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Sample records for aircraft shield test reactor

  1. REACTOR AND SHIELD PHYSICS. Comprehensive Technical Report, General Electric Direct-Air-Cycle, Aircraft Nuclear Propulsion Program.

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, W.E.; Simpson, J.D.

    1962-01-01

    This volume is one of twenty-one summarizing the Aircraft Nuclear Propulsion Program of the General Electric Company. This volume describes the experimental and theoretical work accomplished in the areas of reactor and shield physics.

  2. Aircraft shielding experiments at general dynamics Fort Worth

    International Nuclear Information System (INIS)

    The Nuclear Aircraft Research Facility was established by Convair, Fort Worth, in 1950 under U.S. Air Force auspices to support the Aircraft Nuclear Propulsion Program in the areas of shielding and radiation effects problems affecting the airframe. The company subsequently became General Dynamics, Fort Worth. In 1954, an experimental shielding program was developed by B.P. Leonard and N.M. Schaeffer that incorporated air, ground, and structure scattering experiments with three sources: a large Co source, the gorund test reactor (GTR), and finally, the aircraft shield test reactor (ASTR). Shield penetration measurements were also planned with the GTR. Principal elements of this program are summarized in the paper

  3. Space reactor shield technology

    International Nuclear Information System (INIS)

    The reactor shield mass contributes a large portion (10% to 25%) to the total mass of an unmanned reactor system. Different shield materials are required to attenuate neutrons and gamma rays and still obtain a minimum mass. The shield material selection should also consider structural characteristics, physical and chemical properties, fabricability and availability. Minimum mass is achieved by using a shadow shield. Neutron capture gamma ray and heat generation are extremely important considerations. Lithium hydride was selected for the neutron shield material due to its excellent properties. It has to be canned and may be compartmentalized to reduce the probability of complete shielding effectiveness loss due to meteoroid puncture of the can. The initial shield design was based on previous SNAP shield design experience. The Monte Carlo Neutron Photon code, which includes the radiation scattering with the radiator and power conversion system, was then used to ensure that the design requirements were met. Fabrication of the shield by casting techniques is recommended to maintain shield integrity during vibration and to accommodate complex penetrations. A method for casting full-scale shields is described

  4. Shielding calculations for the Tokamak Fusion Test Reactor neutral beam injectors

    International Nuclear Information System (INIS)

    Two-dimensional discrete-ordinates calculations have been performed to determine the location and thickness of concrete shielding around the Tokamak Fusion Test Reactor neutral beam injectors. Two sets of calculations were performed, one to determine the dose equivalent rate on the roof and wall of the test cell building when no injectors are present, and one to determine the contribution to the dose equivalent rate at these locations from radiation streaming through the injection duct. Shielding the side and rear of the neutral beam injector with 0.305 and 0.61 m of concrete, respectively, and lining the inside of the test cell wall with an additional layer of concrete having a thickness of 0.305 m and a height above the axis of deuteron injection of 3.10 m is sufficient to maintain the biological dose equivalent rate outside the test cell to approx. 1 mrem/D-T pulse

  5. An investigation of the effects of water content on the shielding performance of the primary upper shield in the High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    High Temperature Engineering Test Reactor (HTTR) is the first High Temperature Gas-cooled Reactor in Japan. The plant layout and radiation shielding are designed so that the plant can be operated without any employee receiving a high radiation dose rate. The primary upper shield of HTTR is composed of concrete (grout) and carbon steel. The function of the primary upper shield is to attenuate neutrons and gamma rays generated in the core to satisfy dose rate criterion for the operating floor. Since the HTGR uses high temperature helium as coolant, temperatures of shielding materials could be higher than that of conventional reactors. According to the analytical simulation, the maximum temperature inside of the primary upper shield during full-power operation was estimated to be about 85degC. Since water content in the primary upper shield concrete depends on its temperature, shielding performance of the concrete under HTTR operational conditions need to be confirmed. Thus the water content of the concrete was experimentally investigated by out-of-pile heat-up tests to 175degC. According to these tests, a water release model was developed. The out-of-pile heat-up tests showed that the water content in the concrete was larger than 78 kg/m3 up to 110degC in a 95% confidence limit. (author)

  6. High performance inboard shield design for the compact TIBER-II test reactor: Appendix A-2

    International Nuclear Information System (INIS)

    The compactness of the TIBER-II reactor has placed a premium on the design of a high performance inboard shield to protect the inner legs of the toroidal field (TF) coils. The available space for shield is constrained to 48 cm and the use of tungsten is mandatory to protect the magnet against the 1.53 MW/m2 neutron wall loading. The primary requirement for the shield is to limit the fast neutron fluence to 1019 n/cm2. In an optimization study, the performance of various candidate materials for protecting the magnet was examined. The optimum shield consists of a 40 cm thick W layer, followed by an 8 cm thick H2O/LiNO3 layer. The mechanical design of the shield calls for tungsten blocks within SS stiffened panels. All the coolant channels are vertical with more of them in the front where there is a high heat load. The coolant pressure is 0.2 MPa and the maximum structural surface temperature is 0C. The effects of the detailed mechanical design of the shield and the assembly gaps between the shield sectors on the damage in the magnet were analyzed and peaking factors of ∼2 were found at the hot spots. 2 refs., 6 figs., 2 tabs

  7. Shielding benchmark test

    International Nuclear Information System (INIS)

    Iron data in JENDL-2 have been tested by analyzing shielding benchmark experiments for neutron transmission through iron block performed at KFK using CF-252 neutron source and at ORNL using collimated neutron beam from reactor. The analyses are made by a shielding analysis code system RADHEAT-V4 developed at JAERI. The calculated results are compared with the measured data. As for the KFK experiments, the C/E values are about 1.1. For the ORNL experiments, the calculated values agree with the measured data within an accuracy of 33% for the off-center geometry. The d-t neutron transmission measurements through carbon sphere made at LLNL are also analyzed preliminarily by using the revised JENDL data for fusion neutronics calculation. (author)

  8. Tower Shielding Reactor II design and operation report. Vol. 3. Assembling and testing of the control mechanism assembly

    International Nuclear Information System (INIS)

    The mechanisms that are operated to control the reactivity of the Tower Shielding Reactor II(TSR-II) are mounted on a Control Mechanism Housing (CMH) that is centered inside the reactor core. The information required to procure, fabricate, inspect, and assemble a CMH is contained in the ORNL engineering drawings listed in the appropriate sections. The components are fabricated and inspected from these drawings in accordance with a Quality Assurance Plan and a Manufacturing Plan. The material in this report describes the acceptance and performance tests of CMH subassemblies used ty the Tower Shielding Facility (TSF) staff but it can also be used by personnel fabricating the components. This information which was developed and used before the advent of the formalized QA Program and Manufacturing Plans evolved during the fabrication and testing of the first five CMHs

  9. Space reactor shielding: an assessment of the technology

    International Nuclear Information System (INIS)

    Space power reactor systems require shielding to protect payload and reactor shielding components, and also maintenance and operating personnel. Shield composition, size, and shape are important design considerations, since the shield can dominate the overall weight of the system. Techniques for space reactor shield design analysis and optimization and experimental test facilities are available for design verification. With these tools, a shielding technology in support of current and future space power reactor systems can be developed. Efforts in this direction should begin with a generic shielding program to provide information on materials properties and geometric effects and should be followed by project-specific shielding programs to provide design optimization and prototype shield verification

  10. Radiation shield for nuclear reactors

    International Nuclear Information System (INIS)

    A reusable radiation shield for use in a reactor installation comprises a thin-walled, flexible and resilient container, made of plastic or elastomeric material, containing a hydrogenous fluid with boron compounds in solution. The container can be filled and drained in position and the fluid can be recirculated if required. When not in use the container can be folded and stored in a small space. The invention relates to a shield to span the top of the annular space between a reactor vessel and the primary shield. For this purpose a continuous toroidal container or a series of discrete segments is used. Other forms can be employed for different purposes, e.g. mattress- or blanket-like forms can be draped over potential sources of radiation or suspended from a mobile carrier and placed between a worker and a radiation source. (author)

  11. Water shielding nuclear reactor container

    International Nuclear Information System (INIS)

    The reactor container of the present invention contains a reactor pressure vessel, and has double steel plate walls endurable to elevated inner pressure and keeping airtightness, and shielding water is filled inside from a water injection port. It is endurable to a great inner pressure satisfactorily and keep airtightness by the two spaced relatively thin steel plates. It exhibits radiation shielding effect by filling water substantially the same as that of a conventional reactor container made of iron reinforced concretes. Then, it is no more necessary to use concretes for the construction of the reactor container, which shortens the term of the construction, and saves the construction cost. In addition, a cooling effect for the reactor container is provided. Syphons are disposed contiguously to a water injection port and the top end of the syphon is immersed in an equipment temporarily storage pool, and further, pipelines are connected to the double steel plate walls or the syphons for supplying shielding water to enhance the cooling effect. (N.H.)

  12. Reactor vessel head permanent shield

    International Nuclear Information System (INIS)

    A nuclear reactor is described comprising: a nuclear reactor pressure vessel closure head; control rod drive mechanisms (CRDMs) disposed within the closure head so as to project vertically above the closure head; cooling air baffle means surrounding the control rod drive mechanisms for defining cooling air paths relative to the control rod drive mechanisms; means defined within the periphery of the closure head for accommodating fastening means for securing the closure head to its associated pressure vessel; lifting lugs fixedly secured to the closure head for facilitating lifting and lowering movements of the closure head relative to the pressure vessel; lift rods respectively operatively associated with the plurality of lifting lugs for transmitting load forces, developed during the lifting and lowering movements of the closure head, to the lifting lugs; upstanding radiation shield means interposed between the cooling air baffle means and the periphery of the enclosure head of shielding maintenance personnel operatively working upon the closure head fastening means from the effects of radiation which may emanate from the control rod drive mechanisms and the cooling air baffle means; and connecting systems respectively associated with each one of the lifting lugs and each one of the lifting rods for connecting each one of the lifting rods to a respective one of each one of the lifting lugs, and for simultaneously connecting a lower end portion of the upstanding radiation shield means to each one of the respective lifting lugs

  13. Shielding calculations for ETRR-1 reactor

    International Nuclear Information System (INIS)

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

  14. Development and testing of multigroup library with correction of self-shielding effects in fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    A multigroup library HENDL2.1/SS (Hybrid Evaluated Nuclear Data Library/Self-Shielding) based on ENDF/B-VII.0 evaluate data has been generated using Bondarenko and flux calculator method for the correction of self-shielding effect of neutronics analyses. To validate the reliability of the multigroup library HENDL2.1/SS, transport calculations for fusion-fission hybrid system FDS-I were performed in this paper. It was verified that the calculations with the HENDL2.1/SS gave almost the same results with MCNP calculations and were better than calculations with the HENDL2.0/MG which is another multigroup library without self-shielding correction. The test results also showed that neglecting resonance self-shielding caused underestimation of the Keff, neutron fluxes and waste transmutation ratios in the multigroup calculations of FDS-I.

  15. Development and testing of multigroup library with correction of self-shielding effects in fusion-fission hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zou Jun, E-mail: jzou@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); He Zhaozhong; Zeng Qin; Qiu Yuefeng; Wang Minghuang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China)

    2010-12-15

    A multigroup library HENDL2.1/SS (Hybrid Evaluated Nuclear Data Library/Self-Shielding) based on ENDF/B-VII.0 evaluate data has been generated using Bondarenko and flux calculator method for the correction of self-shielding effect of neutronics analyses. To validate the reliability of the multigroup library HENDL2.1/SS, transport calculations for fusion-fission hybrid system FDS-I were performed in this paper. It was verified that the calculations with the HENDL2.1/SS gave almost the same results with MCNP calculations and were better than calculations with the HENDL2.0/MG which is another multigroup library without self-shielding correction. The test results also showed that neglecting resonance self-shielding caused underestimation of the K{sub eff}, neutron fluxes and waste transmutation ratios in the multigroup calculations of FDS-I.

  16. Concepts and Tests for the Remote-Controlled Dismantling of the Biological Shield and Form work of the KNK Reactor - 13425

    Energy Technology Data Exchange (ETDEWEB)

    Neff, Sylvia; Graf, Anja; Petrick, Holger; Rothschmitt, Stefan [WAK Rueckbau- und Entsorgungs- GmbH, P.O.Box 12 63, 76339 Eggenstein- Leopoldshafen (Germany); Klute, Stefan [Siempelkamp Nukleartechnik GmbH, Am Taubenfeld 25/1, 69123 Heidelberg (Germany); Stanke, Dieter [Siempelkamp NIS Ingenieurgesellschaft mbH, Industriestrasse 13, 63755 Alzenau (Germany)

    2013-07-01

    The compact sodium-cooled nuclear reactor facility Karlsruhe (KNK), a prototype Fast Breeder, is currently in an advanced stage of dismantling. Complete dismantling is based on 10 partial licensing steps. In the frame of the 9. decommissioning permit, which is currently ongoing, the dismantling of the biological shield is foreseen. The biological shield consists of heavy reinforced concrete with built-in steel fitments, such as form-work of the reactor tank, pipe sleeves, ventilation channels, and measuring devices. Due to the activation of the inner part of the biological shield, dismantling has to be done remote-controlled. During a comprehensive basic design phase a practical dismantling strategy was developed. Necessary equipment and tools were defined. Preliminary tests revealed that hot wire plasma cutting is the most favorable cutting technology due to the geometrical boundary conditions, the varying distance between cutter and material, and the heavy concrete behind the steel form-work. The cutting devices will be operated remotely via a carrier system with an industrial manipulator. The carrier system has expandable claws to adjust to the varying diameter of the reactor shaft during dismantling progress. For design approval of this prototype development, interaction between manipulator and hot wire plasma cutting was tested in a real configuration. For the demolition of the concrete structure, an excavator with appropriate tools, such as a hydraulic hammer, was selected. Other mechanical cutting devices, such as a grinder or rope saw, were eliminated because of concrete containing steel spheres added to increase the shielding factor of the heavy concrete. Dismantling of the biological shield will be done in a ring-wise manner due to static reasons. During the demolition process, the excavator is positioned on its tripod in three concrete recesses made prior to the dismantling of the separate concrete rings. The excavator and the manipulator carrier system

  17. Concepts and Tests for the Remote-Controlled Dismantling of the Biological Shield and Form work of the KNK Reactor - 13425

    International Nuclear Information System (INIS)

    The compact sodium-cooled nuclear reactor facility Karlsruhe (KNK), a prototype Fast Breeder, is currently in an advanced stage of dismantling. Complete dismantling is based on 10 partial licensing steps. In the frame of the 9. decommissioning permit, which is currently ongoing, the dismantling of the biological shield is foreseen. The biological shield consists of heavy reinforced concrete with built-in steel fitments, such as form-work of the reactor tank, pipe sleeves, ventilation channels, and measuring devices. Due to the activation of the inner part of the biological shield, dismantling has to be done remote-controlled. During a comprehensive basic design phase a practical dismantling strategy was developed. Necessary equipment and tools were defined. Preliminary tests revealed that hot wire plasma cutting is the most favorable cutting technology due to the geometrical boundary conditions, the varying distance between cutter and material, and the heavy concrete behind the steel form-work. The cutting devices will be operated remotely via a carrier system with an industrial manipulator. The carrier system has expandable claws to adjust to the varying diameter of the reactor shaft during dismantling progress. For design approval of this prototype development, interaction between manipulator and hot wire plasma cutting was tested in a real configuration. For the demolition of the concrete structure, an excavator with appropriate tools, such as a hydraulic hammer, was selected. Other mechanical cutting devices, such as a grinder or rope saw, were eliminated because of concrete containing steel spheres added to increase the shielding factor of the heavy concrete. Dismantling of the biological shield will be done in a ring-wise manner due to static reasons. During the demolition process, the excavator is positioned on its tripod in three concrete recesses made prior to the dismantling of the separate concrete rings. The excavator and the manipulator carrier system

  18. Early test facilities and analytic methods for radiation shielding: Proceedings

    International Nuclear Information System (INIS)

    This report represents a compilation of eight papers presented at the 1992 American Nuclear Society/European Nuclear Society International Meeting. The meeting is of special significance since it commemorates the fiftieth anniversary of the first controlled nuclear chain reaction. The papers contained in this report were presented in a special session organized by the Radiation Protection and Shielding Division in keeping with the historical theme of the meeting. The paper titles are good indicators of their content and are: (1) The origin of radiation shielding research: The Oak Ridge experience, (2) Shielding research at the hanford site, (3) Aircraft shielding experiments at General Dynamics Fort Worth, 1950-1962, (4) Where have the neutrons gone?, a history of the tower shielding facility, (5) History and evolution of buildup factors, (6) Early shielding research at Bettis atomic power laboratory, (7) UK reactor shielding: then and now, (8) A very personal view of the development of radiation shielding theory

  19. Early test facilities and analytic methods for radiation shielding: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, D.T. (comp.) (Oak Ridge National Lab., TN (United States)); Ingersoll, J.K. (comp.) (Tec-Com, Knoxville, TN (United States))

    1992-11-01

    This report represents a compilation of eight papers presented at the 1992 American Nuclear Society/European Nuclear Society International Meeting. The meeting is of special significance since it commemorates the fiftieth anniversary of the first controlled nuclear chain reaction. The papers contained in this report were presented in a special session organized by the Radiation Protection and Shielding Division in keeping with the historical theme of the meeting. The paper titles are good indicators of their content and are: (1) The origin of radiation shielding research: The Oak Ridge experience, (2) Shielding research at the hanford site, (3) Aircraft shielding experiments at General Dynamics Fort Worth, 1950-1962, (4) Where have the neutrons gone , a history of the tower shielding facility, (5) History and evolution of buildup factors, (6) Early shielding research at Bettis atomic power laboratory, (7) UK reactor shielding: then and now, (8) A very personal view of the development of radiation shielding theory.

  20. Shielding design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    This report first describes the basic design philosophy of radiation shields for the fusion experimental reactor (FER) which has been proposed to be the next step machine to JT-60. Next, geometrical models and calculation parameters for shielding calculations were investigated to establish the standard design calculation methods, and accuracy of the calculation was evaluated. Further, irradiation properties of in-vessel components and bulk shielding properties were summarized in the useful form for the future design works. (author)

  1. Technical specifications for the bulk shielding reactor

    International Nuclear Information System (INIS)

    This report provides information concerning the technical specifications for the Bulk Shielding Reactor. Areas covered include: safety limits and limiting safety settings; limiting conditions for operation; surveillance requirements; design features; administrative controls; and monitoring of airborne effluents. 10 refs

  2. Shielding design to obtain compact marine reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, Akio; Sako, Kiyoshi (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment)

    1994-06-01

    The marine reactors equipped in previously constructed nuclear ships are in need of the secondary shield which is installed outside the containment vessel. Most of the weight and volume of the reactor plants are occupied by this secondary shield. An advanced marine reactor called MRX (Marine Reactor X) has been designed to obtain a more compact and lightweight marine reactor with enhanced safety. The MRX is a new type of marine reactor which is an integral PWR (The steam generator is installed in the pressure vessel.) with adopting a water-filled containment vessel and a new shielding design method of no installation of the secondary shield. As a result, MRX is considerably lighter in weight and more compact in size as compared with the reactors equipped in previously constructed nuclear ships. For instance, the plant weight and volume of the containment vessel of MRX are about 50% and 70% of those of the Nuclear Ship MUTSU, in spite of the power of MRX is 2.8 times as large as the MUTSU's reactor. The shielding design calculation was made using the ANISN, DOT3.5, QAD-CGGP2 and ORIGEN codes. The computational accuracy was confirmed by experimental analyses. (author).

  3. Reactor shielding. Report of a panel

    International Nuclear Information System (INIS)

    Reactor shielding is necessary that people may work and live in the vicinity of reactors without receiving detrimental biological effects and that the necessary materials and instrumentation for reactor operation may function properly. Much of the necessary theoretical work and experimental measurement has been accomplished in recent years. Scientists have developed some very sophisticated methods which have contributed to a more thorough understanding of the problems involved and have produced some very reliable results leading to significant reductions in shield configurations. A panel of experts was convened from 9 to 13 March 1964 in Vienna at the Headquarters of the International Atomic Energy Agency to discuss the present status of reactor shielding. The participants were prominent shielding experts from most of the laboratories engaged in this field throughout the world. They presented status reports describing the past history and plans for further development of reactor shielding in their countries and much valuable discussion took place on some of the most relevant aspects of reactor shielding. All this material is presented in this report, together with abstracts of the supporting papers read to the Panel

  4. Planetary surface reactor shielding using indigenous materials

    International Nuclear Information System (INIS)

    The exploration and development of Mars will require abundant surface power. Nuclear reactors are a low-cost, low-mass means of providing that power. A significant fraction of the nuclear power system mass is radiation shielding necessary for protecting humans and/or equipment from radiation emitted by the reactor. For planetary surface missions, it may be desirable to provide some or all of the required shielding from indigenous materials. This paper examines shielding options that utilize either purely indigenous materials or a combination of indigenous and nonindigenous materials

  5. Shield structure for a nuclear reactor

    International Nuclear Information System (INIS)

    An improved nuclear reactor shield structure is described for use where there are significant amounts of fast neutron flux above an energy level of approximately 70 keV. The shield includes structural supports and neutron moderator and absorber systems. A portion at least of the neutron moderator material is magnesium oxide either alone or in combination with other moderator materials such as graphite and iron. (U.K.)

  6. Technical specifications: Tower Shielding Reactor II

    International Nuclear Information System (INIS)

    The technical specifications define the key limitations that must be observed for safe operation of the Tower Shielding Reactor II (TSR-II) and an envelope of operation within which there is reasonable assurance that these limits cannot be exceeded. The specifications were written to satisfy the requirements of the Department of Energy (DOE) Manual Chapter 0540, September 1, 1972

  7. A Monitoring System for the Assessment of Reactor Shield Performance

    International Nuclear Information System (INIS)

    This paper describes the objectives of a shield survey, presents the results of intercomparisons and makes recommendations on the selection of a monitoring system. The performance of a reactor shield must be assessed efficiently to enable immediate repairs to be made to defective shields. Correlation of survey results with shield designer's predictions has proved difficult and the factors contributing to the disagreement have been examined. The shield designer uses flux to dose conversion factors which neglect the radiation equilibrium situation at the shield-air interface and the effect of attenuation and scattering external to the shield. The presence of the operator can significantly influence a measurement. The use of the ''maximum exposure dose'' (MED) concept is recommended for gamma photon predictions and this is compared with absorbed dose measurements in phantoms. Shield survey instruments must have a flat response over a wide range of energies. For example, many gamma dosimeters are designed for use at energies up to about 2 MeV whereas around most reactors a proportion of dose is due to 6 MeV gamma photons, with some contribution from higher energies. Tests were carried out to select suitable detectors using actual reactor operating conditions and simulated conditions at specific energies, notably 6 MeV. In practice discrepancies exceeding a factor of 3 were frequently found. A comparison was made between thermoluminescent dosimeters and film badges for the purpose of fixed position integrating dosimeters. The film dosimeter (AERE/RPS) was found to over-estimate by a factor of 2 for 6 MeV gamma radiation and in practical situations over-estimated by 20 to 80%. Thermoluminescent dosimetry is recommended for shield surveys provided that a build-up cap is used to achieve charged particle equilibrium. (author)

  8. Performance test on shielding concrete

    International Nuclear Information System (INIS)

    The cylinder of the shielding concrete is made from common Portland cement and home-made coarse or fine aggregates. Orthogonal design experiment and regression analysis are adopted to study the effects of the water content, sand percentage and water-cement ratio on the property of shielding concrete and the difference between them. The test shows that the tensile strength is in inverse proportion with water-cement ratio, and the influence is quite significant. Another factor is the type of aggregates. The effect of the age on its density is not obvious. Similarly, the concrete shielding γ rays shares the same influencing factors with that shielding neutron rays on density, slump and tensile strength. And both have the same change rules regarding to mechanical property. (authors)

  9. Nuclear data requirements for fusion reactor shielding

    International Nuclear Information System (INIS)

    The nuclear data requirements for experimental, demonstration and commercial fusion reactors are reviewed. Particular emphasis is given to the shield as well as major reactor components of concern to the nuclear performance. The nuclear data requirements are defined as a result of analyzing four key areas. These are the most likely candidate materials, energy range, types of needed nuclear data, and the required accuracy in the data. Deducing the latter from the target goals for the accuracy in prediction is also discussed. A specific proposal of measurements is recommended. Priorities for acquisition of data are also assigned. (author)

  10. STUDY OF WING SHIELDING EFFECT OF PROPELLER AIRCRAFT

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    The calculation of wing shielding effect starts from solving Ffowcs Williams and Hawkings equation without quadrupole source in time domain. The sound scattering of the wing and fuselage which are surrounded by a multi-propeller sound field is modeled as a second sound source. A program is developed to calculate the acoustical effects of the rigid fuselage as well as wings with arbitrary shape in motion at low Mach number. As an example, the numerical calculation of the wing shielding of Y12 aircraft with an approximate shape is presented. The result manifests clearly the shielding effect of the wing on the fuselage and the approach is more efficient than that published before.

  11. Methodology of shielding calculation for nuclear reactors

    International Nuclear Information System (INIS)

    A methodology of calculation that coupling a serie of computer codes in a net that make the possibility to calculate the radiation, neutron and gamma transport, is described, for deep penetration problems, typical of nuclear reactor shielding. This net of calculation begining with the generation of constant multigroups, for neutrons and gamma, by the AMPX system, coupled to ENDF/B-IV data library, the transport calculation of these radiations by ANISN, DOT 3.5 and Morse computer codes, up to the calculation of absorbed doses and/or equivalents buy SPACETRAN code. As examples of the calculation method, results from benchmark n0 6 of Shielding Benchmark Problems - ORNL - RSIC - 25, namely Neutron and Secondary Gamma Ray fluence transmitted through a Slab of Borated Polyethylene, are presented. (Author)

  12. Structural design of shield-integrated thin-wall vacuum vessel and manufacturing qualification tests for International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Conceptual design of shield-integrated thin-wall vacuum vessel has been done for ITER (International Thermonuclear Experimental Reactor). The vacuum vessel concept is based on a thin-double-wall structure, which consists of inner and outer plates and rib stiffeners. Internal shielding structures, which provide neutron irradiation shielding to protect TF coils, are set up between the inner plate and the outer plate of the vessel to avoid complexity of machine systems such as supporting systems of blanket modules. The vacuum vessel is assembled/disassembled by remote handling, so that welding joints are chosen as on-site joint method from reliability of mechanical strength. From a view point of assembling TF coils, the vacuum vessel is separated at the side of port, and is divided into 32 segments similar to the ITER-CDA reference design. Separatrix sweeping coils are located in the vacuum vessel to reduce heat fluxes onto divertor plates. Here, the coil structure and attachment to the vacuum vessel have been investigated. A sectorized saddle-loop coil is available for assembling and disassembling the coil. To support electromagnetic loads on the coils, they are attached to the groove in the vacuum vessel by welding. Flexible multi-plate supporting structure (compression-type gravity support), which was designed during CDA, is optimized by investigating buckling and frequency response properties, and concept on manufacturing and fabrication of the gravity support are proposed. Partial model of the vacuum vessel is manufactured for trial, so that fundamental data on welding and fabrication are obtained. From mechanical property tests of weldment and partial models, mechanical intensity and behaviors of the weldment are obtained. Informations on FEM-modeling are obtained by comparing analysis results with experimental results. (author)

  13. Design of radiation shields in nuclear reactor core

    International Nuclear Information System (INIS)

    This article consists of designing radiation shields in the core of nuclear reactors to control and restrain the harmful nuclear radiations in the nuclear reactor cores. The radiation shields protect the loss of energy. caused by nuclear radiation in a nuclear reactor core and consequently, they cause to increase the efficiency of the reactor and decrease the risk of being under harmful radiations for the staff. In order to design these shields, by making advantages of the Oppenheim Electrical Networkmethod, the structure of the shields are physically simulated and by obtaining a special algorithm, the amount of optimized energy caused by nuclear radiations, is calculated

  14. Investigation of shielding analysis method for fusion reactors

    International Nuclear Information System (INIS)

    An investigation has been made, at the shielding laboratory, into the status of shielding analysis method for fusion reactor based on conceptual designs of a variety of fusion power reactors and fusion experimental facilities, in cooperation with the Fusion Reactor Shielding Working Group in the Research Committee on Fast Neutron Shielding of the Atomic Energy Society of Japan. The reactors and facilities considered are CULHAM MKII(U.K), SPTR (Japan), TFTR(U.S.A.), STARFIRE(U.S.A.) and INTOR-USA(U.S.A.). (author)

  15. Shielding benchmark test of JENDL-3

    Energy Technology Data Exchange (ETDEWEB)

    Kawai, M. (Toshiba Corp., Kawasaki, Kanagawa (Japan)); Hasegawa, A.; Ueki, K. (and others)

    1990-02-01

    The integral test of JENDL-3 for shielding application was made on a cross sections of carbon, sodium and iron by analyzing the various shielding benchmark experiments: Broomstick experiments at ORNL for iron and sodium, neutron transmission experiments for sodium at ORNL, iron and carbon, ASPIS deep penetration experiments for iron, measurements of leakage spectrum from iron spheres at KfK, angular neutron spectrum measurements in graphite block at RPI. Analyses were made with radiation transport codes ANISN (1D, Sn), DIAC (1D, Sn), DOT-3.5 (2D, Sn) and MCNP (3D, point Monte Carlo). It was observed that revising JENDL-3T iron data resulted in an improvement in reproducing the experimental data, particularly in the MeV neutron energy region. For sodium, JENDL-3 gave better results than JENDL-2 and ENDF/B-IV. For carbon, JENDL-3 gave better agreement, compared to ENDF/B-IV. The conclusion is that JENDL-3 is highly applicable to the reactor shielding analyses. (author).

  16. Shielding design for research and education reactor

    International Nuclear Information System (INIS)

    For the purpose of education and research at the University, 20-KW powered SLOWPOKE-2 research reactor has been chosen as a prototype reactor. In order to study the safety characteristics of the reactor, exposure rate has been estimated at the pool boundary. Reactor core as a radiation source is assumed to be cylindrical volume source. Thus point kernel integration method can be applied to determine the exposure rate. For the sake of simplicity, calculation was done only for the prompt fission gamma rays and fission product gamma rays. As a result, the maximum exposure rate at the pool boundary was estimated to be 18R/min at the same height of the center of the core. In order to examine the accuracy for the point kernel integration method, two shielding experiments were carried out: one for the water tank only and the other for with concrete blocks outside the water tank. Water tank was made of wood pieces which is 13.4cm wide, 1.5cm thick and 2.15m long. Thus the water tank has the total dimension of 1 m radius and 2.1 m height. The experiment was carried out for the radiation source of 0.968 mCi Co-60 at the center of the water tank and the penetrated gamma rays were measured at 5 different detector positions. For the measurement and analysis of the responses, NaI(T1) 3''x3'' detector and 256 channel multichannel analyzer was utilized. To convert pulse height distribution to the exposure rate, Moriuchi conversion factor was adopted. Data from the calculations by point kernel method were well agreed within 10% band with the data from the the experiments. (Author)

  17. Tests of Neutron Spectrum Calculations with the Help of Foil Measurements in a D2O and in an H2O-Moderated Reactor and in Reactor Shields of Concrete an Iron

    International Nuclear Information System (INIS)

    Foil measurements covering the fast, epithermal and thermal neutron energy regions have been made in the centre of the Swedish D2O-moderated reactor R1, in the pool reactor R2-0, and in different positions in reactor shields of iron, magnetite concrete and ordinary concrete. Neutron spectra have also been calculated for most of these positions, often with the help of a numerical integration of the Boltzmann equation. The measurements and the calculated spectra are presented

  18. Weight Assessment for Fuselage Shielding on Aircraft With Open-Rotor Engines and Composite Blade Loss

    Science.gov (United States)

    Carney, Kelly; Pereira, Michael; Kohlman, Lee; Goldberg, Robert; Envia, Edmane; Lawrence, Charles; Roberts, Gary; Emmerling, William

    2013-01-01

    The Federal Aviation Administration (FAA) has been engaged in discussions with airframe and engine manufacturers concerning regulations that would apply to new technology fuel efficient "openrotor" engines. Existing regulations for the engines and airframe did not envision features of these engines that include eliminating the fan blade containment systems and including two rows of counter-rotating blades. Damage to the airframe from a failed blade could potentially be catastrophic. Therefore the feasibility of using aircraft fuselage shielding was investigated. In order to establish the feasibility of this shielding, a study was conducted to provide an estimate for the fuselage shielding weight required to provide protection from an open-rotor blade loss. This estimate was generated using a two-step procedure. First, a trajectory analysis was performed to determine the blade orientation and velocity at the point of impact with the fuselage. The trajectory analysis also showed that a blade dispersion angle of 3deg bounded the probable dispersion pattern and so was used for the weight estimate. Next, a finite element impact analysis was performed to determine the required shielding thickness to prevent fuselage penetration. The impact analysis was conducted using an FAA-provided composite blade geometry. The fuselage geometry was based on a medium-sized passenger composite airframe. In the analysis, both the blade and fuselage were assumed to be constructed from a T700S/PR520 triaxially-braided composite architecture. Sufficient test data on T700S/PR520 is available to enable reliable analysis, and also demonstrate its good impact resistance properties. This system was also used in modeling the surrogate blade. The estimated additional weight required for fuselage shielding for a wing- mounted counterrotating open-rotor blade is 236 lb per aircraft. This estimate is based on the shielding material serving the dual use of shielding and fuselage structure. If the

  19. Fusion reactor blanket/shield design study

    International Nuclear Information System (INIS)

    A joint study of tokamak reactor first-wall/blanket/shield technology was conducted by Argonne National Laboratory (ANL) and McDonnell Douglas Astronautics Company (MDAC). The objectives of this program were the identification of key technological limitations for various tritium-breeding-blanket design concepts, establishment of a basis for assessment and comparison of the design features of each concept, and development of optimized blanket designs. The approach used involved a review of previously proposed blanket designs, analysis of critical technological problems and design features associated with each of the blanket concepts, and a detailed evaluation of the most tractable design concepts. Tritium-breeding-blanket concepts were evaluated according to the proposed coolant. The ANL effort concentrated on evaluation of lithium- and water-cooled blanket designs while the MDAC effort focused on helium- and molten salt-cooled designs. A joint effort was undertaken to provide a consistent set of materials property data used for analysis of all blanket concepts. Generalized nuclear analysis of the tritium breeding performance, an analysis of tritium breeding requirements, and a first-wall stress analysis were conducted as part of the study. The impact of coolant selection on the mechanical design of a tokamak reactor was evaluated. Reference blanket designs utilizing the four candidate coolants are presented

  20. Jet Noise Shielding Provided by a Hybrid Wing Body Aircraft

    Science.gov (United States)

    Doty, Michael J.; Brooks, Thomas F.; Burley, Casey L.; Bahr, Christopher J.; Pope, Dennis S.

