIS)

    Gondard, C.

    1990-01-01

    The Ispra Joint Research Center has entered, since many years a study on fatigue crack propagation in PWR reactor vessels. The objective of this study is to establish a relation between the size and the location of defects and the lifetime of the vessel. For verifying the theoretical models validity a mockup has been built. This document gives the results of CEA for 6 in service inspection during 5 years [fr

  16. Reactor power peaking information display

    International Nuclear Information System (INIS)

    Book, T.L.; Kochendarfer, R.A.

    1986-01-01

    This patent describes a system for monitoring operating conditions within a nuclear reactor. The system consists of a method for measuring the operating parameters within the nuclear reactor, including the position of axial power shaping rods and regulating control rod. It also includes a method for determining from the operating parameters the operating limits before a power peaking condition exists within the nuclear reactor, and a method for displaying the operating limits which consists of a visual display permitting the continuous monitoring of the operating conditions within the nuclear reactor as a graph of the shaping rod position vs the regulating rod position having a permissible area and a restricted area. The permissible area is further divided into a recommended operating area for steady state operation and a cursor located on the graph to indicate the present operating condition of the nuclear reactor to allow an operator to view any need for corrective action based on the movement of the cursor out of the recommended operating area and to take any corrective transient action within the permissible area

  17. Reactor power region measuring device

    International Nuclear Information System (INIS)

    Kashiwa, Takao.

    1996-01-01

    The device of the present invention can rapidly detect abnormality of a local power region monitor (LPRM) even at a low power region caused such as upon start-up of a BWR type reactor. Namely, the present invention comprises (1) an LPRM detector for measuring neutron fluxes in the reactor, (2) a gamma thermo detector for calibrating the sensitivity of the LPRM detector, (3) a comparison circuit for comparing the detected values of the detectors (1) and (2), and (4) an alarm circuit for outputting an alarm when the comparative difference of the output of the circuit (3) exceeds a predetermined value. Signals of an alarm for a lower limit of the LPRM detector have been issued continuously upon start-up and shut down of the reactor since neutron fluxes in the reactor are reduced. However, the gamma thermo detector is always secured in the inside of the reactor different from a travelling-type incore probe monitor (TIP) disposed so far for the same purpose. Accordingly, the alarm generated upon usual start-up can be eliminated by comparing the detected values of the detector (2) and abnormality of the detector (1) can be rapidly detected by judging the abnormality of the comparative difference. (I.S.)

  18. Preparation of 3D Printed Divertor Mock-up Design and Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Park, Sung Dae; Kim, Dong Jun; Kim, Suk Kwon; Lee, Eo Hwak [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The divertor for fusion reactor is known to be able to remove the extreme heat flux up to 10 MW/m2 and the various type of divertors have been developed for enhancing the heat transfer such as hypervapotron, twisted tape insertion, screwed tube, and so on. In order to overcome this limitation, 3D printing method is considered to be used in the fusion reactor divertor design in present study. With the advantages of the 3D printing, the various shapes of the inner divertor cooling tube are investigated to enhance the turbulence of coolant and to reduce the pressure drop. The metallic powder of the fusion reactor candidate material is produced as the preliminary step for using in 3D printer. The material is a reduced activation ferritic-matensitic steel named as ARAA (Advanced Reduced Activation Alloy) which have been independently developed in Korea. Gas atomization method was used to make the spherical particles with average diameter of 100 μm. Several candidates were presented to achieve the excellent heat removal capacity and the low pressure drop. Thermal-hydraulic analysis was performed to confirm the effects of the inner cooling tube geometry with a conventional CFD code, ANSYS-CFX v14.5. The modified screw type called as a rail type twisted tube was presented through the optimization process. This complicated tube could be made by 3D printing technology. (metallic powder). Thermal-hydraulic analysis was conducted to compare the 3 type geometric divertor. A rail type twisted tube has good heat transfer performance in comparison with a conventional twisted tube. The pressure drop of a rail type twisted tube was reduced about 36% compared with a conventional twisted tube.

  19. Preparation of 3D Printed Divertor Mock-up Design and Fabrication

    International Nuclear Information System (INIS)

    Lee, Dong Won; Park, Sung Dae; Kim, Dong Jun; Kim, Suk Kwon; Lee, Eo Hwak

    2016-01-01

    The divertor for fusion reactor is known to be able to remove the extreme heat flux up to 10 MW/m2 and the various type of divertors have been developed for enhancing the heat transfer such as hypervapotron, twisted tape insertion, screwed tube, and so on. In order to overcome this limitation, 3D printing method is considered to be used in the fusion reactor divertor design in present study. With the advantages of the 3D printing, the various shapes of the inner divertor cooling tube are investigated to enhance the turbulence of coolant and to reduce the pressure drop. The metallic powder of the fusion reactor candidate material is produced as the preliminary step for using in 3D printer. The material is a reduced activation ferritic-matensitic steel named as ARAA (Advanced Reduced Activation Alloy) which have been independently developed in Korea. Gas atomization method was used to make the spherical particles with average diameter of 100 μm. Several candidates were presented to achieve the excellent heat removal capacity and the low pressure drop. Thermal-hydraulic analysis was performed to confirm the effects of the inner cooling tube geometry with a conventional CFD code, ANSYS-CFX v14.5. The modified screw type called as a rail type twisted tube was presented through the optimization process. This complicated tube could be made by 3D printing technology. (metallic powder). Thermal-hydraulic analysis was conducted to compare the 3 type geometric divertor. A rail type twisted tube has good heat transfer performance in comparison with a conventional twisted tube. The pressure drop of a rail type twisted tube was reduced about 36% compared with a conventional twisted tube

  20. Time-zero efficiency of European power derivatives markets

    International Nuclear Information System (INIS)

    Peña, Juan Ignacio; Rodriguez, Rosa

    2016-01-01

    We study time-zero efficiency of electricity derivatives markets. By time-zero efficiency is meant a sequence of prices of derivatives contracts having the same underlying asset but different times to maturity which implies that prices comply with a set of efficiency conditions that prevent profitable time-zero arbitrage opportunities. We investigate whether statistical tests, based on the law of one price, and trading rules, based on price differentials and no-arbitrage violations, are useful for assessing time-zero efficiency. We apply tests and trading rules to daily data of three European power markets: Germany, France and Spain. In the case of the German market, after considering liquidity availability and transaction costs, results are not inconsistent with time-zero efficiency. However, in the case of the French and Spanish markets, limitations in liquidity and representativeness are challenges that prevent definite conclusions. Liquidity in French and Spanish markets should improve by using pricing and marketing incentives. These incentives should attract more participants into the electricity derivatives exchanges and should encourage them to settle OTC trades in clearinghouses. Publication of statistics on prices, volumes and open interest per type of participant should be promoted. - Highlights: •We test time-zero efficiency of derivatives power markets in Germany, France and Spain. •Prices in Germany, considering liquidity and transaction costs, are time-zero efficient. •In France and Spain, limitations in liquidity and representativeness prevent conclusions. •Liquidity in France and Spain should improve by using pricing and marketing incentives. •Incentives attract participants to exchanges and encourage them to settle OTC trades in clearinghouses.

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

    International Nuclear Information System (INIS)

    Karsten, G.; Elbel, K.; Dienst, W.; Schaefer, L.

    1978-01-01

    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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Hernández, F., E-mail: francisco.hernandez@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (INR) (Germany); Kolb, M. [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-WPT) (Germany); Annabattula, R. [Indian Institute of Technology Madras (IITM), Department of Mechanical Engineering (India); Weth, A. von der [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (INR) (Germany)

    2014-10-15

    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 (Li{sub 4}SiO{sub 4}) 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 Li{sub 4}SiO{sub 4} 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 Li{sub 4}SiO{sub 4} pebble bed, and (3) in situ measurement of thermal conductivity of the Li{sub 4}SiO{sub 4} 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

  3. Comparison between FEM and high heat flux thermal fatigue testing results of ITER divertor plasma facing mock-ups

    Energy Technology Data Exchange (ETDEWEB)

    Crescenzi, F., E-mail: fabio.crescenzi@enea.it; Roccella, S.; Visca, E.; Moriani, A.

    2014-10-15

    Highlights: • Divertor is an important part of the ITER machine. • Finite element analysis allows designers to explore multiple design options, reducing physical prototypes and optimizing design performance. • The hydraulic thermal-mechanical analysis performed by ANSYS and the test results on small-scale mock-ups manufactured by HRP were compared. • FEA results confirmed many experimental data, then it could be very useful for next design optimization. - Abstract: The divertor is one of the most challenging components of “DEMO” the next step ITER machine, so many tasks regarding modeling and experiments have been made in the past years to assess manufacturing processes, materials and thus the life-time of the components. In this context the finite element analysis (FEA) allows designers to explore multiple design options, to reduce physical prototypes and to optimize design performance. The comparison between the hydraulic thermal-mechanical analysis performed by ANSYS WORKBENCH 14.5 and the test results [1] on small-scale mock-ups manufactured with the Hot Radial Pressing (HRP) [2] technology is presented in this paper. During the thermal fatigue testing in the Efremov TSEFEY facility to assess the heat flux load-carrying capability of the mock-ups, only the surface temperature was measured, so the FEA was important because it allowed to know any other information (temperature inside the materials, local water temperature, local stress, etc.). FEA was performed coupling the thermal-hydraulic analysis, that calculated the temperature distributions on the components and the heat transfer coefficient (HTC) between water and heat sink tube, with the mechanical analysis. The comparison between analysis and testing results was based on the temperature maps of the loaded surface and on number of the cycles supported during the testing and those predicted by the mechanical analysis using the experimental fatigue curves for CuCrZr-IG, that is the structural

  4. Surveillance of nuclear power reactors

    International Nuclear Information System (INIS)

    Marini, J.