    2014-01-01

    One approach toward achieving NASA's aggressive N+2 noise goal of 42 EPNdB cumulative margin below Stage 4 is through the use of novel vehicle configurations like the Hybrid Wing Body (HWB). Jet noise measurements from an HWB acoustic test in NASA Langley's 14- by 22-Foot Subsonic Tunnel are described. Two dual-stream, heated Compact Jet Engine Simulator (CJES) units are mounted underneath the inverted HWB model on a traversable support to permit measurement of varying levels of shielding provided by the fuselage. Both an axisymmetric and low noise chevron nozzle set are investigated in the context of shielding. The unshielded chevron nozzle set shows 1 to 2 dB of source noise reduction (relative to the unshielded axisymmetric nozzle set) with some penalties at higher frequencies. Shielding of the axisymmetric nozzles shows up to 6.5 dB of reduction at high frequency. The combination of shielding and low noise chevrons shows benefits beyond the expected additive benefits of the two, up to 10 dB, due to the effective migration of the jet source peak noise location upstream for increased shielding effectiveness. Jet noise source maps from phased array results processed with the Deconvolution Approach for the Mapping of Acoustic Sources (DAMAS) algorithm reinforce these observations.

  1. Measurement of Neutron Tissue Dose Outside the Reactor Shielding

    International Nuclear Information System (INIS)

    Intermediate neutrons form an important part of the neutron-tissue dose outside the reactor shielding. Equipment developed in the RUS series makes it possible to measure the flux and the tissue dose rate of intermediate neutrons. In experiments on the 1RT-1000 reactor the neutron-dose composition was studied and it was shown that this depends greatly on the composition of the shielding. It was found that the neutron-tissue dose calculated from data obtained by means of RPN-1 apparatus is in reality too low by a factor of up to 1.5 for water shielding and 5 for concrete shielding. (author)

  2. Evaluation of the influence of aircraft shielding on the aircrew exposure through an aircraft mathematical model.

    Science.gov (United States)

    Ferrari, A; Pelliccioni, M; Villari, R

    2004-01-01

    In order to investigate the influence of aircraft shielding on the galactic component of cosmic rays, an aircraft mathematical model has been developed by the combinatorial geometry package of the Monte-Carlo transport code FLUKA. The isotropic irradiation of the aircraft in the cosmic ray environment has been simulated. Effective dose and ambient dose equivalent rates have been determined inside the aircraft at several locations along the fuselage, at a typical civil aviation altitude (10 580 m), for vertical cut-off rigidity of 0.4 GV (poles) and 17.6 GV (equator) and deceleration potential of 465 MV. The values of both quantities were generally lower than those in the free atmosphere. They depend, in an intricate manner, on the location within the aircraft, quantity of fuel, number of passengers, etc. The position onboard of crew members should be taken into account when assessing individual doses. Likewise due consideration must be taken when positioning detectors which are used to measure H*(10). Care would be needed to avoid ambiguity when comparing the results of calculation with the experimental data. PMID:14978289

  3. Evaluation of the influence of aircraft shielding on the aircrew exposure through an aircraft mathematical model

    International Nuclear Information System (INIS)

    In order to investigate the influence of aircraft shielding on the galactic component of cosmic rays, an aircraft mathematical model has been developed by the combinatorial geometry package of the Monte-Carlo transport code FLUKA. The isotropic irradiation of the aircraft in the cosmic ray environment has been simulated. Effective dose and ambient dose equivalent rates have been determined inside the aircraft at several locations along the fuselage, at a typical civil aviation altitude (10 580 m), for vertical cut-off rigidity of 0.4 GV (poles) and 17.6 GV (equator) and deceleration potential of 465 MV. The values of both quantities were generally lower than those in the free atmosphere. They depend, in an intricate manner, on the location within the aircraft, quantity of fuel, number of passengers, etc. The position onboard of crew members should be taken into account when assessing individual doses. Likewise due consideration must be taken when positioning detectors which are used to measure H *(10). Care would be needed to avoid ambiguity when comparing the results of calculation with the experimental data. (authors)

  4. Fusion reactor design towards radwaste minimum with advanced shield material

    International Nuclear Information System (INIS)

    A new concept of fusion reactor design is proposed to minimize the radioactive waste of the reactor. The main point of the concept is to clear massive structural components located outside the neutron shield from regulatory control. The concept requires some reinforcement of shielding with an advanced shield material such as a metal hydride, detriation, and tailoring of a detrimental element from the superconductor. Our assessment confirmed a large impact of the concept on radwaste reduction, in that it reduces the radwaste fraction of a fusion reactor A-SSTR2 from 92 wt.% to 17 wt.%. (author)

  5. Advances in shielding calculations for the PEC reactor

    International Nuclear Information System (INIS)

    In this paper calculations of neutron and gamma streaming through various penetrations in the plug and neutron shield of the sodium cooled fast reactor PEC, currently under construction, are described. The object of the calculations has been to verify the accessibility, 3 days after reactor shut-down, of the area directly above the reactor. (author)

  6. Reactor cavity cleanup system shielded filter installation

    International Nuclear Information System (INIS)

    The Seabrook Station reactor cavity cleanup system provides a flow path for refueling pool purification and drain down during plant refueling evolutions. The original system design included refueling pool surface skimmers and drains, a skimmer pump, an unshielded duplex basket type pump suction strainer and interconnecting stainless steel piping. The piping design utilized socket welded joints in small bore pipe with diaphragm values installed in the horizontal pipe runs downstream of the skimmer pump. The previously installed unshielded strainer in addition to the skimmer pump downstream piping components were determined to be inconsistent with Seabrook's proactive approach to dose reduction. To be consistent with ALARA (As Low As Reasonably Achievable) policy, a plant design change was authorized to install a lead shielded filter unit as a replacement for the existing duplex strainer. This filter unit, which utilizes multiple micron rating disposable basket type cartridges, has a threefold function of protecting the skimmer pump from large solids, providing bulk filtration of activated corrosion products from the refueling water in order to minimize CRUD buildup in downstream components, and enabling retrieval of foreign material drawn into the refueling pool drains

  7. Development of methods for shield and reactor analyses

    International Nuclear Information System (INIS)

    Methods development work in the Division continues in the shielding area (discrete ordinates and Monte Carlo) and in the reactor analysis area, primarily with the VENTURE system. An interesting new effort is directed toward depletion perturbation theory for reactor burnup analysis. A new version of the basic two-dimensional discrete ordinates code, DOT 4.3, has been developed and is now undergoing testing and documentation. This code includes diffusion acceleration, and has improved convergence rates and adaptability to non-IBM computers. An international review of Monte Carlo codes included ORNL experience with the French TRIPOLI code as well as the ORNL MORSE code. The core analysis package based around the VENTURE three-dimensional diffusion theory neutronics code now includes depletion and fuel management capability. The accurate computation of resonance reaction rates in heterogeneous reactor lattices is one of the most challenging problems in LWR physics. Under EPRI sponsorship a new methodology (i.e., OZMA) was developed for solving the point-energy integral or discrete ordinates transport equations on a dense energy grid as required to handle resonance profiles. The new approach eliminates simultaneously most of the approximations involved in simple treatments of resonance shielding, and the statistical problems inevitable in Monte Carlo calculations

  8. Mobile reactor concepts as applied to testing of compact fusion reactors

    International Nuclear Information System (INIS)

    Compact fusion reactor concepts have recently received increased emphasis because of advantages principally related to their low cost and short development time. Physics experiments are underway and test results are sufficiently encouraging to merit consideration of ignition-type experiments. Since experiments of this nature involve radioactivity, the requirement for test facilities which incorporate remote handling capabilities becomes apparent. One approach to a test facility concept which has particularly attractive features is based on the mobile test reactor concept employing facilities such as are found at the Idaho National Engineering Laboratory (INEL). The mobile reactor test concept was developed in the 1950s and was used extensively in the testing of aircraft nuclear propulsion reactors at the INEL. In this instance, test reactors were assembled on a dolly and were transported to and from test facilities on a four-rail track system. Nuclear operations were conducted from heavily shielded underground control rooms and, for major maintenance operations, the reactors were unplugged and returned to a large, centrally located hot shop. A similar concept is envisioned for compact fusion reactor testing

  9. The mobile reactor concept as applied to testing of compact fusion reactors

    International Nuclear Information System (INIS)

    Compact fusion reactor concepts have recently received increased emphasis because of advantages principally related to their low cost and short development time. Physics experiments are underway and test results are sufficiently encouraging to merit consideration of ''ignition''-type experiments. Since experiments of this nature involve radioactivity, the requirement for test facilities which incorporate remote handling capabilities becomes apparent. One approach to a test facility concept which has particularly attractive features is based on the mobile test reactor concept employing facilities such as are found at the Idaho National Engineering Laboratory (INEL). The mobile reactor test concept was developed in the 1950s and was used extensively in the testing of aircraft nuclear propulsion reactors at the INEL. In this instance, test reactors were assembled on a dolly and were transported to and from test facilities on a fourrail track system. Nuclear operations were conducted from heavily shielded underground control rooms and, for major maintenance operations, the reactors were ''unplugged'' and returned to a large, centrally located ''hot'' shop. A similar concept is envisioned for compact fusion reactor testing

  10. Shielding benchmark tests of JENDL-3

    International Nuclear Information System (INIS)

    The integral test of neutron cross sections for major shielding materials in JENDL-3 has been performed by analyzing various shielding benchmark experiments. For the fission-like neutron source problem, the following experiments are analyzed: (1) ORNL Broomstick experiments for oxygen, iron and sodium, (2) ASPIS deep penetration experiments for iron, (3) ORNL neutron transmission experiments for iron, stainless steel, sodium and graphite, (4) KfK leakage spectrum measurements from iron spheres, (5) RPI angular neutron spectrum measurements in a graphite block. For D-T neutron source problem, the following two experiments are analyzed: (6) LLNL leakage spectrum measurements from spheres of iron and graphite, and (7) JAERI-FNS angular neutron spectrum measurements on beryllium and graphite slabs. Analyses have been performed using the radiation transport codes: ANISN(1D Sn), DIAC(1D Sn), DOT3.5(2D Sn) and MCNP(3D point Monte Carlo). The group cross sections for Sn transport calculations are generated with the code systems PROF-GROUCH-G/B and RADHEAT-V4. The point-wise cross sections for MCNP are produced with NJOY. For comparison, the analyses with JENDL-2 and ENDF/B-IV have been also carried out. The calculations using JENDL-3 show overall agreement with the experimental data as well as those with ENDF/B-IV. Particularly, JENDL-3 gives better results than JENDL-2 and ENDF/B-IV for sodium. It has been concluded that JENDL-3 is very applicable for fission and fusion reactor shielding analyses. (author)

  11. Shielding benchmark tests of JENDL-3

    Energy Technology Data Exchange (ETDEWEB)

    Kawai, Masayoshi [Toshiba Corp., Kawasaki, Kanagawa (Japan); Hasegawa, Akira; Ueki, Kohtaro; Yamano, Naoki; Sasaki, Kenji; Matsumoto, Yoshihiro; Takemura, Morio; Ohtani, Nobuo; Sakurai, Kiyoshi

    1994-03-01

    The integral test of neutron cross sections for major shielding materials in JENDL-3 has been performed by analyzing various shielding benchmark experiments. For the fission-like neutron source problem, the following experiments are analyzed: (1) ORNL Broomstick experiments for oxygen, iron and sodium, (2) ASPIS deep penetration experiments for iron, (3) ORNL neutron transmission experiments for iron, stainless steel, sodium and graphite, (4) KfK leakage spectrum measurements from iron spheres, (5) RPI angular neutron spectrum measurements in a graphite block. For D-T neutron source problem, the following two experiments are analyzed: (6) LLNL leakage spectrum measurements from spheres of iron and graphite, and (7) JAERI-FNS angular neutron spectrum measurements on beryllium and graphite slabs. Analyses have been performed using the radiation transport codes: ANISN(1D Sn), DIAC(1D Sn), DOT3.5(2D Sn) and MCNP(3D point Monte Carlo). The group cross sections for Sn transport calculations are generated with the code systems PROF-GROUCH-G/B and RADHEAT-V4. The point-wise cross sections for MCNP are produced with NJOY. For comparison, the analyses with JENDL-2 and ENDF/B-IV have been also carried out. The calculations using JENDL-3 show overall agreement with the experimental data as well as those with ENDF/B-IV. Particularly, JENDL-3 gives better results than JENDL-2 and ENDF/B-IV for sodium. It has been concluded that JENDL-3 is very applicable for fission and fusion reactor shielding analyses. (author).

  12. Experimental Evaluation of the Thermal Performance of a Water Shield for a Surface Power Reactor

    International Nuclear Information System (INIS)

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 deg. C. The CFD model with 1/6-g predicts a maximum water temperature of 88 deg. C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield

  13. EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    REID, ROBERT S. [Los Alamos National Laboratory; PEARSON, J. BOSIE [Los Alamos National Laboratory; STEWART, ERIC T. [Los Alamos National Laboratory

    2007-01-16

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

  14. Thermal top shield for gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Proposed is a thermal top shield for gas-cooled nuclear reactors which together with the thermal side and bottom shield forms an almost gas-tight room for taking up the core structure and which protects the top of the concrete vessel sufficiently against overheating. The thermal top shield consists of top shield elements put closely together, which are made of at least two horizontal metal layers and at least one moderator layer located between the metal layers and which are fixed to the top liner by means of drawbars. (orig.)

  15. The optimum shielding for a power reactor using local components

    International Nuclear Information System (INIS)

    Some local concrete mixtures have been picked out (selected) to be studied as shielding concrete for prospective nuclear power reactor in Syria. This research has interested in the attenuation of gamma radiation and neutron fluxes by these local concretes in the ordinary conditions. In addition to the heat effect on the shielding and physical properties of local concrete. Furthermore the neutron activation of the elements of the local concrete mixtures have been studied that for selection the low-activation materials (low dose rate and short half life radioisotopes). In this way biological shielding for nuclear reactor can be safe during operation of nuclear power reactor, in addition to be low radioactive waste after decommissioning the reactor. (author)

  16. SP-100 GES/NAT radiation shielding systems design and development testing

    International Nuclear Information System (INIS)

    Advanced Energy Systems (AES) of Westinghouse Electric Corporation is under subcontract to the General Electric Company to supply nuclear radiation shielding components for the SP-100 Ground Engineering System (GES) Nuclear Assembly Test to be conducted at Westinghouse Hanford Company at Richland, Washington. The radiation shielding components are integral to the Nuclear Assembly Test (NAT) assembly and include prototypic and non-prototypic radiation shielding components which provide prototypic test conditions for the SP-100 reactor subsystem and reactor control subsystem components during the GES/NAT operations. W-AES is designing three radiation shield components for the NAT assembly; a prototypic Generic Flight System (GFS) shield, the Lower Internal Facility Shield (LIFS), and the Upper Internal Facility Shield (UIFS). This paper describes the design approach and development testing to support the design, fabrication, and assembly of these three shield components for use within the vacuum vessel of the GES/NAT. The GES/NAT shields must be designed to operate in a high vacuum which simulates space operations. The GFS shield and LIFS must provide prototypic radiation/thermal environments and mechanical interfaces for reactor system components. The NAT shields, in combination with the test facility shielding, must provide adequate radiation attenuation for overall test operations. Special design considerations account for the ground test facility effects on the prototypic GFS shield. Validation of the GFS shield design and performance will be based on detailed Monte Carlo analyses and developmental testing of design features. Full scale prototype testing of the shield subsystems is not planned

  17. ANS shielding standards for light-water reactors

    International Nuclear Information System (INIS)

    The purpose of the American Nuclear Society Standards Subcommittee, ANS-6, Radiation Protection and Shielding, is to develop standards for radiation protection and shield design, to provide shielding information to other standards-writing groups, and to develop standard reference shielding data and test problems. A total of seven published ANS-6 standards are now current. Additional projects of the subcommittee, now composed of nine working groups, include: standard reference data for multigroup cross sections, gamma-ray absorption coefficients and buildup factors, additional benchwork problems for shielding problems and energy spectrum unfolding, power plant zoning design for normal and accident conditions, process radiation monitors, and design for postaccident radiological conditions

  18. Radiological shielding of low power compact reactor: calculation and design

    International Nuclear Information System (INIS)

    The development of compact reactors becoming a technology that offers great projection and innumerable use possibilities, both in electricity generation and in propulsion.One of the requirements for the operation of this type of reactor is that it must include a radiological shield that will allow for different types of configurations and that, may be moved with the reactor if it needs to be transported.The nucleus of a reactor emits radiation, mainly neutrons and gamma rays in the heat of power, and gamma radiation during the radioactive decay of fission products.This radiation must be restrained in both conditions of operation to avoid it affecting workers or the public.The combination of different materials and properties in layers results in better performance in the form of a decrease in radiation, hence causing the dosage outside the reactor, whether in operation or shut down, to fall within the allowed limits.The calculations and design of radiological shields is therefore of paramount importance in reactor design.The choice of material and the design of the shield have a strong impact on the cost and the load capacity, the latter being one of the characteristics to optimize.The imposed condition of design is that the reactor can be transported together with the decay shield in a standard container of 40 foot

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

    International Nuclear Information System (INIS)

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

  20. Shielding benchmark test for JENDL-3T

    Energy Technology Data Exchange (ETDEWEB)

    Hasegawa, Akira (Japan Atomic Energy Research Inst., Tokai, Ibaraki. Tokai Research Establishment)

    1988-03-01

    The results of the shielding benchmark tests for JENDL-3T (testing stage version of JENDL-3), performed by JNDC Shielding Sub-working group, are summarized. Especially, problems of total cross-section in MeV range for O, Na, Fe, revealed from the analysis of the Broomstick's experiment, are discussed in details. For the deep penetration profiles of Fe, which is very important feature in shielding calculation, ASPIS benchmark experiment is analysed and discussed. From the study overall applicability of JENDL-3T data for the shielding calculation is confirmed. At the same time some problems still remained are also pointed out. By the reflection of this feedback information applicability of JENDL-3, forth coming official version, will be greatly improved.

  1. Integral test of JENDL-3.3 with shielding benchmarks

    International Nuclear Information System (INIS)

    Integral tests of neutron and gamma-ray production data for cross-section libraries based on the Japanese Evaluated Nuclear Data Library, Version 3.3 (JENDL-3.3) have been performed by using shielding benchmarks. An evaluation scheme for shielding benchmark analysis established in Japanese Nuclear Data Committee (JNDC) was applied to the integral test for medium-heavy nuclei such as Oxygen, Sodium, Aluminum, Silicon, Titanium, Vanadium, Chromium, Iron, Cobalt, Nickel, Copper, Zirconium, Niobium, Molybdenum, Tungsten and Mercury. Calculations were made based on a continuous-energy Monte Carlo code MCNP4C and multi-group discrete ordinates codes ANISN, DORT and TORT. Calculations with JENDL-3.2, ENDF/B-VI, EFF-2, FENDL-1 and FENDL-2 were also made for comparison. The results of JENDL-3.3 were generally satisfactory and the cross-section libraries generated with JENDL-3.3 were verified to shielding applications for fission and fusion reactors. (author)

  2. Activation of TRIGA Mark II research reactor concrete shield

    International Nuclear Information System (INIS)

    To determine neutron activation inside the TRIGA research reactor concrete body a special sample-holder for irradiation inside horizontal channel was developed and tested. In the sample-holder various samples can be irradiated at different concrete shielding depths. In this paper the description of the sample-holder, experiment conditions and results of long-lived activation measurements are given. Long-lived neutron-induced gamma-ray-emitting radioactive nuclides in the samples were measured with HPGe detector. The most active long-lived radioactive nuclides in ordinary concrete samples were found to be 60Co and 152Eu and in barytes concrete samples 60Co, 152Eu and 133Ba. Measured activity density of all nuclides was found to decrease almost linearly with depth in logarithmic scale. (author)

  3. Design, fabrication, and testing of gadolinium-shielded metal fuel samples in the hydraulic tube of the high flux isotope reactor

    International Nuclear Information System (INIS)

    The use of hydraulic rabbit capsules inserted into and ejected from the core of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) during full power operation allows for precise control of the neutron fluence in fueled experiments. Rabbit capsules with strong thermal neutron absorbers must be used to screen out thermal neutrons, thereby reducing the heat generation rate while maintaining the fast neutron flux that produces displacement damage similar to fast reactor type conditions. However, rapid insertion and ejection of rabbit capsules containing a strong neutron absorber causes a reactivity response in the reactor that has the potential to engage the HFIR safety response system which could result in an unplanned shutdown. Therefore, a set of tests were performed to provide the data needed to establish limits on the reactivity worth that can be ejected from the hydraulic facility without causing a reactor shutdown. This paper will describe the design, operation, and results of the reactivity measurements undertaken to understand the reactor response to insertion of the gadolinium-lined rabbit capsules. (author)

  4. A device for vertically positioning a nuclear-reactor shield

    International Nuclear Information System (INIS)

    Description is given of a device for vertically positioning a shield over a nuclear reactor, comprising a closing-head mounted on a pressurized container of the reactor, a number of hoisting-rods distributed about said closing-head, fixed to the latter, at one extremity, and leaving their other extremities protruding through openings in the shield so as to support the latter on said hoisting-rods. The latter are stationary and each of them supports members comprising a tubular element carried by a shoulder of said stationary hoisting-rods and a device driven on the tubular element cooperating with it so that the tubular element is capable of vertically moving the shield towards a new position

  5. Influence of aircraft impact on seismic isolated SMR reactor

    International Nuclear Information System (INIS)

    In the past decades a lot of effort has been done to increase the reliability of NPP, particularly against the earthquakes effects, adopting the highly attractive strategy of the seismic isolation. Isolator bearings seem able to increase the safety margin/integrity of the safety relevant nuclear structures and to enable the standardization of the reactor design to be deployed across a wider range of sites. However in principle the design of a nuclear power plant depends on the safety aspects related also to other type of external events, like the aircraft impact that was/is of relevant importance for NPP safety (especially after the Sept. 2001) and must be considered in the design of both Generation III+ and IV reactors. This paper is related to a preliminary study of the global response of a seismically isolated reactor building subjected to a vicious commercial aircraft impact. In this framework the effects of impulsive loading due to the progressive aircraft crashing were evaluated, considering the potential for structural failure of the external building walls due to shearing and bending dynamic loads, with reference to the effects of the structure perforation, including concrete spalling of the internal surfaces and propagation of dynamic waves that could affect NPP safety systems and structures. To the purpose a rather refined numerical methodology was employed; three-dimensional models (FEM approach) of a reference SMR reactor containment and possible realistic aircraft structures were set up and used in the performed analyses, taking also into account suitable materials behaviour and constitutive laws. The structural analysis of the reference NPP internal components was carried out to appropriately check mainly the containment strength margin in the case of the considered accident and test the chosen models and numerical calculation approach. (author)

  6. Shielding design calculation of a 50 MW research reactor

    International Nuclear Information System (INIS)

    The computer code ANISN/PC has been applied to calculate the group flux distribution across different shield layers of a 50 MW light water research reactor. The code has been run in P3 approximation and S8 discrete ordinates. The calculated group fluxes multiplied by appropriate flux-to-dose rate conversion factors have been used to give the dose distribution across the shield layers. The thickness of the concrete shield has been determined to give the dose rate at the outer surface of the shield as 0.5 nSv/sec. The same calculation have been also performed in axial direction to determine the thickness of water needed above the core to reduce the dose level to 25 nSv/sec. The result of calculation shows that the contribution of capture gamma rays to the total dose at the outer surface of the shield is more than 50 percent. This simplifies the calculations to determine the shield layer thickness, especially in preliminary stages of the shield design. (author)

  7. ORNL fusion reactor shielding integral experiments

    International Nuclear Information System (INIS)

    Integral experiments that measure the neutron and gamma-ray energy spectra resulting from the attenuation of approx. 14 MeV T(D,n) 4He reaction neutrons in laminated slabs of stainless steel type 304, borated polyethylene, and a tungsten alloy (Hevimet) and from neutrons streaming through a 30-cm-diameter iron duct (L/D = 3) imbedded in a concrete shield have been performed. The facility, the NE-213 liquid scintillator detector system, and the experimental techniques used to obtain the measured data are described. The two-dimensional discrete ordinates radiation transport codes, calculational models, and nuclear data used in the analysis of the experiments are reviewed

  8. Operating manual for the Bulk Shielding Reactor

    International Nuclear Information System (INIS)

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxillary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supercedes all previous operating manuals for the BSR

  9. Operating manual for the Bulk Shielding Reactor

    International Nuclear Information System (INIS)

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR

  10. Neutronics analysis of inboard shielding capability for a DEMO fusion reactor CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Songlin; Li, Jiangang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Zheng, Shanliang, E-mail: shanliang.zheng@ccfe.ac.uk [Culham Centre for Fusion Energy, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Mitchell, Neil [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    The inboard shielding of a fusion reactor can be a crucial issue due to the limited space available in a tokamak configuration. It is necessary to assess the inboard shielding capability of DEMO for its initial design. In this paper, 1D and 3D neutronics models were developed based on a reference design of the Chinese Fusion Engineering Testing Reactor (CFETR). The neutron wall load (NWL) is in the range of 1.5–3 MW/m{sup 2} and the inboard shielding thickness is constrained within 40–70 cm in order to achieve the tritium self-sufficiency of the reactor. Referring to the detailed design of the ITER Toroidal Field Coils (TFCs) and using radiation hardening technology developed for ITER, the inboard blanket shielding capability and nuclear responses of the TFC are investigated for both FLiBe and Li{sub 4}SiO{sub 4} breeding blanket concepts. The impact of the gaps on shielding performance is discussed. Some suggestions on improving the inboard shielding performance for DEMO are also proposed.

  11. 14 CFR 21.127 - Tests: aircraft.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Tests: aircraft. 21.127 Section 21.127 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT CERTIFICATION PROCEDURES FOR PRODUCTS AND PARTS Production Under Type Certificate Only § 21.127 Tests: aircraft. (a)...

  12. Analysis and Testing of a Composite Fuselage Shield for Open Rotor Engine Blade-Out Protection

    Science.gov (United States)

    Pereira, J. Michael; Emmerling, William; Seng, Silvia; Frankenberger, Charles; Ruggeri, Charles R.; Revilock, Duane M.; Carney, Kelly S.

    2015-01-01

    The Federal Aviation Administration is working with the European Aviation Safety Agency to determine the certification base for proposed new engines that would not have a containment structure on large commercial aircraft. Equivalent safety to the current fleet is desired by the regulators, which means that loss of a single fan blade will not cause hazard to the Aircraft. The NASA Glenn Research Center and The Naval Air Warfare Center (NAWC), China Lake, collaborated with the FAA Aircraft Catastrophic Failure Prevention Program to design and test lightweight composite shields for protection of the aircraft passengers and critical systems from a released blade that could impact the fuselage. In the test, two composite blades were pyrotechnically released from a running engine, each impacting a composite shield with a different thickness. The thinner shield was penetrated by the blade and the thicker shield prevented penetration. This was consistent with pre-test predictions. This paper documents the live fire test from the full scale rig at NAWC China Lake and describes the damage to the shields as well as instrumentation results.

  13. Shielding considerations for advanced space nuclear reactor systems

    International Nuclear Information System (INIS)

    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO2) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications

  14. Shielding analyses of the IFMIF test cell

    International Nuclear Information System (INIS)

    Full 3-D shielding calculations of the IFMIF test cell were performed using a computational scheme for coupled Monte Carlo/deterministic transport calculations that enables the use of a detailed geometry model of the test cell in the Monte Carlo calculation and is suitable, at the same time, to handle the deep penetration transport through the thick surrounding concrete walls. Calculations for the test cell cover, which includes numerous penetrations through which neutrons stream, were performed by the Monte Carlo method. The results demonstrate that the dose rate limit for work personnel access to the access/maintenance room can be safely met during IFMIF operation assuming the test modules are surrounded by a horseshoe shield and the back heavy concrete wall is no less than 250 cm thick. No work personnel access to the room above the cover will be permitted during IFMIF operation due to the strong neutron streaming through the cover penetrations

  15. EMC Test Report Electrodynamic Dust Shield

    Science.gov (United States)

    Carmody, Lynne M.; Boyette, Carl B.

    2014-01-01

    This report documents the Electromagnetic Interference E M I evaluation performed on the Electrodynamic Dust Shield (EDS) which is part of the MISSE-X System under the Electrostatics and Surface Physics Laboratory at Kennedy Space Center. Measurements are performed to document the emissions environment associated with the EDS units. The purpose of this report is to collect all information needed to reproduce the testing performed on the Electrodynamic Dust Shield units, document data gathered during testing, and present the results. This document presents information unique to the measurements performed on the Bioculture Express Rack payload; using test methods prepared to meet SSP 30238 requirements. It includes the information necessary to satisfy the needs of the customer per work order number 1037104. The information presented herein should only be used to meet the requirements for which it was prepared.

  16. Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors.

    Science.gov (United States)

    Mohammadi, A; Hassanzadeh, M; Gharib, M

    2016-02-01

    In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified. PMID:26720262

  17. Neutron activation measurements in research reactor concrete shield

    International Nuclear Information System (INIS)

    The results of activation measurement inside TRIGA research reactor concrete shielding are given. Samples made of ordinary and barytes concrete together with gold and nickel foils were irradiated in the reactor body. Long-lived neutron-induced gamma-ray-emitting radioactive nuclides in the samples were measured with HPGe detector. The most active longlived radioactive nuclides in ordinary concrete samples were found to be 60Co and 152Eu and in barytes concrete samples 60Co, 152Eu and 133Ba. Measured activity density of all nuclides was found to decrease almost linearly with depth in logarithmic scale.(author)

  18. Embrittlement of the Shippingport reactor neutron shield tank

    International Nuclear Information System (INIS)

    The irradiation embrittlement of the Shippingport neutron shield tank material (A212 Grade B steel) has been characterized. Irradiation increases the Charpy transition temperature (CTT) by 23--28 degree C (41--50 degree F) and decreases the upper shelf energy. The shift in CTT is not as severe as that observed in the HFIR surveillance specimens. However, the actual value of CTT is higher than that for the HFIR data and the toughness at service temperature is low. The increase in yield stress is 51 MPa (7.4 ksi), which is comparable to the HFIR data. The results also indicate that the material is weaker in the TL orientation than LT orientation. Some effects of the location across the thickness of the wall are also observed; CTT is slightly greater for the specimens from the inner region of the wall. The data agree well with results from high-flux test reactors. Annealing studies indicate complete recovery of embrittlement after a 2-h anneal at 400 degree C. Although the weld metal is significantly tougher than the base metal, the shifts in CTT are comparable. The weld metal shows a strong affect of location across the thickness of the wall; only the inner regions of the weld show embrittlement. 11 refs., 14 figs., 3 tabs

  19. Irradiation induced embrittlement of steels used as reactor pressure vessel and end shields

    International Nuclear Information System (INIS)

    A review of the various factors that influence the irradiation induced changes in mechanical properties of reactor pressure vessel and end shield steel materials have been brought out in this report followed by some typical results on 3 w/o nickel steels and A 302B type of pressure vessel steels. These are the type of steels used in operating reactors in India. In this report the irradiation induced mechanical property changes have been analysed from reported results on nil ductility transition temperature changes (NDTT) from impact tests. The report also brings out the importance of carrying out reactor surveillance programme. (author)

  20. Shielding integrity testing of radioactive material transport packaging

    International Nuclear Information System (INIS)

    Although this Code of Practice is intended primarily to cover shielding integrity test requirements for off-site shielded radioactive material transport packaging, it may also be partly applicable to containers and specialised handling equipment (e.g. fuelling machines) used only on site, and to radiation shielding generally. The code is not concerned with proving adequacy of shielding design or with its absolute shielding value. (author)

  1. Where are we and where are we going in reactor shielding

    International Nuclear Information System (INIS)

    We can note substantial recent progress in most all aspects of reactor shielding and we anticipate continued gains. Methods and data may tend to assume greater stability in the future although the expected concern with new reactor systems such as those based on fusion make this less likely. Recent problems with neutron streaming may aid in gaining consideration of shielding earlier in the reactor design process. A set of possible challenges in reactor shielding is given below. (orig.)

  2. Radiation shielding for the ITER neutral beam test facility

    International Nuclear Information System (INIS)

    The NB system for the International Thermonuclear Experimental Reactor (ITER) consists of two heating and current drive (H and CD) NB injectors and a diagnostic neutral beam (DNB) injector. The NB accelerates negative deuterium ions with maximum energy of 1 MeV and maximum beam current of 40 A. The ITER (H and CD) NB will be tested in the Neutral Beam Test Facility (NBTF) that will be located in Italy, near Padua. The performance test will be based on different operation phases starting with low energy hydrogen beam. In the initial testing phase for many months the machine will operate with hydrogen only and with deuteron at a reduced intensity suggesting the possibility of hosting the device in a light shielding room/area. In the paper the study performed to evaluate the minimum shielding needed in connection with the different operation phases is shown. The source terms were calculated starting from neutron source characterisation and then assessing article transport in the ITER NB structure with a mathematical model of the components geometry that was implemented into MCNP computer code. The neutron source definition was outlined considering both D-D and D-T neutron production. Shielding was assessed for hydrogen operation only and for 20, 60, 100 and 1000 kV (full energy) deuteron acceleration, accounting for the associated beam current intensity. Related results are presented and discussed in the paper. (author)

  3. Feasibility study for increasing reflector or integrated thermal shielding in reactor

    International Nuclear Information System (INIS)

    In this paper, the structure of pressurized-water reactor (PWR) nuclear power plant reactor (the second generation addition) is analyzed. This paper introduces a conception of an additional reflective layer or the overall heat shielding plate, without changing the basic size of the reactor pressure vessel (RPV). Through designing three programs, the structural changes and the required certification tests caused by each program are analyzed. The structure can be used to the reactor pressure vessel under construction and follow-up by the model test, to ensure that reduce the integral fast flux of the reactor pressure vessel surface effectively, and reduce the pressure vessel materials irradiation damage, for the purpose of extending the service lifetime of the nuclear power plant. (authors)

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

    International Nuclear Information System (INIS)

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

  5. FFTF reactor-characterization program: gamma-ray measurements and shield characterization

    Energy Technology Data Exchange (ETDEWEB)

    Bunch, W.L.; Moore, F.S. Jr.