    1983-01-01

    Surveillance of nuclear power reactors is now a necessity imposed by such regulatory documents as USNRC Regulatory Guide 1.133. In addition to regulatory requirements, however, nuclear reactor surveillance offers plant operators significant economic advantages insofar as a single day's outage is very costly. The economic worth of a reactor surveillance system can be stated in terms of the improved plant availability provided through its capability to detect incidents before they occur and cause serious damage. Furthermore, the TMI accident has demonstrated the need for monitoring certain components to provide operators with clear information on their functional status. In response to the above considerations, Framatome has developed a line of products which includes: pressure vessel leakage detection systems, loose part detection systems, component vibration monitoring systems, and, crack detection and monitoring systems. Some of the surveillance systems developed by Framatome are described in this paper

  5. Thermal and mechanical behaviour of an experimental mock-up of a nuclear containment

    International Nuclear Information System (INIS)

    Chauvel, D.; Barre, F.

    2007-01-01

    In order to better understand the behaviour of a reactor containment submitted to combined pressure and temperature loads by means of studies of the concrete permeability and the state of cracking evolution, EDF and its French partners have built a prestressed concrete test model which represents a PWR containment typical section. The monitoring system was designed to follow the evolution of strains, temperature and state of cracking of the concrete wall from construction stage to air and steam tests. The measurements results as well as their comparison with theoretical laws or data and calculated values, allow to determine the main thermal and mechanical characteristics of the concrete, to analyse the thermo-mechanical behaviour of the structure and also to check the design criteria of prestressed concrete containments. (authors)

  6. Advances in small zero-leak valves point to better nuclear power-plant reliability

    Energy Technology Data Exchange (ETDEWEB)

    Eacott, K B; Kin, J C; Hotta, Y [Dresser Japan, Ltd.

    1978-04-01

    In the selection of small valves less than two inches used for nuclear power plants, sufficient consideration must be given to the reliability to radioactive material, the easy operability, and the significant function, especially zero leak. These valves are classified into bellows and diaphragm seal types which must satisfy zero leak, 4000 cycles life test and good maintainability. Welded bellows, formed bellows, and metal diaphragms are actually used for these requirements. The construction of these types are shown. The requirements and principal specifications for these small valves are explained, and some examples are given. These zero leak valves are installed in reactor coolant loop system, borated water from B. A. system, pressurizer instrument system, containment spray system, high head system and off gas system for PWRS, and main steam line system, diesel generator cooling water system, re-circulation system, clean up water system, etc. for BWRS.

  7. Measurement of the physics properties of gas-cooled fast reactors in the zero energy reactor PROTEUS and analysis of the results

    International Nuclear Information System (INIS)

    Richmond, R.

    1982-12-01

    The main aim of the fast reactor physics measurements carried out in the zero energy reactor PROTEUS was to check the performance of data sets and calculation methods used in the design of fast breeder reactors. This allowed the accuracy of the power reactor calculations to be determined and enabled an assessment to be made of whether this accuracy would be sufficient to allow the design, construction and licensing of the GCFR power reactor. In order to carry out the physics measurements an existing zero energy reactor was converted to a form in which a central fast reactor lattice was surrounded by thermal zones to drive the reactor critical. One of the most important measuring techniques used to check the performance of data sets and calculation methods was the determination of reaction rate ratios and, by using an appropriate range of nuclides, it was possible to obtain a detailed picture covering 70% of reactions taking place in the central part of the fast reactor zone and with an accuracy of +-1.5% in a typical ratio. A further technique used during the work on GCFR-PROTEUS was the measurement of neutron spectrum which was carried out in a wide range of environments and, in the later stages of the work, covered the energy range from 9 keV to 2.3 MeV. These measurements, in particular, indicated significant errors in the FGL4 scattering cross-sections. A third technique, which was developed to a high degree of accuracy, was the measurement of reactivity worths. This was used in measurements of the worths of small samples and also in the application of the null reactivity technique to determine k-infinity and hence the absorption cross-sections of reactor structural materials. (Auth.)

  8. Fabrication of full-size mock-up for 10° section of ITER vacuum vessel thermal shield

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dong Kwon [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Nam, Kwanwoo, E-mail: kwnam@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Kang, Kyoung-O; Noh, Chang Hyun; Chung, Wooho [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Lim, Kisuk; Kang, Youngkil [SFA Engineering Corp., Asan-si, Chungcheongnam-do 336-873 (Korea, Republic of); Hamlyn-Harris, Craig; Her, Namil; Robby, Hicks [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2015-10-15

    In this paper, a full-scale prototype fabrication for vacuum vessel thermal shield (VVTS) of ITER tokamak is described and test results are reported. All the manufacturing processes except for silver coating were performed in the fabrication of 10° section of VVTS. Pre-qualification test was conducted to compare the vertical and the horizontal welding positions. After shell welding, shell distortion was measured and adjusted. Shell thickness change was also measured after buffing process. Specially, VVTS ports need large bending and complex welding of shell and flange. Bending method for the complex and long cooling tube layout especially for the VVTS ports was developed in detail. Dimensional inspection of the fabricated mock-up was performed with a 3D laser scanner and the scanning data was analyzed.

  9. Examination of C/C flat tile mock-ups with hypervapotron cooling after high heat flux testing

    International Nuclear Information System (INIS)

    Schedler, B.; Friedrich, T.; Traxler, H.; Eidenberger, E.; Scheu, C.; Clemens, H.; Pippan, R.; Escourbiac, F.

    2007-01-01

    Two C/C flat tile mock-ups with a hypervapotron cooling concept, have been successfully tested beyond ITER specification (3000 cycles at 15 MW/m 2 , 300 cycles at 20 MW/m 2 and 800-1000 cycles at 25 MW/m 2 ) in two electron beam testing facilities [F. Escourbiac, et al., Experimental simulation of cascade failure effect on tungsten and CFC flat tile armoured HHF components, Fusion Eng. Des., submitted for publication; F. Escourbiac, et al., A mature industrial solution for ITER divertor plasma facing components: hypervapotron cooling concept adapted to Tore Supra flat tile technology, Fusion Eng. Des. 75-79 (2005) 387-390]. Both mock-ups provide a SNECMA SEPCARB NS31 armour, which has been joined onto the CuCrZr heat sink by active metal casting (AMC) and electron beam welding (EBW). No tile detachment or sudden loss of single tiles has been observed; a cascade-like failure of flat tile armours was impossible to generate. At the maximum cyclic heat flux load of 25 MW/m 2 all tested tiles performed well except one, which revealed already a clear indication in the thermographic examination at the end of the manufacture. Visual examination and analysis of metallographic cuts of the remaining tiles demonstrated that the interface has not been altered. In addition, the shear strength of the C/C to copper joints measured after the high heat flux (HHF) test has been found to be still above the interlamellar shear strength of the used C/C material. The high resistance of the interface is explained by a modification of the C/C to copper joint interface due to silicon originating from the used C/C material

  10. Examination of C/C flat tile mock-ups with hypervapotron cooling after high heat flux testing

    Energy Technology Data Exchange (ETDEWEB)

    Schedler, B. [Technology Centre of PLANSEE SE, A-6600 Reutte (Austria)], E-mail: bertram.schedler@plansee.com; Friedrich, T.; Traxler, H. [Technology Centre of PLANSEE SE, A-6600 Reutte (Austria); Eidenberger, E.; Scheu, C.; Clemens, H. [Department of Physical Metallurgy and Materials Testing, University of Leoben, A-8700 Leoben (Austria); Pippan, R. [Austrian Academy of Sciences, Erich-Schmid-Institute of Material Science, A-8700 Leoben (Austria); Escourbiac, F. [Association EURATOM-CEA, DSM/DRFC, CEA Cadarache, F-13108 St. Paul Lez Durance (France)

    2007-04-15

    Two C/C flat tile mock-ups with a hypervapotron cooling concept, have been successfully tested beyond ITER specification (3000 cycles at 15 MW/m{sup 2}, 300 cycles at 20 MW/m{sup 2} and 800-1000 cycles at 25 MW/m{sup 2}) in two electron beam testing facilities [F. Escourbiac, et al., Experimental simulation of cascade failure effect on tungsten and CFC flat tile armoured HHF components, Fusion Eng. Des., submitted for publication; F. Escourbiac, et al., A mature industrial solution for ITER divertor plasma facing components: hypervapotron cooling concept adapted to Tore Supra flat tile technology, Fusion Eng. Des. 75-79 (2005) 387-390]. Both mock-ups provide a SNECMA SEPCARB NS31 armour, which has been joined onto the CuCrZr heat sink by active metal casting (AMC) and electron beam welding (EBW). No tile detachment or sudden loss of single tiles has been observed; a cascade-like failure of flat tile armours was impossible to generate. At the maximum cyclic heat flux load of 25 MW/m{sup 2} all tested tiles performed well except one, which revealed already a clear indication in the thermographic examination at the end of the manufacture. Visual examination and analysis of metallographic cuts of the remaining tiles demonstrated that the interface has not been altered. In addition, the shear strength of the C/C to copper joints measured after the high heat flux (HHF) test has been found to be still above the interlamellar shear strength of the used C/C material. The high resistance of the interface is explained by a modification of the C/C to copper joint interface due to silicon originating from the used C/C material.

  11. Power reactor pressure vessel benchmarks

    International Nuclear Information System (INIS)

    Rahn, F.J.

    1978-01-01

    A review is given of the current status of experimental and calculational benchmarks for use in understanding the radiation embrittlement effects in the pressure vessels of operating light water power reactors. The requirements of such benchmarks for application to pressure vessel dosimetry are stated. Recent developments in active and passive neutron detectors sensitive in the ranges of importance to embrittlement studies are summarized and recommendations for improvements in the benchmark are made. (author)

  12. Synthesis and characterisation of SiC{sub f}/Cu matrix composites and their application in a divertor flat-tile mock-up; Synthese und Charakterisierung von SiC{sub f}/Cu-Matrix-Verbundwerkstoffen und ihre Anwendung in einem Modell einer Divertor-Komponente

    Energy Technology Data Exchange (ETDEWEB)

    Paffenholz, Verena

    2010-06-30

    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/m{sup 2} 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, SiC{sub f}/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 SiC{sub f}/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

  13. The Optimization of power reactor control system

    International Nuclear Information System (INIS)

    Danupoyo, S.D.

    1997-01-01

    A power reactor is an important part in nuclear powered electrical plant systems. Success in controlling the power reactor will establish safety of the whole power plant systems. Until now, the power reactor has been controlled by a classical control system that was designed based on output feedback method. To meet the safety requirements that are now more restricted, the recently used power reactor control system should be modified. this paper describes a power reactor control system that is designed based on a state feedback method optimized with LQG (Linear-quadrature-gaussian) method and equipped with a state estimator. A pressurized-water type reactor has been used as the model. by using a point kinetics method with one group delayed neutrons. the result of simulation testing shows that the optimized control system can control the power reactor more effective and efficient than the classical control system

  14. Power reactors in Member States. 1979 edition

    International Nuclear Information System (INIS)

    1979-01-01

    This is the fifth issue of a periodic computer-based listing of nuclear power reactors, presenting the situation as of 1 May 1979. The basic design data for all reactors in operation, under construction, planned and shut down have been included. The following two tables are included to give a general picture of the overall situation: Table I: Reactor types and net electrical power. Table II: Reactor units and net electrical powered by country cummulated by year

  15. Computerized reactor power regulation with logarithmic controller

    International Nuclear Information System (INIS)

    Gossanyi, A.; Vegh, E.

    1982-11-01

    A computerized reactor control system has been operating at a 5 MW WWR-SM research reactor in the Central Research Institute for Physics, Budapest, for some years. This paper describes the power controller used in the SPC operating mode of the system, which operates in a 5-decade wide power range with +-0.5% accuracy. The structure of the controller easily limits the minimal reactor period and produces a reactor transient with constant period if the power demand changes. (author)

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

    International Nuclear Information System (INIS)

    Moriot, J.; Gastaldi, O.; Maxit, L.; Guyader, J-L.; Perisse, J.; Migot, B.

    2013-06-01

    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

  17. Integration of coal gasification and packed bed CLC for high efficiency and near-zero emission power generation

    NARCIS (Netherlands)

    Spallina, V.; Romano, M.C.; Chiesa, P.; Gallucci, F.; Sint Annaland, van M.; Lozza, G.

    2014-01-01

    A detailed thermodynamic analysis has been carried out of large-scale coal gasification-based power plant cycles with near zero CO2 emissions, integrated with chemical looping combustion (CLC). Syngas from coal gasification is oxidized in dynamically operated packed bed reactors (PBRs), generating a

  18. 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.

  19. Operational power reactor health physics

    International Nuclear Information System (INIS)

    Watson, B.A.

    1987-01-01

    Operational Health Physics can be comprised of a multitude of organizations, both corporate and at the plant sites. The following discussion centers around Baltimore Gas and Electric's (BG and E) Calvert Cliffs Nuclear Power Plant, located in Lusby, Maryland. Calvert Cliffs is a twin Combustion Engineering 825 MWe pressurized water reactor site with Unit I having a General electric turbine-generator and Unit II having a Westinghouse turbine-generator. Having just completed each Unit's ten-year Inservice Inspection and Refueling Outge, a total of 20 reactor years operating health physics experience have been accumulated at Calvert Cliffs. Because BG and E has only one nuclear site most health physics functions are performed at the plant site. This is also true for the other BG and E nuclear related organizations, such as Engineering and Quality Assurance. Utilities with multiple plant sites have corporate health physics entity usually providing oversight to the various plant programs

  20. Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showing agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm2 and temporary heat flux limit of 600 W/cm2. This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.

  1. Power Reactor Embrittlement Data Base

    International Nuclear Information System (INIS)

    Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

    1990-01-01

    Regulatory and research evaluations of embrittlement prediction models and of vessel integrity under load can be greatly expedited by the use of a well designed, computerized embrittlement data base. The Power Reactor Embrittlement Data Base (PR-EDB) is a comprehensive collection of data from surveillance reports and other published reports of commercial nuclear reactors. The uses of the data base require that as many different data as available are collected from as many sources as possible with complete references and that subsets of relevant data can be easily retrieved and processed. The objectives of this NRC-sponsored program are the following: (1) to compile and to verify the quality of the PR-EDB; (2) to provide user-friendly software to access and process the data; (3) to explore or confirm embrittlement prediction models; and (4) to interact with standards organizations to provide the technical bases for voluntary consensus standards that can be used in regulatory guides, standard review plans, and codes. To achieve these goals, the data base architecture was designed after much discussion and planning with prospective users, namely, material scientists and members of the research staff. The current compilation of the PR-EDB (Version 1) contains results from surveillance capsule reports of 78 reactors with 381 data points for 110 different irradiated base materials and 161 data points for 79 different welds. Results from heat-affected zone materials are also listed. The time and effort required to process and evaluate different types of data in the PR-EDB have been drastically reduced from previous data bases. The Electric Power Research Institute (EPRI), reactor vendors, and utilities are in the process of providing back-up quality assurance checks of PR-EDB and will be supplementing the data base with additional data and documentation

  2. Reactor Division semestrial progress report July - December 1987

    International Nuclear Information System (INIS)

    1987-01-01

    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)

  3. Reactor Division semestrial progress report January - June 1987

    International Nuclear Information System (INIS)

    1987-01-01

    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)

  4. Reactor power reduction system and method

    International Nuclear Information System (INIS)

    Bruno, S.J.; Dunn, S.A.; Raber, M.

    1978-01-01

    A method of operating a nuclear power reactor is disclosed which enables an accelerated power reduction of the reactor without completely shutting the reactor down. The method includes monitoring the incidents which, upon their occurrence, would require an accelerated power reduction in order to maintain the reactor in a safe operation mode; calculating the power reduction required on the occurrence of such an incident; determining a control rod insertion sequence for the normal operation of the reactor, said sequence being chosen to optimize reactor power capability; selecting the number of control rods necessary to respond to the accelerated power reduction demand, said selection being made according to a priority determined by said control rod insertion sequence; and inserting said selected control rods into the reactor core. 11 claims, 13 figures

  5. Polymorphism in Br2 clathrate hydrates.

    Science.gov (United States)

    Goldschleger, I U; Kerenskaya, G; Janda, K C; Apkarian, V A

    2008-02-07

    The structure and composition of bromine clathrate hydrate has been controversial for more than 170 years due to the large variation of its observed stoichiometries. Several different crystal structures were proposed before 1997 when Udachin et al. (Udachin, K. A.; Enright, G. D.; Ratcliffe, C. I.; Ripmeester, J. A. J. Am. Chem. Soc. 1997, 119, 11481) concluded that Br2 forms only the tetragonal structure (TS-I). We show polymorphism in Br2 clathrate hydrates by identifying two distinct crystal structures through optical microscopy and resonant Raman spectroscopy on single crystals. After growing TS-I crystals from a liquid bromine-water solution, upon dropping the temperature slightly below -7 degrees C, new crystals of cubic morphology form. The new crystals, which have a limited thermal stability range, are assigned to the CS-II structure. The two structures are clearly distinguished by the resonant Raman spectra of the enclathrated Br2, which show long overtone progressions and allow the extraction of accurate vibrational parameters: omega(e) = 321.2 +/- 0.1 cm(-1) and omega(e)x(e) = 0.82 +/- 0.05 cm(-1) in TS-I and omega(e) = 317.5 +/- 0.1 cm(-1) and omega(e)x(e) = 0.70 +/- 0.1 cm(-1) in CS-II. On the basis of structural analysis, the discovery of the CS-II crystals implies stability of a large class of bromine hydrate structures and, therefore, polymorphism.

  6. Some properties of zero power neutron noise in a time-varying medium with delayed neutrons

    International Nuclear Information System (INIS)

    Kitamura, Y.; Pal, L.; Pazsit, I.; Yamamoto, A.; Yamane, Y.

    2008-01-01

    The temporal evolution of the distribution of the number of neutrons in a time-varying multiplying system, producing only prompt neutrons, was treated recently with the master equation technique by some of the present authors. Such a treatment gives account of both the so-called zero power reactor noise and the power reactor noise simultaneously. In particular, the first two moments of the neutron number, as well as the concept of criticality for time-varying systems, were investigated and discussed. The present paper extends these investigations to the case when delayed neutrons are also taken into account. Due to the complexity of the description, only the expectation of the neutron number is calculated. The concept of criticality of a time-varying system is also generalized to systems with delayed neutrons. The temporal behaviour of the expectation of the number of neutrons and its asymptotic properties are displayed and discussed

  7. Physical protection of power reactors

    International Nuclear Information System (INIS)

    Darby, J.L.

    1979-01-01

    Sandia Laboratories has applied a systematic approach to designing physical protection systems for nuclear facilities to commercial light-water reactor power plants. A number of candidate physical protection systems were developed and evaluated. Focus is placed on the design of access control subsystems at each of three plant layers: the protected area perimeter, building surfaces, and vital areas. Access control refers to barriers, detectors, and entry control devices and procedures used to keep unauthorized personnel and contraband out of the plant, and to control authorized entry into vital areas within the plant

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

    2013-01-01

    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.)