    1983-02-01

    A series of experiments is to be made during the acceptance test program of the Fast Flux Test Facility (FFTF) to measure the gamma ray characteristics of the Fast Test Reactor (FTR) and to establish the performance characteristics of the reactor shield. These measurements are a part of the FFTF Reactor Characterization Program (RCP). Detailed plans have been developed for these experiments. During the initial phase of the Characteristics Program, which will be carried out in the In-Reactor Thimble (IRT), both active and passive measurement methods will be employed to obtain as much information concerning the gamma ray environment as is practical. More limited active gamma ray measurements also will be made in the Vibration Open Test Assembly (VOTA).

  6. High altitude aircraft flight tests

    Science.gov (United States)

    Helmken, Henry; Emmons, Peter; Homeyer, David

    1996-03-01

    In order to make low earth orbit L-band propagation measurements and test new voice communication concepts, a payload was proposed and accepted for flight aboard the COMET (now METEOR) spacecraft. This Low Earth Orbiting EXperiment payload (LEOEX) was designed and developed by Motorola Inc. and sponsored by the Space Communications Technology Center (SCTC), a NASA Center for the Commercial Development of Space (CCDS) located at Florida Atlantic University. In order to verify the LEOEX payload for satellite operation and obtain some preliminary propagation data, a series of 9 high altitude aircraft (SR-71 and ER-2) flight tests were conducted. These flights took place during a period of 7 months, from October 1993 to April 1994. This paper will summarize the operation of the LEOEX payload and the particular configuration used for these flights. The series of flyby tests were very successful and demonstrated how bi-directional, Time Division Multiple Access (TDMA) voice communication will work in space-to-ground L-band channels. The flight tests also acquired propagation data which will be representative of L-band Low Earth Orbiting (LEO) communication systems. In addition to verifying the LEOEX system operation, it also uncovered and ultimately aided the resolution of several key technical issues associated with the payload.

  7. Design and Test of a Blast Shield for Boeing 737 Overhead Compartment

    Directory of Open Access Journals (Sweden)

    Xinglai Dang

    2006-01-01

    Full Text Available This work demonstrates the feasibility of using a composite blast shield for hardening an overhead bin compartment of a commercial aircraft. If a small amount of explosive escapes detection and is brought onboard and stowed in an overhead bin compartment of a passenger aircraft, the current bins provide no protection against a blast inside the compartment. A blast from the overhead bin will certainly damage the fuselage and likely lead to catastrophic inflight structural failure. The feasibility of using an inner blast shield to harden the overhead bin compartment of a Boeing 737 aircraft to protect the fuselage skin in such a threat scenario has been demonstrated using field tests. The blast shield was constructed with composite material based on the unibody concept. The design was carried out using LS-DYNA finite element model simulations. Material panels were first designed to pass the FAA shock holing and fire tests. The finite element model included the full coupling of the overhead bin with the fuselage structure accounting for all the different structural connections. A large number of iterative simulations were carried out to optimize the fiber stacking sequence and shield thickness to minimize weight and achieve the design criterion. Three designs, the basic, thick, and thin shields, were field-tested using a frontal fuselage section of the Boeing 737–100 aircraft. The basic and thick shields protected the integrity of the fuselage skin with no skin crack. This work provides very encouraging results and useful data for optimization implementation of the blast shield design for hardening overhead compartments against the threat of small explosives.

  8. Test reactor technology

    International Nuclear Information System (INIS)

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

  9. Impedance change measurements of a superconducting shielded-core reactor

    International Nuclear Information System (INIS)

    A device was constructed using a stack of superconducting rings surrounding a ferrite rod, with the assembly inserted in a high turns count solenoid. Superconducting end pieces were also placed at either end of the rod to minimize flux leakage to the ferrite rod. The superconducting rings act as a magnetic shield to the ferrite, effectively eliminating the low reluctance path the ferrite offers. At a specific field the superconductor will be fully penetrated, placing the ferrite in the magnetic circuit and reducing the reactance offered by the solenoidal winding. In this mode of operation the shielded core reactor can be applied as a current limiting device. Results included in this paper, indicate that in the best design achieved leakage to the ferrite core could not be eliminated. The superconducting current induced by this leakage eliminated the low reluctance path of the ferrite by producing a counter-flux in the core exactly opposing the applied field. Shielding currents set up by penetration of the externally applied field were found to be minimal compared to the induced currents caused by leakage flux in the ferrite core

  10. Gas-cooled fast breeder reactor shielding benchmark calculation

    Energy Technology Data Exchange (ETDEWEB)

    Rouse, C.A.; Mathews, D.R.; Koch, P.K.

    1977-01-01

    This report summarizes the results of a shielding benchmark calculation performed by General Atomic (GA) and Oak Ridge National Laboratory (ORNL). The problem analyzed was a neutron-coupled gamma ray transport calculation of the core blanket shield of the 300-MW(e) gas-cooled fast breeder reactor (GCFR). Comparison of the initial GA and ORNL results indicated good agreement for fast fluxes (E greater than 0.9 MeV and E greater than 0.086 MeV) but poor agreement for epithermal and thermal neutron fluxes. Examination of the results revealed that a deficiency in the GA fine-group cross section preparation code was responsible for the differences in the GA and ORNL iron cross sections. Modification of the GA cross sections to include self-shielding was accomplished, and the updated GA benchmark calculation performed with the self-shielded iron cross sections was in excellent agreement with the ORNL results for fast neutron fluxes with E greater than 0.9 MeV and E greater than 0.086 MeV and in good agreement for epithermal and thermal fluxes. The agreement of the gamma heating rates also improved significantly. Thus, it was concluded that the good agreement of the GA and ORNL neutron-coupled gamma ray transport calculation indicates that (1) the methods and cross sections used by both laboratories were compatible and consistent and (2) the use of 24 neutron energy groups and 15 gamma energy groups by GA was adequate compared with the use of 51 neutron energy groups and 25 gamma energy groups by ORNL.

  11. Simulation of nuclear reactor shielding experiment using the Sn code dot 3.5

    International Nuclear Information System (INIS)

    A large scale test facility representing the shielding arrangement of a fast breeder reactor has been simulated on the computer. The transport discrete discrete ordinate dot 3.5 has been used for this purpose. The shielding arrangement is a typical three dimensional geometry. A special strategy is developed to enable the simulation using two dimensional model. The neutron and gamma ray dose rates along the cavity, representing a reactor coolant channel, are compared with published experimental results and theoretical calculations using the Sn code twotran II. The calculation scheme adopted in the present work achieves better agreement with experiment than previous calculation- using the same code-with less computer time. 6 fig

  12. RSMASS: A preliminary reactor/shield mass model for SDI applications

    International Nuclear Information System (INIS)

    A simple mathematical model (RSMASS) has been developed to provide rapid estimates of reactor and shield masses for space-based reactor power systems. Approximations are used rather than correlations or detailed calculations to estimate the reactor fuel mass and the masses of the moderator, structure, reflector, pressure vessel, miscellaneous components, and the reactor shield. The fuel mass is determined either by neutronics limits, specific power limits, or fuel burnup limits - whichever yields the largest mass. RSMASS requires the reactor power and energy, 24 reactor parameters, and 20 shield parameters to be specified. This parametric approach should provide good mass estimates for a very broad range of reactor types. Reactor and shield masses calculated by RSMASS were found to be in good agreement with the masses obtained from detailed calculations

  13. Nondestructive testing of aging aircraft

    International Nuclear Information System (INIS)

    Aircraft fleet in the US military is getting old, averaging over 40 years. These old planes are planned to be used for additional 20-30 years. Some commercial fleets are getting older as well, though not on the same level. Many NDT methods are in practice and new ones being developed. Corrosion and fatigue are the two main sources of damage to aircraft structures and require cost-effective NDT methods to detect and characterize the damage. Current approaches to this difficult task reviewed.

  14. A mathematical model of aircraft for evaluating the effects of shielding structure on aircrew exposure

    International Nuclear Information System (INIS)

    To investigate the influence of the aircraft structures and contents on the exposure of aircrew to the galactic component of cosmic rays, a mathematical model of an aeroplane has been developed. The irradiation of the mathematical model in the cosmic ray environment has been simulated using the Monte Carlo transport code FLUKA. Effective dose and ambient dose-equivalent rates have been determined inside the aircraft at several locations along the fuselage at a typical civil aviation altitude. A significant effect of the shielding of aircraft structures has been observed on the ambient dose-equivalent rates, while the impact on the effective dose rates seems to be minor. Care should be taken in positioning the detectors onboard when the measurements are aimed at validating the codes. (authors)

  15. Modular Electric Propulsion Test Bed Aircraft Project

    Data.gov (United States)

    National Aeronautics and Space Administration — A hybrid electric aircraft simulation system and test bed is proposed to provide a dedicated development environment for the rigorous study and advancement of...

  16. Modular Electric Propulsion Test Bed Aircraft Project

    Data.gov (United States)

    National Aeronautics and Space Administration — An all electric aircraft test bed is proposed to provide a dedicated development environment for the rigorous study and advancement of electrically powered...

  17. Shielding analyses for design of the upgraded JRR-3 research reactor, 1

    International Nuclear Information System (INIS)

    Shielding analyses for design of the upgraded JRR-3 research reactor have been performed. In the report described are the design principles and the overall analytical procedures. In addition, described are the results of shielding analyses of reactor, canal, spent fuel storage pond and so on. (author)

  18. Up-dating of the RA-0 reactor shielding. Gamma and neutron isodoses

    International Nuclear Information System (INIS)

    A comparative analysis of the historical shielding configurations of the RA-0 reactor is performed and the comparison methodology is described. The gamma and neutron dose mapping of the last two stages of the reactor shielding has been carried out and the results are analysed

  19. Beam removal block and shielding resign for the MARS neutron therapy reactor

    International Nuclear Information System (INIS)

    The beam removal block and shielding design for the MARS neutron therapy reactor are described. The requirements to the beams' characteristics, filters, collimator and reactor shielding are formulated. Radiation field levels in medical box are analyzed for beams' different operation conditions. It is stated that the removal block and shutter compositions meet necessary conditions in radiation treatment and emergency evacuation

  20. Engineering design and development for prototype fast breeder reactor (PFBR) shielding experiments at Apsara

    International Nuclear Information System (INIS)

    Prototype fast breeder reactor (PFBR) houses radial shields inside the reactor vessel which consists of many layers of steel and borated graphite within sodium coolant so as to reduce the neutron flux impingement on Intermediate Heat Exchanger (IHX) (also located inside the reactor vessel) to an acceptable limit. In order to cross check the uncertainties involved in theoretical shielding calculations and neutron cross-section data used, IGCAR proposed to carry out various shielding experiments at Apsara reactor to simulate the theoretical shielding configuration. The experiments would also provide bias factors for detailed shielding design calculations. The shielding experiments were planned to be carried out at Apsara shielding corner with reactor core brought to C-dash (C) position. The neutron flux intensity in the shielding corner was inadequate for the purpose of carrying out experiments. Hence the neutron flux level was enhanced to the order of 1010 n/cm2/s by replacing the water column between the core edge and SS liner of Apsara pool on the shielding corner side with an air filled leak tight aluminium box. The fuel loading in the reactor core was also modified to increase neutron flux intensity towards aluminium box. The neutron flux emerging out of the pool into the shielding corner is essentially a thermal neutron spectrum, which was converted into a typical fast reactor leakage neutron spectrum with the help of converter assemblies (CAs ). The converter assemblies were made of depleted uranium and the assemblies were installed on a CA trolley. The CA trolley was positioned outside Apsara pool in the shielding corner. The models of proposed shields manufactured from various shielding materials viz. sodium, steel, borated graphite and boron carbide were installed on a shield model (SM) trolley. The SM trolley was positioned behind CA trolley. Shield models had provisions for irradiating in any foils which were used for measuring the neutron attenuation

  1. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.

  2. RSMASS-D models: An improved method for estimating reactor and shield mass for space reactor applications

    International Nuclear Information System (INIS)

    Three relatively simple mathematical models have been developed to estimate minimum reactor and radiation shield masses for liquid-metal-cooled reactors (LMRs), in-core thermionic fuel element (TFE) reactors, and out-of-core thermionic reactors (OTRs). The approach was based on much of the methodology developed for the Reactor/Shield Mass (RSMASS) model. Like the original RSMASS models, the new RSMASS-derivative (RSMASS-D) models use a combination of simple equations derived from reactor physics and other fundamental considerations, along with tabulations of data from more detailed neutron and gamma transport theory computations. All three models vary basic design parameters within a range specified by the user to achieve a parameter choice that yields a minimum mass for the power level and operational time of interest. The impact of critical mass, fuel damage, and thermal limitations are accounted for to determine the required fuel mass. The effect of thermionic limitations are also taken into account for the thermionic reactor models. All major reactor component masses are estimated, as well as instrumentation and control (I ampersand C), boom, and safety system masses. A new shield model was developed and incorporated into all three reactor concept models. The new shield model is more accurate and simpler to use than the approach used in the original RSMASS model. The estimated reactor and shield masses agree with the mass predictions from separate detailed calculations within 15 percent for all three models

  3. RSMASS-D models: An improved method for estimating reactor and shield mass for space reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, A.C.

    1997-10-01

    Three relatively simple mathematical models have been developed to estimate minimum reactor and radiation shield masses for liquid-metal-cooled reactors (LMRs), in-core thermionic fuel element (TFE) reactors, and out-of-core thermionic reactors (OTRs). The approach was based on much of the methodology developed for the Reactor/Shield Mass (RSMASS) model. Like the original RSMASS models, the new RSMASS-derivative (RSMASS-D) models use a combination of simple equations derived from reactor physics and other fundamental considerations, along with tabulations of data from more detailed neutron and gamma transport theory computations. All three models vary basic design parameters within a range specified by the user to achieve a parameter choice that yields a minimum mass for the power level and operational time of interest. The impact of critical mass, fuel damage, and thermal limitations are accounted for to determine the required fuel mass. The effect of thermionic limitations are also taken into account for the thermionic reactor models. All major reactor component masses are estimated, as well as instrumentation and control (I&C), boom, and safety system masses. A new shield model was developed and incorporated into all three reactor concept models. The new shield model is more accurate and simpler to use than the approach used in the original RSMASS model. The estimated reactor and shield masses agree with the mass predictions from separate detailed calculations within 15 percent for all three models.

  4. Inhibited Shaped Charge Launcher Testing of Spacecraft Shield Designs

    Science.gov (United States)

    Grosch, Donald J.

    1996-01-01

    This report describes a test program in which several orbital debris shield designs were impact tested using the inhibited shaped charge launcher facility at Southwest Research Institute. This facility enables researchers to study the impact of one-gram aluminum projectiles on various shielding designs at velocities above 11 km/s. A total of twenty tests were conducted on targets provided by NASA-MSFC. This report discusses in detail the shield design, the projectile parameters and the test configuration used for each test. A brief discussion of the target damage is provided, as the detailed analysis of the target response will be done by NASA-MSFC.

  5. Applications of point kernel estimates to the fuel conditioning facility shield test program

    International Nuclear Information System (INIS)

    Use of a multigroup point kernel gamma ray attenuation program helped Argonne National Laboratory complete a project that verified the integrity of the shields in the Fuel Conditioning Facility (FCF). Test procedures were developed based on predictions of dose equivalents as functions of source strengths, source-to-detector distances, and thickness of various shield materials. Tables and plots of these data were used to select and position the test sources and to compare as-built shields to their design thicknesses. Part of the program involved a study of the penetration of photons from spent fuel as a function of cooling time. Such information is important to estimate the effectiveness of FCF shields on mixed fission product sources from various reactors

  6. Engineering and Fabrication Considerations for Cost-Effective Space Reactor Shield Development

    International Nuclear Information System (INIS)

    Investment in developing nuclear power for space missions cannot be made on the basis of a single mission. Current efforts in the design and fabrication of the reactor module, including the reactor shield, must be cost-effective and take into account scalability and fabricability for planned and future missions. Engineering considerations for the shield need to accommodate passive thermal management, varying radiation levels and effects, and structural/mechanical issues. Considering these challenges, design principles and cost drivers specific to the engineering and fabrication of the reactor shield are presented that contribute to lower recurring mission costs

  7. Fault current limiter-predominantly resistive behavior of a BSCCO shielded-core reactor

    International Nuclear Information System (INIS)

    Tests were conducted to determine the electrical and magnetic characteristics of a superconductor shielded core reactor (SSCR). The results show that a closed-core SSCR is predominantly a resistive device and an open-core SSCR is a hybrid resistive/inductive device. The open-core SSCR appears to dissipate less than the closed-core SSCR. However, the impedance of the open-core SSCR is less than that of the closed-core SSCR. Magnetic and thermal diffusion are believed to be the mechanism that facilitates the penetration of the superconductor tube under fault conditions

  8. Shielded canister transporter equipment acceptance test operations

    International Nuclear Information System (INIS)

    The defense waste processing facility (DWPF) processes high level waste at the Savannah River Plant (SRP) by vitrifying the waste and placing it in stainless stell canisters for long term storage. The shielded canister transporter (SCT) is a diesel powered mobile rubber tired self-propelled vehicle which transports the canisters from the DWPF processing facility to the on-site waste storage building. The SCT has a system of automatic programmable logic controls (PLC) which provides operational handling control with a shielded transfer cask and associated canister positional equipment

  9. Experimental Tests of Neutron Shielding for the ATLAS Forward Region

    CERN Document Server

    Pospísil, S; Cechák, T; Cermák, P; Jakubek, J; Kluson, J; Konícek, J; Kubasta, J; Linhart, V; Sinor, M; Leroy, C; Dolezal, Z; Leitner, R; Lukianov, G A; Soustruznik, K; Lokajícek, M; Némécek, S; Pálla, G; Sodomka, J

    1999-01-01

    Experimental tests devoted to the optimization of the neutron shielding for the ATLAS forward region were performed at the CERN-PS with a 4 GeV/c proton beam. Spectra of fast neutrons, slow neutrons and gamma rays escaping a block of iron (40$\\times$40$\\times$80 cm$^3$) shielded with different types of neutron and gamma shields (pure polyethylene - PE, borated polyethylene - BPE, lithium filled polyethylene - LiPE, lead, iron) were measured by means of plastic scintillators, a Bonner spectrometer, a HPGe detector and a slow neutron detector. Effectiveness of different types of shielding agaisnt neutrons and $\\gamma$-rays were compared. The idea of a segmented outer layer shielding (iron, BPE, iron, LiPE) for the ATLAS Forward Region was also tested.

  10. Homogeneity test on heavy concrete shield wall for ACP facility

    International Nuclear Information System (INIS)

    The hot cell facility for research activities related to the electrolytic reduction of spent fuel, which is designed to permit a safe handling of radioactive materials up to 1,385 TBq, is scheduled to be constructed in 2005. The design features of the radiation safety are reviewed for the shield wall, rear door, shielding window, penetrations, toboggan, and the storage vault. The calculations by QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts and the gamma scanning test is described to examine the integrity of the shielding structure for the hot cell. The gamma scanning test is especially good at detecting any void and cracks in a heavy concrete wall and finding crevices between the wall and the devices frames. The shielding effectiveness and homogeneity of the hot cell wall, shield window, rear door etc., shall be measured by reading the activity level of the radiation

  11. Electronic materials testing in commercial aircraft engines

    Science.gov (United States)

    Brand, Dieter

    A device for the electronic testing of materials used in commercial aircraft engines is described. The instrument can be used for ferromagnetic, ferrimagnetic, and nonferromagnetic metallic materials, and it functions either optically or acoustically. The design of the device is described and technical data are given. The device operates under the principle of controlled self-inductivity. Its mode of operation is described.

  12. Production of a datolite-based heavy concrete for shielding nuclear reactors and megavoltage radiotherapy rooms

    International Nuclear Information System (INIS)

    Biological shielding of nuclear reactors has always been a great concern and decreasing the complexity and expense of these installations is of great interest. In this study, we used datolite and galena minerals for production of a high performance heavy concrete. Materials and Methods: Datolite and galena minerals which can be found in many parts of Iran were used in the concrete mix design. To measure the gamma radiation attenuation of the Datolite and galena concrete samples, they were exposed to both narrow and wide beams of gamma rays emitted from a cobalt-60 radiotherapy unit. An Am-Be neutron source was used for assessing the shielding properties of the samples against neutrons. To test the compression strengths, both types of concrete mixes (Datolite and galena and ordinary concrete) were investigated. Results: The concrete samples had a density of 4420-4650 kg/m3 compared to that of ordinary concrete (2300-2500 kg/m3) or barite high density concrete (up to 3500 kg/m3). The measured half value layer thickness of the Datolite and galena concrete samples for cobalt-60 gamma rays was much less than that of ordinary concrete (2.56 cm compared to 6.0 cm). Furthermore, the galena concrete samples had a significantly higher compressive strength as well as 20% more neutron absorption. Conclusion: The Datolite and galena concrete samples showed good shielding/engineering properties in comparison with other reported samples made, using high-density materials other than depleted uranium. It is also more economic than the high-density concretes. Datolite and galena concrete may be a suitable option for shielding nuclear reactors and megavoltage radiotherapy rooms.

  13. Shielding analyses for design of the upgraded JRR-3 research reactor, 2

    International Nuclear Information System (INIS)

    Shielding analyses of neutron beam holes have been presented for the shield design of the upgraded JRR-3 research reactor. Description is given about the calculational procedures and results for the standard beam hole, the beam hole for neutron radiography and the guide tunnels. The streaming analyses are made by using the MORSE-CG and DOT 3.5 codes. (author)

  14. Reactor Simulator Testing

    Science.gov (United States)

    Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise J.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  15. Structural Integrity Assessment of Reactor Containment Subjected to Aircraft Crash

    International Nuclear Information System (INIS)

    When an accident occurs at the NPP, containment building which acts as the last barrier should be assessed and analyzed structural integrity by internal loading or external loading. On many occasions that can occur in the containment internal such as LOCA(Loss Of Coolant Accident) are already reflected to design. Likewise, there are several kinds of accidents that may occur from the outside of containment such as earthquakes, hurricanes and strong wind. However, aircraft crash that at outside of containment is not reflected yet in domestic because NPP sites have been selected based on the probabilistic method. After intentional aircraft crash such as World Trade Center and Pentagon accident in US, social awareness for safety of infrastructure like NPP was raised world widely and it is time for assessment of aircraft crash in domestic. The object of this paper is assessment of reactor containment subjected to aircraft crash by FEM(Finite Element Method). In this paper, assessment of structural integrity of containment building subjected to certain aircraft crash was carried out. Verification of structure integrity of containment by intentional severe accident. Maximum stress 61.21MPa of horizontal shell crash does not penetrate containment. Research for more realistic results needed by steel reinforced concrete model

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

    International Nuclear Information System (INIS)

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

  17. Design of the shield door and transporter for the Culham Conceptual Tokamak Reactor Mark II

    International Nuclear Information System (INIS)

    In the Culham Conceptual Tokamak Reactor MK II access to the interior for blanket maintenance is through large openings in the fixed shield structure closed by removable shield doors when the reactor is operational. This report describes the design of the 200 tonne doors and the associated special-purpose remote operating transporter manipulator. The design, which has not been optimised, generally uses available commercial equipment and state-of-the-art techniques. (U.K.)

  18. Magnetic shielding tests for MFTF-B neutral beamlines

    International Nuclear Information System (INIS)

    A test program to determine the effectiveness of various magnetic shielding designs for MFTF-B beamlines was established at Lawrence Livermore National Laboratory (LLNL). The proposed one-tenth-scale shielding-design models were tested in a uniform field produced by a Helmholtz coil pair. A similar technique was used for the MFTF source-injector assemblies, and the model test results were confirmed during the Technology Demonstration in 1982. The results of these tests on shielding designs for MFTF-B had an impact on the beamline design for MFTF-B. The iron-core magnet and finger assembly originally proposed were replaced by a simple, air-core, race-track-coil, bending magnet. Only the source injector needs to be magnetically shielded from the fields of approximately 400 gauss

  19. A Cylindrical Shielding Design Concept for the Prototype Gen-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    In the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), a metal fueled, blanket-free, pool type SFR concept is adopted to acquire the inherent safety characteristics and high proliferation-resistance. In the pool type fast reactor, the intermediate heat exchangers (IHXs), which transfer heat from the primary sodium pool to a secondary sodium loop, are placed inside of the reactor vessel. Hence, secondary sodium passing the IHXs can be radioactivated by a 23Na(n,g)24Na reaction, and radioactivated secondary sodium causes a significant dose in the Steam Generator Building (SGB). Therefore, a typical core of a pool type fast reactor is usually surrounded by a massive quantity of shields. In addition, the blanket composed of depleted uranium plays a role as superior shielding material; a significant increase in shields is required in the blanket-free pool type SFR. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR. In a conventional shielding design, massive axial shields are required to prevent irradiation of secondary sodium passing IHXs and they should be replaced according to the subassembly replacement in spite of negligible depletion of the shielding material. The proposed shielding design concept minimizes the quantity of shields without their replacement. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR such as a PGSFR. The proposed design concept satisfied the dose limit in the steam generator building successfully without introducing a large quantity of B4C shielding inside the subassembly

  20. Gamma-ray shielding design and performance test of WASTEF

    International Nuclear Information System (INIS)

    The Waste Safety Testing Facility (WASTEF) was planned in 1978 to test the safety performance of HLW vitrified forms under the simulated conditions of long term storage and disposal, and completed in August 1981. The designed feature of the facility is to treat the vitrified forms contain actual high-level wastes of 5 x 104 Ci in maximum with 5 units of concrete shilded hot cells (3 units : Bate-Gamma cells, 2 units : Alpha-Gamma cells) and one units of Alpha-Gamma lead shielded cell, and to store radioactivity of 106 Ci in maximum. The safety performance of this facility is fundamentally maintained with confinement of radioactivity and shielding of the radiation. This report describes the method of gamma-ray shielding design, evaluation of the shielding test performed by using sealded gamma-ray sources(Co-60). (author)

  1. The self-shielding factors in activation detectors used in the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    This work aims to obtain self-shielding factors (G) to several activation foils, such as gold, cobalt, scandium, magnesium, uranium, thorium and indium foils used at measurements of the neutron spectrum energy in the IPEN/MB-01 reactor core. The knowledge of the self-shielding factors allows obtaining of precise nuclear reaction rates without the effects of neutron flux depression inside of activation foils. This study is carried out in two parts. First are determined the self-shielding factors for bare foils (without cadmium covered) and after the self-shielding factors to foils with cadmium covered. (author)

  2. Radiation shielding for the Fermilab Vertical Cavity Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Ginsburg, Camille; Rakhno, Igor; /Fermilab

    2010-03-01

    The results of radiation shielding studies for the vertical test cryostat VTS1 at Fermilab performed with the codes FISHPACT and MARS15 are presented and discussed. The analysis is focused on operations with two RF cavities in the cryostat. The vertical cavity test facility (VCTF) for superconducting RF cavities in Industrial Building 1 at Fermilab has been in operation since 2007. The facility currently consists of a single vertical test cryostat VTS1. Radiation shielding for VTS1 was designed for operations with single 9-cell 1.3 GHz cavities, and the shielding calculations were performed using a simplified model of field emission as the radiation source. The operations are proposed to be extended in such a way that two RF cavities will be in VTS1 at a time, one above the other, with tests for each cavity performed sequentially. In such a case the radiation emitted during the tests from the lower cavity can, in part, bypass the initially designed shielding which can lead to a higher dose in the building. Space for additional shielding, either internal or external to VTS1, is limited. Therefore, a re-evaluation of the radiation shielding was performed. An essential part of the present analysis is in using realistic models for cavity geometry and spatial, angular and energy distributions of field-emitted electrons inside the cavities. The calculations were performed with the computer codes FISHPACT and MARS15.

  3. 76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-01-20

    ... COMMISSION 10 CFR Part 52 RIN 3150-AI84 U.S. Advanced Boiling Water Reactor Aircraft Impact Design... the U.S. Advanced Boiling Water Reactor (ABWR) standard plant design to comply with the NRC's aircraft...--Design Certification Rule for the U.S. Advanced Boiling Water Reactor IV. Section-by-Section Analysis...

  4. Decommissioning of the ASTRA research reactor: Dismantling of the biological shield

    Directory of Open Access Journals (Sweden)

    Meyer Franz

    2006-01-01

    Full Text Available The paper describes the dismantling of the inactive and activated areas of the biological shield of the ASTRA research reactor at the Austrian Research Center in Seibersdorf. The calculation of the parameters determining the activated areas at the shield (reference nuclide, nuclide vector in the barite concrete and horizontal and vertical reduction behaviors of activity concentration and the activation profiles within the biological shield for unrestricted release, release restricted to permanent deposit and radioactive waste are presented. Considerations of located activation anomalies in the shield, e.g. in the vicinities of the beam-tubes, were made according to the reactor's operational history. Finally, an overview of the materials removed from the biological shield is given.

  5. Decommissioning of the ASTRA research reactor - dismantling of the biological shield

    International Nuclear Information System (INIS)

    The paper describes the dismantling of the inactive and activated areas of the biological shield of the ASTRA research reactor at the Austrian Research Center in Seibersdorf. The calculation of the parameters determining the activated areas at the shield (reference nuclide, nuclide vector in the barite concrete and horizontal and vertical reduction behaviors of activity concentration) and the activation profiles within the biological shield for unrestricted release, release restricted to permanent deposit and radioactive waste are presented. Considerations of located activation anomalies in the shield, e. g. in the vicinities of the beam-tubes, were made according to the reactor's operational history. Finally, an overview of the materials removed from the biological shield is given. (author)

  6. Reactor Simulator Testing Overview

    Science.gov (United States)

    Schoenfeld, Michael P.

    2013-01-01

    Test Objectives Summary: a) Verify operation of the core simulator, the instrumentation & control system, and the ground support gas and vacuum test equipment. b) Examine cooling & heat regeneration performance of the cold trap purification. c) Test the ALIP pump at voltages beyond 120V to see if the targeted mass flow rate of 1.75 kg/s can be obtained in the RxSim. Testing Highlights: a) Gas and vacuum ground support test equipment performed effectively for operations (NaK fill, loop pressurization, and NaK drain). b) Instrumentation & Control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings and ramped within prescribed constraints. It effectively interacted with reactor simulator control model and defaulted back to temperature control mode if the transient fluctuations didn't dampen. c) Cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the minimum temperature indicating the design provided some heat regeneration. d) ALIP produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  7. Research Reactor Spent Fuel Transfer/Storage Cask with Application to TRIGA Fuel - Designed Cask Shielding Independent Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Margeanu, C.A.; Iorgulis, C. [Reactor Physics, Nuclear Fuel Performances and Nuclear Safety Department, Institute for Nuclear Research Pitesti, Campului Street, no.1, 115400 Mioveni (Romania); Ciocanescu, M. [Institute for Nuclear Research Pitesti, Campului Street, no.1, 115400 Mioveni (Romania); Prava, M. [Design Department, Institute for Nuclear Research Pitesti, Campului Str, no.1, 115400 Mioveni (Romania); Margeanu, S. [Radiation Protection Department, Institute for Nuclear Research Pitesti, Campului Street, no.1, 115400 Mioveni (Romania)

    2011-07-01

    Institute for Nuclear Research (INR) Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies (dual-core concept involves independent operation of TRIGA 14 MW Steady-State Reactor and TRIGA Annular-Core Pulsing Reactor at each end of a large pool). In May 2006, TRIGA 14 MW SSR core was fully converted to Low Enriched Uranium (LEU 20 wt% {sup 235}U) fuel, according to Reduced Enrichment for Research and Test Reactors agreements and current worldwide non-proliferation efforts. Paper presents a shielding independent evaluation applied to designed transfer/ storage cask for TRIGA INR spent fuel, a mandatory step in preparation of the documentation required for spent fuel transfer/storage cask authorisation process. Fuel elements irradiation was modelled by assuming constant power for entire residence time inside reactor core, for 14 MW reactor operation power and two different scenarios characteristic for accident calculations according to TRIGA 14 MW SSR safety report and reactor operation experience. The discharged spent LEU fuel was cooled down for 2 and 5 years, respectively. Source term assessment and spent fuel characteristic parameters estimation were done by means of ORIGEN-S burn-up code (included in Oak Ridge National Laboratory's SCALE6 package) with specific cross-sections libraries, updating data for each burn-up step. For the transfer/storage cask shielding analysis, two different cases have been considered, the main difference residing in TRIGA fuel elements loading. The radiation dose rates to the transfer/storage cask wall and in air at different distances from the cask have been estimated by means of MAVRIC/Monaco shielding 3D Monte Carlo code included in ORNL's SCALE6 package. (author)

  8. Research Reactor Spent Fuel Transfer/Storage Cask with Application to TRIGA Fuel - Designed Cask Shielding Independent Evaluation

    International Nuclear Information System (INIS)

    Institute for Nuclear Research (INR) Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies (dual-core concept involves independent operation of TRIGA 14 MW Steady-State Reactor and TRIGA Annular-Core Pulsing Reactor at each end of a large pool). In May 2006, TRIGA 14 MW SSR core was fully converted to Low Enriched Uranium (LEU 20 wt% 235U) fuel, according to Reduced Enrichment for Research and Test Reactors agreements and current worldwide non-proliferation efforts. Paper presents a shielding independent evaluation applied to designed transfer/ storage cask for TRIGA INR spent fuel, a mandatory step in preparation of the documentation required for spent fuel transfer/storage cask authorisation process. Fuel elements irradiation was modelled by assuming constant power for entire residence time inside reactor core, for 14 MW reactor operation power and two different scenarios characteristic for accident calculations according to TRIGA 14 MW SSR safety report and reactor operation experience. The discharged spent LEU fuel was cooled down for 2 and 5 years, respectively. Source term assessment and spent fuel characteristic parameters estimation were done by means of ORIGEN-S burn-up code (included in Oak Ridge National Laboratory's SCALE6 package) with specific cross-sections libraries, updating data for each burn-up step. For the transfer/storage cask shielding analysis, two different cases have been considered, the main difference residing in TRIGA fuel elements loading. The radiation dose rates to the transfer/storage cask wall and in air at different distances from the cask have been estimated by means of MAVRIC/Monaco shielding 3D Monte Carlo code included in ORNL's SCALE6 package. (author)

  9. Investigations of cracks in the shielding concrete of a TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Cracks in the reactor shielding concrete of the TRIGA Mark II reactor, Vienna, caused an experimental and theoretical program to investigate the crack reason. After the investigation of the mechanical concrete data, the crack motion was measured as a function of various environmental temperatures. The temperature stress in the concrete was calculated analytically and with the finite-elements method and good accordance with the actual crack distribution was found. Finally some possibilities to avoid concrete cracks in future research reactor shielding construction are outlined. (orig.)