  10. Nuclear reactor power supply system

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

    The redundant signals from the sensor assemblies measuring the process parameters of a nuclear reactor power supply are transmitted each in its turn to a protection system which operates to actuate the protection apparatus for signals indicating off-process conditions. Each sensor assembly includes a number of like sensors measuring the same parameters. The sets of process signals derived from the sensor assemblies are each in its turn transmitted from the protection system to the control system which impresses control signals on the reactor or its components to counteract the tendency for conditions to drift off-normal status requiring operation of the protection system. A parameter signal selector prevents a parameter signal which differs from the other parameter signals of the set by more than twice the allowable variation from passing to the control system. Test signals are periodically impressed by a test unit on a selected pair of a selection unit and control channels. This arrangement eliminates the possibility that a single component failure which may be spurious will cause an inadvertent trip of the reactor during test. (author)

  11. Nuclear power plant with several reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grishanin, E I; Ilyunin, V G; Kuznetsov, I A; Murogov, V M; Shmelev, A N

    1972-05-10

    A design of a nuclear power plant suggested involves several reactors consequently transmitting heat to a gaseous coolant in the joint thermodynamical circuit. In order to increase the power and the rate of fuel reproduction the low temperature section of the thermodynamical circuit involves a fast nuclear reactor, whereas a thermal nuclear reactor is employed in the high temperature section of the circuit for intermediate heating and for over-heating of the working body. Between the fast nuclear and the thermal nuclear reactors there is a turbine providing for the necessary ratio between pressures in the reactors. Each reactor may employ its own coolant.

  12. Microstructure and mechanical properties of reactor pressure vessel mock-up material treated by intercritical heat treatment

    International Nuclear Information System (INIS)

    Kim, M. C.; Lee, B. S.; Hong, J. H.; Lee, H. J.; Park, S. D.; Kim, K. B.; Yoon, J. H.; Kim, J. S.; Oh, J. M.

    2003-12-01

    The mechanical properties and microstructures of base metal and weld HAZ (Heat-Affected Zone) of a Mn-Mo-Ni low alloy steels treated by intercritical heat treatment were investigated to evaluate effects of intercritical heat treatment on mechanical properties. In order to clarify the effects of intercritical heat treatment, two types of specimen were prepared by CHT(Conventional Heat Treatment) and IHT(CHT+Intercritical Heat Treatment). Tensile test, charpy impact test and vickers hardness test were carried out to evaluate the mechanical properties. It is found that impact toughness and hardness were improved by intercritical heat treatment. Mean size of precipitates and effective grain were quantitatively analysed as microstructural factors. It is found that precipitate size was decreased and shape of precipitate was spherodized by intercritical heat treatment and grain size was also decreased. So, it is thought that these microstructural changes cause the improvement of mechanical properties by intercritical heat treatment. The simulated specimen using a Gleeble thermal simulator system was used to evaluate the mechanical properties of HAZ. It is well known that IRHAZ and SRHAZ have lower toughness than base metal. However, in the case of IHT, impact toughness of IRHAZ and SRHAZ were slightly higher than that of base metal. It is obvious that this improvement of fracture toughness in IRHAZ and SRHAZ region was closely related to the microstructural changes, such as spheroidization of precipitate and decreases of precipitate size and grain size

  13. Nuclear reactor instrumentation power monitor

    International Nuclear Information System (INIS)

    Suzuki, Shigeru.

    1989-01-01

    The present invention concerns a nuclear reactor instrumentation power monitor that can be used in, for example, BWR type nuclear power plants. Signals from multi-channel detectors disposed on field units are converted respectively by LPRM signal circuits. Then, the converted signals are further converted by a multiplexer into digital signals and transmitted as serial data to a central monitor unit. The thus transmitted serial data are converted into parallel data in the signal processing section of the central monitor unit. Then, LPRM signals are taken out from each of channel detectors to conduct mathematical processing such as trip judgment or averaging. Accordingly, the field unit and the central monitor unit can be connected by way of only one data transmission cable thereby enabling to reduce the number of cables. Further, since the data are transmitted on digital form, it less undergoes effect of noises. (I.S.)

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

    International Nuclear Information System (INIS)

    Messiaen, A.; Vervier, M.; Dumortier, P.; Grine, D.; Lamalle, P.U.; Durodie, F.; Koch, R.; Louche, F.; Weynants, R.

    2009-01-01

    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.

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

    International Nuclear Information System (INIS)

    Youchison, D.L.; Guiniiatouline, R.; Watson, R.D.

    1994-01-01

    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/m 2 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/m 2 and surface temperatures up to 690 degrees C before debonding at 10 MW/m 2 . A third tile debonded after 9200 thermal fatigue cycles at 5 MW/m 2 , 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

  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. Elimination mechanisms of Br2+ and Br+ in photodissociation of 1,1- and 1,2-dibromoethylenes using velocity imaging technique

    International Nuclear Information System (INIS)

    Hua Linqiang; Zhang Bing; Lee, Wei-Bin; Chao, Meng-Hsuan; Lin, King-Chuen

    2011-01-01

    Elimination pathways of the Br 2 + and Br + ionic fragments in photodissociation of 1,2- and 1,1-dibromoethylenes (C 2 H 2 Br 2 ) at 233 nm are investigated using time-of-flight mass spectrometer equipped with velocity ion imaging. The Br 2 + fragments are verified not to stem from ionization of neutral Br 2 , that is a dissociation channel of dibromoethylenes reported previously. Instead, they are produced from dissociative ionization of dibromoethylene isomers. That is, C 2 H 2 Br 2 is first ionized by absorbing two photons, followed by the dissociation scheme, C 2 H 2 Br 2 + + hv→Br 2 + + C 2 H 2 . 1,2-C 2 H 2 Br 2 gives rise to a bright Br 2 + image with anisotropy parameter of -0.5 ± 0.1; the fragment may recoil at an angle of ∼66 deg. with respect to the C = C bond axis. However, this channel is relatively slow in 1,1-C 2 H 2 Br 2 such that a weak Br 2 + image is acquired with anisotropy parameter equal to zero, indicative of an isotropic recoil fragment distribution. It is more complicated to understand the formation mechanisms of Br + . Three routes are proposed for dissociation of 1,2-C 2 H 2 Br 2 , including (a) ionization of Br that is eliminated from C 2 H 2 Br 2 by absorbing one photon, (b) dissociation from C 2 H 2 Br 2 + by absorbing two more photons, and (c) dissociation of Br 2 + . Each pathway requires four photons to release one Br + , in contrast to the Br 2 + formation that involves a three-photon process. As for 1,1-C 2 H 2 Br 2 , the first two pathways are the same, but the third one is too weak to be detected.

  18. Reactor power control method upon accidents of electrical power system

    International Nuclear Information System (INIS)

    Hirose, Masao.

    1983-01-01

    Purpose: To enable to continue the operation of a BWR type reactor by avoiding the scram while suppressing the reactor power, just after the external disturbance such as earth-trouble in power-transmission network. Method: Steep power drop of an electrical generator is to be detected not only by a current-type power-load-unbalance relay but also with a power-type power-load-unbalance-relay. If steep power-drop was detected by the latter relay, a previously selected control rod is rapidly inserted into the reactor. In this way, in the case where there is a possibility of the reactor scram, the scram can be avoided by suppressing the reactor power, thus the reactor operation can be continued. (Kamimura, M.)

  19. Irradiation tests in BR2 of miniature fission chambers in pulse, Campbelling and current mode

    Energy Technology Data Exchange (ETDEWEB)

    Vermeeren, L. [SCK.CEN, Boeretang 200, B-2400 Mol (Belgium); Geslot, B.; Breaud, S.; Filliatre, P.; Jammes, C. [CEA/DEN/SPEx/LDCI, Centre de Cadarache, F-13109 Saint-Paul-lez-Durance (France); Legrand, A. [CEA/DEN/DRSN/SIREN/LASPI Saclay, F-91191 Gif sur Yvette Cedex (France); Barbot, L. [CEA/DEN/SPEx/LDCI, Centre de Cadarache, F-13109 Saint-Paul-lez-Durance (France)

    2011-07-01

    The FNDS system ('Fast Neutron Detection System') for the on-line in-pile detection of the fast neutron flux in the presence of a significant thermal neutron flux and a high gamma dose rate is being developed in the framework of the SCK.CEN-CEA Laboratoire Commun. The system has been patented in 2008. The system consists of a miniature Pu-242 fission chamber as main detector, complemented by a U-235 fission chamber or a rhodium Self-Powered Neutron Detector (SPND) for thermal neutron flux monitoring and a dedicated acquisition system that also takes care of the processing of the signals from both detectors to extract fast neutron flux data. This paper describes a FNDS qualification experiment in the SCK.CEN BR2 reactor, with experimental results on a large set of fission chambers in current and Campbelling mode. (authors)

  20. Irradiation tests in BR2 of miniature fission chambers in pulse, Campbelling and current mode

    International Nuclear Information System (INIS)

    Vermeeren, L.; Geslot, B.; Breaud, S.; Filliatre, P.; Jammes, C.; Legrand, A.; Barbot, L.

    2011-01-01

    The FNDS system ('Fast Neutron Detection System') for the on-line in-pile detection of the fast neutron flux in the presence of a significant thermal neutron flux and a high gamma dose rate is being developed in the framework of the SCK.CEN-CEA Laboratoire Commun. The system has been patented in 2008. The system consists of a miniature Pu-242 fission chamber as main detector, complemented by a U-235 fission chamber or a rhodium Self-Powered Neutron Detector (SPND) for thermal neutron flux monitoring and a dedicated acquisition system that also takes care of the processing of the signals from both detectors to extract fast neutron flux data. This paper describes a FNDS qualification experiment in the SCK.CEN BR2 reactor, with experimental results on a large set of fission chambers in current and Campbelling mode. (authors)

  1. BN-1200 Reactor Power Unit Design Development

    International Nuclear Information System (INIS)

    Vasilyev, B.A.; Shepelev, S.F.; Ashirmetov, M.R.; Poplavsky, V.M.

    2013-01-01

    Main goals of BN-1200 design: • Develop a reliable new generation reactor plant for the commercial power unit with fast reactor to implement the first-priority objectives in changing over to closed nuclear fuel cycle; • Improve technical and economic indices of BN reactor power unit to the level of those of Russian VVER of equal power; • Enhance the safety up to the level of the requirements for the 4th generation RP

  2. Manufacturing and testing of W/Cu mono-block small scale mock-up for EAST by HIP and HRP technologies

    Energy Technology Data Exchange (ETDEWEB)

    Li, Qiang [Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), Hefei, Anhui (China); Qin, Sigui [Advanced Technology and Materials Co., Ltd, Beijing (China); Wang, Wanjing; Qi, Pan [Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), Hefei, Anhui (China); Roccella, Selanna; Visca, Eliseo [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Liu, Guohui [Advanced Technology and Materials Co., Ltd, Beijing (China); Luo, Guang-Nan, E-mail: liqiang577@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), Hefei, Anhui (China)

    2013-10-15

    ITER-like W/Cu mono-block plasma-facing components (PFCs) will be used in vertical target regions of the experimental advanced superconducting tokamak (EAST) divertor. The first W/Cu mono-block small scale mock-up with five W mono-blocks has been manufactured successfully by technological combination of hot isostatic pressing (HIP) and hot radial pressing (HRP). The joining of a W mono-block and a pure copper interlayer was achieved by means of HIP technology and the bonding strength was over 150 MPa. The good bonding between the pure copper interlayer and a CuCrZr cooling tube was obtained by means of HRP technology. In order to understand deeply the process of HRP, the stress distribution of the mock-up during HRP process was simulated using ANSYS code. Ultrasonic Nondestructive Testing (NDT) of the W/Cu and Cu/CuCrZr interfaces was performed, showing that excellent bonding of the W/Cu and Cu/CuCrZr interfaces. The thermal cycle fatigue testing of the mock-up has been carried out by means of an e-beam device in Southwest Institute of Physics, Chengdu (SWIP) and the mock-up withstood 1000 cycles of heat loads up to 8.4 MW/m{sup 2} with the cooling water of 2 m/s, 20 °C, 0.2 MPa.

  3. Mathematical game type optimization of powerful fast reactors

    International Nuclear Information System (INIS)

    Pavelesku, M.; Dumitresku, Kh.; Adam, S.

    1975-01-01

    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)

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

    International Nuclear Information System (INIS)

    2015-01-01

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

  5. Further analysis of the zero-energy experiment on the Dragon reactor

    International Nuclear Information System (INIS)

    Woloch, F.; Neuberger, W.

    1978-01-01

    The analysis of the Zero-Energy Experiments performed on the Dragon reactor, a high-temperature reactor of the Organization for Economic Cooperation and Development, has been continued. The first analysis established the main route of calculations within the WIMS-E scheme and was reported elsewhere. This Note presents further calculations showing the merits of a refinement in the number of neutron energy groups, of the use of different condensation spectra, and of transport calculations

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

    International Nuclear Information System (INIS)

    Powell, J.R.

    1977-01-01

    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-He 3 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

  7. Power reactor embrittlement data base

    International Nuclear Information System (INIS)

    Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

    1989-01-01

    Regulatory and research evaluations of embrittlement prediction models and of vessel integrity under load can be greatly expedited by the use of a well-designed, computerized embrittlement data base. The Power Reactor Embrittlement Data Base (PR-EDB) is a comprehensive collection of data from surveillance reports and other published reports of commercial nuclear reactors. The uses of the data base require that as many different data as available are collected from as many sources as possible with complete references and that subsets of relevant data can be easily retrieved and processed. The objectives of this NRC-sponsored program are the following: to compile and to verify the quality of the PR-EDB; to provide user-friendly software to access and process the data; to explore or confirm embrittlement prediction models; and to interact with standards organizations to provide the technical bases for voluntary consensus standards that can be used in regulatory guides, standard review plans, and codes. 9 figs

  8. Performance indicators for power reactors

    International Nuclear Information System (INIS)

    Gillies, C.; White, M.

    1995-11-01

    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

  9. Transmutation of actinides in power reactors.

    Science.gov (United States)

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

  10. Power controlling method for BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1983-01-01

    Purpose: To enable reactor operation exactly following after an aimed curve in the high power resuming and maintaining period without failures in cladding tubes. Method: Upon recovery of the reactor power to a high power level after changing the reactor power from the high power to the low power level, control rod is operated under such conditions that the linear power density after operation of the control rod does not exceed the PC envelope in the low power period, and the core flow rate is coordinated to the control rod operation. The linear power density can be suppressed within an allowable linear power density by the above operation during high power resuming and maintaining period and, as the result, PCI failures can be prevented. (Kamimura, M.)

  11. Validation of MCNP and ORIGEN-S 3-D computational model for reactivity predictions during BR2 operation

    International Nuclear Information System (INIS)

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

    2005-01-01

    The Belgian Material Test Reactor (MTR) BR2 is strongly heterogeneous high flux engineering test reactor at SCK-CEN (Centre d'Etude de l'energie Nucleaire) in Mol at a thermal power 60 to 100 MW. It deploys highly enriched uranium, water cooled concentric plate fuel elements, positioned inside a beryllium reflector with complex hyperboloid arrangement of test holes. The objective of this paper is the validation of a MCNP and ORIGEN-S 3D model for reactivity predictions of the entire BR2 core during reactor operation. We employ the Monte Carlo code MCNP-4C for evaluating the effective multiplication factor k eff and 3D space dependent specific power distribution. The 1D code ORIGEN-S is used for calculation of isotopic fuel depletion versus burn up and preparation of a database (DB) with depleted fuel compositions. The approach taken is to evaluate the 3D power distribution at each time step and along with DB to evaluate the 3D isotopic fuel depletion at the next step and to deduce the corresponding shim rods positions of the reactor operation. The capabilities of the both codes are fully exploited without constraints on the number of involved isotope depletion chains or increase of the computational time. The reactor has a complex operation, with important shutdowns between cycles, and its reactivity is strongly influenced by poisons, mainly 3 He and 6 Li from the beryllium reflector, and burnable absorbers 149 Sm and 10 B in the fresh UAlx fuel. Our computational predictions for the shim rods position at various restarts are within 0.5$ (β eff =0.0072). (author)

  12. CFD prediction of mixing in a steam generator mock-up: Comparison between full geometry and porous medium approaches

    International Nuclear Information System (INIS)

    Dehbi, A.; Badreddine, H.

    2013-01-01

    Highlights: • CFD is used to simulate single phase mixing in a model steam generator. • Motive of the work is to compare porous media approach with full geometry representation of tubes. • Porous media approach is found to compare favorably with full representation in steady states. - Abstract: In CFD simulations of single phase flow mixing in a steam generator (SG) during a station blackout severe accident, 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 that is treated as a single porous medium which replicates the pressure loss and heat transfer characteristics of the real tube. This approach significantly 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 a series of simulations. We first validate the porous medium approach against results of the 1/7th scale Westinghouse SG-S3 test. In a second step, we make two separate simulations of flow in the PSI SG mock-up, i.e. one in which the porous medium model is used for the tube bundle, and another in which the full geometry is represented. In all simulations, the Reynolds Stress (RSM) model of turbulence is used. We show that in steady state conditions, the porous medium treatment yields results which are comparable to those of the full geometry representation (temperature distribution, recirculation ratio, hot plume spread, etc.). Hence, the porous medium approach can be extended with a good degree of confidence to model single phase mixing in the full scale SG

  13. Investigation on water content in fresco mock-ups in the microwave and near-IR spectral regions

    International Nuclear Information System (INIS)

    Magrini, Donata; Riminesi, Cristiano; Cucci, Costanza; Olmi, Roberto; Picollo, Marcello

    2017-01-01

    Water diffusion inside masonry is responsible for the majority of the decay phenomena observed in wall paintings and frescos. Thus, the diagnostics of moisture and water content and their monitoring represent a key issue. In order to preserve the integrity of surfaces of artistic interest, investigations by means of non-destructive techniques (NDT) are preferred over others. The aim of this research is to determine methodologies to quantify the moisture content (MC) of frescos by means of the integrated use of two non-invasive techniques, namely fiber optic reflectance spectroscopy (FORS) in the near-IR region and evanescent field dielectrometry (EFD) in the microwave range. The FORS technique has been employed in order to assess the amount of water adsorbed from the surface by means of an analysis of the reflectance spectra in the Vis–NIR (350-2200 nm) range. This technique investigates the electronic and vibrational transitions that are characteristic of each compound and enables their identification. The water content is evaluated on the basis of the 1920 nm and 1450 nm absorption bands. The EFD system consists of a resonant probe connected to a network analyzer. The resonance frequency of the cavity under different moisture-content conditions of frescos is in the 1.0–1.5 GHz range. The device makes it possible to compute, in real time, the MC from a measurement of the transmission coefficient (amplitude versus frequency) through the probe. Fresco mock-ups have been prepared in collaboration with the Opificio delle Pietre Dure in order to recreate most of the possible chromatic shades obtained by mixing iron oxides and hydroxide-based pigments. Measurements were performed by employing both techniques on fresco models after wet-dry cycles obtained by means of poultices with a known water content. The results obtained with these two techniques were compared, and cross relationships between the EFD and FORS data were defined. (paper)

  14. A WIMS E analysis of zero energy experiments performed on the Dragon reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lancefield, M. J.; Broadhouse, B.; Woloch, F.

    1974-10-15

    UKAEA methods embodied in the WINS-E modular scheme of codes are described in their application to the analysis of zero energy experiments performed on the DRAGON reactor. Measured reactivity and reaction rate distributions are compared with the predictions of the analysis.