  10. 78 FR 12259 - Unmanned Aircraft System Test Site Program

    Science.gov (United States)

    2013-02-22

    ... of provisions pertaining to integration of unmanned aircraft systems (UAS) into the National Airspace... Federal Aviation Administration 14 CFR Part 91 Unmanned Aircraft System Test Site Program AGENCY: Federal... be levied on the Unmanned Aircraft Systems Test Site operators, but prior to the close of the...

  11. Attenuation of reactor thermal neutrons in a bulk shield of ordinary concrete

    International Nuclear Information System (INIS)

    This work is concerned with the study of the distribution attenuation of doses of thermal neutrons emitted directly from the core of research reactor in ordinary concrete shield. In practice it is not possible to identify the reactor thermal neutrons in the emitted continuos neutron spectrum. Therefore, measurement was carried out by using a direct and cadmium filtered beam of reactor neutrons. All measurements were performed using Li2B4O7:Mn thermoluminescent dosimeters. The data obtained were analyzed and the dose distributions of reactor thermal neutrons were evaluated. A group of isodose curves constructed which give directly the shape and thickness of the shield required to attenuate the intensity of doses of reactor thermal neutrons to specific values. In addition, the thermal neutron relaxation lengths in ordinary concrete were derived for disc-collimated beam and infinite plane mono-directional sources

  12. Radiation shielding considerations for the repair and maintenance of a swimming pool-type tokamak reactor

    International Nuclear Information System (INIS)

    The radiation shielding relevant to the repair and maintenance of a swimming pool-type tokamak reactor is considered. The dose rate during the reactor operation can be made low enough for personnel access into the reactor room if a 2m thick water layer is installed above the magnet cryostat. The dose rate 24 h after shutdown is such that the human access is allowed above the magnet cryostat. Sufficient water layer thickness is provided in the inboard space for the operation of automatic welder/cutter while retaining the magnet shielding capability. Some forced cooling is required for the decay heat removal in the first wall. The penetration shield thickness around the neutral beam injector port is estimated to be barely sufficient in terms of the magnet radiation damage. (orig.)

  13. Simulator for materials testing reactors

    International Nuclear Information System (INIS)

    A real-time simulator for both reactor and irradiation facilities of a materials testing reactor, “Simulator of Materials Testing Reactors”, was developed for understanding reactor behavior and operational training in order to utilize it for nuclear human resource development and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR (Japan Materials Testing Reactor), and it simulates operation, irradiation tests and various kinds of anticipated operational transients and accident conditions caused by the reactor and irradiation facilities. The development of the simulator was sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. This report summarizes the simulation components, hardware specification and operation procedure of the simulator. (author)

  14. Temperature distribution due to the heat generation in nuclear reactor shielding

    International Nuclear Information System (INIS)

    A study is performed for calculating nuclear heating due to the interaction of neutrons and gamma-rays with matter. Modifications were implemented in the ANISN and DOT 3.5 codes, that solve the transport equation using the discrete ordinate method, in one two-dimensions respectively, to include nuclear heating calculations in these codes. In order to determine the temperature distribution, using the finite difference method, a numerical model was developed for solving the heat conduction equation in one-dimension, in plane, cylindrical and spherical geometries, and in two-dimensions, X-Y and R-Z geometries. Based on these models, computer programs were developed for calculating the temperature distribution. Tests and applications of the implemented modifications were performed in problems of nuclear heating and temperature distribution due to radiation energy deposition in fission and fusion reactor shields. (Author)

  15. Discrete nodal integral transport-theory method for multidimensional reactor physics and shielding calculations

    International Nuclear Information System (INIS)

    A coarse-mesh discrete nodal integral transport theory method has been developed for the efficient numerical solution of multidimensional transport problems of interest in reactor physics and shielding applications. The method, which is the discrete transport theory analogue and logical extension of the nodal Green's function method previously developed for multidimensional neutron diffusion problems, utilizes the same transverse integration procedure to reduce the multidimensional equations to coupled one-dimensional equations. This is followed by the conversion of the differential equations to local, one-dimensional, in-node integral equations by integrating back along neutron flight paths. One-dimensional and two-dimensional transport theory test problems have been systematically studied to verify the superior computational efficiency of the new method

  16. Discrete nodal integral transport-theory method for multidimensional reactor physics and shielding calculations

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, R.D.; Dorning, J.J.

    1980-01-01

    A coarse-mesh discrete nodal integral transport theory method has been developed for the efficient numerical solution of multidimensional transport problems of interest in reactor physics and shielding applications. The method, which is the discrete transport theory analogue and logical extension of the nodal Green's function method previously developed for multidimensional neutron diffusion problems, utilizes the same transverse integration procedure to reduce the multidimensional equations to coupled one-dimensional equations. This is followed by the conversion of the differential equations to local, one-dimensional, in-node integral equations by integrating back along neutron flight paths. One-dimensional and two-dimensional transport theory test problems have been systematically studied to verify the superior computational efficiency of the new method.

  17. Radiation embrittlement of the neutron shield tank from the Shippingport reactor

    International Nuclear Information System (INIS)

    The irradiation embrittlement of neutron shield tank (NST) material (A212 Grade B steel) from the Shippingport reactor has been characterized. Irradiation increases the Charpy transition temperature (CTT) by 23--28 degrees C (41--50 degrees F) and decreases the upper-shelf energy. The shift in CTT is not as severe as that observed in high-flux isotope reactor (HFIR) surveillance specimens. However, the actual value of the CTT is higher than that for the HFIR data. The increase in yield stress is 51 MPa (7.4 ksi), which is comparable to HFIR data. The NST material is weaker in the transverse orientation than in the longitudinal orientation. Some effects of position across the thickness of the wall are also observed; the CTT shift is slightly greater for specimens from the inner region of the wall. Annealing studies indicate complete recovery from embrittlement after 1 h at 400 degrees C (752 degrees F). Although the weld metal is significantly tougher than the base metal, the shifts in CTT are comparable. The shifts in CTT for the Shippingport NST are consistent with the test and Army reactor data for irradiations at 8 n/cm2·s and at the low operating temperatures of the Shippingport NST, i.e., 55 degrees C (130 degrees F). This suggests that the accelerated embrittlement of HFIR surveillance samples is most likely due to the relatively higher proportion of thermal neutrons in the HFIR spectrum compared to that for the test reactors. 28 refs., 25 figs

  18. Design for shielding kit for local irradiation in mice and test of its shielding effect

    International Nuclear Information System (INIS)

    In order to fulfill the immediate requirement for experimental studies on biological effect of low dose radiation. A shielding kit for local irradiation in mice is designed. Its advantages are: (1) several mice can be irradiated at the same time; (2) the radiation condition is identical; (3) it is easy to control and perform; (4) during irradiation, the animals don't need any special treatments such as anaesthesia. It was proved by TLD test, that absorbed dose in areas of spleen and head in mice after shielding has decreased to 0.85% and 0.5% of the original dose in the center of radiation field respectively. The results suggest that the kit was able to satisfy the needs of the experimental studies on radiation biology

  19. Aircraft fuel tank slosh and vibration test

    Science.gov (United States)

    Zimmermann, H.

    1981-12-01

    A dynamic qualification test for a subsonic and a supersonic external drop tank for a European fighter is presented. The test rig and the specimens are described and the measuring results are discussed. It is shown that for the supersonic tank as well as for the subsonic tank a certain slosh angle an eigenfrequency of the rig increases the amplitudes at the excitation position and the accelerations on the tank. For the subsonic tank it seems that an eigenfrequency is excited for the nose down position of the tank. The qualification requirements are examined. It is proposed that instead of using an arbitrary vibration amplitude and frequency for excitation, frequency ranges and amplitudes which are averaged out of flight measurements at the tank attachment points on the aircraft be used and that the demand for a certain input amplitude at the top of the attachment bulkheads and an output amplitude at the bottom of the attachment bulkheads be deleted.

  20. Shielding analyses for the reactor pressure vessel of SMART-P

    International Nuclear Information System (INIS)

    In Korea, an advanced reactor system of 330 MWt power called SMART is being developed by KAERI to supply energy for seawater desalination as well as electricity generation. As the basic design of SMART has recently been completed, a national project to construct a scaled-down reactor of the SMART system called 'SMART-P' for demonstrating the safety and performance of the SMART system is underway. The SMART-P is a small-sized advanced integral PWR that produces a thermal energy of 65.5 MWt under full power operating conditions. A shielding design for the SMART-P reactor assembly is established by two-dimensional discrete ordinates radiation transport analyses. The DORT two-dimensional discrete ordinates transport code is used to evaluate the shielding designs for the reactor assembly of the SMART-P. The SMART-P reactor assembly is divided into three separate calculation models, each of which is modeled by R-Z geometry. The results indicate that the maximum neutron fluence at the bottom of the reactor vessel is 2.0x1018 n/cm2 and that on the radial surface of the reactor vessel is 4.0x1017 n/cm2. These results meet the requirements, 1.0x1020 n/cm2 and the integrity of the SMART-P reactor vessel during the lifetime of the reactor, is confirmed. (author)

  1. On an optimized neutron shielding for an advanced molten salt fast reactor design

    International Nuclear Information System (INIS)

    The molten salt reactor technology has gained renewed interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner core vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all core internal structures. On the basis of this new geometry a model for neutron physics calculation is presented and applied for a shielding optimization. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system has to be significantly increased and will finally be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem. (author)

  2. Resonance self-shielding in the blanket of a hybrid reactor

    International Nuclear Information System (INIS)

    Three sets of energy group cross sections were obtained using various approximations for resonance self shielding. The three models used in obtaining the cross sections were: (a) infinitely dilute model, (b) homogeneous-medium resonance self shielding, and (c) heterogeneous-medium resonance self shielding. The effects on the blanket performance of fusion--fission hybrid reactors, and in particular, on the performance of the current reference Westinghouse Demonstration Tokamak Hybrid Reactor blanket, were compared and analyzed for a variety of fuel-coolant combinations. It has been concluded that (1) the infinitely dilute cross sections can be used to produce preliminary crude estimates for beginning-of-life (BOL) only, (2) the resonance absorber finite dilution should be considered for BOL, poorly moderated blankets and well moderated blankets with low fissile material content situations, and (3) the spacial details should be considered in high fissile content, well moderated blanket situations

  3. Aircraft

    Science.gov (United States)

    Hibbs, Bart D.; Lissaman, Peter B. S.; Morgan, Walter R.; Radkey, Robert L.

    1998-01-01

    This disclosure provides a solar rechargeable aircraft that is inexpensive to produce, is steerable, and can remain airborne almost indefinitely. The preferred aircraft is a span-loaded flying wing, having no fuselage or rudder. Travelling at relatively slow speeds, and having a two-hundred foot wingspan that mounts photovoltaic cells on most all of the wing's top surface, the aircraft uses only differential thrust of its eight propellers to turn. Each of five sections of the wing has one or more engines and photovoltaic arrays, and produces its own lift independent of the other sections, to avoid loading them. Five two-sided photovoltaic arrays, in all, are mounted on the wing, and receive photovoltaic energy both incident on top of the wing, and which is incident also from below, through a bottom, transparent surface. The aircraft is capable of a top speed of about ninety miles per hour, which enables the aircraft to attain and can continuously maintain altitudes of up to sixty-five thousand feet. Regenerative fuel cells in the wing store excess electricity for use at night, such that the aircraft can sustain its elevation indefinitely. A main spar of the wing doubles as a pressure vessel that houses hydrogen and oxygen gasses for use in the regenerative fuel cell. The aircraft has a wide variety of applications, which include weather monitoring and atmospheric testing, communications, surveillance, and other applications as well.

  4. A decade of radiological and shielding experience at the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    The Fast Flux Test Facility (FFTF) was designed to permit irradiation testing of fuels and materials to support the commercial development of liquid-metal-cooled fast reactors (LMRs). A secondary objective was to gain experience in the design, construction, and operation of a relatively large LMR. The radiological experience gained from the operation of the facility as it applies to the area of radiation protection and shielding is presented. Experience from 8 yr of FFTF operation has demonstrated that radiological safety can be achieved in large LMRs. Layout of plant equipment in shielded compartments, careful operational planning, and adherence to procedures have combined to minimize personnel doses at FFTF and the release of radioactivity to the environment. The experience derived form the design, construction, and operation of FFTF should be of inestimable value in supporting future LMR development

  5. Advanced Neutron Source Reactor zoning, shielding, and radiological optimization guide

    International Nuclear Information System (INIS)

    In the design of major nuclear facilities, it is important to protect both humans and equipment excessive radiation dose. Past experience has shown that it is very effective to apply dose reduction principles early in the design of a nuclear facility both to specific design features and to the manner of operation of the facility, where they can aid in making the facility more efficient and cost-effective. Since the appropriate choice of radiological controls and practices varies according to the case, each area of the facility must be analyzed for its radiological impact, both by itself and in interactions with other areas. For the Advanced Neutron Source (ANS) project, a large relational database will be used to collect facility information by system and relate it to areas. The database will also hold the facility dose and shielding information as it is produced during the design process. This report details how the ANS zoning scheme was established and how the calculation of doses and shielding are to be done

  6. A decade of radiological and shielding experience at the fast flux test facility

    International Nuclear Information System (INIS)

    This paper reports on which the Fast Flux Test Facility (FFTF) which has operated for almost a decade after first going critical during February 1980. Based on about 2,000 effective full-power days of operation, it is concluded that radiological safety can be achieved in large liquid metal-cooled fast reactors. The collective dose equivalents received by operating personnel are significantly lower than those received at commercial light water reactors. No major contamination problems have been encountered in operating and maintaining the plant, and release of radioactive materials to the environment has been well below acceptable limits. All shields have performed satisfactorily and in agreement with design calculations. The experience derived from the design, construction, and operation of the FFTF should be of inestimable value in supporting future development of liquid metal reactors

  7. Long-lived activation products in TRIGA Mark II research reactor concrete shield: calculation and experiment

    Science.gov (United States)

    Žagar, Tomaž; Božič, Matjaž; Ravnik, Matjaž

    2004-12-01

    In this paper, a process of long-lived activity determination in research reactor concrete shielding is presented. The described process is a combination of experiment and calculations. Samples of original heavy reactor concrete containing mineral barite were irradiated inside the reactor shielding to measure its long-lived induced radioactivity. The most active long-lived (γ emitting) radioactive nuclides in the concrete were found to be 133Ba, 60Co and 152Eu. Neutron flux, activation rates and concrete activity were calculated for actual shield geometry for different irradiation and cooling times using TORT and ORIGEN codes. Experimental results of flux and activity measurements showed good agreement with the results of calculations. Volume of activated concrete waste after reactor decommissioning was estimated for particular case of Jožef Stefan Institute TRIGA reactor. It was observed that the clearance levels of some important long-lived isotopes typical for barite concrete (e.g. 133Ba, 41Ca) are not included in the IAEA and EU basic safety standards.

  8. The IBR-2 test reactor

    International Nuclear Information System (INIS)

    Major design criteria, specifications and potential fields of application of the IBR-2 pulsed test reactor (now under construction in Dubna, USSR) are described. The pulsed power bursts will be due to fast periodic reactivity changes by a rotating reflector. The frequency of approximately 100 μs pulsed may be 5, 12.5 or 50 Hz. The IBR-2 reactor will be mostly profitable for slow neutron experiments when investigating solids, nuclei or neutrons themselves using spectroscopic methods. Due to the high peak flux of thermal neutrons (1016-1017 n/cm2xs) the reactor will be superior (for the sort of experiments) to the currently operating SM-2 and HFR high flux steady-state test reactors for many times

  9. Numerical investigation of particle deposition inside aero-shielded solar cyclone reactor: A promising solution for reactor clogging

    International Nuclear Information System (INIS)

    Highlights: ► CFD analysis predicting particle deposition inside a solar reactor is given. ► “Aero-shielded solar cyclone reactor” is a promising solution to reactor clogging. ► Particle deposition on reactor walls can be eliminated via laminar “Wall screening”. ► Argon is the best window and wall screening gas to protect reactor from deposition. ► Conical section adjustment avoids deposition by wider flow and particle distribution. -- Abstract: Solar cracking of methane is considered to be an attractive option due to its CO2 free hydrogen production process. Carbon particle deposition on the reactor window, walls and exit is a major obstacle to achieve continuous operation of methane cracking solar reactors. As a solution to this problem a novel “aero-shielded solar cyclone reactor” was created. In this present study the prediction of particle deposition at various locations for the aero-shielded reactor is numerically investigated by a Lagrangian particle dispersion model. A detailed three dimensional computational fluid dynamic (CFD) analysis for carbon deposition at the reactor window, walls and exit is presented using a Discrete Phase Model (DPM). The flow field is based on a RNG k–ε model and species transport with methane as the main flow and argon/ hydrogen as window and wall screening fluid. Flow behavior and particle deposition have been observed with the variation of main flow rates from 10–20 L/min and with carbon particle mass flow rate of 7 × 10−6 and 1.75 × 10−5 kg/s. In this study the window and wall screening flow rates have been considered to be 1 L/min and 10 L/min by employing either argon or hydrogen. Also, to study the effect of particle size simulations have also been carried out (i) with a variation of particle diameter with a size distribution of 0.5–234 μm and (ii) by taking 40 μm mono sized particles which is the mean value for the considered size distribution. Results show that by appropriately

  10. Radiation shielding calculations for Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Radiation shielding calculations have been performed for PARR-1 LEU (low enriched uranium) core at 10 MW operation. The radiation include fast neutrons, fission gammas, fission products decay gammas and activation products decay gammas. Dose rates have been calculated at various locations, including pool water surface and around experimental facilities at beam port floor. The results indicate that sodium 24 activity is the main concentration to the pool water surface dose. Its saturation activity came out to be 29 mR/hr, which is almost 85% of the total activity. Dose rate at pool wall outer surface was found to be around 0.5 mR/hr except at beam tube plug surface, where the dose rate was calculated to be 80 mR/hr. (author)

  11. 77 FR 14319 - Unmanned Aircraft System Test Sites

    Science.gov (United States)

    2012-03-09

    ... can be found in the Federal Register published on April 11, 2000 (65 FR 19477-19478), as well as at... Federal Aviation Administration 14 CFR Part 91 Unmanned Aircraft System Test Sites AGENCY: Federal... test ranges/sites to integrate unmanned aircraft systems (UAS) into the National Airspace System...

  12. Shielding assessment of the radiotherapy room of the second egyptian research reactor (ET-RR-2)

    International Nuclear Information System (INIS)

    one of the applications of the ET-RR-2 multipurpose reactor is boron capture therapy. The reactor is provided with a radiotherapy room for patients to be irradiated with thermal neutrons emerging from one of its irradiation tubes. This room has special shield arrangements. The present work is aimed to assess the shielding performance of the above room in the abnormal situation i.e. when the irradiation tube is completely opened (without any collimation of the irradiation beam). One-dimensional ANISN and two-dimensional DOT 3.5 transport codes were used to calculate neutrons as well as primary and secondary gamma fluxes and doses. The cross sections utilized in the ANISN and the dot 3.5 calculations originated the multigroup cross library DLC2 (VITAMIN-C Library) for coupled neutrons and gamma rays. Results were obtained at different distances from the core center up to the front wall of the tumor irradiation room passing through the beam port, the room entrance and around the shielding door and its access entrance. calculations were also done across the room shielding materials to assess its performance. The present results were compared with the design calculations and the actual doses measured during reactor operation at full power

  13. Effect of Self-Shielding on Burn-Up Calculation of ETRR-2 Reactor

    International Nuclear Information System (INIS)

    There exist two approaches for burn-up calculation. The first on is to use cell parameters generated using cell calculation code at different degrees of burn-up. The other is to use microscopic cross sections with self-shielding in order to compensate for the variation of spectrum at different degree of burn-up. The effect of using different forms of self-shielding factors on burn-up calculation for ETRR-2 reactor has been determined. The results of the two approaches are inter-compared up to 50% burn-up

  14. Convergence of finite element calculations of shielding factors in multigroup reactor analysis

    International Nuclear Information System (INIS)

    Finite Element-Spherical Harmonics Transport calculations can lead to accurate values of group shielding factors in reactor analysis. A benchmark problem in resonance absorption in a lattice unit cell is considered to study the convergence of the results obtained by this approach. The treatment refers to a representative symmetry portion of the unit cell for which the boundary conditions can be handled precisely for any specified order of the angular expansion of the neutron flux. Complementary approaches using variational methods are available to bracket desired results of the analysis, such as shielding factors or effective resonance integrals, between upper and lower bounds. This convergence is discussed

  15. Design and Testing of Improved Spacesuit Shielding Components

    International Nuclear Information System (INIS)

    In prior studies of the current Shuttle Spacesuit (SSA), where basic fabric lay-ups were tested for shielding capabilities, it was found that the fabric portions of the suit give far less protection than previously estimated due to porosity and non-uniformity of fabric and LCVG components. In addition, overall material transmission properties were less than optimum. A number of alternate approaches are being tested to provide more uniform coverage and to use more efficient materials. We will discuss in this paper, recent testing of new material lay-ups/configurations for possible use in future spacesuit designs

  16. Design and Testing of Improved Spacesuit Shielding Components

    Energy Technology Data Exchange (ETDEWEB)

    Ware, J.; Ferl, J.; Wilson, J.W.; Clowdsley, M.S.; DeAngelis, G.; Tweed, J.; Zeitlin, C.J.

    2002-05-08

    In prior studies of the current Shuttle Spacesuit (SSA), where basic fabric lay-ups were tested for shielding capabilities, it was found that the fabric portions of the suit give far less protection than previously estimated due to porosity and non-uniformity of fabric and LCVG components. In addition, overall material transmission properties were less than optimum. A number of alternate approaches are being tested to provide more uniform coverage and to use more efficient materials. We will discuss in this paper, recent testing of new material lay-ups/configurations for possible use in future spacesuit designs.

  17. Radiation Resistance Test of Wireless Sensor Node and the Radiation Shielding Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Li, Liqan; Sur, Bhaskar [Atomic Energy of Canada Limited, Ontario (Canada); Wang, Quan [University of Western Ontario, Ontario (Canada); Deng, Changjian [The University of Electronic Science and Technology, Chengdu (China); Chen, Dongyi; Jiang, Jin [Applied Physics Branch, Ontario (Korea, Republic of)

    2014-08-15

    A wireless sensor network (WSN) is being developed for nuclear power plants. Amongst others, ionizing radiation resistance is one essential requirement for WSN to be successful. This paper documents the work done in Chalk River Laboratories of Atomic Energy of Canada Limited (AECL) to test the resistance to neutron and gamma radiation of some WSN nodes. The recorded dose limit that the nodes can withstand before being damaged by the radiation is compared with the radiation environment inside a typical CANDU (CANada Deuterium Uranium) power plant reactor building. Shielding effects of polyethylene, cadmium and lead to neutron and gamma radiations are also analyzed using MCNP simulation. The shielding calculation can be a reference for the node case design when high dose rate or accidental condition (like Fukushima) is to be considered.

  18. Advanced test reactor. Testing capabilities and plans

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. For future research, some ATR modifications and enhancements are currently planned. In 2007 the US Department of Energy designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR for material testing research by a broader user community. This paper provides more details on some of the ATR capabilities, key design features, experiments, and plants for the NSUF. (author)

  19. Analysis of crack-formation in the shielding concrete of a TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Within a short time after the start-up of the reactor several cracks appeared at the concrete surface and the number and width of the cracks had grown till now. Experimental and theoretical analysis were made in order to investigate the origin of the cracks and to prevent further crack increase. Crack movement was measured by inductive gages and simultaneously the temperature of the cooling water in the reactor tank at the top and at the bottom as well as the air and the concrete temperature were recorded. The calculations of the thermal stresses were made in two independent ways: 1. Analytically, simulating the shielding concrete as an infinite hollow cylinder of constant thickness and 2. Using the Finite Element method, for a better description of the geometry. It was concluded that the cracks of the shielding concrete are exclusively caused by the thermal stresses. The thermal insulation at the lower part of the shielding is not effective. The structural system of the shielding concrete as a monolithic block without joints produces automatically tensile stresses

  20. RADHEAT, Transport, Heat Generator, Radiation Damage Cross-Sections in Reactor and Shield

    International Nuclear Information System (INIS)

    1 - Nature of physical problem solved: Multi-group cross sections for neutron and photon are generated by using the evaluated nuclear data, ENDF/B, JENDL and DLC-15 for transport, heat generation, and radiation damage calculations in nuclear reactors and shields. The secondary gamma-ray production cross section by the induced reaction and the Bremsstrahlung effect is also generated. The radiation transport problems for shielding designs and safety analyses of fusion and fission reactors are solved. 2 - Method of solution: A point-wise processing technique and the Bondarenko-type resonance self-shielding factors are adopted to generate multi-group cross sections. One- and two-dimensional Sn transport methods and three-dimensional Monte Carlo method are used for shielding calculation. 3 - Restrictions on the complexity of the problem: The maximum number of neutron and photon fine-groups is 200 and 50, respectively. The maximum number of angular meshes in the macroscopic cross section is 64. The energy ranges from 2.61 E-5 eV to 16.74 MeV for neutrons and from 1 KeV to 20 MeV for photons are available

  1. Determination of boron in Jabroc wood used as a shielding material in nuclear reactors

    International Nuclear Information System (INIS)

    Jabroc are non-impregnated, densified wood laminates developed commercially for a wide range of industrial applications. Jabroc can be used with other neutron shielding materials such as Lead to form complex shielding structures. Its relative light weight and cleanliness in handling are additional features that make it a suitable candidate for the standard design of neutron shielding equipment. Jabroc can also be impregnated with Boron up to a maximum of 4% to be used in areas where Gamma radiation produced on Neutron capture reaches unacceptable dose rates. Boron impregnated Jabroc wood finds application in TAPS 3 and 4 as a shielding material for the Ion Chambers and the Horizontal Flux Units (HFU). The shielding property of this material is optimized by incorporating requisite amount of boron in wood. Boron content in this material has to be determined accurately prior to its use in the nuclear reactors. In this work a method was standardized to determine boron in Jabroc wood samples to check for conformance to specifications. The wood sample flakes were wetted with saturated barium hydroxide solution and dries under IR. The sample was ashed in a muffle furnace at 600℃ for 2 h

  2. Shielding Calculations for The New Spent Fuel Storage Pool of Etrr1 Reactor

    International Nuclear Information System (INIS)

    MCNP code was used to model and simulate the new spent fuel storage pool of Etrr1 research reactor. Shielding calculations for the pool were performed to calculate the radiation dose through different pool layers. Radiation sources for photons and neutrons inside the pool were determined under different conditions. Key parameters that affect the radiation dose outside the pool were studied. Comparison with the designer values was performed, agreement and disagreement were investigated. Radiation safety of the pool has been verified

  3. Collimator and shielding design for boron neutron capture therapy (BNCT) facility at TRIGA MARK II reactor

    International Nuclear Information System (INIS)

    The geometry of reactor core, thermal column, collimator and shielding system for BNCT application of TRIGA MARK II Reactor were simulated with MCNP5 code. Neutron particle lethargy and dose were calculated with MCNPX code. Neutron flux in a sample located at the end of collimator after normalized to measured value (Eid Mahmoud Eid Abdel Munem, 2007) at 1 MW power was 1.06 x 108 n/ cm2/ s. According to IAEA (2001) flux of 1.00 x 109 n/ cm2/ s requires three hours of treatment. Few modifications were needed to get higher flux. (Author)

  4. Conceptual design of neutron shield for ECH launcher on D-T fusion reactors

    International Nuclear Information System (INIS)

    Conceptual design of Electron Cyclotron Heating and Current Drive (ECH/ECCD) launcher for fusion reactors is described. The ECH injection power of 20∼25 MW per a port and the shielding capability to protect superconducting magnets and ECH torus windows from radiation damages are required for the ECH launcher in deuterium - tritium (D-T) fusion reactors. The conceptual design study and the nuclear analysis (2D) for the ECH launcher to qualify the design specification were carried out. The guideline of the detail design for the ECH launcher was obtained. (author)

  5. PITR: Princeton Ignition Test Reactor

    International Nuclear Information System (INIS)

    The principal objectives of the PITR - Princeton Ignition Test Reactor - are to demonstrate the attainment of thermonuclear ignition in deuterium-tritium, and to develop optimal start-up techniques for plasma heating and current induction, in order to determine the most favorable means of reducing the size and cost of tokamak power reactors. This report describes the status of the plasma and engineering design features of the PITR. The PITR geometry is chosen to provide the highest MHD-stable values of beta in a D-shaped plasma, as well as ease of access for remote handling and neutral-beam injection

  6. PITR: Princeton Ignition Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-01

    The principal objectives of the PITR - Princeton Ignition Test Reactor - are to demonstrate the attainment of thermonuclear ignition in deuterium-tritium, and to develop optimal start-up techniques for plasma heating and current induction, in order to determine the most favorable means of reducing the size and cost of tokamak power reactors. This report describes the status of the plasma and engineering design features of the PITR. The PITR geometry is chosen to provide the highest MHD-stable values of beta in a D-shaped plasma, as well as ease of access for remote handling and neutral-beam injection.

  7. Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending December 31, 1956

    Energy Technology Data Exchange (ETDEWEB)

    NA, NA [ORNL

    1957-03-12

    This quarterly progress report of the Aircraft Nuclear Propulsion Project at ORNL records the technical progress of research on circulating-fuel reactors and other ANP research at the Laboratory. The report is divided into five major parts: 1) Aircraft Reactor Engineering, 2) Chemistry, and 3) Metallurgy, 4) Heat Transfer and Physical Properties, Radiation Damage, and Fuel Recovery and Reprocessing, and 5) Reactor Shielding.

  8. Seismic analysis of the mirror fusion test facility shielding vault

    International Nuclear Information System (INIS)

    This report presents a seismic analysis of the vault in Building 431 at Lawrence Livermore National Laboratory which houses the mirror Fusion Test Facility. The shielding vault structure is approximately 120 ft long by 80 ft wide and is constructed of concrete blocks approximately 7 x 7 x 7 ft. The north and south walls are approximately 53 ft high and the east wall is approximately 29 ft high. These walls are supported on a monolithic concrete foundation that surrounds a 21-ft deep open pit. Since the 53-ft walls appeared to present the greatest seismic problem they were the first investigated

  9. Space Environmental Testing of the Electrodynamic Dust Shield Technology

    Science.gov (United States)

    Calle, Carlos I.; Mackey, P. J.; Hogue, M. D.; Johansen, M .R.; Yim, H.; Delaune, P. B.; Clements, J. S.

    2013-01-01

    NASA's exploration missions to Mars and the moon may be jeopardized by dust that will adhere to surfaces of (a) Optical systems, viewports and solar panels, (b) Thermal radiators, (c) Instrumentation, and (d) Spacesuits. We have developed an active dust mitigation technology, the Electrodynamic Dust Shield, a multilayer coating that can remove dust and also prevents its accumulation Extensive testing in simulated laboratory environments and on a reduced gravity flight shows that high dust removal performance can be achieved Long duration exposure to the space environment as part of the MISSE-X payload will validate the technology for lunar missions.

  10. A Review on the Production Methods and Testing of Textiles for Electro Magnetic Interference (EMI shielding

    Directory of Open Access Journals (Sweden)

    Bagavathi M,

    2015-02-01

    Full Text Available The need of the present generation to protect themselves from electromagnetic radiation due the various technological developments has paved way to the birth of EMI shielding of textiles. The shielding effectiveness of the developed fabric will vary depending upon the fabric or the coating constituents. The shielding requirements for different applications vary widely which has resulted in the development of wide variety of shielding mechanisms and materials which can be used in the production of shielding equipment and work wear. In addition to their production, testing of shielding gears involves various methods to be adopted depending on the application.

  11. Design and Test of a Blast Shield for Boeing 737 Overhead Compartment

    OpenAIRE

    Xinglai Dang; Philemon C. Chan

    2006-01-01

    This work demonstrates the feasibility of using a composite blast shield for hardening an overhead bin compartment of a commercial aircraft. If a small amount of explosive escapes detection and is brought onboard and stowed in an overhead bin compartment of a passenger aircraft, the current bins provide no protection against a blast inside the compartment. A blast from the overhead bin will certainly damage the fuselage and likely lead to catastrophic inflight structural failure. The feasibil...

  12. Improvements at the biological shielding of BNCT research facility in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    The technique of neutron capture in boron is a promising technique in cancer treatment, it uses the high LET particles from the reaction 10B (n, α) 7Li to destroy cancer cells.The development of this technique began in the mid-'50s and even today it is the object of study and research in various centers around the world, Brazil has built a facility that aims to conduct research in BNCT, this facility is located next to irradiation channel number three at the research nuclear reactor IEA-R1 and has a biological shielding designed to meet the radiation protection standards. This biological shielding was developed to allow them to conduct experiments with the reactor at maximum power, so it is not necessary to turn on and off the reactor to irradiate samples. However, when the channel is opened for experiments the background radiation in the experiments salon increases and this background variation makes it impossible to perform measurements in a neutron diffraction research that utilizes the irradiation channel number six. This study aims to further improve the shielding in order to minimize the variation of background making it possible to perform the research facility in BNCT without interfering with the action of the research group of the irradiation channel number six. To reach this purpose, the code MCNP5, dosimeters and activation detectors were used to plan improvements in the biological shielding. It was calculated with the help of the code an improvement that can reduce the average heat flow in 71.2% ± 13 and verified experimentally a mean reduce of 70 ± 9% in dose due to thermal neutrons. (author)

  13. Analysis of the conceptual shielding design for the upflow Gas-Cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Slater, C.O.; Reed, D.A.; Cramer, S.N.; Emmett, M.B.; Tomlinson, E.T.

    1981-01-01

    Conceptual Shielding Configuration III for the Gas-Cooled Fast Breeder Reactor (GCFR) was analyzed by performing global calculations of neutron and gamma-ray fluences and correcting the results as appropriate with bias factors from localized calculations. Included among the localized calculations were the radial and axial cell streaming calculations, plus extensive preliminary calculations and three final confirmation calculations of the plenum flow-through shields. The global calculations were performed on the GCFR mid-level and the lower and upper plenum regions. Calculated activities were examined with respect to the design constraint, if any, imposed on the particular activity. The spatial distributions of several activities of interest were examined with the aid of isoplots (i.e., symbols are used to describe a surface on which the activity level is everywhere the same). In general the results showed that most activities were below the respective design constraints. Only the total neutron fluence in the core barrel appeared to be marginal with the present reactor design. Since similar results were obtained for an earlier design, it has been proposed that the core barrel be cooled with inlet plenum gas to maintain it at a temperature low enough that it can withstand a higher fluence limit. Radiation levels in the prestressed concrete reactor vessel (PCRV) and liner appeared to be sufficiently below the design constraint that expected results from the Radial Shield Heterogeneity Experiment should not force any levels above the design constraint. A list was also made of a number of issues which should be examined before completion of the final shielding design.