  15. Power generator in BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1984-01-01

    Purpose: To enable to perform stable and dynamic conditioning operation for nuclear fuels in BWR type reactors. Constitution: The conditioning operation for the nuclear fuels is performed by varying the reactor core thermal power in a predetermined pattern by changing the predetermined power changing pattern of generator power, the rising rate of the reactor core thermal power and the upper limit for the rising power of the reactor core thermal power are calculated and the power pattern for the generator is corrected by a power conditioning device such that the upper limit for the thermal power rising rate and the upper limit for the thermal power rising rate are at the predetermined levels. Thus, when the relation between the reactor core thermal power and the generator electrical power is fluctuated, the fluctuation is detected based on the variation in the thermal power rising rate and the limit value for the thermal power rising rate, and the correction is made to the generator power changing pattern so that these values take the predetermined values to thereby perform the stable conditioning operation for the nuclear fuels. (Moriyama, K.)

  16. Integral data for fast reactors

    International Nuclear Information System (INIS)

    Collins, P.J.; Poenitz, W.P.; McFarlane, H.F.

    1988-01-01

    Requirements at Argonne National Laboratory to establish the best estimates and uncertainties for LMR design parameters have lead to an extensive evaluation of the available critical experiment database. Emphasis has been put upon selection of a wide range of cores, including both benchmark, assemblies covering a range of spectra and compositions and power reactor mock-up assemblies with diverse measured parameters. The integral measurements have been revised, where necessary, using the most recent reference data and a covariance matrix constructed. A sensitivity database has been calculated, embracing all parameters, which enables quantification of the relevance of the integral data to parameters calculated with ENDF/B-V.2 cross sections

  17. Impacts on power reactor health physics programs

    International Nuclear Information System (INIS)

    Meyer, B.A.

    1991-01-01

    The impacts on power reactor health physics programs form implementing the revised 10 CFR Part 20 will be extensive and costly. Every policy, program, procedure and training lesson plan involving health physics will require changes and the subsequent retraining of personnel. At each power reactor facility, hundreds of procedures and thousands of people will be affected by these changes. Every area of a power reactor health physics program will be affected. These areas include; ALARA, Respiratory Protection, Exposure Control, Job Coverage, Dosimetry, Radwaste, Effluent Accountability, Emergency Planning and Radiation Worker Training. This paper presents how power reactor facilities will go about making these changes and gives possible examples of some of these changes and their impact on each area of power reactor health physics program

  18. Investigation of zero-release cycle using fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    The task force was organized for the main purpose of offering quantitative basic data to the study group on nuclear fuel cycle in February, 1997. The effect of so-called frontier technologies such as the isotope separation by laser method, the FP annihilation with electron beam accelerators and so on in the FBR cycle based on MOX fuel and PUREX reprocessing method was expected. It is aimed at to recycle the total amount of minor actinides. The object of recycling is the nuclides which contribute largely to toxicity, namely 11 elements, 12 nuclides. The preconditions and the target to be attained of the investigation are explained. As the results of investigation, the amount of reloading MA and FP into a reactor, squeezing the recycling scenario, the effect of reducing toxicity and the subject of the countermeasures to the nuclides with long half-life which cannot be reloaded are reported. As the technical evaluation required for realizing the concept, the concept of the core which excludes recriticality, the advance of reprocessing technology, isotope separation, the fabrication into the optimal form for recycling and so on are discussed. The economical efficiency of the recycling based on MOX and PUREX and the proposal of the development scenario are described. (K.I.)

  19. Fast reactors: potential for power

    International Nuclear Information System (INIS)

    1983-02-01

    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.)

  20. Brominated methanes as photoresponsive molecular storage of elemental Br2.

    Science.gov (United States)

    Kawakami, Kazumitsu; Tsuda, Akihiko

    2012-10-01

    The photochemical generation of elemental Br(2) from brominated methanes is reported. Br(2) was generated by the vaporization of carbon oxides and HBr through oxidative photodecomposition of brominated methanes under a 20 W low-pressure mercury lamp, wherein the amount and situations of Br(2) generation were photochemically controllable. Liquid CH(2)Br(2) can be used not only as an organic solvent but also for the photoresponsive molecular storage of Br(2), which is of great technical benefit in a variety of organic syntheses and in materials science. By taking advantage of the in situ generation of Br(2) from the organic solvent itself, many organobromine compounds were synthesized in high practical yields with or without the addition of a catalyst. Herein, Br(2) that was generated by the photodecomposition of CH(2)Br(2) retained its reactivity in solution to undergo essentially the same reactions as those that were carried out with solutions of Br(2) dissolved in CH(2)Br(2) that were prepared without photoirradiation. Furthermore, HBr, which was generated during the course of the photodecomposition of CH(2)Br(2), was also available for the substitution of the OH group for the Br group and for the preparation of the HBr salts of amines. Furthermore, the photochemical generation of Br(2) from CH(2)Br(2) was available for the area-selective photochemical bleaching of natural colored plants, such as red rose petals, wherein Br(2) that was generated photochemically from CH(2)Br(2) was painted onto the petal to cause radical oxidations of the chromophoric anthocyanin molecules. Copyright © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  1. Thermionic reactors for space nuclear power

    Science.gov (United States)

    Homeyer, W. G.; Merrill, M. H.; Holland, J. W.; Fisher, C. R.; Allen, D. T.

    1985-01-01

    Thermionic reactor designs for a variety of space power applications spanning the range from 5 kWe to 3 MWe are described. In all of these reactors, nuclear heat is converted directly to electrical energy in thermionic fuel elements (TFEs). A circulating reactor coolant carries heat from the core of TFEs directly to a heat rejection radiator system. The recent design of a thermionic reactor to meet the SP-100 requirements is emphasized. Design studies of reactors at other power levels show that the same TFE can be used over a broad range in power, and that design modifications can extend the range to many megawatts. The design of the SP-100 TFE is similar to that of TFEs operated successfully in test reactors, but with design improvements to extend the operating lifetime to seven years.

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

    Energy Technology Data Exchange (ETDEWEB)

    Henry, H.G. [Virginia Power, Mineral, VA (United States); Reilly, B.P. [Bechtel Power Corp., Gaithersburg, MD (United States)

    1995-03-01

    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.

  3. Specific power of liquid-metal-cooled reactors

    International Nuclear Information System (INIS)

    Dobranich, D.

    1987-10-01

    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

  4. Method and device for controlling reactor power

    International Nuclear Information System (INIS)

    Oohashi, Masahisa; Masuda, Hiroyuki.

    1982-01-01

    Purpose: To enable load following-up operation of a reactor adapted to perform power conditioning by the control of the liquid poison density in the core and by the control rods. Constitution: In a case where the reactor power is repeatedly changed in a reactor having a liquid poison density control device and control rods, the time period for the power control is divided depending on the magnitude of the change with time in the reactivity and the optimum values are set for the injection and removal amount of the liquid poison within the divided period. Then, most parts of the control required for the power change are alloted to the liquid poison that gives no effect on the power distribution while minimizing the movement of the control rods, whereby the power change in the reactor as in the case of the load following-up operation can be practiced with ease. (Kawakami, Y.)

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

    International Nuclear Information System (INIS)

    1961-01-01

    Pursuant to Section 15 of the 'Contract for the Lease of Enriched Uranium' for the NORA reactor, the United States Atomic Energy Commission has reduced, with effect from 1 January 1962, the rates of the Use Charge specified in Section II(a) and of the Consumption Charge specified in Section II(b)

  6. Nuclear power reactors of new generation

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoi, N.N.; Slesarev, I.S.

    1988-01-01

    The paper presents discussions on the following topics: fuel supply for nuclear power; expansion of the sphere of nuclear power applications, such as district heating; comparative estimates of power reactor efficiencies; safety philosophy of advanced nuclear plants, including passive protection and inherent safety concepts; nuclear power unit of enhanced safety for the new generation of nuclear power plants. The emphasis is that designers of new generation reactors face a complicated but technically solvable task of developing highly safe, efficient, and economical nuclear power sources having a wide sphere of application

  7. Addition reaction of adamantylideneadamantane with Br2 and 2Br2: a computational study.

    Science.gov (United States)

    Islam, Shahidul M; Poirier, Raymond A

    2008-01-10

    Ab initio calculations were carried out for the reaction of adamantylideneadamantane (Ad=Ad) with Br2 and 2Br2. Geometries of the reactants, transition states, intermediates, and products were optimized at HF and B3LYP levels of theory using the 6-31G(d) basis set. Energies were also obtained using single point calculations at the MP2/6-31G(d)//HF/6-31G(d), MP2/6-31G(d)//B3LYP/6-31G(d), and B3LYP/6-31+G(d)//B3LYP/6-31G(d) levels of theory. Intrinsic reaction coordinate (IRC) calculations were performed to characterize the transition states on the potential energy surface. Only one pathway was found for the reaction of Ad=Ad with one Br2 producing a bromonium/bromide ion pair. Three mechanisms for the reaction of Ad=Ad with 2Br2 were found, leading to three different structural forms of the bromonium/Br3- ion pair. Activation energies, free energies, and enthalpies of activation along with the relative stability of products for each reaction pathway were calculated. The reaction of Ad=Ad with 2Br2 was strongly favored over the reaction with only one Br2. According to B3LYP/6-31G(d) and single point calculations at MP2, the most stable bromonium/Br3- ion pair would form spontaneously. The most stable of the three bromonium/Br3- ion pairs has a structure very similar to the observed X-ray structure. Free energies of activation and relative stabilities of reactants and products in CCl4 and CH2ClCH2Cl were also calculated with PCM using the united atom (UA0) cavity model and, in general, results similar to the gas phase were obtained. An optimized structure for the trans-1,2-dibromo product was also found at all levels of theory both in gas phase and in solution, but no transition state leading to the trans-1,2-dibromo product was obtained.

  8. First steps in designing an all-in-one ICT-based device for persons with cognitive impairment: evaluation of the first mock-up.

    Science.gov (United States)

    Boman, Inga-Lill; Persson, Ann-Christine; Bartfai, Aniko

    2016-03-07

    This project Smart Assisted Living involving Informal careGivers++ (SALIG) intends to develop an ICT-based device for persons with cognitive impairment combined with remote support possibilities for significant others and formal caregivers. This paper presents the identification of the target groups' needs and requirements of such device and the evaluation of the first mock-up, demonstrated in a tablet. The inclusive design method that includes end-users in the design process was chosen. First, a scoping review was conducted in order to examine the target group's need of an ICT-based device, and to gather recommendations regarding its design and functionalities. In order to capture the users' requirements of the design and functionalities of the device three targeted focus groups were conducted. Based on the findings from the publications and the focus groups a user requirement specification was developed. After that a design concept and a first mock-up was developed in an iterative process. The mock-up was evaluated through interviews with persons with cognitive impairment, health care professionals and significant others. Data were analysed using content analysis. Several useful recommendations of the design and functionalities of the SALIG device for persons with cognitive impairment were identified. The main benefit of the mock-up was that it was a single device with a set of functionalities installed on a tablet and designed for persons with cognitive impairment. An additional benefit was that it could be used remotely by significant others and formal caregivers. The SALIG device has the potentials to facilitate everyday life for persons with cognitive impairment, their significant others and the work situation for formal caregivers. The results may provide guidance in the development of different types of technologies for the target population and for people with diverse disabilities. Further work will focus on developing a prototype to be empirically tested

  9. A SCALE-UP Mock-Up: Comparison of Student Learning Gains in High- and Low-Tech Active-Learning Environments.

    Science.gov (United States)

    Soneral, Paula A G; Wyse, Sara A

    2017-01-01

    Student-centered learning environments with upside-down pedagogies (SCALE-UP) are widely implemented at institutions across the country, and learning gains from these classrooms have been well documented. This study investigates the specific design feature(s) of the SCALE-UP classroom most conducive to teaching and learning. Using pilot survey data from instructors and students to prioritize the most salient SCALE-UP classroom features, we created a low-tech "Mock-up" version of this classroom and tested the impact of these features on student learning, attitudes, and satisfaction using a quasi--experimental setup. The same instructor taught two sections of an introductory biology course in the SCALE-UP and Mock-up rooms. Although students in both sections were equivalent in terms of gender, grade point average, incoming ACT, and drop/fail/withdraw rate, the Mock-up classroom enrolled significantly more freshmen. Controlling for class standing, multiple regression modeling revealed no significant differences in exam, in-class, preclass, and Introduction to Molecular and Cellular Biology Concept Inventory scores between the SCALE-UP and Mock-up classrooms. Thematic analysis of student comments highlighted that collaboration and whiteboards enhanced the learning experience, but technology was not important. Student satisfaction and attitudes were comparable. These results suggest that the benefits of a SCALE-UP experience can be achieved at lower cost without technology features. © 2017 P. A. G. Soneral and S. A. Wyse. CBE—Life Sciences Education © 2017 The American Society for Cell Biology. This article is distributed by The American Society for Cell Biology under license from the author(s). It is available to the public under an Attribution–Noncommercial–Share Alike 3.0 Unported Creative Commons License (http://creativecommons.org/licenses/by-nc-sa/3.0).

  10. Results of Koo measurements of HTGR lattice by oscillated zero reactivity technique using the AGIP-NUCLEARE RB-2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Benedetti, F; Brighenti, G.; Chiodi, P. L.; Ghilardotti, G.; Giuliani, C.

    1974-10-15

    This paper describes k-infinity measurements conducted using an assembly of loose HTGR coated particles in the BR-2 reactor by means of null reactivity oscillating method comparing the effect of poisoned and unpoisoned lattices like tests performed in the Physical Constants Test Reactor (PCTR) at Hanford. The RB-2 reactor was the property of the Italian firm AGIP NUCLEARE and operated at the Montecuccolino Center in Bologna.

  11. Power supplyer for reactor coolant recycling pump

    International Nuclear Information System (INIS)

    Nara, Hiroshi; Okinaka, Yo.

    1991-01-01

    The present invention concerns a variable voltage/variable frequency static power source (static power source) used as a power source for a coolants recycling pump motor of a nuclear power plant. That is, during lower power operation such as start up or shutdown in which stoppage of the power source gives less effect to a reactor core, power is supplied from a power system, a main power generator connected thereto or a high voltage bus in the plant or a common high voltage bus to the static power source. However, during rated power operation, power is supplied from the output of an axially power generator connected with a main power generator having an extremely great inertia moment to the static power device. With such a constitution, the static power device is not stopped by the lowering of the voltage due to a thunderbolt falling accident or the like to a power-distribution line suddenly occurred in the power system. Accordingly, reactor core flowrate is free from rapid decrease caused by the reduction of rotation speed of the recycling pump. Accordingly, disadvantgages upon operation control in the reactor core is not caused. (I.S.)

  12. Small and medium power reactors 1987

    International Nuclear Information System (INIS)

    1987-12-01

    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

  13. Small and medium power reactors 1987

    Science.gov (United States)

    1987-12-01

    This TECDOC follows the publication of TECDOC-347: Small and Medium Power Reactors (SMPR) Project Initiation Study, Phase 1, 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 program. 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 1 of the above mentioned TECDOC-347.

  14. 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)

    Kodeli, I.

    2007-01-01

    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)

  15. 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.

  16. A SCALE-UP Mock-Up: Comparison of Student Learning Gains in High- and Low-Tech Active-Learning Environments

    Science.gov (United States)

    Soneral, Paula A. G.; Wyse, Sara A.

    2017-01-01

    Student-centered learning environments with upside-down pedagogies (SCALE-UP) are widely implemented at institutions across the country, and learning gains from these classrooms have been well documented. This study investigates the specific design feature(s) of the SCALE-UP classroom most conducive to teaching and learning. Using pilot survey data from instructors and students to prioritize the most salient SCALE-UP classroom features, we created a low-tech “Mock-up” version of this classroom and tested the impact of these features on student learning, attitudes, and satisfaction using a quasi-­experimental setup. The same instructor taught two sections of an introductory biology course in the SCALE-UP and Mock-up rooms. Although students in both sections were equivalent in terms of gender, grade point average, incoming ACT, and drop/fail/withdraw rate, the Mock-up classroom enrolled significantly more freshmen. Controlling for class standing, multiple regression modeling revealed no significant differences in exam, in-class, preclass, and Introduction to Molecular and Cellular Biology Concept Inventory scores between the SCALE-UP and Mock-up classrooms. Thematic analysis of student comments highlighted that collaboration and whiteboards enhanced the learning experience, but technology was not important. Student satisfaction and attitudes were comparable. These results suggest that the benefits of a SCALE-UP experience can be achieved at lower cost without technology features. PMID:28213582

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

    International Nuclear Information System (INIS)

    Banetta, S.; Zacchia, F.; Lorenzetto, P.; Bobin-Vastra, I.; Boireau, B.; Cottin, A.; Mitteau, R.; Eaton, R.; Raffray, R.

    2014-01-01

    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/m 2 ) 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

  18. 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)

    Granger, L.; Rieg, C.Y.; Touret, J.P.; Fleury, F.; Nahas, G.; Danisch, R.; Brusa, L.; Millard, A.; Laborderie, C.; Ulm, F.; Contri, P.; Schimmelpfennig, K.; Barre, F.; Firnhaber, M.; Gauvain, J; Coulon, N.; Dutton, L.M.C.; Tuson, A.

    2001-01-01

    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

  19. Multiple microprocessor based nuclear reactor power monitor

    International Nuclear Information System (INIS)

    Lewis, P.S.; Ethridge, C.D.

    1979-01-01

    The reactor power monitor is a portable multiple-microprocessor controlled data acquisition device being built for the International Atomic Energy Association. Its function is to measure and record the hourly integrated operating thermal power level of a nuclear reactor for the purpose of detecting unannounced plutonium production. The monitor consists of a 3 He proportional neutron detector, a write-only cassette tape drive and control electronics based on two INTEL 8748 microprocessors. The reactor power monitor operates from house power supplied by the plant operator, but has eight hours of battery backup to cover power interruptions. Both the hourly power levels and any line power interruptions are recorded on tape and in memory. Intermediate dumps from the memory to a data terminal or strip chart recorder can be performed without interrupting data collection

  20. Power distribution forecasting device for reactors

    International Nuclear Information System (INIS)

    Tsukii, Makoto

    1981-01-01

    Purpose: To save expensive calculations on the forecasting of reactor power distribution. Constitution: Core status (CSD) such as entire coolant flow rate, pressures in the reactor, temperatures at the outlet and inlet and positions for control rods are inputted into a power distribution calculation device to calculate the power distribution based on physical models intermittently. Further, present power distribution is calculated based on in-core neutron flux measured values and CSD in a process control computer. Further, the ratio of the calculation results of the latter to those of the former is calculated, stored and inputted into a correction device to correct the forecast power distribution obtained by the power distribution calculation device. This enables to forecast the power distribution with excellent responsivity in the reactor site. (Furukawa, Y.)

  1. Power distribution monitor in a nuclear reactor

    International Nuclear Information System (INIS)

    Uematsu, Hitoshi

    1983-01-01

    Purpose: To enable accurate monitoring for the reactor power distribution within a short time in a case where abnormality occurs in in-core neutron monitors or in a case where the reactor core state changes after the calibration for the neutron monitors. Constitution: The power distribution monitor comprises a power distribution calculator adapted to be inputted counted values from a reactor core present state data instruments and calculate the neutron flux distribution in the reactor core and the power distribution based on previously incorporated physical models, an RCF calculator adapted to be inputted with the counted values from the in-core neutron monitors and the neutron flux distribution and the power distribution calculated in the power distribution calculator and compensate the counted errors included in the counted values form the in-core neutron monitors and the calculation errors included in the power distribution calculated in the power distribution calculator to thereby calculate the power distribution within the reactor core, and an input/output device for the input of the data required for said power distribution calculator and the display for the calculation result calculated in the RCF calculator. (Ikeda, J.)