  14. Analysis of the conceptual shielding design for the upflow Gas-Cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Conceptual Shielding Configuration III for the Gas-Cooled Fast Breeder Reactor (GCFR) was analyzed by performing global calculations of neutron and gamma-ray fluences and correcting the results as appropriate with bias factors from localized calculations. Included among the localized calculations were the radial and axial cell streaming calculations, plus extensive preliminary calculations and three final confirmation calculations of the plenum flow-through shields. The global calculations were performed on the GCFR mid-level and the lower and upper plenum regions. Calculated activities were examined with respect to the design constraint, if any, imposed on the particular activity. The spatial distributions of several activities of interest were examined with the aid of isoplots (i.e., symbols are used to describe a surface on which the activity level is everywhere the same). In general the results showed that most activities were below the respective design constraints. Only the total neutron fluence in the core barrel appeared to be marginal with the present reactor design. Since similar results were obtained for an earlier design, it has been proposed that the core barrel be cooled with inlet plenum gas to maintain it at a temperature low enough that it can withstand a higher fluence limit. Radiation levels in the prestressed concrete reactor vessel (PCRV) and liner appeared to be sufficiently below the design constraint that expected results from the Radial Shield Heterogeneity Experiment should not force any levels above the design constraint. A list was also made of a number of issues which should be examined before completion of the final shielding design

  15. Testing the reactor charging machine

    International Nuclear Information System (INIS)

    One of the main objective of the R - D technological engineering program devoted to the Fuel Handling System is domestic production of equipment and technology for testing the ends of the reactor charging machine (MID) destined to Cernavoda NPP, beginning with Unit 2. To achieve the objective based on an own design, a bench-scale testing stand of MIDs which can simulate the pressure, flow-rate, and temperature conditions proper to fuel channels in operating CANDU 600 reactors. The main components of this testing facility are: - fuel channels, cold also test sections, allowing the coupling of MID end upwardly and downwardly, corresponding to the direction of the water flow through the channel; - technological installation feeding with light water the testing sections of the facility in thermohydraulic conditions, similar to those in the reactor, allowing the cold and hot testings, respectively, of the MID end; - cold testing installation, water supply and oil control panel, feeding the hydraulic drives of the MID's end during the testings; - fixed bridge and mobile carrier for MID's end positioning against testing sections; - installation for functional testing of MID thrusters, before pre-admission and reception tests; - dedicated tools and devices; - raising and transport mechanical devices for handling and positioning the MID's end upon the carrier; - automation panel for controlling the stand equipment and MID's end; - process computer for conducting on-line tests. MID's end testing implies mainly the following operations: - regulation, calibration and functional testing of the MID thrusters carried out independently on a specialised stand; - regulation and calibration of MID's end sub-assemblages; - carrying out the cold and hot pre-admission tests consisting in automatic performing, without operator intervention, of 12 fuel changes, two of which being successive; - performing the cold and hot reception tests, consisting in automatic accomplishment of 4

  16. A study on radiation shielding analysis for toroidal field coils of a tokamak-type fusion reactor

    International Nuclear Information System (INIS)

    A study on the radiation shield for toroidal field (TF) coils of a tokamak type fusion device is reported. The study was performed to provide the design data base for the radiation shielding analysis for TF coils which can be commonly used for other systems, and to produce some universal recommendation about the neutron flux attenuation in the shield of a fusion reactor. Some simple estimation procedure instead of difficult and expensive neutron calculation can be carried out in this case on the basis of the fundamental knowledge on neutron behavior. The present studies are composed of the fundamentals required for shield estimation, the analysis of shield effectiveness, the analysis of the shielding performance of blankets, the analysis of radiation permeating through the inhomogeneous blanket and shield of a fusion reactor, and the analysis of the TF coil shield of the International Thermonuclear Experimental Reactor (ITER) on the basis of the results of the ITER conceptual design activities for three years. The methodological recommendation was developed for the ANISN and the DOT3 codes. (K.I.)

  17. 76 FR 45011 - Control of Air Pollution From Aircraft and Aircraft Engines; Proposed Emission Standards and Test...

    Science.gov (United States)

    2011-07-27

    ... Procedures for Aircraft;'' Final Rule, 38 FR 19088, July 17, 1973. \\12\\ U.S. EPA, ``Control of Air Pollution from Aircraft and Aircraft Engines; Emission Standards and Test Procedures;'' Final Rule, 62 FR 25356... Engines; Emission Standards and Test Procedures;'' Final Rule, 70 FR 2521, November 17, 2005. E....

  18. Neutronic shielding analysis of the water-cooled lithium lead test blanket module in the ITER machine

    International Nuclear Information System (INIS)

    During the operations of the next experimental fusion machine three breeding test blanket modules (TBM) for a power reactor will be inserted in the horizontal ports and their performance examined. The insertion will change the overall shielding capability of the structure and thus the regular operability of the machine could be affected. In this paper, I report the Monte Carlo simulations made to account for the water-cooled lithium lead TBM insertion in the international thermonuclear experimental reactor (ITER) machine. A 9 deg. torus sector of ITER is modelled comprising a detailed description of the TBM located in position, with an additional shield in the back. Results show that the present project of the WCLL TBM, with an additional backshield, is suitable for testing in ITER and does not interfere with the regular operations of the machine

  19. Set of benchmark experiments on slit shielding compositions of thermonuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Andreev, M.I.; Afanasiev, V.V.; Belevitin, A.G.; Karaulov, A.V.; Romodanov, V.L. E-mail: rom@lng.mephi.msk.su; Sakharov, V.K.; Tikhomirov, G.V.; Vasiliev, A.P.; Kandiev, Ya. Z.; Lyutov, V.D.; Sokolov, Yu. A.; Terekhin, V.A.; Shmakov, V.M.; Androsenko, P.A.; Semenov, V.P.; Trykov, L.A.; Lopatkin, A.V.; Muratov, V.G

    2001-09-01

    The paper is based on the results of the ISTC project no. 180 that has recently been completed. The aim of the project was the development of methodical, hardware and design basis to carry out computational and experimental research on non-uniform shieldings of thermonuclear reactors. As a result a set of benchmark experiments were created. On their basis verification of the domestic and foreign computational codes with the nuclear data estimated was realized. For these purposes the iron hollow slits shielding compositions irradiated with 14.8 MeV energy neutrons were studied. The experimental installations allowed research of the shielding compositions with the following characteristics: a solid structure, a structure with one slit of a central symmetry, and the structures with asymmetric slits and with two slits. The thickness of shielding compositions in this research was 500 mm. The results of experiments were compared to the results of calculations by means of the MCNP-4a and PRIZMA computing codes with use of the FENDL-1.1, FENDL-2, JENDL-3.2 and BAS-78 libraries of nuclear data. The results of comparison made it possible to obtain the recommendations for use of these nuclear data.

  20. Fuel Burnup and Fuel Pool Shielding Analysis for Bushehr Nuclear Reactor VVER-1000

    Science.gov (United States)

    Hadad, Kamal; Ayobian, Navid

    Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of γ rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that γ rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.

  1. Characterisation of the inventory of radioisotopes induced in the biological shield a WWER-440 reactor

    International Nuclear Information System (INIS)

    A significant part of the radwaste originating from the decommissioning of NPPs is made up of the activated concrete and steel components of the biological shield. The paper presents the results of studies aimed at the determination of the amount of radionuclides accumulating in the serpentinous and ordinary concrete shield around the WWER-440 reactors of the Paks NPP. For the calculations, the reactor, vessel and shield were modelled in detail both in terms of geometry and material composition. The spatial and energy distribution of the activating neutron spectrum was determined by certain modules of SCALE 4.3 and the code TORT in two and three dimensions, while the activation was calculated using ORIGEN-S for 22 geometrical regions. The results showed that the activity of the concrete structures at final shutdown after 30 years of operation is approximately 50 TBq, which decreases to 20, 12, 1.1 TBq and 27 GBq after 1 month, 1 year, 10 and 100 years, respectively (Authors)

  2. Preliminary optimization analysis of the radiation shielding of the China Lead-based Research Reactor

    International Nuclear Information System (INIS)

    Accelerator Driven subcritical System (ADS) is recognized as an efficient nuclear waste transmutation device. Supported by the Strategic Priority Research Program of 'the Future Advanced Nuclear Fission Energy-ADS transmutation system', the China LEAd-based Research Reactor (CLEAR-I) is proposed. Along with the approaching of the CLEAR-I design, the radiation shielding for CLEAR-I is updated and optimized step by step to meet with new shielding requirements. Employing the modeling program MCAM and calculation system VisualBUS developed by FDS Team, the shielding capability was verified using Monte Carlo method. As shown from the results, the fast neutron flux for components in reactor vessel is under the limitation and the neutron radiation for mechanism in containing room has been as low as possible. After shutdown for 7 days, the dose rate in most area of containing room is lower than 100 μSv/hr, allowing hands on operation. Replacement of components such as the spallation target in containing room is possible. (author)

  3. Development of a computer code for optimization of nuclear reactor radiation shield using non-linear programming

    International Nuclear Information System (INIS)

    The attenuation of nuclear radiation emanating from the reactor core and secondary radiation produced in the structural material is the main task of radiation shield in nuclear reactors. In this regard one can use a shield with optimum mixture of materials or multi-layer shields with optimum thickness. The objective function in shield optimization is generally volume, cost or weight, and the most important constraint is the reduction of the radiation field intensity less than a predetermined value. In this project the computer program SHLDOPT has been developed to find the optimum thicknesses of the shield layers. In the program, dose distribution in various shield layers up to the outer surface of the shield has been calculated using three dimensional linear interpolation, assuming dose in each layer varies exponentially with attenuation coefficients depending on the thicknesses of the desired layer and preceding and following layers. Regarding the speed and accuracy of the dose calculation in the program, the rejection with complete search method has been used as the optimization scheme. Dose attenuation factors used in the input file of SHLDOPT have calculated by the computer code ANISN in P3 and S 8 approximation. In order to rely on the results of SHLDOPT program the calculated dose distribution in different layers of the Bushehr reactor shield have been compared with ANISN code results. The reasonable agreement shows that the program and its dose attenuation coefficient library work properly. Therefore SHKOPT has been applied to find the optimum shield layer thicknesses of Bushehr nuclear reactor for three objectives of volume mass and cost. Dose rate at the outermost shield surface have been considered to be less than 50 n Sv/sec. The calculated thicknesses have been compared with designed values of Kraft We rk Union. The results show that the radiation shield of Bushehr nuclear reactor is not necessarily optimized according to mass or volume, and it is more

  4. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, Steven [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-29

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  5. FASTER test reactor preconceptual design report summary

    International Nuclear Information System (INIS)

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  6. Preliminary Design of a LSA Aircraft Using Wind Tunnel Tests

    Directory of Open Access Journals (Sweden)

    Norbert ANGI

    2015-12-01

    Full Text Available This paper presents preliminary results concerning the design and aerodynamic calculations of a light sport aircraft (LSA. These were performed for a new lightweight, low cost, low fuel consumption and long-range aircraft. The design process was based on specific software tools as Advanced Aircraft Analysis (AAA, XFlr 5 aerodynamic and dynamic stability analysis, and Catia design, according to CS-LSA requirements. The calculations were accomplished by a series of tests performed in the wind tunnel in order to assess experimentally the aerodynamic characteristics of the airplane.

  7. Design of a management information system for the Shielding Experimental Reactor ageing management

    International Nuclear Information System (INIS)

    The problem of nuclear reactor ageing is a topic of increasing importance in nuclear safety recent years. Ageing management is usually implemented for reactors maintenance. In the practice, a large number of data and records need to be processed. However, there are few professional software applications that aid reactor ageing management, especially for research reactors. This paper introduces the design of a new web-based management information system (MIS), named the Shielding Experimental Reactor Ageing Management Information System (SERAMIS). It is an auxiliary means that helps to collect data, keep records, and retrieve information for a research reactor ageing management. The Java2 Enterprise Edition (J2EE) and network database techniques, such as three-tiered model, Model-View-Controller architecture, transaction-oriented operations, and JavaScript techniques, are used in the development of this system. The functionalities of the application cover periodic safety review (PSR), regulatory references, data inspection, and SSCs classification according to ageing management methodology. Data and examples are presented to demonstrate the functionalities. For future work, techniques of data mining will be employed to support decision-making.

  8. Design of a management information system for the Shielding Experimental Reactor ageing management

    Energy Technology Data Exchange (ETDEWEB)

    He Jie, E-mail: hejiejoe@163.co [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Xu Xianhong [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)

    2010-01-15

    The problem of nuclear reactor ageing is a topic of increasing importance in nuclear safety recent years. Ageing management is usually implemented for reactors maintenance. In the practice, a large number of data and records need to be processed. However, there are few professional software applications that aid reactor ageing management, especially for research reactors. This paper introduces the design of a new web-based management information system (MIS), named the Shielding Experimental Reactor Ageing Management Information System (SERAMIS). It is an auxiliary means that helps to collect data, keep records, and retrieve information for a research reactor ageing management. The Java2 Enterprise Edition (J2EE) and network database techniques, such as three-tiered model, Model-View-Controller architecture, transaction-oriented operations, and JavaScript techniques, are used in the development of this system. The functionalities of the application cover periodic safety review (PSR), regulatory references, data inspection, and SSCs classification according to ageing management methodology. Data and examples are presented to demonstrate the functionalities. For future work, techniques of data mining will be employed to support decision-making.

  9. Calculation of neutron fluxes in biological shield of the TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    The complete calculation of neutron fluxes in biological shield and verification with experimental results is presented. Calculated results are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Experimental results used for comparison are available from irradiation experiment with selected type of concrete and other materials in irradiation channel 4 in TRIGA Mark II reactor. These experimental results were used as a benchmark. Homogeneous type of problem (without inserted irradiation channel) and problem with asymmetry (inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. Deviation from material data set up as original parameters is also considered (first of all presence of water in concrete and density of concrete) for type of concrete in biological shield and for selected type of concrete in irradiation channel. BUGLE-96 (47 neutron energy groups) library is used. Excellent agreement between calculated and experimental results for reaction rate is received.(author)

  10. Methods and experimental coefficients used in the computation of reactor shielding

    International Nuclear Information System (INIS)

    1. The concept of an effective removal cross section has been developed in order more easily to compute reactor shielding thicknesses. We have built an experimental facility for the purpose of measuring effective removal cross sections, the value of which had not been published at that time. The first part of this paper describes the device or facility used, the computation method applied, and the results obtained. 2. Starting from this concept, we endeavored to define a removal cross section as a function of energy. This enabled us to use the method for computations bearing on the attenuation of fast neutrons of any spectrum. An experimental verification was carried out for the case of fission neutrons filtered by a substantial thickness of graphite. 3. Finally, we outline a computation method enabling us to determine the sources of captured gamma rays by the age theory and we give an example of the application in a composite shield. (author)

  11. Measured and calculated fast neutron spectra in a depleted uranium and lithium hydride shielded reactor

    Science.gov (United States)

    Lahti, G. P.; Mueller, R. A.

    1973-01-01

    Measurements of MeV neutron were made at the surface of a lithium hydride and depleted uranium shielded reactor. Four shield configurations were considered: these were assembled progressively with cylindrical shells of 5-centimeter-thick depleted uranium, 13-centimeter-thick lithium hydride, 5-centimeter-thick depleted uranium, 13-centimeter-thick lithium hydride, 5-centimeter-thick depleted uranium, and 3-centimeter-thick depleted uranium. Measurements were made with a NE-218 scintillation spectrometer; proton pulse height distributions were differentiated to obtain neutron spectra. Calculations were made using the two-dimensional discrete ordinates code DOT and ENDF/B (version 3) cross sections. Good agreement between measured and calculated spectral shape was observed. Absolute measured and calculated fluxes were within 50 percent of one another; observed discrepancies in absolute flux may be due to cross section errors.

  12. Graphite reflecting characteristics and shielding factors for Miniature Neutron Source Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Albarhoum, M., E-mail: pscientific1@aec.org.s [Department of Nuclear Engineering, Atomic Energy Commission, P.O. Box 6091, Damascus (Syrian Arab Republic)

    2011-01-15

    The usability of graphite as a reflector for MNSRs is investigated in this paper. Its use is optimized and shielding factors are calculated. Graphite seems to be compatible with liquid water. As a reflector, graphite proves to be usable as well, but it decreases the fuel cycle lifetime by about 7%. To optimize its use the average worth reactivity of the unit volume was assessed for the different modes of filling the shim tray of the reactor with graphite which were: RIOS, RIOC, ROIS, and ROIC modes for the radial direction, and ASM, and ACM modes for the axial one. This quantity was found to be maximum for the ROIC mode reaching more than 0.01 mk/cm{sup 3}. The shielding factors for the radial and axial filling modes were found to be 0.7101 and 0.6266, respectively.

  13. Graphite reflecting characteristics and shielding factors for Miniature Neutron Source Reactors

    International Nuclear Information System (INIS)

    The usability of graphite as a reflector for MNSRs is investigated in this paper. Its use is optimized and shielding factors are calculated. Graphite seems to be compatible with liquid water. As a reflector, graphite proves to be usable as well, but it decreases the fuel cycle lifetime by about 7%. To optimize its use the average worth reactivity of the unit volume was assessed for the different modes of filling the shim tray of the reactor with graphite which were: RIOS, RIOC, ROIS, and ROIC modes for the radial direction, and ASM, and ACM modes for the axial one. This quantity was found to be maximum for the ROIC mode reaching more than 0.01 mk/cm3. The shielding factors for the radial and axial filling modes were found to be 0.7101 and 0.6266, respectively.

  14. Bibliography, subject index, and author index of the literature examined by the Radiation Shielding Information Center (Reactor and Weapons Radiation Shielding)

    International Nuclear Information System (INIS)

    An indexed bibliography is presented of literature selected by the Radiation Shielding Information Center since the previous volume was published in 1974 in the area of radiation transport and shielding against radiation from nuclear reactors, x-ray machines, radioisotopes, nuclear weapons (including fallout), and low-energy accelerators (e.g., neutron generators). In addition to lists of literature titles by subject categories (accessions 3501-4950), author and keyword indexes are given. Most of the literature selected for Vol. V was published in the years 1973 to 1976

  15. Bibliography, subject index, and author index of the literature examined by the radiation shielding information center. Volume 6. Reactor and weapons radiation shielding

    International Nuclear Information System (INIS)

    An indexed bibliography is presented of literature selected by the Radiation Shielding Information Center since the previous volume was published in 1978 in the area of radiation transport and shielding against radiation from nuclear reactors, x-ray machines, radioisotopes, nuclear weapons (including fallout), and low energy accelerators (e.g., neutron generators). The bibliography was typeset from data processed by computer from magnetic tape files. In addition to lists of literature titles by subject categories (accessions 4951-6200), an author index is given

  16. Elastic response of a German HTR-Modul reactor building to aircraft impact

    International Nuclear Information System (INIS)

    The paper presents the results of a study carried out for a German HTR-Modul reactor building when subjected to aircraft impact. The structural analysis has been carried out using finite element package program PERMAS. The soil-structure interaction has also been considered using springs. Eigenmode values have been calculated. The elastic response of the building to aircraft impact has been studied by analysing for the stresses and the deflections. (author)

  17. Reactor Simulator Integration and Testing

    Science.gov (United States)

    Schoenfield, M. P.; Webster, K. L.; Pearson, J. B.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator (RxSim) test loop was designed and built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing were to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V because the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This Technical Memorandum summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained, which was lower than the predicted 750 K but 156 K higher than the cold temperature, indicating the design provided some heat regeneration. The annular linear induction pump tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  18. Development of the breeding blanket and shield model for the fusion power reactors system SYCOMORE

    International Nuclear Information System (INIS)

    SYCOMORE, a fusion reactor system code based on a modular approach is under development at CEA. Within this framework, this paper describes the relevant sub-modules which have been implemented to model the main outputs of the breeding blanket and shield block of the system code: tritium breeding ratio, peak energy deposition in toroidal field coils, reactor layout and power deposition, blanket pressure drops and materials inventory. Blanket and shield requirements are calculated by several sub-modules: the blanket assembly and layout sub-module, the neutronic sub-module, the blanket design sub-module (thermal hydraulic and thermo-mechanic pre-design tool). A power flow module has also been developed which is directly linked to the blanket thermo-dynamic performances, which is not described in this paper. For the blanket assembly and layout and the blanket module design sub-modules, explicit analytic models have been developed and implemented; for the neutronic sub-module neural networks that replicate the results of appropriate simplified 1D and 2D neutronic simulations have been built. Presently, relevant model for the Helium Cooled Lithium Lead is available. Sub-modules have been built in a way that they can run separately or coupled into the breeding blanket and shield module in order to be integrated in SYCOMORE. In the paper, the objective and main input/output parameters of each sub-module are reported and relevant models discussed. The application to previous studied reactor models (PPCS model AB, DEMO-HCLL 2006–2007 studies) is also presented

  19. Development of the breeding blanket and shield model for the fusion power reactors system SYCOMORE

    Energy Technology Data Exchange (ETDEWEB)

    Li-Puma, Antonella, E-mail: antonella.lipuma@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Jaboulay, Jean-Charles, E-mail: Jean-Charles.jaboulay@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Martin, Brunella, E-mail: brunella.martin@gmail.com [Incka, 19-21 Rue du 8 mai 1945, F-94110 Arcueil (France)

    2014-10-15

    SYCOMORE, a fusion reactor system code based on a modular approach is under development at CEA. Within this framework, this paper describes the relevant sub-modules which have been implemented to model the main outputs of the breeding blanket and shield block of the system code: tritium breeding ratio, peak energy deposition in toroidal field coils, reactor layout and power deposition, blanket pressure drops and materials inventory. Blanket and shield requirements are calculated by several sub-modules: the blanket assembly and layout sub-module, the neutronic sub-module, the blanket design sub-module (thermal hydraulic and thermo-mechanic pre-design tool). A power flow module has also been developed which is directly linked to the blanket thermo-dynamic performances, which is not described in this paper. For the blanket assembly and layout and the blanket module design sub-modules, explicit analytic models have been developed and implemented; for the neutronic sub-module neural networks that replicate the results of appropriate simplified 1D and 2D neutronic simulations have been built. Presently, relevant model for the Helium Cooled Lithium Lead is available. Sub-modules have been built in a way that they can run separately or coupled into the breeding blanket and shield module in order to be integrated in SYCOMORE. In the paper, the objective and main input/output parameters of each sub-module are reported and relevant models discussed. The application to previous studied reactor models (PPCS model AB, DEMO-HCLL 2006–2007 studies) is also presented.

  20. Optimized 4 pi spherical shell depleted uranium-water shield weights for 200 to 550-megawatt reactors

    Science.gov (United States)

    Wohl, M. L.; Celnik, J.; Schamberger, R. D.

    1972-01-01

    Optimization calculations to determine minimum 4 pi spherical-shell weights were performed at 200-, 375-, and 550-megawatt-thermal reactor power levels. Monte Carlo analyses were performed for a reactor power level corresponding to 375 megawatts. Power densities for the spherical reactor model used varied from 64.2 to 256 watts per cubic centimeter. The dose rate constraint in the optimization calculations was 0.25 mrem per hour at 9.14 meters from the reactor center. The resulting shield weights were correlated with the reactor power levels and power densities by a regression analysis. The optimum shield weight for a 375-megawatt, 160-watt-per-cubic-centimeter reactor was 202,000 kilograms.

  1. Neutron beam experiments using nuclear research reactors: honoring the retirement of professor Bernard W. Wehring -II. 7. Redesign of the University of Texas Thermal Neutron Imaging Facility Shielding

    International Nuclear Information System (INIS)

    A thermal neutron imaging facility (TNIF) was developed at the University of Texas Nuclear Engineering Teaching Laboratory from 1994 to 1998 using a 1-MW TRIGA reactor. Currently, neutron radiography is being investigated as a method to detect flaws in large carbon composite flywheels using the TNIF. Thermal neutrons have successfully been used to detect flaws in thin carbon composites (60% of the neutrons that enter the shield walls are reflected back into the experimental area. MCNP calculations indicate that the addition of a 1.25-cm Boral liner on the inner wall is sufficient to lower the external dose to acceptable levels and reduce the percentage of neutrons reflected back into the experimental area to <2%. MCNP simulations have been a valuable tool to test shielding configurations before construction. The redesigned shutter is composed of aluminum, lead, and boron carbide. MCNP simulations for the external shielding have shown that the addition of a Boral liner on the inner shield wall is sufficient to reduce external radiation exposure to acceptable levels. The Boral liner also greatly reduces the amount of neutrons reflected back into the experimental region. The implementation of the redesigned neutron shutter and external shielding should greatly enhance the TNIF capabilities and overall usability. The new neutron shutter will allow work to be performed inside the shielding cave while the reactor is at power. The improved external shielding will enable radiographs to be taken at higher flux levels, which will be beneficial when imaging thick carbon composites. The reduction of neutron scattering within the experimental area will also enhance image quality and improve the TNIF resolution. (authors)

  2. Mesos-scale modeling of irradiation in pressurized water reactor concrete biological shields

    Energy Technology Data Exchange (ETDEWEB)

    Le Pape, Yann [ORNL; Huang, Hai [Idaho National Laboratory (INL)

    2016-01-01

    Neutron irradiation exposure causes aggregate expansion, namely radiation-induced volumetric expansion (RIVE). The structural significance of RIVE on a portion of a prototypical pressurized water reactor (PWR) concrete biological shield (CBS) is investigated by using a meso- scale nonlinear concrete model with inputs from an irradiation transport code and a coupled moisture transport-heat transfer code. RIVE-induced severe cracking onset appears to be triggered by the ini- tial shrinkage-induced cracking and propagates to a depth of > 10 cm at extended operation of 80 years. Relaxation of the cement paste stresses results in delaying the crack propagation by about 10 years.

  3. Shielding experiments

    International Nuclear Information System (INIS)

    Shielding mock-up experiments for Prototype Fast Breeder Reactor (PFBR) and Advanced Heavy Water Reactor (AHWR) are carried out in shielding corner facility of APSARA reactor, to assess the overall accuracy of the codes and nuclear data used in reactor shield design. As APSARA is a swimming pool-type thermal reactor, for fast reactor experiments, typical fast reactor shielding facility was created by using uranium assemblies as spectrum converter. The flux was also enhanced by replacing water by air. Experiments have been carried out to study neutron attenuation through typical fast reactor radial and axial bulk shielding materials such as steel, sodium, graphite, borated graphite and boron carbide. A large number of reaction rates, sensitive to different regions of the neutron energy spectrum, were measured using foil activation and Solid State Nuclear Track Detector (SSNTD) techniques. These experimental results were analysed using computational tools normally used in design calculations, viz., discrete ordinate transport codes with multigroup cross section sets. Comparison of measured reaction rates with calculations provided suitable bias factors for parameters relevant to shield design, such as sodium activation, fast neutron fluence, fission equivalent fluxes etc. The measured neutron spectrum on the incident face of shield model compares well with the calculated fast reactor blanket leakage neutron spectrum. The comparison of calculated reaction rates within shield model indicate that the calculations suffer from considerable uncertainties, in shield models with boron carbide/borated graphite. For AHWR shielding experiments, no spectrum converter was used as it is also a thermal reactor. Radiation streaming studies through penetrations/ducts of various shapes and sizes relevant to AHWR shielding were carried out. (author)

  4. Real time simulator for material testing reactor

    International Nuclear Information System (INIS)

    Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

  5. Importance of self-shielding for improving sensitivity coefficients in light water nuclear reactors

    International Nuclear Information System (INIS)

    Highlights: • A new method has been developed for calculating sensitivity coefficients. • This method is based on the use of infinite dilution cross-sections instead of effective cross-sections. • The change of self-shielding factor due to cross-section perturbation has been considered. • SRAC and SAINT codes are used for calculating improved sensitivities, while MCNP code has been used for verification. - Abstract: In order to perform sensitivity analyzes in light water reactors where self-shielding effect becomes important, a new method has been developed for calculating sensitivity coefficient of core characteristics relative to the infinite dilution cross-sections instead of the effective cross-sections. This method considers the change of the self-shielding factor due to cross-section perturbation for different nuclides and reactions. SRAC and SAINT codes are used to calculate the improved sensitivity; while the accuracy of the present method has been verified by MCNP code and good agreement has been found

  6. Study on shielding design methods for fusion reactors using benchmark experiments

    International Nuclear Information System (INIS)

    In this study, a series of engineering benchmark experiments have been performed on the critical issues of shielding designs for DT fusion reactors. Based on the experiments, calculational accuracy of shielding design methods used in the ITER conceptual design, discrete ordinates code DOT3.5 and Monte Carlo code MCNP-3, have been estimated, and difficulties on calculational methods have been revealed. Furthermore, the feasibility for shielding designs have been examined with respect to a discrete ordinates code system BERMUDA which is developed to attain high accuracy of calculation. As for neutron streaming in an off-set narrow gap experimental assembly made of stainless steel, DOT3.5 and MCNP-3 codes reproduced the experiments within the accuracy presumed in the ITER conceptual design. DOT3.5 and MCNP-3 codes are available for secondary γ ray nuclear heating in a type 316L stainless steel assembly and neutron streaming in a multi-layered slit experimental assembly, respectively. Moreover, BERMUDA-2DN code is an effective tool as to neutron deep penetration in a type 316L stainless steel assembly and the neutron behavior in a large cavity experimental assembly. (author)

  7. Study on shielding design methods for fusion reactors using benchmark experiments

    Energy Technology Data Exchange (ETDEWEB)

    Nakashima, Hiroshi (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment)

    1992-02-01

    In this study, a series of engineering benchmark experiments have been performed on the critical issues of shielding designs for DT fusion reactors. Based on the experiments, calculational accuracy of shielding design methods used in the ITER conceptual design, discrete ordinates code DOT3.5 and Monte Carlo code MCNP-3, have been estimated, and difficulties on calculational methods have been revealed. Furthermore, the feasibility for shielding designs have been examined with respect to a discrete ordinates code system BERMUDA which is developed to attain high accuracy of calculation. As for neutron streaming in an off-set narrow gap experimental assembly made of stainless steel, DOT3.5 and MCNP-3 codes reproduced the experiments within the accuracy presumed in the ITER conceptual design. DOT3.5 and MCNP-3 codes are available for secondary {gamma} ray nuclear heating in a type 316L stainless steel assembly and neutron streaming in a multi-layered slit experimental assembly, respectively. Moreover, BERMUDA-2DN code is an effective tool as to neutron deep penetration in a type 316L stainless steel assembly and the neutron behavior in a large cavity experimental assembly. (author).

  8. Reactor vessel dismantling at the high flux materials testing reactor Petten

    International Nuclear Information System (INIS)

    The project of replacing the reactor vessel of the high flux materials testing reactor (HFR) originated in 1974 when results of several research programs confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report describes the dismantling philosophy and organisation, the design of special underwater equipment, the dismantling of the reactor vessel and thermal column, and the conditioning and shielding activities resulting in a working area for the installation of the new vessel with no access limitations due to radiation. Finally an overview of the segmentation, waste disposal and radiation exposure is given. The total dismantling, segmentation and conditioning activities resulted in a total collective radiation dose of 300 mSv. (orig.)

  9. Application of SCALE 6.1 MAVRIC Sequence for Activation Calculation in Reactor Primary Shield Concrete

    International Nuclear Information System (INIS)

    Activation calculation requires flux information at desired location and reaction cross sections for the constituent elements to obtain production rate of activation products. Generally it is not an easy task to obtain fluxes or reaction rates with low uncertainties in a reasonable time for deep penetration problems by using standard Monte Carlo methods. The MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations) sequence in SCALE 6.1 code package is intended to perform radiation transport on problems that are too challenging for standard, unbiased Monte Carlo methods. And the SCALE code system provides plenty of ENDF reaction types enough to consider almost all activation reactions in the nuclear reactor materials. To evaluate the activation of the important isotopes in primary shield, SCALE 6.1 MAVRIC sequence has been utilized for the KSNP reactor model and the calculated results are compared to the isotopic activity concentration of related standard. Related to the planning for decommission, the activation products in concrete primary shield such as Fe-55, Co-60, Ba-133, Eu-152, and Eu-154 are identified as important elements according to the comparisons with related standard for exemption. In this study, reference data are used for the concrete compositions in the activation calculation to see the applicability of MAVRIC code to the evaluation of activation inventory in the concrete primary shield. The composition data of trace elements as shown in Table 1 are obtained from various US power plant sites and accordingly they have large variations in quantity due to the characteristics of concrete composition. In practical estimation of activation radioactivity for a specific plant related to decommissioning, rigorous chemical analysis of concrete samples of the plant would first have to be performed to get exact information for compositions of concrete. Considering the capability of solving deep penetration transport problems and richness

  10. Neutronics shielding analysis for the end plug of a tandem mirror fusion reactor

    Science.gov (United States)

    Ragheb, Magdi M. H.; Maynard, Charles W.

    1981-10-01

    A neutronics analysis using the Monte Carlo method is carried out for the end-plug penetration and magnet system of a tandem mirror fusion reactor. Detailed penetration and the magnets' three-dimensional configurations are modeled. A method of position dependent angular source biasing is developed to adequately sample the DT fusion source in the central cell region and obtain flux contributions at the penetration components. To assure cryogenic stability, the barrier cylindrical solenoid is identified as needing substantial shielding of about 1 m of a steel-lead-boron-carbide-water mixture. Heating rates there would require a thermal-hydraulic design similar to that in the central cell blanket region. The transition coils, however, need a minimal 0.2 m thickness shield. The leakage neutron flux at the direct converters is estimated at 1.3×1015 n/(m2·s), two orders of magnitude lower than that reported at the neutral beam injectors for tokamaks around 1017 n/(m2·s) for a 1 MW/m2 14 MeV neutron wall loading. This result is obtained through a coupling between the nuclear and plasma physics designs in which hydrogen ions rather than deuterium atoms are used for energy injection at the end plug, to avoid creating a neutron source there. This lower and controllable radiation leakage problem is perceived as a potential major advantage of tandem mirrors compared to tokamaks and laser reactor systems.

  11. Aircraft-crash-protected steel reactor building roof structure for the European market

    International Nuclear Information System (INIS)

    This paper recommends the use of all steel roof structures for the reactor building of European Boiling Water Reactor (BWR) plants. This change would make the advanced US BWR designs more compatible with European requirements. Replacement of the existing concrete roof slab with a sufficiently thick steel plate would eliminate the concrete spelling resulting from a postulated aircraft crash, potentially damaging the drywell head or the spent fuel pool

  12. Materials development for ITER shielding and test blanket in China

    International Nuclear Information System (INIS)

    China is a member of the ITER program and is developing her own materials for its shielding and test blanket modules. The materials include vacuum-hot-pressing (VHP) Be, CuCrZr alloy, 316L(N) and China low activation ferritic/martensitic (CLF-1) steels. Joining technologies including Be/Cu hot isostatic pressing (HIP) and electron beam (EB) weldability of 316L(N) were investigated. Chinese VHP-Be showed good properties, with BeO content and ductility that satisfy the ITER requirements. Be/Cu mock-ups were fabricated for Be qualification tests at simulated ITER vertical displacement event (VDE) and heat flux cycling conditions. Fine microstructure and good mechanical strength of the CuCrZr alloy were achieved by a pre-forging treatment, while the weldability of 316L(N) by EB was demonstrated for welding depths varying from 5 to 80 mm. Fine microstructure, high strength, and good ductility were achieved in CLF-1 steel by an optimized normalizing, tempering and aging procedure.

  13. Pilot tests for dismantling by blasting of the biological shield of a shut down nuclear power station

    International Nuclear Information System (INIS)

    Following free-field tests on concrete blocks the feasibility of explosive dismantling of the biological shield of nuclear power stations has been succesfully tested at the former hotsteam reaction in Karlstein/Main Germany. For this purpose a model shield of scale 1:2 was embedded into the reactor structure at which bore-hole blasting tests employing up to about 15 kg of explosive were performed. An elaborate measurement system allowed to receive detailed information on the blast side-effects: Special emphasis was focussed on the quantitative registration of the dynamic blast loads; data for the transfer of the dismantling method to the removal of real ractor structures were obtained. (orig.)

  14. Structural evaluation of the Shippingport Reactor Pressure Vessel and Neutron Shield Tank package for impact and puncture loads

    International Nuclear Information System (INIS)

    A structural evaluation of Shippingport Reactor Pressure Vessel and Neutron Shield Tank package for impact and puncture loads under the normal and hypothetical accident conditions of 10 CFR 71 was performed. Component performance criteria for the Shippingport package and the corresponding structural acceptance criteria for these components were developed based on a review of the package geometry, the planned transport environment, and the external radiation standards and dispersal limits of 10 CFR 71. The evaluation was performed using structural analysis methods. A demonstration combining simplified model tests and nonlinear finite element analyses was made to substantiate the structural analysis methods used to evaluate the Shippingport package. The package was analyzed and the results indicate that the package meets external radiation standards and release limits of 10 CFR 71. 13 refs., 50 figs., 19 tabs

  15. Carbon Nanostructures for Electromagnetic Shielding and Lightning Strike Protection Applications in Aircraft

    Science.gov (United States)

    Shah, T.; Jones, M.; Alberding, M.; Laszewski, M.

    2012-05-01

    Applied NanoStructured Solutions, LLC (ANS) has developed a unique Chemical Vapor Deposition (CVD) process for the growth of Carbon Nanotubes (CNT) onto various fiber substrates including carbon, glass, ceramics and aramids. This process is continuous and operates at atmospheric pressures enabling high volume/low cost manufacturing. This process infuses conductive CNTs in a highly entangled form referred to as Carbon Nanostructures (CNS) onto the surface of the normally insulative fiber making it highly conductive overall. Composites made from this CNS-infused filler then have unique Electromagnetic Interference (EMI) shielding and Lightening Strike Protection (LSP) properties.

  16. Ground test facility for nuclear testing of space reactor subsystems

    International Nuclear Information System (INIS)

    Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs

  17. Neutron flux measurements at the TRIGA reactor in Vienna for the prediction of the activation of the biological shield.

    Science.gov (United States)

    Merz, Stefan; Djuricic, Mile; Villa, Mario; Böck, Helmuth; Steinhauser, Georg

    2011-11-01

    The activation of the biological shield is an important process for waste management considerations of nuclear facilities. The final activity can be estimated by modeling using the neutron flux density rather than the radiometric approach of activity measurements. Measurement series at the TRIGA reactor Vienna reveal that the flux density next to the biological shield is in the order of 10(9)cm(-2)s(-1) at maximum power; but it is strongly influenced by reactor installations. The data allow the estimation of the final waste categorization of the concrete according to the Austrian legislation. PMID:21646026

  18. Startup operational tests of fast reactors

    International Nuclear Information System (INIS)

    This paper is mainly concerned with the experiences of the two main phases of startup operational tests of fast reactors: (1) The general tests and Sodium filling before core loading. (2) The core loading,approach to criticality and power build up operational tests, taking for example a large and middle demonstrating integrated-type fast reactor. (author)

  19. Where have the neutrons gone: A history of the Tower Shielding Facility

    International Nuclear Information System (INIS)

    In the early 1950's, the concept of the unit shield for the nuclear powered aircraft reactor changed to one of the divided shield concept where the reactor and crew compartment shared the shielding load. Design calculations for the divided shield were being made based on data obtained in studies for the, unit shield. It was believed that these divided shield designs were subject to error, the magnitude of which could not be estimated. This belief led to the design of the Tower Shielding Facility where divided-shield-type measurements could be made without interference from ground or structural scattering. This paper discusses that facility, its reactors, and some chosen experiments from the list of many that were performed at that facility during the past 38 years

  20. Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

    Energy Technology Data Exchange (ETDEWEB)

    Ware, A.G.; Longhurst, G.R.

    1981-12-01

    This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available.

  1. Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

    International Nuclear Information System (INIS)

    This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available

  2. The Creation Of A Formal Test Flying System Within The British Microlight Aircraft Association And A Discussion Of The Spin Testing Of Microlight Aircraft

    OpenAIRE

    Gratton, GB; Porteous, TC

    2000-01-01

    The UK microlight aircraft community, under the guidance of the British Microlight Aircraft Association (BMAA), has developed a formalised system for the training and qualification of civil test pilots on this class of aircraft. This system is unique in Britain where most of the rest of the industry relies upon a pool of military-trained test aircrew, most of whom have no experience of microlight aircraft. This paper describes the system operated by the BMAA for the training and qualifica...

  3. Developmental testing of partially volatile neutron shields for high-performance shipping casks

    International Nuclear Information System (INIS)

    Results of the phase one tests have demonstrated that the neutron-shielding concept described in this paper is a viable design option for spent fuel shipping casks. The tests have shown that the Boro-silicone 236 shield is superior to the other shield materials considered. Repeated TGA, aging and fire tests demonstrated the reliability of the data. A second phase of the test program is now being pursued where the Boro-silicone 236 is injected into all-steel slab sections, and cured in place. 5 tables

  4. Acoustic and Thermal Testing of an Integrated Multilayer Insulation and Broad Area Cooling Shield System

    Science.gov (United States)

    Wood, Jessica J.; Foster, Lee W.

    2013-01-01

    A Multilayer Insulation (MLI) and Broad Area Cooling (BAC) shield thermal control system shows promise for long-duration storage of cryogenic propellant. The NASA Cryogenic Propellant Storage and Transfer (CPST) project is investigating the thermal and structural performance of this tank-applied integrated system. The MLI/BAC Shield Acoustic and Thermal Test was performed to evaluate the MLI/BAC shield's structural performance by subjecting it to worst-case launch acoustic loads. Identical thermal tests using Liquid Nitrogen (LN2) were performed before and after the acoustic test. The data from these tests was compared to determine if any degradation occurred in the thermal performance of the system as a result of exposure to the acoustic loads. The thermal test series consisted of two primary components: a passive boil-off test to evaluate the MLI performance and an active cooling test to evaluate the integrated MLI/BAC shield system with chilled vapor circulating through the BAC shield tubes. The acoustic test used loads closely matching the worst-case envelope of all launch vehicles currently under consideration for CPST. Acoustic test results yielded reasonable responses for the given load. The thermal test matrix was completed prior to the acoustic test and successfully repeated after the acoustic test. Data was compared and yielded near identical results, indicating that the MLI/BAC shield configuration tested in this series is an option for structurally implementing this thermal control system concept.

  5. ZZ ABBN, 26 Group Cross-Sections and Self Shielding Factors for Fast Reactors

    International Nuclear Information System (INIS)

    1 - Description of program or function: Format: special format; Number of groups: 26 group X-section and resonance self-shielding factor library. Nuclides: H, D, Li-6, Li-7, Be, B-10, B-11, C, N, O, Na, Mg, Al, Si, K, Ca, Ti, V, Cr, Fe, Ni, Cu, Zr, Nb, Mo, Ta, W, Re, Pb, Bi, Th-232, U-233, U-234, U-235, U-236, U-238, Pu-239, Pu-240, Pu-241, Pu-242, FP-U-233, FP-U-235, FP-Pu-239. Origin: Multiple experimental sources; Weighting spectrum: yes; 26 group cross section and resonance self-shielding factor library for the following materials: H, D, Li-6, Li-7, Be, B-10, B-11, C, N, O, Na, Mg, Al, Si, K, Ca, Ti, V, Cr, Fe, Ni, Cu, Zr, Nb, Mo, Ta, W, Re, Pb, Bi, Th-232, U-233, U-234, U-235, U-236, U-238, Pu-239, Pu-240, Pu-241, Pu-242, FP-U-233, FP-U-235, FP-Pu-239. 2 - Restrictions on the complexity of the problem: This group cross section library has been developed for fast and intermediate reactors

  6. ITER [International Thermonuclear Experimental Reactor] shield and blanket work package report

    International Nuclear Information System (INIS)

    This report summarizes nuclear-related work in support of the US effort for the International Thermonuclear Experimental Reactor (ITER) Study. The purpose of this work was to prepare for the first international ITER workshop devoted to defining a basic ITER concept that will serve as a basis for an indepth conceptual design activity over the next 2-1/2 years. Primary tasks carried out during the past year included: design improvements of the inboard shield developed for the TIBER concept, scoping studies of a variety of tritium breeding blanket options, development of necessary design guidelines and evaluation criteria for the blanket options, further safety considerations related to nuclear components and issues regarding structural materials for an ITER device. 44 refs., 31 figs., 29 tabs

  7. Activation of the biological shield of the shut-down Gundremmingen block A reactor

    International Nuclear Information System (INIS)

    For the dismantling planning of a nuclear reactor, it is important to know the depth of the activation of the biological shield. With an important sampling and measurement program to support activity computer calculations, data have been obtained and hypothesis defined to avoid in the future high-cost measurement program. Measurement results agree with calculations. Some provisional results have been used as well to correct measurement results, doing new measurements, as to correct enter data, more particularly for what concerns the weight proportions. It is shown that a calculation of the activity in the median plane of the core is sufficient to determine the field from which concrete is only weakly activated. For the A-block of the RWE-Bayerwerk nuclear power plant, this field is before the external layer (primary concrete). Only, the inner (secondary) concrete is activated, separated from the first one by a layer of styropore

  8. Calculation of neutron and gamma ray energy spectra for fusion reactor shield design: comparison with experiment

    International Nuclear Information System (INIS)

    Integral experiments that measure the transport of approx. 14 MeV D-T neutrons through laminated slabs of proposed fusion reactor shield materials have been carried out. Measured and calculated neutron and gamma ray energy spectra are compared as a function of the thickness and composition of stainless steel type 304, borated polyethylene, and Hevimet (a tungsten alloy), and as a function of detector position behind these materials. The measured data were obtained using a NE-213 liquid scintillator using pulse-shape discrimination methods to resolve neutron and gamma ray pulse height data and spectral unfolding methods to convert these data to energy spectra. The calculated data were obtained using two-dimensional discrete ordinates radiation transport methods in a complex calculational network that takes into account the energy-angle dependence of the D-T neutrons and the nonphysical anomalies of the S/sub n/ method

  9. Production of an economic high-density concrete for shielding megavoltage radiotherapy rooms and nuclear reactors

    International Nuclear Information System (INIS)

    In megavoltage radiotherapy rooms, ordinary concrete is usually used due to its low construction costs, although higher density concrete are sometimes used, as well. The use of high-density concrete decreases the required thickness of the concrete barrier; hence, its disadvantage is its high cost. In a nuclear reactor, neutron radiation is the most difficult to shield. A method for production of economic high-density concrete witt, appropriate engineering properties would be very useful. Materials and Methods: Galena (Pb S) mineral was used to produce of a high-density concrete. Galena can be found in many parts of Iran. Two types of concrete mixes were produced. The water-to-concrete (w/c) ratios of the reference and galena concrete mixes were 0.53 and 0.25, respectively. To measure the gamma radiation attenuation of Galena concrete samples, they were exposed to a narrow beam of gamma rays emitted from a cobalt-60 therapy unit. Results: The Galena mineral used in this study had a density of 7400 kg/m3. The concrete samples had a density of 4800 kg/m3. The measured half value layer thickness of the Galena concrete samples for cobalt 60 gamma rays was much less than that of ordinary concrete (2.6 cm compared to 6.0 cm). Furthermore, the galena concrete samples had significantly higher compressive strength (500 kg/cm2 compared to 300 kg/cm2). Conclusion: The Galena concrete samples made in our laboratories had showed good shielding/engineering properties in comparison with all samples made by using high-density materials other than depleted uranium. Based on the preliminary results, Galena concrete is maybe a suitable option where high-density concrete is required in megavoltage radiotherapy rooms as well as nuclear reactors

  10. Simulation of radiation dose distribution and thermal analysis for the bulk shielding of an optimized molten salt reactor

    Institute of Scientific and Technical Information of China (English)

    张志宏; 夏晓彬; 蔡军; 王建华; 李长园; 葛良全; 张庆贤

    2015-01-01

    The Chinese Academy of Science has launched a thorium-based molten-salt reactor (TMSR) research project with a mission to research and develop a fission energy system of the fourth generation. The TMSR project intends to construct a liquid fuel molten-salt reactor (TMSR-LF), which uses fluoride salt as both the fuel and coolant, and a solid fuel molten-salt reactor (TMSR-SF), which uses fluoride salt as coolant and TRISO fuel. An optimized 2 MWth TMSR-LF has been designed to solve major technological challenges in the Th-U fuel cycle. Preliminary conceptual shielding design has also been performed to develop bulk shielding. In this study, the radiation dose and temperature distribution of the shielding bulk due to the core were simulated and analyzed by performing Monte Carlo simulations and computational fluid dynamics (CFD) analysis. The MCNP calculated dose rate and neutron and gamma spectra indicate that the total dose rate due to the core at the external surface of the concrete wall was 1.91 µSv/h in the radial direction, 1.16 µSv/h above and 1.33 µSv/h below the bulk shielding. All the radiation dose rates due to the core were below the design criteria. Thermal analysis results show that the temperature at the outermost surface of the bulk shielding was 333.86 K, which was below the required limit value. The results indicate that the designed bulk shielding satisfies the radiation shielding requirements for the 2 MWth TMSR-LF.

  11. Damage evaluation of 500 MWe Indian Pressurized Heavy Water Reactor nuclear containment for aircraft impact

    International Nuclear Information System (INIS)

    Safety assessment of Indian nuclear containments has been carried out for aircraft impact. The loading time history for Boeing and Airbus categories of aircrafts is generated based on the principle of momentum transfer of crushable aircrafts. The case studies include the analysis of BWR Mark III containment as a benchmark problem and analyses of Pressurised Heavy Water Reactor containment (inner and outer containment) for impulsive loading due to aircraft impact. Initially, the load is applied on outer containment wall model and subsequently the load is transferred to inner containment after the local perforation of the outer containment wall is noticed in the transient simulation. The analysis methodology evolved in the present work would be useful for studying the behavior of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due to the commercial aircrafts operated in India

  12. Erosion Testing of Coatings for V-22 Aircraft Applications

    Directory of Open Access Journals (Sweden)

    G. Y. Richardson

    2003-01-01

    Full Text Available High-velocity (183 m/sec sand erosion tests in a wind tunnel were conducted to evaluate developmental coatings from three separate companies under funding by the Navy's phase I small business innovative research program. The purpose of the coatings was to address a particular problem the V-22 tilt-rotor aircraft (Osprey was having with regard to ingestion of sand particles by a titanium impeller that was associated with the aircraft's environmental control system. The three coatings that were deposited on titanium substrates and erosion-tested included (1 SixCy/DLC multilayers deposited by chemical vapor deposition (CVD; (2 WC/TaC/TiC processed by electrospark deposition; and (3 polymer ceramic mixtures applied by means of an aqueous synthesis. The erosion test results are presented; they provided the basis for assessing the suitability of some of these coatings for the intended application.

  13. Shielding designs and tests of a new exclusive ship for transporting spent nuclear fuels

    International Nuclear Information System (INIS)

    The Rokuei-Maru, a ship built specially for the transport of spent nuclear fuels in casks, was launched April in 1996. She is the first ship to comply with special Japanese regulations, KAISA 520, based on the INF code. DOT3.5 and MCNP-4A were used for the evaluation of dose equivalent rates of her shielding structures. On-board gamma-ray shielding tests were executed to confirm the effectiveness of the ship's shielding performance. The tests confirmed that effective shielding has been achieved and the dose equivalent rate in the accommodation and other inhabited spaces is sufficiently lower than the regulated limitations. This was achieved by employing the appropriate calculation methods and shielding materials. (author)

  14. Removal of the Plutonium Recycle Test Reactor - 13031

    International Nuclear Information System (INIS)

    The 309 Facility housed the Plutonium Recycle Test Reactor (PRTR), an operating test reactor in the 300 Area at Hanford, Washington. The reactor first went critical in 1960 and was originally used for experiments under the Hanford Site Plutonium Fuels Utilization Program. The facility was decontaminated and decommissioned in 1988-1989, and the facility was deactivated in 1994. The 309 facility was added to Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) response actions as established in an Interim Record of Decision (IROD) and Action Memorandum (AM). The IROD directs a remedial action for the 309 facility, associated waste sites, associated underground piping and contaminated soils resulting from past unplanned releases. The AM directs a removal action through physical demolition of the facility, including removal of the reactor. Both CERCLA actions are implemented in accordance with U.S. EPA approved Remedial Action Work Plan, and the Remedial Design Report / Remedial Action Report associated with the Hanford 300-FF-2 Operable Unit. The selected method for remedy was to conventionally demolish above grade structures including the easily distinguished containment vessel dome, remove the PRTR and a minimum of 300 mm (12 in) of shielding as a single 560 Ton unit, and conventionally demolish the below grade structure. Initial sample core drilling in the Bio-Shield for radiological surveys showed evidence that the Bio-Shield was of sound structure. Core drills for the separation process of the PRTR from the 309 structure began at the deck level and revealed substantial thermal degradation of at least the top 1.2 m (4LF) of Bio-Shield structure. The degraded structure combined with the original materials used in the Bio-Shield would not allow for a stable structure to be extracted. The water used in the core drilling process proved to erode the sand mixture of the Bio-Shield leaving the steel aggregate to act as ball bearings against the

  15. Experimental investigation on time-domain testing of shielding effectiveness of composite materials against EMP

    International Nuclear Information System (INIS)

    The shielding effectiveness (SE) of materials with different thickness is experimentally investigated using an improved time-domain testing system of SE and a window on a shielding enclosure in a GTEM cell. Experimental results show that the proper test position of transient electric probe (E-probe) within the enclosure is determined to make the system capable of finding out the maximum peak value of pulsed electric field inside enclosure. The SE test experiments for several shielding composite materials fabricated using nickel-plated carbon fiber proved that the SE of materials for electromagnetic pulse (EMP) is highly dependent on the thickness of materials and the proposed testing method can obtain high repeatability, and thus it can be used as a standard test system to assess the shielding effectiveness of materials for pulsed electric field.

  16. The design and testing of subscale smart aircraft wing bolts

    International Nuclear Information System (INIS)

    Presently costly periodic inspection is vital in guaranteeing the structural integrity of aircraft. This investigation assesses the potential for significantly reducing aircraft maintenance costs without modification of aircraft structures by implementing smart wing bolts, manufactured from TRIP steel, which can be monitored for damage in situ. TRIP steels undergo a transformation from paramagnetic austenite to ferromagnetic martensite during deformation. Subscale smart aircraft wing bolts were manufactured from hot rolled TRIP steel. These wing bolts were used to demonstrate that washers incorporating embedded inductance coils can be utilized to measure the martensitic transformation occurring in the TRIP steel during bolt deformation. Early in situ warning of a critical bolt stress level was thereby facilitated, potentially reducing the costly requirement for periodic wing bolt removal and inspection. The hot rolled TRIP steels that were utilized in these subscale bolts do not however exhibit the mechanical properties required of wing bolt material. Thus warm rolled TRIP steel alloys were also investigated. The mechanical properties of the best warm rolled TRIP steel alloy tested almost matched those of AISI 4340. The warm rolled alloys were also shown to exhibit transformation before yield, allowing for earlier warning when overload occurs. Further work will be required relating to fatigue crack detection, environmental temperature fluctuation and more thorough material characterization. However, present results show that in situ early detection of wing bolt overload is feasible via the use of high alloy warm rolled TRIP steel wing bolts in combination with inductive sensor embedded washers. (paper)

  17. The design and testing of subscale smart aircraft wing bolts

    Science.gov (United States)

    Vugampore, J. M. V.; Bemont, C.

    2012-07-01

    Presently costly periodic inspection is vital in guaranteeing the structural integrity of aircraft. This investigation assesses the potential for significantly reducing aircraft maintenance costs without modification of aircraft structures by implementing smart wing bolts, manufactured from TRIP steel, which can be monitored for damage in situ. TRIP steels undergo a transformation from paramagnetic austenite to ferromagnetic martensite during deformation. Subscale smart aircraft wing bolts were manufactured from hot rolled TRIP steel. These wing bolts were used to demonstrate that washers incorporating embedded inductance coils can be utilized to measure the martensitic transformation occurring in the TRIP steel during bolt deformation. Early in situ warning of a critical bolt stress level was thereby facilitated, potentially reducing the costly requirement for periodic wing bolt removal and inspection. The hot rolled TRIP steels that were utilized in these subscale bolts do not however exhibit the mechanical properties required of wing bolt material. Thus warm rolled TRIP steel alloys were also investigated. The mechanical properties of the best warm rolled TRIP steel alloy tested almost matched those of AISI 4340. The warm rolled alloys were also shown to exhibit transformation before yield, allowing for earlier warning when overload occurs. Further work will be required relating to fatigue crack detection, environmental temperature fluctuation and more thorough material characterization. However, present results show that in situ early detection of wing bolt overload is feasible via the use of high alloy warm rolled TRIP steel wing bolts in combination with inductive sensor embedded washers.

  18. Integral transport solutions to multi-dimensional shielding and reactor problems

    International Nuclear Information System (INIS)

    The so-called ''singularity-subtracting'' technique, which relieves the singularity usually present in the integral transport equation and allows direct application of the Gaussian quadrature formula, has been shown to be very accurate and effective in solving shielding problems. In the work, it is combined with the iteration method and extended to explore multi-dimensional reactor problems. In addition, a general method was developed for evaluating the additional integral resulted from singularity subtraction, especially on 2-D and 3-D problems. One-dimensional and two-dimensional problems were defined and solved by the proposed method. The accuracy of these results was verified by comparing them with those obtained from the discrete ordinates method (S/sub n/), or Monte-Carlo method (MORSE). A scheme of approximation was designed for the 3-D calculation. When the flux at some point of interest is being calculated, the integrals are evaluated by integrating over only the vicinity region, which has essential influence on that point, instead of integrating over the entire region. By so doing, a great savings in the computing time can be expected. Application of this approximation to a 3-D reactor problem has shown a 10-fold reduction in the computing time while the accuracy is sacrificed by less than 1%. Most calculations in this work were based on two-energy group and isotropic scattering assumptions

  19. 78 FR 18932 - Public Meeting: Unmanned Aircraft Systems Test Site Program; Privacy Approach

    Science.gov (United States)

    2013-03-28

    ... privacy policy approach for the unmanned aircraft systems (UAS) test site program. The FAA is seeking the... operation of unmanned aircraft systems within the test site program (78 FR 12259). The proposed privacy... Federal Aviation Administration 14 CFR Part 91 Public Meeting: Unmanned Aircraft Systems Test Site......

  20. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    International Nuclear Information System (INIS)

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior

  1. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    Energy Technology Data Exchange (ETDEWEB)

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior.

  2. Reduction of the external dose rate of a research nuclear reactor shielding utilizing concrete with higher incorporated water quantity

    International Nuclear Information System (INIS)

    The ordinary concrete used in constructions hold about 4 wt% to 6 wt% of water content in its mixtures under two distinct fractions: free water fraction and chemically constituted water fraction. Studies have been showing that is possible and convenient to increase 3 to 4 wt% the water content in the special concrete mixtures used in nuclear reactor shielding, increasing the free water fraction and yet considering an adequate concrete composition of the components in order not to compromise the structural resistance with the increased water to cement ratio. This is technically viable, because these structures have dimensions much superior than those imposed by structural calculations since they are determined as a function of the shielding design itself. The use of a concrete with larger quantity of incorporated water assures a more efficient neutron capture process in the foremost concrete shielding layers leaving a great portion of material thickness to attenuate the high-energy gamma radiation resulting from these neutron captures. This in turn reduces the dose rate in the external shielding wall satisfying the ALARA principle, and assures a less radioactive environment in the reactor surroundings, which is important for the procedures of radiological protection of the installation. This work evaluated the reduction of the dose rate in the outer wall of the shielding of a reactor type pool of 10 MWth due to variations of the water fraction content in concrete, in a range from 4 to 7 wt%. This variation range was chosen conservatively and took into account the free water fraction incorporated into an ordinary concrete and experimentally estimated through neutron radiography techniques carried using the irradiation channel of the Argonauta reactor installed at the Instituto de Engenharia Nuclear. The discrete ordinate code ANISN, the EURLIB4 library of neutrons and gammas, and heavy concrete of different densities were used in the calculations. (author)

  3. Reduction of the external dose rate of a research nuclear reactor shielding utilizing concrete with higher incorporated water quantity

    International Nuclear Information System (INIS)

    The ordinary concrete used in constructions hold about 4 wt % to 6 wt % of water content in its mixtures under two distinct fractions: free water fraction and chemically constituted water fraction. Studies have been showing that is possible and convenient to increase 3 to 4 wt % the water content in the special concrete mixtures used in nuclear reactor shielding, increasing the free water fraction and yet considering an adequate concrete composition of the components in order not to compromise the structural resistance with the increased water to cement ratio. This is technically viable, because these structures have dimensions much superior than those imposed by structural calculations since they are determined as a function of the shielding design itself. The use of a concrete with larger quantity of incorporated water assures a more efficient neutron capture process in the foremost concrete shielding layers leaving a great portion of material thickness to attenuate the high energy gamma radiation resulting from these neutron captures. This in turn reduces the dose rate in the external shielding wall satisfying the ALARA principle, and assures a less radioactive environment in the reactor surroundings, which is important for the procedures of radiological protection of the installation. This work evaluated the reduction of the dose rate in the outer wall of the shielding of a reactor type pool of 10 M Wth due to variations of the water fraction content in concrete, in a range from 4 to 7 wt %. This variation range was chosen conservatively and took into account the free water fraction incorporated into an ordinary concrete and experimentally estimated through neutron radiography techniques carried using the irradiation channel of the Argonauta reactor installed at the Instituto de Engenharia Nuclear. The discrete ordinate code ANISN, the EURLIB4 library of neutrons and gammas, and heavy concrete of different densities were used in the calculations. (author)

  4. Material test reactor fuel research at the BR2 reactor

    International Nuclear Information System (INIS)

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

  5. Microprocessor based scanner for scanning of reactor shields (Paper No. 034)

    International Nuclear Information System (INIS)

    A microprocessor based scanner was developed to help the experimental physicists move either a gamma or neutron detector along the shielding surface which is in the inaccessible shielding corner at APSARA. (author). 4 figs

  6. Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  7. Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  8. Structural testing of concorde aircraft: Further report on United Kingdom tests

    Science.gov (United States)

    Harpur, N.

    1972-01-01

    A summary of tests conducted on the Concorde aircraft nacelle structure is presented. The tests were conducted as a part of the structural development and certification program. The nacelle structural specimens are described. The problems associated with the intake testing and engine-bay and nozzle testing are discussed.

  9. Early Test Facilities and Analytic Methods for Radiation Shielding

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, D.T.

    1992-01-01

    This report represents a compilation of eight papers presented at the 1992 American Nuclear Society/European Nuclear Society International Meeting held in Chicago, Illinois on November 15 20,1992. The meeting is of special significance since it commemorates the 50th anniversary of the first controlled nuclear chain reaction, which occurred, not coincidentally, in Chicago. The papers contained in this report were presented in a special session organized by the Radiation Protection and Shielding Division in keeping with the historical theme of the meeting.

  10. Hydramite II screening tests of potential bremsstrahlung converter debris shield materials

    International Nuclear Information System (INIS)

    Results of a brief test series aimed at screening a number of potential bremsstrahlung converter debris shield materials are reported. These tests were run on Sandia National Laboratories' Hydramite II accelerator using a diode configuration which produces a pinched electron beam. The materials tested include: (1) laminated Kevlar 49/polyester and E-glass/polyester composites, (2) a low density laminated Kevlar 49 composite, and (3) two types of through-the-thickness reinforced Kevlar 49 composites. As expected, tests using laminated Kevlar 49/polyester shields showed that shield permanent set (i.e., permanent deflection) increased with increasing tantalum conversion foil thickness and decreased with increasing shield thickness. The through-the-thickness reinforced composites developed localized, but severe, back surface damage. The laminated composites displayed little back surface damage, although extensive internal matrix cracking and ply delaminations were generated. Roughly the same degree of permanent set was produced in shields made from the low density Kevlar 49 composite and the Kevlar 49/polyester. The E-glass reinforced shields exhibited relatively low levels of permanent set

  11. Design and Testing of a Flight Control System for Unstable Subscale Aircraft

    OpenAIRE

    Sobron, Alejandro

    2015-01-01

    The primary objective of this thesis was to study, implement, and test low-cost electronic flight control systems (FCS) in remotely piloted subscale research aircraft with relaxed static longitudinal stability. Even though this implementation was carried out in small, simplified test-bed aircraft, it was designed with the aim of being installed later in more complex demonstrator aircraft such as the Generic Future Fighter concept demonstrator project. The recent boom of the unmanned aircraft ...

  12. A review of candidate ceramic materials for use as heat shield tiles in a supercritical-water-cooled-reactor

    International Nuclear Information System (INIS)

    The proposed Canadian supercritical-water-cooled reactor (SCWR) utilizes a reactor shell made of a zirconium alloy insulated with a ceramic tile heat shield. The main consideration in the selection of a tile material will be resistance to corrosion in supercritical water and long term microstructure stability, in addition to thermal conductivity. This paper provides a review of the literature on corrosion behaviours of ceramic materials in supercritical water and ranks candidate ceramic materials accordingly. Materials reviewed include alumina, zirconia, silica glasses, silicon carbide, silicon nitride, sialon, mullite, and aluminum nitride. (author)

  13. Test Facility for SMART Reactor Flow Distribution

    International Nuclear Information System (INIS)

    A Reactor Flow Distribution Test Facilities for SMART, named SCOP (SMART Core Flow and Pressure Test Facility), were designed in order to simulate the distributions of (1) core flow and (2) reactor sectional flow resistance and flow rates. SCOP facility was designed based on the linear scaling law in order to preserve the flow characteristics of the prototype system, which are distributions of flow rate and pressure drop. The reduced scale was selected as a 1/5 of prototype length scale. The nominal flow condition was designed to be similar based on the velocity as that of the SMART reactor, which can minimize the flow distortion in the reduced scale of test facility by maintaining high Re number flow. Test facility includes fluid system, control/instrumentation system, data acquisition system, power system, which were designed to meet the requirement for each system. This report describes the details of the scaling and design features for the test facility

  14. NASA Hybrid Wing Aircraft Aeroacoustic Test Documentation Report

    Science.gov (United States)

    Heath, Stephanie L.; Brooks, Thomas F.; Hutcheson, Florence V.; Doty, Michael J.; Bahr, Christopher J.; Hoad, Danny; Becker, Lawrence; Humphreys, William M.; Burley, Casey L.; Stead, Dan; Pope, Dennis S.; Spalt, Taylor B.; Kuchta, Dennis H.; Plassman, Gerald E.; Moen, Jaye A.

    2016-01-01

    This report summarizes results of the Hybrid Wing Body (HWB) N2A-EXTE model aeroacoustic test. The N2A-EXTE model was tested in the NASA Langley 14- by 22-Foot Subsonic Tunnel (14x22 Tunnel) from September 12, 2012 until January 28, 2013 and was designated as test T598. This document contains the following main sections: Section 1 - Introduction, Section 2 - Main Personnel, Section 3 - Test Equipment, Section 4 - Data Acquisition Systems, Section 5 - Instrumentation and Calibration, Section 6 - Test Matrix, Section 7 - Data Processing, and Section 8 - Summary. Due to the amount of material to be documented, this HWB test documentation report does not cover analysis of acquired data, which is to be presented separately by the principal investigators. Also, no attempt was made to include preliminary risk reduction tests (such as Broadband Engine Noise Simulator and Compact Jet Engine Simulator characterization tests, shielding measurement technique studies, and speaker calibration method studies), which were performed in support of this HWB test. Separate reports containing these preliminary tests are referenced where applicable.

  15. Reliability tests for reactor internals replacement technology

    International Nuclear Information System (INIS)

    Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for replacing reactor internals, which was directed at preventive maintenance before damage and repair after damage for the aging degradation. The project has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995, and it follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the whole test plans and the test results for the BWR reactor internals replacement methods; core shroud, ICM housing, and CRD Housing and stub tube. The test results have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

  16. MFTF-α + T shield design

    International Nuclear Information System (INIS)

    MFTF-α+T is a DT upgrade option of the Tandem Mirror Fusion Test Facility (MFTF-B) to study better plasma performance, and test tritium breeding blankets in an actual fusion reactor environment. The central cell insert, designated DT axicell, has a 2-MW/m2 neutron wall loading at the first wall for blanket testing. This upgrade is completely shielded to protect the reactor components, the workers, and the general public from the radiation environment during operation and after shutdown. The shield design for this upgrade is the subject of this paper including the design criteria and the tradeoff studies to reduce the shield cost

  17. Radiation shielding calculations for MuCool Test Area at Fermilab

    CERN Document Server

    Rakhno, I

    2004-01-01

    The MuCool Test Area (MTA) is an intense primary beam facility derived directly from the Fermilab Linac to test heat deposition and other technical concerns associated with the liquid hydrogen targets being developed for cooling intense muon beams. In this shielding study the results of Monte Carlo radiation shielding calculations performed using the MARS14 code for the MuCool Test Area and including the downstream portion of the target hall and berm around it, access pit, service building, and parking lot are presented and discussed within the context of the proposed MTA experimental configuration.