  2. Power conditioning system for a nuclear reactor

    International Nuclear Information System (INIS)

    Higashigawa, Yuichi; Joge, Toshio.

    1981-01-01

    Purpose: To provide a power conditioning system for a BWR type reactor which has a function to be automatically operated within a range that the relationship between the heat power of the reactor and the electric power of an electric generator does not lose the safety of fuel by eliminating the unnecessary fluctuation of the power of the reactor. Constitution: A load request error signal fed from a conventional turbine control system to recirculation flow regulator is eliminated, and a reactor power conditioning system is newly provided, to which an electric generator power signal, a reactor average power area monitor signal and a load request signal are inputted. Thus, the load request signal is compared directly with the electric power of the electric generator, the recirculation flow rate is controlled by the compared result, and whether the correlation between the heat power of the reqctor and the electric power of the generator satisfies the correlation determined to prove the safety of fuel or not is checked. If this correlation is satisfied, the recirculation flow rate is merely automatically controlled. (Yoshino, Y.)

  3. Nuclear power reactors: reactor safety and military and civil defence

    International Nuclear Information System (INIS)

    Hvinden, T.

    1976-01-01

    The formation of fission products and plutonium in reactors is briefly described, followed by a short general discussion of reactor safety. The interaction of reactor safety and radioactive release considerations with military and civil defence is thereafter discussed. Reactors and other nuclear plants are factors which must be taken into account in the defence of the district around the site, and as potential targets of both conventional and guerilla attacks and sabotage, requiring special defence. The radiological hazards arising from serious damage to a power reactor by conventional weapons are briefly discussed, and the benefits of underground siting evaluated. Finally the author discusses the significance of the IAEA safeguards work as a preventive factor. (JIW)

  4. Identification of fast power reactivity effect in nuclear power reactor

    International Nuclear Information System (INIS)

    Efanov, A.I.; Kaminskas, V.A.; Lavrukhin, V.S.; Rimidis, A.P.; Yanitskene, D.Yu.

    1987-01-01

    A nuclear power reactor is an object of control with distributed parameters, characteristics of which vary during operation time. At the same time the reactor as the object of control has internal feedback circuits, which are formed as a result of the effects of fuel parameters and a coolant (pressure, temperature, steam content) on the reactor breeding properties. The problem of internal feedback circuit identification in a nuclear power reactor is considered. Conditions for a point reactor identification are obtained and algorithms of parametric identification are constructed. Examples of identification of fast power reactivity effect for the RBMK-1000 reactor are given. Results of experimental testing have shown that the developed method of fast power reactivity effect identification permits according to the data of normal operation to construct adaptive models for the point nuclear reactor, designed for its behaviour prediction in stationary and transition operational conditions. Therefore, the models considered can be used for creating control systems of nuclear power reactor thermal capacity (of RBMK type reactor, in particular) which can be adapted to the change in the internal feedback circuit characteristics

  5. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    Science.gov (United States)

    Balent, R.

    1963-03-12

    This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

  6. A nuclear power reactor concept for Brazil

    International Nuclear Information System (INIS)

    Sefidvash, F.

    1980-01-01

    For the purpose of developing an independent national nuclear technology and effective manner of transferring such a technology, as well as developing a modern reactor, a new nuclear power reactor concept is proposed which is considered as a suitable and viable project for Brazil to support its development and finally construct its prototype as an indigeneous venture. (Author) [pt

  7. SEISMIC DESIGN CRITERIA FOR NUCLEAR POWER REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, R. A.

    1963-10-15

    The nature of nuclear power reactors demands an exceptionally high degree of seismic integrity. Considerations involved in defining earthquake resistance requirements are discussed. Examples of seismic design criteria and applications of the spectrum technique are described. (auth)

  8. Reactor power control method and device

    International Nuclear Information System (INIS)

    Fushimi, Atsushi; Ishii, Yoshihiko; Miyamoto, Yoshiyuki; Ishii, Kazuhiko; Kiyoharu, Norihiko; Aizawa, Yuko.

    1997-01-01

    The present invention provides a method and a device suitable to rise the temperature and increase the pressure of the reactor to an aimed pressure in accordance with an aimed value for a reactor water temperature changing rate in the course of rising temperature and increasing pressure of the reactor upon start up of a BWR type power plant. Namely, neutron fluxes in the reactor and the temperature of reactor water are detected respectively. The maximum value among the detected values for the neutron fluxes is detected. The reactor water temperature changing rate is calculated based on the detected values of the reactor water temperature, from which the maximum value of the reactor water temperature changing rate is detected. An aimed value for the neutron flux is calculated in accordance with both detected maximum values and the aimed value of the reactor water temperature changing rate. The position of control rods is adjusted in accordance with the aimed value for the calculated neutron flux. Then, an aimed value for the neutron flux for realizing the aimed value for the reactor water temperature changing rate can be obtained accurately with no influence of the sensitivity of the detected values of the neutron fluxes and the time delay of the reactor water temperature changing rate. (I.S.)

  9. 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

  10. Low power reactor for remote applications

    International Nuclear Information System (INIS)

    Meier, K.L.; Palmer, R.G.; Kirchner, W.L.

    1985-01-01

    A compact, low power reactor is being designed to provide electric power for remote, unattended applications. Because of the high fuel and maintenance costs for conventional power sources such as diesel generators, a reactor power supply appears especially attractive for remote and inaccessible locations. Operating at a thermal power level of 135 kWt, the power supply achieves a gross electrical output of 25 kWe from an organic Rankine cycle (ORC) engine. By intentional selection of design features stressing inherent safety, operation in an unattended mode is possible with minimal risk to the environment. Reliability is achieved through the use of components representing existing, proven technology. Low enrichment uranium particle fuel, in graphite core blocks, cooled by heat pipes coupled to an ORC converter insures long-term, virtually maintenance free, operation of this reactor for remote applications. 10 refs., 7 figs., 3 tabs

  11. MIT research reactor. Power uprate and utilization

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Lin-Wen [Nuclear Reactor Laboratory, Massachusetts Inst. of Technology, Cambridge, MA (United States)

    2012-03-15

    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)

  12. Neutron energy spectra calculations in the low power research reactor

    International Nuclear Information System (INIS)

    Omar, H.; Khattab, K.; Ghazi, N.

    2011-01-01

    The neutron energy spectra have been calculated in the fuel region, inner and outer irradiation sites of the zero power research reactor using the MCNP-4C code and the combination of the WIMS-D/4 transport code for generation of group constants and the three-dimensional CITATION diffusion code for core analysis calculations. The neutron energy spectrum has been divided into three regions and compared with the proposed empirical correlations. The calculated thermal and fast neutron fluxes in the low power research reactor MNSR inner and outer irradiation sites have been compared with the measured results. Better agreements have been noticed between the calculated and measured results using the MCNP code than those obtained by the CITATION code. (author)

  13. WWER-440 control assembly local power peaking investigation on LR-0 reactor

    International Nuclear Information System (INIS)

    Mikus, J.

    2002-01-01

    This paper presents information concerning the local power peaking problem induced by the WWER-440 control assembly and the investigation possibilities on the light water, zero power reactor LR-0 at the Nuclear Research Institute (NRI) Rez plc. A brief description is given about the disposable control assembly model, experimental arrangement and conditions on the LR-0 reactor with regard to the earlier performed investigations as well as to the relevant measurements to be realized in the near future.(abstract)

  14. Gaseous fuel reactors for power systems

    International Nuclear Information System (INIS)

    Helmick, H.H.; Schwenk, F.C.

    1978-01-01

    The Los Alamos Scientific Laboratory is participating in a NASA-sponsored program to demonstrate the feasibility of a gaseous uranium fueled reactor. The work is aimed at acquiring experimental and theoretical information for the design of a prototype plasma core reactor which will test heat removal by optical radiation. The basic goal of this work is for space applications, however, other NASA-sponsored work suggests several attractive applications to help meet earth-bound energy needs. Such potential benefits are small critical mass, on-site fuel processing, high fuel burnup, low fission fragment inventory in reactor core, high temperature for process heat, optical radiation for photochemistry and space power transmission, and high temperature for advanced propulsion systems. Low power reactor experiments using uranium hexafluoride gas as fuel demonstrated performance in accordance with reactor physics predictions. The final phase of experimental activity now in progress is the fabrication and testing of a buffer gas vortex confinement system

  15. reactor power control using fuzzy logic

    International Nuclear Information System (INIS)

    Ahmed, A.E.E.

    2001-01-01

    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

  16. Reactor Power Meter type SG-8

    Energy Technology Data Exchange (ETDEWEB)

    Glowacki, S W

    1981-01-01

    The report describes the principle and electronic circuits of the Reactor Power Meter type SG-8. The gamma radiation caused by the activity of the reactor first cooling circuit affectes the ionization chamber being the detector of the instrument. The output detector signal direct current is converted into the frequency of electric pulses by means of the current-to-frequency converter. The output converter frequency is measured by the digital frequency meter: the number of measured digits in time unit is proportional to the reactor power.

  17. New generation of reactors for space power

    International Nuclear Information System (INIS)

    Boudreau, J.E.; Buden, D.

    1982-01-01

    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

  18. Compact reactor/ORC power source

    International Nuclear Information System (INIS)

    Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

    1986-01-01

    A compact power source that combines an organic Rankine Cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for component development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500 0 C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370 0 C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analysis have aided in the power source design. The analyses have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high. 10 refs

  19. A compact reactor/ORC power source

    International Nuclear Information System (INIS)

    Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

    1986-01-01

    A compact power source that combines an organic Rankine cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for componenet development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500 0 C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370 0 C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analyses combined with a finite element thermal analysis have aided in the power source design. The analysis have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high

  20. Thorium utilization in power reactors

    International Nuclear Information System (INIS)

    Saraceno; Marcos.

    1978-10-01

    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) [es