  18. Radiation shielding calculations for MuCool test area at Fermilab

    Energy Technology Data Exchange (ETDEWEB)

    Igor Rakhno; Carol Johnstone

    2004-05-26

    The MuCool Test Area (MTA) is an intense primary beam facility derived directly from the Fermilab Linac to test heat deposition and other technical concerns associated with the liquid hydrogen targets being developed for cooling intense muon beams. In this shielding study the results of Monte Carlo radiation shielding calculations performed using the MARS14 code for the MuCool Test Area and including the downstream portion of the target hall and berm around it, access pit, service building, and parking lot are presented and discussed within the context of the proposed MTA experimental configuration.

  19. Startup testing of Romania dual-core test reactor

    International Nuclear Information System (INIS)

    Late in 1979 both the Annular Core Pulsed Reactor (ACPR) and the 14-MW steady-state reactor (SSR) were loaded to critical. The fuel loading in both was then carried to completion and low-power testing was conducted. Early in 1980 both reactors successfully underwent high-power testing. The ACPR was operated for several hours at 500 kW and underwent pulse tests culminating in pulses with reactivity insertions of $4.60, peak power levels of about 20,000 MW, energy releases of 100 MW-sec, and peak measured fuel temperatures of 830 deg. C. The SSR was operated in several modes, both with natural convection and forced cooling with one or more pumps. The reactor successfully completed a 120-hr full-power test. Subsequent fuel element inspections confirmed that the fuel has performed without fuel damage or distortion. (author)

  20. Composite Structures Materials Testing for the Orion Crew Vehicle Heat Shield

    Science.gov (United States)

    Khemani, Farah N.

    2011-01-01

    As research is being performed for the new heat shield for the Orion capsule, National Aeronautics and Space Administration (NASA) is developing the first composite heat shield. As an intern of the Structures Branch in the Engineering Directorate (ES 2), my main task was to set up a test plan to determine the material properties of the honeycomb that will be used on the Orion Crew Module heat shield to verify that the composite is suitable for the capsule. Before conducting composite shell tests, which are performed to simulate the crush performance of the heat shield on the capsule, it is necessary to determine the compression and shear properties of the composite used on the shell. During this internship, I was responsible for developing a test plan, designing parts for the test fixtures as well as getting them fabricated for the honeycomb shear and compression testing. This involved work in Pro/Engineer as well as coordinating with Fab Express, the Building 9 Composite Shop and the Structures Test Laboratory (STL). The research and work executed for this project will be used for composite sandwich panel testing in the future as well. As a part of the Structures Branch, my main focus was to research composite structures. This involves system engineering and integration (SE&I) integration, manufacturing, and preliminary testing. The procedures for these projects that were executed during this internship included design work, conducting tests and performing analysis.

  1. Safety assessment of A92 reactor building for large commercial aircraft crash

    International Nuclear Information System (INIS)

    The current paper presents key elements of the comprehensive analyses of the effects due to a large aircraft collision with the reactor building of Belene NPP in Bulgaria. The reactor building is a VVER A92; it belongs to the third+ generation and includes structural measures for protection against an aircraft impact as standard design. The A92 reactor building implements a double shell concept and is composed of thick RC external walls and an external shell which surrounds an internal pre-stressed containment and the internal walls of the auxiliary building. The malevolent large aircraft impact is considered as a beyond design base accident (Design Extended Conditions, DEC). The main issues under consideration are the structural integrity, the equipment safety due to the induced vibrations, and the fire safety of the entire installation. Many impact scenarios are analyzed varying both impact locations and loading intensity. A large number of non-linear dynamic analyses are used for assessment of the structural response and capacity, including different type of structural models, different finite element codes, and different material laws. The corresponding impact loadings are represented by load time functions calculated according to three different approaches, i.e. loading determined by Riera's method (Riera, 1968), load time function calculated by finite element analysis (Henkel and Klein, 2007), and coupled dynamic analysis with dynamic interaction between target and projectile. Based on the numerical results and engineering assessments the capacity of the A92 reactor building to resist a malevolent impact of a large aircraft is evaluated. Significant efforts are spent on safety assessment of equipment by using an evaluation procedure based on damage indicating parameters. As a result of these analyses several design modifications of structure elements are performed. There are changes of the layout of reinforcement, special arrangements and spatial

  2. Measurement of dose rate profile and spectra through a cylindrical duct vis-a-vis Monte Carlo simulation studies for optimisation of reactor shield design

    International Nuclear Information System (INIS)

    In the design of a nuclear reactor, penetrations are provided in the top shield to carry out some essential operations. Radiation streaming is envisaged through such penetrations. To avoid radiation streaming, complementary shielding is provided. Optimisation of complementary shielding is carried out by performing calculations using MCNP code. Uncertainties in the calculations are taken care of by incorporating a safety factor. The assumption of the safety factor, while designing the reactor shielding, has been validated by undertaking experimental measurements on a similar geometry vis-a-vis the computed values obtained using MCNP code. The results of the present work agree with the safety factor of two assumed during the shield design. The details of gamma spectral measurements carried out with high purity germanium detector to understand the pattern of the scattered spectrum are also presented

  3. Reactor group constants and benchmark test

    International Nuclear Information System (INIS)

    The evaluated nuclear data files such as JENDL, ENDF/B-VI and JEF-2 are validated by analyzing critical mock-up experiments for various type reactors and assessing applicability for nuclear characteristics such as criticality, reaction rates, reactivities, etc. This is called Benchmark Testing. In the nuclear calculations, the diffusion and transport codes use the group constant library which is generated by processing the nuclear data files. In this paper, the calculation methods of the reactor group constants and benchmark test are described. Finally, a new group constants scheme is proposed. (author)

  4. Operating experience of Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt / 13.2 MWe sodium cooled, loop type mixed carbide fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors and to serve as an irradiation facility for development of fuel and structural material for future fast reactors. The reactor achieved first criticality in October 1985 with small indigenously designed and fabricated Mark I core (70% PuC-30% UC). The reactor power was subsequently raised in steps to 17.4 MWt by addition of Mark II fuel subassemblies (55% PuC-45% UC) and with the Mark I fuel operating at the designed linear heat rating of 400 W/cm. The turbo-generator was synchronized with the grid in July 1997. The achieved peak burn-up is 137 000 MWd/t so far without any fuel-clad failure. Presently the reactor is being operated at a nominal power of 15.7 MWt for irradiation of a test fuel subassembly of the Prototype Fast Breeder Reactor, which is coming up at Kalpakkam. It is also planned to irradiate test subassemblies made of metallic fuel for future fast reactor program. Being a small reactor, all feed back coefficients of reactivity including void coefficient are negative and hence the reactor is inherently safe. This was confirmed by carrying out physics tests. The capability to remove decay heat under various incidental conditions including natural convection was demonstrated by carrying out engineering tests. Thermo couples are provided for on-line monitoring of fuel SA outlet temperature by dedicated real time computer and processed to generate trip signals for the reactor in case of power excursion, increase in clad hot spot temperature and subassembly flow blockage. All pipelines and capacities in primary main circuit are provided with segmented outer envelope to minimize and contain radioactive sodium leak while ensuring forced cooling through reactor to remove decay heat in case of failure of primary boundary. In secondary circuit, provision is

  5. Shielding research at the Hanford Site

    International Nuclear Information System (INIS)

    The original three plutonium production reactors (B, D, and F) constructed at the Hanford Site in 1943--1944 had shields consisting of alternate layers of iron and a high-density pressed-wood product called Masonite *. This design was the engineering response to the scientific request for a mixture of iron and hydrogen. The design mix was based on earlier studies using iron and water or iron and paraffin; however, these materials did not have satisfactory structural characteristics. Although the shields performed satisfactorily, the fabrication cost was high. Each piece had to be machined precisely to fit within structural webs, so as not to introduce cracks through the shield. Before 1950, two additional reactors (DR and H) were built using the same shield design. At the request of R.L. Dickeman, an experimental facility was included in the top of the DR Reactor to permit evaluation of shield materials. Concurrent with the measurement of attenuation properties of materials in this facility, a program was undertaken to investigate the structural characteristics of various high-density Portland cement concretes. This research effort continued for over a decade, and led to the use of these concretes in subsequent reactor shields at the Hanford Site and elsewhere with significant savings in construction costs. Completion of the attenuation and structural measurements on the various high-density concretes provided a database that could be used in the design of shields for new reactors. At the Hanford Site, the top shield of the C Reactor was constructed of concrete, whereas the sides were constructed of iron-Masonite. As more and more data were acquired, the later rectors, KE, KW, and NPR, had shields of various tested concretes. Using concrete in these shields materially reduced the cost of the facilities. Additionally, studies on heat damage to the masonite resulted in changes that permitted increases in production, while at the same time maintaining shield integrity

  6. Large Field Photogrammetry Techniques in Aircraft and Spacecraft Impact Testing

    Science.gov (United States)

    Littell, Justin D.

    2010-01-01

    The Landing and Impact Research Facility (LandIR) at NASA Langley Research Center is a 240 ft. high A-frame structure which is used for full-scale crash testing of aircraft and rotorcraft vehicles. Because the LandIR provides a unique capability to introduce impact velocities in the forward and vertical directions, it is also serving as the facility for landing tests on full-scale and sub-scale Orion spacecraft mass simulators. Recently, a three-dimensional photogrammetry system was acquired to assist with the gathering of vehicle flight data before, throughout and after the impact. This data provides the basis for the post-test analysis and data reduction. Experimental setups for pendulum swing tests on vehicles having both forward and vertical velocities can extend to 50 x 50 x 50 foot cubes, while weather, vehicle geometry, and other constraints make each experimental setup unique to each test. This paper will discuss the specific calibration techniques for large fields of views, camera and lens selection, data processing, as well as best practice techniques learned from using the large field of view photogrammetry on a multitude of crash and landing test scenarios unique to the LandIR.

  7. Nuclear Safety Research Reactor (NSRR) as a facility for reactor safety research and its modification for the future test plan

    International Nuclear Information System (INIS)

    The NSRR is a modified TRIGA-ACPR (annular core pulse reactor), and attained the initial criticality in May, 1975. It was built for studying reactor fuel behavior under a reactivity-initiated accident condition. The reactor is installed in a pool of 3.6 m width, 4.5 m length and 9 m depth, and water above the reactor core serves as a radiation shield. The reactor core contains 149 driver fuel rods, 6 regulating rods, 2 safety rods and 3 transient rods. An arbitrary reactivity up to 4.67 $ can be set up almost instantaneously in the reactor core. The pulse power generation is terminated by the large negative reactivity induced by prompt temperature feedback without inserting the control rods. This is brought about by an excellent property of the driver fuel which contains 12 wt.% U-ZrH enriched to 20 wt.% U-235. As a unique feature, the NSRR is equipped with a big experimental cavity through the center of the reactor core. It has the diameter of 220 mm, and is called loading tube. It is branched into a vertical loading tube and an offset loading tube. The characteristics of the pulse operation in the NSRR, the outline of fuel irradiation experiment, the future test plan and the modification of the NSRR are described. (Kako, I.)

  8. The forced vibration test and observation of earthquake response at PWR (3loop) reactor building

    International Nuclear Information System (INIS)

    At SN nuclear power plant, the dynamic properties of reactor building have been studied by forced vibration test and earthquake measurement in order to obtain practical data for the confirmation of the seismic safety of the power plant facilities and for more rational earthquake resistant design. The forced vibration test was carried out from January to February in 1983, when the reactor building and other surrounding buildings had been almost completed. It was aimed at the investigation of the translational and vertical vibration mode (beam and vertical mode) of outer shield building and inner concrete structure as well as the oval vibration mode (oval mode) of outer shield building. The earthquake response were observed three times in the next year 1984. They were April 17, August 7, and August 15. The observed results were simulated by analyses. (orig.)

  9. The forced vibration test and observation of earthquake response at PWR (3 loop) reactor building

    International Nuclear Information System (INIS)

    At SN nuclear power plant, the dynamic properties of reactor building have been studied by forced vibration test and earthquake measurement in order to obtain practical data for the confirmation of the seismic safety of the power plant facilities and for more rational earthquake resistant design. The forced vibration test was carried out from January to February in 1983, when the reactor building and other surrounding buildings had been almost completed. It is aimed at the investigation of the translational and vertical vibration mode (beam and vertical mode) of outer shield building and inner concrete structure as well as the oval vibration mode (oval mode) of outer shield building. The earthquake response were observed three times in the next year 1984. They were April 17, August 7, and August 15. The observed results were simulated by analyses

  10. Design, fabrication and testing results of vacuum vessel, thermal shield and cryostat of EAST

    International Nuclear Information System (INIS)

    The EAST (Experimental Advanced Superconducting Tokamak) is an advanced steady-state plasma physics experimental device, which has been approved by the Chinese government and being constructed as the Chinese national nuclear fusion Research project. The vacuum vessel as one of the key components for this device can provide ultra-high vacuum and cleanly location of plasma operation. It is a torus with 'D' shaped cross-section, double wall, upper vertical ports, lower vertical ports, horizontal ports and flexible supports. The cryostat is a large single walled vessel surrounding the entire basic machine with central cylindrical section and two end enclosures, a flat base structure with external reinforcements and dome-shaped lid structure. It provides the thermal barrier with the base pressure of 5x10-4 Pa between the ambient temperature testing hall and the liquid helium cooled superconducting magnet. The thermal shields comprise the vacuum vessel thermal shield (VVTS), between the vacuum vessel and the cold Toroidal Field (TF) coil structures, the cryostat thermal shield (CTS), covering the walls of the cryostat, thereby preventing direct line of sight of the room temperature walls to the cold structures, the vacuum port thermal shields (VPTS) that enclose the port connection ducts. This paper is a report of the structure design and stress analyses for the vacuum vessel, thermal shield and cryostat. And also some key R and D and testing results for these components have been presented. (author)

  11. BWR spent fuel storage cask performance test. Volume 2. Pre- and post-test decay heat, heat transfer, and shielding analyses

    International Nuclear Information System (INIS)

    This report describes the decay heat, heat transfer, and shielding analyses conducted in support of performance testing of a Ridhihalgh, Eggers and Associates REA 2033 boiling water reactor (BWR) spent fuel storage cask. The cask testing program was conducted for the US Department of Energy (DOE) Commercial Spent Fuel Management Program by the Pacific Northwest Laboratory (PNL) and by General Electric at the latters' Morris Operation (GE-MO) as reported in Volume I. The analyses effort consisted of performing pretest calculations to (1) select spent fuel for the test; (2) symmetrically load the spent fuel assemblies in the cask to ensure lateral symmetry of decay heat generation rates; (3) optimally locate temperature and dose rate instrumentation in the cask and spent fuel assemblies; and (4) evaluate the ORIGEN2 (decay heat), HYDRA and COBRA-SFS (heat transfer), and QAD and DOT (shielding) computer codes. The emphasis of this second volume is on the comparison of code predictions to experimental test data in support of the code evaluation process. Code evaluations were accomplished by comparing pretest (actually pre-look, since some predictions were not completed until testing was in progress) predictions with experimental cask testing data reported in Volume I. No attempt was made in this study to compare the two heat transfer codes because results of other evaluations have not been completed, and a comparison based on one data set may lead to erroneous conclusions

  12. Investigation on radiation shielding parameters of oxide dispersion strengthened steels used in high temperature nuclear reactor applications

    International Nuclear Information System (INIS)

    Highlights: • Radiation shielding properties of oxide dispersion strengthened (ODS) steels were investigated. • μ/ρ, Zeff, Nel and λ were calculated. • Investigation was carried to explore the shielding properties of ODS. - Abstract: Oxide dispersion strengthened (ODS) based steels are being investigated for reactor/nuclear applications. Therefore, a better understanding of the interaction of different gamma ray energies with ODS materials is required. We use WinXCom program to calculate the mass attenuation coefficient for different compositions of ODS at photon energies (1 keV–100 GeV). The attenuation coefficient data were then used to obtain the effective atomic numbers (Zeff), effective electron densities (Nel) and photon mean free path (λ) of the investigated materials. The investigation was carried out to explore the radiation attenuation parameters of these important materials

  13. Wind-tunnel tests of the XV-15 tilt rotor aircraft

    Science.gov (United States)

    Weiberg, J. A.; Maisel, M. D.

    1980-01-01

    The XV-15 aircraft was tested in the Ames 40 by 80 Foot Wind Tunnel for preliminary evaluation of aerodynamic and aeroelastic characteristics prior to flight. The tests were undertaken to investigate the aircraft performance, stability, control and structural loads for flight modes from helicopter through transition and airplane mode up to the tunnel capability of 170 knots. Results from these tests are presented.

  14. Specification and testing for power by wire aircraft

    Science.gov (United States)

    Hansen, Irving G.; Kenney, Barbara H.

    1993-01-01

    A power by wire aircraft is one in which all active functions other than propulsion are implemented electrically. Other nomenclature are 'all electric airplane,' or 'more electric airplane.' What is involved is the task of developing and certifying electrical equipment to replace existing hydraulics and pneumatics. When such functions, however, are primary flight controls which are implemented electrically, new requirements are imposed that were not anticipated by existing power system designs. Standards of particular impact are the requirements of ultra-high reliability, high peak transient bi-directional power flow, and immunity to electromagnetic interference and lightning. Not only must the electromagnetic immunity of the total system be verifiable, but box level tests and meaningful system models must be established to allow system evaluation. This paper discusses some of the problems, the system modifications involved, and early results in establishing wiring harness and interface susceptibility requirements.

  15. Specification and testing for power by wire aircraft

    Science.gov (United States)

    Hansen, Irving G.; Kenney, Barbara H.

    1993-08-01

    A power by wire aircraft is one in which all active functions other than propulsion are implemented electrically. Other nomenclature are 'all electric airplane,' or 'more electric airplane.' What is involved is the task of developing and certifying electrical equipment to replace existing hydraulics and pneumatics. When such functions, however, are primary flight controls which are implemented electrically, new requirements are imposed that were not anticipated by existing power system designs. Standards of particular impact are the requirements of ultra-high reliability, high peak transient bi-directional power flow, and immunity to electromagnetic interference and lightning. Not only must the electromagnetic immunity of the total system be verifiable, but box level tests and meaningful system models must be established to allow system evaluation. This paper discusses some of the problems, the system modifications involved, and early results in establishing wiring harness and interface susceptibility requirements.

  16. Present status of Japan materials testing reactor

    International Nuclear Information System (INIS)

    The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

  17. Implementation intentions and test anxiety: Shielding academic performance from distraction

    OpenAIRE

    Parks-Stamm, Elizabeth J.; Gollwitzer, Peter M.; Oettingen, Gabriele

    2010-01-01

    College students whose test anxiety was measured completed a working memory-intensive math exam with televised distractions. Students were provided with implementation intentions (if-then plans; Gollwitzer, 1999) designed to either help them ignore the distractions (i.e., temptation-inhibiting plans) or focus more intently on the math exam (i.e., task-facilitating plans). Regression analyses showed that as test anxiety increased, the effectiveness of temptation-inhibiting implementation inten...

  18. Improvement of numerical analytical model for temperature of primary biological shielding toward HTTR-LOFC test with VCS inactive

    International Nuclear Information System (INIS)

    Toward the loss of forced cooling (LOFC) test that is performed at HTTR, this study improved an axisymmetric numerical analysis model. The heat from the reactor pressure vessel (RPV) of HTTR is transferred to the concrete-made primary radiation shielding body that is positioned around the vessel cooling system (VCS), through the radiant heat transfer and natural convection from the fins between the water-cooled tubes adjacent to VCS, and through heat conduction from the metal support that supports VCS. This study aims to properly predict this hear transmission. In this numerical analysis model, the water-cooled tubes of VCS water-cooled panel are simulated as a horizontal arrangement, and the fins are simulated as the shape supported by the metal support. Numerical analysis results were in good agreement with the test results of the temperatures of each structure. From the numerical analysis results, it was clarified that the primary shielding body temperature locally rises in the vicinity of VCS support, but the support vicinity can be maintained at low temperatures. (A.O.)

  19. Effectiveness of Shield Termination Techniques Tested with TEM Cell and Bulk Current Injection

    Science.gov (United States)

    Bradley, Arthur T.; Hare, Richard J.

    2009-01-01

    This paper presents experimental results of the effectiveness of various shield termination techniques. Each termination technique is evaluated by two independent noise injection methods; transverse electromagnetic (TEM) cell operated from 3 MHz 400 MHz, and bulk current injection (BCI) operated from 50 kHz 400 MHz. Both single carrier and broadband injection tests were investigated. Recommendations as to how to achieve the best shield transfer impedance (i.e. reduced coupled noise) are made based on the empirical data. Finally, the noise injection techniques themselves are indirectly evaluated by comparing the results obtained from the TEM Cell to those from BCI.

  20. Heavy Water Components Test Reactor Decommissioning

    International Nuclear Information System (INIS)

    The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D and D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment

  1. FOCUS, Neutron Transport System for Complex Geometry Reactor Core and Shielding Problems by Monte-Carlo

    International Nuclear Information System (INIS)

    1 - Description of problem or function: FOCUS enables the calculation of any quantity related to neutron transport in reactor or shielding problems, but was especially designed to calculate differential quantities, such as point values at one or more of the space, energy, direction and time variables of quantities like neutron flux, detector response, reaction rate, etc. or averages of such quantities over a small volume of the phase space. Different types of problems can be treated: systems with a fixed neutron source which may be a mono-directional source located out- side the system, and Eigen function problems in which the neutron source distribution is given by the (unknown) fundamental mode Eigen function distribution. Using Monte Carlo methods complex 3- dimensional geometries and detailed cross section information can be treated. Cross section data are derived from ENDF/B, with anisotropic scattering and discrete or continuous inelastic scattering taken into account. Energy is treated as a continuous variable and time dependence may also be included. 2 - Method of solution: A transformed form of the adjoint Boltzmann equation in integral representation is solved for the space, energy, direction and time variables by Monte Carlo methods. Adjoint particles are defined with properties in some respects contrary to those of neutrons. Adjoint particle histories are constructed from which estimates are obtained of the desired quantity. Adjoint cross sections are defined with which the nuclide and reaction type are selected in a collision. The energy after a collision is selected from adjoint energy distributions calculated together with the adjoint cross sections in advance of the actual Monte Carlo calculation. For multiplying systems successive generations of adjoint particles are obtained which will die out for subcritical systems with a fixed neutron source and will be kept approximately stationary for Eigen function problems. Completely arbitrary problems can

  2. In-reactor testing of ionic thermometers

    International Nuclear Information System (INIS)

    Ionic thermometers have been tested in a nuclear reactor with attention to the steepness of the ionic conductivity jump and the influence of a glass container on the accuracy of the temperature measurements. It was found that, at the neutron fluxes up to 1.5 x 1018 m-2 s-1 (thermal) and 3 x 1018 m-2 s-1 (fast) in a light water reactor, the change of conductivity jump slope is negligible or nil for an ionic thermometer filled by HgI2, i.e., at 256.0 +- 0.2 0C. The need to use boron-free glass was confirmed. The impact on the accuracy of the temperature point indication in a nuclear reactor core is discussed, as well as obvious inertness of the melting process mechanism to the intense irradiation field

  3. Effects of drop testing on scale model shipping containers shielded with depleted uranium

    International Nuclear Information System (INIS)

    Three scale model shipping containers shielded with depleted uranium were dropped onto an essentially unyielding surface from various heights to determine their margins to failure. This report presents the results of a thorough posttest examination of the models to check for basic structural integrity, shielding integrity, and deformations. Because of unexpected behavior exhibited by the depleted uranium shielding, several tests were performed to further characterize its mechanical properties. Based on results of the investigations, recommendations are made for improved container design and for applying the results to full-scale containers. Even though the specimens incorporated specific design features, the results of this study are generally applicable to any container design using depleted uranium

  4. Neutronics model of the bulk shielding reactor (BSR): validation by comparison of calculations with the experimental measurements

    International Nuclear Information System (INIS)

    A neutronics model for the Oak Ridge National Laboratory Bulk Shielding Reactor (ORNL-SAR) was developed and verified by experimental measurements. A cross-section library was generated from the 218 group Master Library using the AMPX Block Code system. A series of one-, two-, and three-dimensional neutronics calculations were performed utilizing both transport and diffusion theory. Spectral comparison was made with 58Ni(n,p) reaction. The results of the comparison between the calculational model and other experimental measurements showed agreement within 10% and therefore the model was determined to be adequate for calculating the neutron fluence for future irradiation experiments in the ORNL-BSR

  5. Neutronics model of the bulk shielding reactor (BSR): validation by comparison of calculations with the experimental measurements

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J.O.; Miller, L.F.; Kam, F.B.K.

    1981-05-01

    A neutronics model for the Oak Ridge National Laboratory Bulk Shielding Reactor (ORNL-SAR) was developed and verified by experimental measurements. A cross-section library was generated from the 218 group Master Library using the AMPX Block Code system. A series of one-, two-, and three-dimensional neutronics calculations were performed utilizing both transport and diffusion theory. Spectral comparison was made with /sup 58/Ni(n,p) reaction. The results of the comparison between the calculational model and other experimental measurements showed agreement within 10% and therefore the model was determined to be adequate for calculating the neutron fluence for future irradiation experiments in the ORNL-BSR.

  6. Testing and Commissioning of a Multifunctional Tool for the Dismantling of the Activated Internals of the KNK Reactor Shaft - 13524

    International Nuclear Information System (INIS)

    The Compact Sodium Cooled Reactor Facility Karlsruhe (KNK), a prototype reactor to demonstrate the Fast Breeder Reactor Technology in Germany, was in operation from 1971 to 1991. The dismantling activities started in 1991. The project aim is the green field in 2020. Most of the reactor internals as well as the primary and secondary cooling loops are already dismantled. The total contaminated sodium inventory has already been disposed of. Only the high activated reactor vessel shielding structures are remaining. Due to the high dose rates these structures must be dismantled remotely. For the dismantling of the primary shielding of the reactor vessel, 12 stacked cast iron blocks with a total mass of 90 Mg and single masses up to 15.5 Mg, a remote-controlled multifunctional dismantling device (HWZ) was designed, manufactured and tested in a mock-up. After successful approval of the test sequences by the authorities, the HWZ was implemented into the reactor building containment for final assembling of the auxiliary equipment and subsequent hot commissioning in 2012. Dismantling of the primary shielding blocks is scheduled for early 2013. (authors)

  7. Testing and Commissioning of a Multifunctional Tool for the Dismantling of the Activated Internals of the KNK Reactor Shaft - 13524

    Energy Technology Data Exchange (ETDEWEB)

    Rothschmitt, Stefan; Graf, Anja [WAK Rueckbau- und Entsorgungs- GmbH, P.O.Box 12 63, 76339 Eggenstein-Leopoldshafen (Germany); Bauer, Stefan; Klute, Stefan; Koselowski, Eiko [Siempelkamp Nukleartechnik GmbH, Am Taubenfeld 25/1, 69123 Heidelberg (Germany); Hendrich, Klaus [Ingenieurbuero Hendrich, Moerikeweg 14, 75015 Bretten (Germany)

    2013-07-01

    The Compact Sodium Cooled Reactor Facility Karlsruhe (KNK), a prototype reactor to demonstrate the Fast Breeder Reactor Technology in Germany, was in operation from 1971 to 1991. The dismantling activities started in 1991. The project aim is the green field in 2020. Most of the reactor internals as well as the primary and secondary cooling loops are already dismantled. The total contaminated sodium inventory has already been disposed of. Only the high activated reactor vessel shielding structures are remaining. Due to the high dose rates these structures must be dismantled remotely. For the dismantling of the primary shielding of the reactor vessel, 12 stacked cast iron blocks with a total mass of 90 Mg and single masses up to 15.5 Mg, a remote-controlled multifunctional dismantling device (HWZ) was designed, manufactured and tested in a mock-up. After successful approval of the test sequences by the authorities, the HWZ was implemented into the reactor building containment for final assembling of the auxiliary equipment and subsequent hot commissioning in 2012. Dismantling of the primary shielding blocks is scheduled for early 2013. (authors)

  8. Implementation Intentions and Test Anxiety: Shielding Academic Performance from Distraction

    Science.gov (United States)

    Parks-Stamm, Elizabeth J.; Gollwitzer, Peter M.; Oettingen, Gabriele

    2010-01-01

    College students whose test anxiety was measured completed a working memory-intensive math exam with televised distractions. Students were provided with implementation intentions (if-then plans; Gollwitzer, 1999) designed to either help them ignore the distractions (i.e., temptation-inhibiting plans) or focus more intently on the math exam (i.e.,…

  9. High-Temperature Engineering Test Reactor door valve monitor system

    International Nuclear Information System (INIS)

    This manual describes the detector design features, performance, and operating characteristics of the High-Temperature Engineering Test Reactor (HTTR) Door Valve Monitor System spent-fuel monitor. The HTTR Door Valve Monitor System (HDVM) is installed in the HTTR door valve to provide unattended monitoring data for the transfer of spent fuel through the door valve on the top of the reactor. The system includes a pair of detectors to provide direction of travel and redundancy. The fission product gamma rays are measured using ion chambers (ICs) and the curium neutrons are measured using shielded 3He detectors. There are two ICs and one 3He tube inside each detector package. Gamma-ray and neutron detector (GRAND) electronics supply power to the ICs and 3He tubes, and the data are collected in the GRAND and the Field Works computer. The system is designed to operate unattended with data pickup by the inspectors on a 90-day period. This manual gives the performance and calibration procedures

  10. Self-shielding in the NET fusion reactor blanket and effects on uncertainty calculations

    International Nuclear Information System (INIS)

    In this report the results are presented of an analysis of a NET iron/water inboard shielding blanket using energy self-shielded cross sections. Coupled (n,γ) transport calculations have been performed in an S8P8 approximation with the code ANISN using cross sections in the 121-group GAM-II structure. Basic cross sections were obtained from the 217-group MAT175 library, which is based on JEF-1 and EFF-1. Energy self-shielding was taken into account using the Bondarenko method. The results of this analysis are compared with those obtained in report ECN-C--90-034, in which a similar analysis had been presented using infinite dilute cross sections. It is shown that the effect of energy self-shielding on the neutron flux in the coils of the NET design is considerable, whereas the γ-flux is hardly influenced. Also the effect of using energy self-shielded cross sections in sensitivity and uncertainty analyses was studied. In these analyses the sensitivity of the total nuclear heating in the innermost interval of the inboard coils to the total cross sections of Fe, Cr and Ni has been studied. The analyses have been performed using an ECN-modified version of the code SENSIT. Due to the effect of self-shielding not only the value of the response parameter changes (the total nuclear heating in the coil increases with 13%), but also the associated relative uncertainty (the relative uncertainty due to uncertainties in the total cross sections of Fe, Cr and Ni decreases with 8%; the absolute value of the uncertainty increases however). The conclusion is that for reliable calculations of the nuclear heating the effects of self-shielding should be taken into account; for the uncertainty estimates this is less important. (author). 15 refs.; 15 figs.; 6 tabs

  11. Fabrication, testing, and qualification of reactor graphites

    International Nuclear Information System (INIS)

    The work performed under the HBK project for development and testing of reactor graphites could have recourse to results and experience already gained in Great Britain, in the F.R.G., the USA, and the Netherlands. The specific problems to be tackled by the HBK project activities result from the particularly exacting requirements with regard to behaviour under irradiation that are to be met by the graphite reflector for the THTR follower plant. From a great number of candidate graphites, selected for testing and evaluation, the extensive irradiation experiments revealed a variety of graphites best suited to the various tasks in mind, as defined by the operational conditions. The tests examined radiation-induced changes of linear dimension, E-module, thermal expansion, and heat conductivity, as well as radiation-induced creep and corrosion in reactor graphites under specified normal and under accident conditions. The work performed also includes tests for defining design criteria for reactor graphite components. The goals have been achieved, but further work will be necessary, as new requirements are taking shape in the course of current THTR follower plant development. (orig.)

  12. Design, fabrication and testing results of vacuum vessel, thermal shield and cryostat of EAST

    International Nuclear Information System (INIS)

    The EAST (Experimental Advanced Superconducting Tokamak) is an advanced steady-state plasma physics experimental device, which has been approved by the Chinese government and being constructed as the Chinese national nuclear fusion Research project. Vacuum vessel is the location for the operation of plasma as one of the key component for EAST device. During it operation the vacuum vessel will not only endure the electromagnetic force due to the plasma disruption and Halo current but also the pressure of boride water and the thermal stress owing to the 250 deg C baking out by the hot pressure nitrogen gas or the 100 deg C hot wall during plasma operation. The cryostat is a large single walled vessel surrounding the entire Basic Machine with central cylindrical section and two end enclosures, a flat base structure with external reinforcements and dome-shaped lid structure. It provides the thermal barrier with the base pressure of 5x10-4 Pa between the ambient temperature testing hall and the liquid helium cooled superconducting magnet. The thermal shields comprise the vacuum vessel thermal shield (VVTS), between the vacuum vessel and the cold TF coil structures, the cryostat thermal shield (CTS), covering the walls of the cryostat, thereby preventing direct line of sight of the room temperature walls to the cold structures, the vacuum port thermal shields (VPTS) that enclose the port connection ducts. The thermal shields are made of double-wall panels, sandwich structure consist of two stainless steel panels and weld quadrate cooling pipe in between the total surface of the thermal shields is about 351 m2. This paper is a report of the structure design and mechanical analyses on the vacuum vessel, thermal shield and cryostat. According to the allowable stress criteria of ASME, the maximum integrated stress intensity on these key components is less than the allowable design stress intensity 3 Sm. The fabrication for these components was completed in 2004 and has been

  13. Neutron streaming measurements at an 850-MWe pressurized water reactor and subsequent shielding recommendations

    International Nuclear Information System (INIS)

    The St. Lucie Nuclear Power Plant has an annular gap of 2.5 ft to allow for a vent path during a loss of coolant accident. At a late stage in construction of the plant it was discovered that the gap would allow neutrons from the core midplane area to scatter and reflect from the roof of the containment building. During startup a series of gamma and neutron dose rate surveys confirmed this neutron streaming problem. Efforts were made to design and fabricate a neutron shield before startup but unfortunately Permali, a laminated wood material chosen for the shield, was rejected by the NRC because of the fire hazard and the possibility that the shield might become a missile in a loss of coolant accident. An alternate design using nylon and neoprene covered bags holding water has been considered. The water bag shield obviates the fire and missile hazards. Although it has not yet been approved by the NRC similar shields have been approved and used successfully at other plants. (author)

  14. Accuracy evaluation of the current data and method applied to shielding design of the Fusion Experimental Reactor (FER)

    International Nuclear Information System (INIS)

    Shielding benchmarking study of the current data and method applied to the Fusion Experimental Reactor (FER) was performed. First, neutron and gamma ray fluxes were calculated by the one-dimensional SN code using various cross section libraries and the continuous energy Monte Carlo code. The results were compared in terms of the SN/MC ratio. The worst ratios are about 0.5 and 0.25 for neutron flux and gamma ray flux, respectively. Next, the analytical calculations of the iron sphere transmission experiment of 14 MeV neutrons were performed to examine the accuracy of cross section data of iron, which is the most important material of shield. The E/C ratio is larger than 2 even if the continuous energy Monte Carlo code was used. Thirdly, the influence of geometrical representation of the shield was investigated by comparing the homogeneous model and the heterogeneous model (alternating layers of SS316 and water). As a result, it was made clear that the homogeneous model underestimates neutron flux by a factor of 2. Finally, the necessity of benchmark experiment and improvement of cross section library was pointed out as the further R and D issues. (author)

  15. Experience in Remote Demolition of the Activated Biological Shielding of the Multi Purpose Research Reactor (MZFR) on the German Karlsruhe Site - 12208

    Energy Technology Data Exchange (ETDEWEB)

    Eisenmann, Beata; Fleisch, Joachim; Prechtl, Erwin; Suessdorf, Werner; Urban, Manfred [WAK Rueckbau- und Entsorgungs- GmbH, P.O.Box 12 63, 76339 Eggenstein-Leopoldshafen (Germany)

    2012-07-01

    In 2009, WAK Decommissioning and Waste Management GmbH (WAK) became owner and operator of the waste treatment facilities of Karlsruhe Institute of Technology (KIT) as well as of the prototype reactors, the Compact Sodium-Cooled Fast Reactor (KNK) and Multi-Purpose Reactor (MZFR), both being in an advanced stage of dismantling. Together with the dismantling and decontamination activities of the former WAK reprocessing facility since 1990, the envisaged demolishing of the R and D reactor FR2 and a hot cell facility, all governmentally funded nuclear decommissioning projects on the Karlsruhe site are concentrated under the WAK management. The small space typical of prototype research reactors represented a challenge also during the last phase of activated dismantling, dismantling of the activated biological shield of the MZFR. Successful demolition of the biological shield required detailed planning and extensive testing in the years before. In view of the limited space and the ambient dose rate that was too high for manual work, it was required to find a tool carrier system to take up and control various demolition and dismantling tools in a remote manner. The strategy formulated in the concept of dismantling the biological shield by means of a modified electro-hydraulic demolition excavator in an adaptable working scaffolding turned out to be feasible. The following boundary conditions were essential: - Remote exchange of the dismantling and removal tools in smallest space. - Positioning of various supply facilities on the working platform. - Avoiding of interfering edges. - Optimization of mass flow (removal of the dismantled mass from the working area). - Maintenance in the surroundings of the dismantling area (in the controlled area). - Testing and qualification of the facilities and training of the staff. Both the dismantling technique chosen and the proceeding selected proved to be successful. Using various designs of universal cutters developed on the basis of

  16. Instrumentation to Enhance Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  17. Instrumentation to Enhance Advanced Test Reactor Irradiations

    International Nuclear Information System (INIS)

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  18. Hypervelocity impact testing of advanced materials and structures for micrometeoroid and orbital debris shielding

    Science.gov (United States)

    Ryan, Shannon; Christiansen, Eric L.

    2013-02-01

    A series of 66 hypervelocity impact experiments have been performed to assess the potential of various materials (aluminium, titanium, copper, stainless steel, nickel, nickel/chromium, reticulated vitreous carbon, silver, ceramic, aramid, ceramic glass, and carbon fibre) and structures (monolithic plates, open-cell foam, flexible fabrics, rigid meshes) for micrometeoroid and orbital debris (MMOD) shielding. Arranged in various single-, double-, and triple-bumper configurations, screening tests were performed with 0.3175 cm diameter Al2017-T4 spherical projectiles at nominally 6.8 km/s and normal incidence. The top performing shields were identified through target damage assessments and their respective weight. The top performing candidate shield at the screening test condition was found to be a double-bumper configuration with a 0.25 mm thick Al3003 outer bumper, 6.35 mm thick 40 PPI aluminium foam inner bumper, and 1.016 mm thick Al2024-T3 rear wall (equal spacing between bumpers and rear wall). In general, double-bumper candidates with aluminium plate outer bumpers and foam inner bumpers were consistently found to be amongst the top performers. For this impact condition, potential weight savings of at least 47% over conventional all-aluminium Whipple shields are possible by utilizing the investigated materials and structures. The results of this study identify materials and structures of interest for further, more in-depth, impact investigations.

  19. Testing of the Micro-Reactor System

    Czech Academy of Sciences Publication Activity Database

    Krystyník, Pavel; Beneš, Ondřej; Klusoň, Petr; Šolcová, Olga

    Praha: Česká společnost průmyslové chemie, 2015, s. 30 /p104./. ISBN 978-80-86238-73-9. [mezinárodní chemicko-technologická konference (ICCT 2015) /3./. Mikulov (CZ), 13.04.2015-15.04.2015] R&D Projects: GA ČR GA15-14228S Institutional support: RVO:67985858 Keywords : micro-reactor technology * heat transfer * testing Subject RIV: CI - Industrial Chemistry, Chemical Engineering

  20. Testing of the Micro-Reactor System

    Czech Academy of Sciences Publication Activity Database

    Beneš, Ondřej; Hanková, Libuše; Klusoň, Petr; Šolcová, Olga

    Bratislava: Slovak Society of Chemical Engineering, 2015 - (Markoš, J.), s. 40 ISBN 978-80-89475-14-8. [International Conference of Slovak Society of Chemical Engineering /42./. Tatranské Matliare (SK), 25.05.2015-29.05.2015] R&D Projects: GA ČR GA15-14228S Institutional support: RVO:67985858 Keywords : micro-reactor technology * testing * partial oxidation Subject RIV: CI - Industrial Chemistry, Chemical Engineering

  1. Design concepts to minimize the activation of the biological shield of light-water reactors

    International Nuclear Information System (INIS)

    An investigation, concentrating on the nuclear aspects, has been made into the concept of minimizing the activation of the biological shield by substituting the material concrete with other neutron-shielding materials. This work was for nuclear plant designs which have a non-supporting inner shield wall such as that in the General Electric BWR/6 and the Kraftwerk Union PWR. The attenuation performance and activation levels have been analysed. Based on this analysis the performance of the materials in relation to that of concrete was assessed. Other non-nuclear properties were considered but the engineering problems were not addressed. The conclusion reached was that the concept was credible but would require a more rigorous examination in terms of structural design, economics and licensability

  2. Advanced burner test reactor preconceptual design report.

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an

  3. Unusual occurrences in fast breeder test reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled mixed carbide fuelled reactor. Its main aim is to generate experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for the development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 85 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 93 and power was raised to 10.5 MWt in Dec 93. Turbine generator was synchronised to the grid in Jul 97. The indigenously developed mixed carbide fuel has achieved a burnup of 44,000 MW-d/t max at a linear heat rating of 320 W/cm max without any fuel clad failure. The commissioning and operation of sodium systems and components have been smooth and performance of major components, viz., sodium pumps, intermediate heat exchangers and once through sodium heated steam generators (SG) have been excellent. There have been three minor incidents of Na/NaK leaks during the past 14 years, which are described in the paper. There have been no incident of a tube leak in SG. However, three incidents of water leaks from water / steam headers have been detailed. The plant has encountered some unusual occurrences, which were critically analysed and remedial measures, in terms of system and procedural modifications, incorporated to prevent recurrence. This paper describes unusual occurrences of fuel handling incident of May 1987, main boiler feed pump seizure in Apr 1992, reactivity transients in Nov 1994 and Apr 1995, and malfunctioning of the core cover plate mechanism in Jul 1995. These incidents have resulted in long plant shutdowns. During the course of investigation, various theoretical and experimental studies were carried out for better understanding of the phenomena and several inspection techniques and tools were developed resulting in enriching the technology of sodium cooled reactors. FBTR has 36 neutronic and process

  4. Experience in the maintenance of sodium systems of fast breeder test reactor

    International Nuclear Information System (INIS)

    The fast breeder test reactor (FBTR) is a loop type sodium cooled fast reactor located at Kalpakkam in India and that has been operating for 25 years. The reactor has been operated up to a power level of 18.6 MWt with a sodium outlet temperature of 482 C. degrees. Several modifications were carried out in the sodium systems to improve the plant performance. During the course of operation of the reactor, a number of sodium laden components like pumps, valves, cold traps, rupture disks, level probes, shielding plugs, control rod drive mechanisms, experimental assemblies, piping... were removed for various maintenance, modification and replacement jobs which has given the operators a valuable experience in handling large scale sodium systems. This paper details the special procedures followed during the handling of active and inactive sodium laden components

  5. Tissue Equivalent Proportional Counter Microdosimetry Measurements Utilized Aboard Aircraft and in Accelerator Based Space Radiation Shielding Studies

    Science.gov (United States)

    Gersey, Brad B.; Wilkins, Richard T.

    2010-01-01

    This slide presentation reviews the Tissue Equivalent Proportional Counter (TEPC), a description of the spatially restricted LET Model, high energy proton TEPC and the results of modeling, the study of shielding and the results from the flight exposures with the TEPC.

  6. Comparison of neutron fluxes obtained by 2-D and 3-D geometry with different shielding libraries in biological shield of the TRIGA MARK II reactor

    International Nuclear Information System (INIS)

    Neutron fluxes in different spatial locations in biological shield are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Libraries used with TORT code were BUGLE-96 library (coupled library with 47 neutron groups and 20 gamma groups) and VITAMIN-B6 library (coupled library with 199 neutron groups and 42 gamma groups). BUGLE-96 library is derived from VITAMIN-B6 library. 2-D and 3-D models for homogeneous type of problem (without inserted beam port 4) and problem with asymmetry (non-homogeneous problem; inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. The main purpose is to verify the possibility for using 2-D approximation model instead of large 3-D model in some calculations. Another purpose of this paper was to compare neutron spectral constants obtained from neutron fluxes (3-D model) determined with smaller BUGLE-96 library with new constants obtained from fluxes calculated with bigger VITAMIN-B6 library. These neutron spectral constants are used in isotopic calculation with SCALE code package (ORIGEN-S). In past only neutron spectral constants determined by neutron fluxes from BUGLE-96 library were used. Experimental results used for isotopic composition comparison are available from irradiation experiment with selected type of concrete and other materials in beam port 4 (irradiation channel 4) in TRIGA Mark II reactor. These experimental results were used as a benchmark in this paper. (author)

  7. Experience with a servo-hydraulic mechanical testing machine installed in a new shielded active facility at Windscale Nuclear Power Development Laboratories

    International Nuclear Information System (INIS)

    An Instron model 1273 servo-hydraulic machine has been installed within a lead-shielded cell at Windscale in order to provide a facility capable of performing a wide range of mechanical tests on nuclear reactor structural materials and fuel assembly components. This particular type of machine was chosen because it has design features associated with the load frame, location of the actuator and adjustment and clamping of the cross-head that are especially well suited to remote operation within a shielded cell. The design of the testing facility is described and the programmes of work that have been completed over the past 11/2 years of operation are reviewed. (author)

  8. Nuclear Rocket Test Facility Decommissioning Including Controlled Explosive Demolition of a Neutron-Activated Shield Wall

    Energy Technology Data Exchange (ETDEWEB)

    Michael Kruzic

    2007-09-01

    Located in Area 25 of the Nevada Test Site, the Test Cell A Facility was used in the 1960s for the testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program. The facility was decontaminated and decommissioned (D&D) in 2005 using the Streamlined Approach For Environmental Restoration (SAFER) process, under the Federal Facilities Agreement and Consent Order (FFACO). Utilities and process piping were verified void of contents, hazardous materials were removed, concrete with removable contamination decontaminated, large sections mechanically demolished, and the remaining five-foot, five-inch thick radiologically-activated reinforced concrete shield wall demolished using open-air controlled explosive demolition (CED). CED of the shield wall was closely monitored and resulted in no radiological exposure or atmospheric release.

  9. Nuclear Rocket Test Facility Decommissioning Including Controlled Explosive Demolition of a Neutron-Activated Shield Wall

    International Nuclear Information System (INIS)

    Located in Area 25 of the Nevada Test Site, the Test Cell A Facility was used in the 1960s for the testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program. The facility was decontaminated and decommissioned (D and D) in 2005 using the Streamlined Approach For Environmental Restoration (SAFER) process, under the Federal Facilities Agreement and Consent Order (FFACO). Utilities and process piping were verified void of contents, hazardous materials were removed, concrete with removable contamination decontaminated, large sections mechanically demolished, and the remaining five-foot, five-inch thick radiologically-activated reinforced concrete shield wall demolished using open-air controlled explosive demolition (CED). CED of the shield wall was closely monitored and resulted in no radiological exposure or atmospheric release

  10. Final Physics Report for the Engineering Test Reactor

    International Nuclear Information System (INIS)

    This volume continues with discussions of shielding provided for the heat exchanger building, concrete biological shield, top area and bottom area shielding, canal shielding, water shielding requirements during fuel element exchanges, and supplementary shielding requirements

  11. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

  12. Performance tests for integral reactor nuclear fuel

    International Nuclear Information System (INIS)

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34∼38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc

  13. Summary of power ascension test of experimental fast reactor 'JOYO' MK-I

    International Nuclear Information System (INIS)

    On April 24th, 1977, the initial criticality of JOYO was achieved and on July 5th, 1978, the reactor output reached rated power of 50 MW for the first time. The 75MW power ascension test was started in July, 1979, followed by two cycles of rated power operations, and the 100 hour nominal power continuous operation was completed in February, 1980. Through the tests for the core, plant it self, radiation shield and plant monitoring, the results proved satisfactory operation characteristics at 75MW. This report presents the summary of all the results obtained in the Test of MK-I core. (author)

  14. Developing the MAPLE materials test reactor concept

    International Nuclear Information System (INIS)

    MAPLE-MTR is a new multipurpose research facility being planned by AECL Research as a possible replacement for the 35-year-old NRU reactor. In developing the MAPLE-MTR concept, AECL is starting from the recent design and licensing experience with the MAPLE-X10 reactor. By starting from technology developed to support the MAPLE-X10 design and adapting it to produce a concept that satisfies the requirements of fuel channel materials testing and fuel irradiation programs, AECL expects to minimize the need for major advances in nuclear technology (e.g., fuel, heat transfer). Formulation of the MAPLE-MTR concept is at an early stage. This report describes the irradiation requirements of the research areas, how these needs are translated into design criteria for the project and elements of the preliminary design concept

  15. Decommissioning of the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola

    2003-10-28

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.

  16. Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D and D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D and D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D and D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget

  17. TRIGA reactor dynamics: Frequency response tests

    International Nuclear Information System (INIS)

    In this work, the results of frequency response tests conducted on ITU TRIGA Reactor are presented. To conduct the experiments, a special 'micro control rod' and its submersible stepping-motor drive mechanism was designed and constructed. The experiments cover a frequency range of 0.002 - 2 Hz., and 0.02, 4, 200 kW nominal power levels. Zero-power and at-power reactivity to % power transfer functions are presented as gain, and phase shift vs. frequency diagrams. Low power response is in close agreement with the point reactor zero-power transfer function. Response at 200 kW is studied with the help of a Nyquist diagram, and found to be stable. An elaboration on the main features of the feedback mechanism is also given. Power to reactivity feedback was measured to be just about 1.5 cent / % power change. (authors)

  18. Standard Test Method for Preparing Aircraft Cleaning Compounds, Liquid Type, Water Base, for Storage Stability Testing

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

    1.1 This test method covers the determination of the stability in storage, of liquid, water-base chemical cleaning compounds, used to clean the exterior surfaces of aircraft. 1.2 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  19. Fuel irradiation test plan at the Japan materials testing reactor

    International Nuclear Information System (INIS)

    Development of high performance fuels, which enables burnup extension and high duty uses of light water reactors (LWRs) by means of power up rates and flexible operating cycles, is one of key technical issues for extending the uses for longer periods. Introduction of new design fuel rods with new cladding alloys and wider utilization of mixed oxide fuels is expected in Japan. Fuel irradiation tests for development and safety demonstration are quite important, in order to realize theses progress. Operational management on water chemistry, minimizing the long term degradation of reactor components, could have unfavorable influence on the integrity of the fuel rods. Japanese government and the Japan Atomic Energy Agency have decided to re new the Japan Materials Testing Reactor (JMTR) and to install new test rigs, in order to play an active role solving the issues on the development and the safety of the fuel and the plant aging. Fuel integrity under abnormal transient conditions will be investigated using a special capsule type test rig, which has its own power control system under simulated LWR cooling conditions. Water loops for simulation of high duty operation, e.g. high power, high burnup and high rod internal pressure conditions, are proposed for the development and safety examination of the high performance fuels. Combination of the JMTR tests with simulated reactivity initiated accident tests in the Nuclear Safety Research Reactor and loss of coolant accident tests in hot laboratories would provide a comprehensive data for safety evaluation and design progress of the high performance fuels at extended burnups, covering from the normal to the accident conditions, including abnormal transients

  20. Standard Test Method to Determine Color Change and Staining Caused by Aircraft Maintenance Chemicals upon Aircraft Cabin Interior Hard Surfaces

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2001-01-01

    1.1 This test method covers the determination of color change and staining from liquid solutions, such as cleaning or disinfecting chemicals or both, on painted metallic surfaces and nonmetallic surfaces of materials being used inside the aircraft cabin. The effects upon the exposed specimens are measured with the AATCC Gray Scale for Color Change and AATCC Gray Color Scale for Staining. Note 1—This test method is applicable to any colored nonmetallic hard surface in contact with liquids. The selected test specimens are chosen because these materials are present in the majority of aircraft cabin interiors. 1.2This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  1. Experimental flight test vibration measurements and nondestructive inspection on a USCG HC-130H aircraft

    Energy Technology Data Exchange (ETDEWEB)

    Moore, D.G.; Jones, C.R. [Sandia National Labs., Albuquerque, NM (United States). FAA Airworthiness Assurance NDI Validation Center; Mihelic, J.E.; Barnes, J.D. [Coast Guard Aircraft Repair and Supply Center, Elizabeth City, NC (United States)

    1998-08-01

    This paper presents results of experimental flight test vibration measurements and structural inspections performed by the Federal Aviation Administration`s Airworthiness Assurance NDI Validation Center (AANC) at Sandia National Laboratories and the US Coast Guard Aircraft Repair and Supply Center (ARSC). Structural and aerodynamic changes induced by mounting a Forward Looking Infrared (FLIR) system on a USCG HC-130H aircraft are described. The FLIR adversely affected the air flow characteristics and structural vibration on the external skin of the aircraft`s right main wheel well fairing. Upon initial discovery of skin cracking and visual observation of skin vibration in flight by the FLIR, a baseline flight without the FLIR was conducted and compared to other measurements with the FLIR installed. Nondestructive inspection procedures were developed to detect cracks in the skin and supporting structural elements and document the initial structural condition of the aircraft. Inspection results and flight test vibration data revealed that the FLIR created higher than expected flight loading and was the possible source of the skin cracking. The Coast Guard performed significant structural repair and enhancement on this aircraft, and additional in-flight vibration measurements were collected on the strengthened area both with and without the FLIR installed. After three months of further operational FLIR usage, the new aircraft skin with the enhanced structural modification was reinspected and found to be free of flaws. Additional US Coast Guard HC-130H aircraft are now being similarly modified to accommodate this FLIR system. Measurements of in-flight vibration levels with and without the FLIR installed, and both before and after the structural enhancement and repair were conducted on the skin and supporting structure in the aircraft`s right main wheel fairing. Inspection results and techniques developed to verify the aircraft`s structural integrity are discussed.

  2. Off reactor testings. Technological engineering applicative research

    International Nuclear Information System (INIS)

    By the end of year 2000 over 400 nuclear electro-power units were operating world wide, summing up a 350,000 MW total capacity, with a total production of 2,300 TWh, representing 16% of the world's electricity production. Other 36 units, totalizing 28,000 MW, were in construction, while a manifest orientation towards nuclear power development was observed in principal Asian countries like China, India, Japan and Korea. In the same world's trend one find also Romania, the Cernavoda NPP Unit 1 generating electrical energy into the national system beginning with 2 December 1996. Recently, the commercial contract was completed for finishing the Cernavoda NPP Unit 2 and launching it into operation by the end of year 2004. An important role in developing the activity of research and technological engineering, as technical support for manufacturing the CANDU type nuclear fuel and supplying with equipment the Cernavoda units, was played by the Division 7 TAR of the INR Pitesti. Qualification testings were conducted for: - off-reactor CANDU type nuclear fuel; - FARE tools, pressure regulators, explosion proof panels; channel shutting, as well as functional testing for spare pushing facility as a first step in the frame of the qualification tests for the charging/discharging machine (MID) 4 and 5 endings. Testing facilities are described, as well as high pressure hot/cool loops, measuring chains, all of them fulfilling the requirements of quality assurance. The nuclear fuel off-reactor tests were carried out to determine: strength; endurance; impact, pressure fall and wear resistance. For Cernavoda NPP equipment testings were carried out for: the explosion proof panels, pressure regulators, behaviour to vibration and wear of the steam generation tubings, effects of vibration upon different electronic component, channel shutting (for Cernavoda Unit 2), MID operating at 300 and 500 cycles. A number of R and D programs were conducted in the frame of division 7 TAR of INR

  3. On the response of a reactor building and its equipment to aircraft crash

    International Nuclear Information System (INIS)

    The present study investigates the dynamic response of the ASEA-ATOM BWR 75 reactor building in terms of response spectra at significant locations considering various aircraft and points of load application. In the first part of the study a total of 21 forcing functions, most of them from the open literature and including the commonly used standard functions, have been studied with respect to documentation, consistency and frequency content. Since none of the forcing functions have been experimentally verified, their validity must be assessed mainly by judging the structural models and assumptions used in their derivation and by checking their consistency. In the second part, linear dynamical models of various degrees of detailedness have been investigated regarding their capacity to describe the behavior of the reactor building under this high frequency loading. The most detailed model consists of plane stress finite elements for every significant wall and floor. In the third part of the study the effects of a number of parameters on the response of the building are investigated. The parameters include the points of attack, damping values, soil spring stiffness as well as different forcing functions of various frequency contents. The reponse is displayed as response spectra and member forces for characteristic locations. The results serve as a basis for development of standardized design floor response spectra and for the structural verification of the bui

  4. Radiation shielding for superconducting RF cavity test facility at A0

    International Nuclear Information System (INIS)

    The results of Monte Carlo radiation shielding study performed with the MARS15 code for the vertical test facility at the A0 north cave enclosure at Fermilab are presented and discussed. The vertical test facility at the A0 north cave is planned to be used for testing 1.3 GHz single-cell superconducting RF cavities with accelerating length of 0.115 m. The operations will be focused on high accelerating gradients--up to 50 MV/m. In such a case the facility can be a strong radiation source (1). When performing a radiation shielding design for the facility one has to take into account gammas generated due to interactions of accelerated electrons with cavity walls and surroundings (for example, range of 3.7-MeV electrons in niobium is approximately 3.1 mm while the thickness of the niobium walls of such RF cavities is about 2.8 mm). The electrons are usually the result of contamination in the cavity. The radiation shielding study was performed with the MARS15 Monte Carlo code (2). A realistic model of the source term has been used that describes spatial, energy and angular distributions of the field-emitted electrons inside the RF cavities. The results of the calculations are normalized using the existing experimental data on measured dose rate in the vicinity of such RF cavities

  5. Radiation shielding for superconducting RF cavity test facility at A0

    Energy Technology Data Exchange (ETDEWEB)

    Dhanaraj, N.; Ginsburg, C.; Rakhno, I.; Wu, G.; /Fermilab

    2008-11-01

    The results of Monte Carlo radiation shielding study performed with the MARS15 code for the vertical test facility at the A0 north cave enclosure at Fermilab are presented and discussed. The vertical test facility at the A0 north cave is planned to be used for testing 1.3 GHz single-cell superconducting RF cavities with accelerating length of 0.115 m. The operations will be focused on high accelerating gradients--up to 50 MV/m. In such a case the facility can be a strong radiation source [1]. When performing a radiation shielding design for the facility one has to take into account gammas generated due to interactions of accelerated electrons with cavity walls and surroundings (for example, range of 3.7-MeV electrons in niobium is approximately 3.1 mm while the thickness of the niobium walls of such RF cavities is about 2.8 mm). The electrons are usually the result of contamination in the cavity. The radiation shielding study was performed with the MARS15 Monte Carlo code [2]. A realistic model of the source term has been used that describes spatial, energy and angular distributions of the field-emitted electrons inside the RF cavities. The results of the calculations are normalized using the existing experimental data on measured dose rate in the vicinity of such RF cavities.

  6. Construction and performance test of radiation shielding for 300 keV/20 mA electron beam machine

    International Nuclear Information System (INIS)

    The construction and performance test of radiation shielding for 300 keV/20 mA electron beam machine (EBM) has been done. Radiation shielding is used for reduce X-ray radiation which is generated by operation of the EBM, so it is not harmful for people who work in that environment. Radiation shielding plates made of bars of lead (Pb) with a length of 135 cm, 10 cm of width, 2.5 cm of thick and composed a EBM radiation shielding. The plates are made of lead by way of casting and finished mechanically by machine, then installed manually on a frame to form a EBM radiation shielding. In the calculation of thick radiation shield already qualified dose rate limit is set by BAPETEN ≤ 2.5 mrem/hr. The results of the initial test radiation shielding is functioning, it is shown by the results of measurements of the maximum dose rate 0.26 mrem/hr at the operating conditions of EBM with voltage 209 kV and 50 mA of electron beam current. Based on the results test of the construction of radiation shielding are qualified dose rate limit set by BAPETEN. (author)

  7. Thermal Hydraulic Tests for Reactor Core Safety

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

    2007-06-15

    The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

  8. Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Indah Rosidah, M., E-mail: indah.maymunah@gmail.com; Suud, Zaki, E-mail: szaki@fi.itb.ac.id [Department of Nuclear Physics, Faculty of Mathematic and Natural Sciences, Institut Teknologi Bandung (Indonesia); Yazid, Putranto Ilham [Research and Development of Nuclear Association (Indonesia)

    2015-09-30

    The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With the tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature

  9. Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

    International Nuclear Information System (INIS)

    The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With the tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature

  10. Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

    Science.gov (United States)

    Indah Rosidah, M.; Suud, Zaki; Yazid, Putranto Ilham

    2015-09-01

    The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With the tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature

  11. SRS reactor stack plume marking tests

    International Nuclear Information System (INIS)

    Tests performed in 105-K in 1987 and 1988 demonstrated that the stack plume can successfully be made visible (i.e., marked) by introducing smoke into the stack breech. The ultimate objective of these tests is to provide a means during an emergency evacuation so that an evacuee can readily identify the stack plume and evacuate in the opposite direction, thus minimizing the potential of severe radiation exposure. The EPA has also requested DOE to arrange for more tests to settle a technical question involving the correct calculation of stack downwash. New test canisters were received in 1988 designed to produce more smoke per unit time; however, these canisters have not been evaluated, because normal ventilation conditions have not been reestablished in K Area. Meanwhile, both the authorization and procedure to conduct the tests have expired. The tests can be performed during normal reactor operation. It is recommended that appropriate authorization and procedure approval be obtained to resume testing after K Area restart

  12. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    M.E. Lumia; C.A. Gentile

    2002-01-18

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed.

  13. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed

  14. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    CERN Document Server

    Lumia, M E

    2002-01-01

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed.

  15. NEA organised validation of multi-dimensional transport codes and cross-section sets used presently in reactor shielding

    International Nuclear Information System (INIS)

    A Task Force of the OECD/NEA Nuclear Science Committee performed benchmark calculations to validate two- and three-dimensional transport codes and cross-section sets used for radiation shielding. Based on careful measurements from the VENUS critical facility of SCK/CEN-Mol/Belgium, the Task Force started with two-dimensional benchmark calculations representing the arrangement of VENUS-1, which is a mock-up of the Westinghouse three loop reactor. About 20 different calculations were presented. Several 2D-SN-codes were used including some in-house codes and the Monte Carlo program MCNP-4A. The transport cross-sections were taken from ENDF/B-VI or JEF 2.2 and the dosimetry data mainly from IRDF-90v2. The most challenging task was the validation of the latest versions of three-dimensional transport codes with the complex arrangement of VENUS-3. Approximately 14 independent benchmark calculations were contributed world-wide. Several versions of 3D-SN-codes and the Monte Carlo code MCNP were applied. In conclusion the results of the three-dimensional benchmark (VENUS-3) are in general much closer to the experimental values than for the two-dimensional benchmark (VENUS-1). This shows the inherent deficiencies of two-dimensional transport calculations, due to approximations such as diffusion based axial buckling corrections and improper averaging of source and shielding materials, since real configurations are always three-dimensional. (author)

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

    International Nuclear Information System (INIS)

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

  17. Source terms and shielding studies related to the spent fuel storage of the R-A reactor

    International Nuclear Information System (INIS)

    The paper presents the methods developed and calculations carried out to analyse the photon source term and equivalent dose rates of the Vinca R-A reactor spent fuel elements load inside the aluminium barrels. An analysis methodology for the radiological characteristics evolution, based on the SAS2H, MCNP-4C/ORIGEN2.1 and KENOV. a/ORIGEN2.1 utility codes is presented along with information representing the validation of the methods and geometrical models. The sequence SAS4 with the MORSE-SGC code, and the MCNP-4C, a Monte Carlo code has been used for solving the shielding problem. Gamma dose rates estimated in radial direction from the aluminium barrels with spent fuel elements are presented and discussed. (author)

  18. On The Deign And Construction Of A Radiation Shielding System For Development Of Neutron Beams Based On The Horizontal Channel No.2 Of Dalat Reactor

    International Nuclear Information System (INIS)

    An optimal structural system of filtered neutron beam and radiation shielding has been designed and calculated using the Monte-Carlo code MCNP5. The system was constructed and installed into the horizontal channel No. 2 of the Dalat reactor. The neutron beam is applied for experimental studies on nuclear physics, nuclear data measurements, and personal training. (author)

  19. Calculation of self-shielding coefficients, flux depression and cadmium factor for thermal neutron flux measurement of the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    A calculation methodology of Flux Depression, Self-Shielding and Cadmium Factors is presented, using the ANISN code, for experiments conducted at the IPEN/MB-01 Research Reactor. The correction factors were determined considering thermal neutron flux and 0.125 e 0.250 mm diameter of 197 Au wires. (author)

  20. Elimination of the coil shielding for MFE-Reactors through a liquid-protected first-wall

    International Nuclear Information System (INIS)

    The idea of a protective, flowering liquid zone to protect the first wall of magnetic fusion energy (MFE) reactor from the direct exposure of the fusion reaction product is not new. This could extend the lifetime of the first wall to the lifetime of the fusion power plant, namely to 30 years. The present work discusses the possibility that such a liquid zone could lead also to the elimination of the magnetic coil shielding for MFE reactors. Contrary to a related previous work, the liquid wall is now placed at the outermost periphery of the plasma chamber, in order to leave a greater space for the fusion plasma volume and consequently to lead the higher fusion power with the same plasma parameters and consequently to lead to higher fusion power with same plasma parameters. In this work, SS-304 type steel, SiC and graphite are selected as structural materials. Different types of liquid coolant with tritium breeding capabilities (Flibe, Li17Pb83, natural lithium, all with natural lithium component) are investigated to protect the first wall from neutron- and bremsstrahlung-radiation and fusion reaction debris. The calculations are conducted for a power generation of 1GWe1 over 30 years of reactor operation with a thermodynamically conversion efficiency of 35% leading to 2.857GWth by a capacity factor of 100%. The most important improvement through the placement of the protective liquid wall at the outer periphery in the new blanket can be cited as follows. Such a blanket (1) would in practice not necessitate extra shielding for superconducting coils around the fusion plasma chamber. (2) Would open the possibility of utilization of conventional stainless steel for fusion reactors due to the sufficiently low residual radioactivity in the structural materials after decommissioning of the plant. Research efforts and costs, involved in searching new alternative ceramic structural materials, such as SiC and graphite, based on unproven technology can be saved. (3) And would

  1. Design of the precast, post-tensioned concrete shielding structure for the TFTR neutral beam test cell

    International Nuclear Information System (INIS)

    At the TFTR facility, the Neutral Beam Test Cell is a room separated from the TFTR Cell by a 4-foot-thick concrete wall and devoted to testing the neutral beam injector. The function of the shielding structure is to protect personnel from radiation casued by pulsing the injector. The distance from the TFTR device to the injector is large enough to permit use of magnetic materials in the shielding structure, and the neutron flux levels are small enough so that ordinary concrete of moderate thickness may be employed. Radiation considerations are not discussed in this paper, which is devoted to a description of the structural design of the shield

  2. High field, low current operation of engineering test reactors

    International Nuclear Information System (INIS)

    Steady state engineering test reactors with high field, low current operation are investigated and compared to high current, lower field concepts. Illustrative high field ETR parameters are R = 3 m, α ∼ 0.5 m, B ∼ 10 T, β = 2.2% and I = 4 MA. For similar wall loading the fusion power of an illustrative high field, low current concept could be about 50% that of a lower field device like TIBER II. This reduction could lead to a 50% decrease in tritium consumption, resulting in a substantial decrease in operating cost. Furthermore, high field operation could lead to substantially reduced current drive requirements and cost. A reduction in current drive source power on the order of 40 to 50 MW may be attainable relative to a lower field, high current design like TIBER II implying a possible cost savings on the order of $200 M. If current drive is less efficient than assumed, the savings could be even greater. Through larger β/sub p/ and aspect ratio, greater prospects for bootstrap current operation also exist. Further savings would be obtained from the reduced size of the first wall/blanket/shield system. The effects of high fields on magnet costs are very dependent on technological assumptions. Further improvements in the future may lie with advances in superconducting and structural materials

  3. Non-Parametric, Closed-Loop Testing of Autonomy in Unmanned Aircraft Systems Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The proposed Phase I program aims to develop new methods to support safety testing for integration of Unmanned Aircraft Systems into the National Airspace (NAS)...

  4. A New Method for Predicting the Penetration and Slowing-Down of Neutrons in Reactor Shields

    International Nuclear Information System (INIS)

    A new approach is presented in the formulation of removal-diffusion theory. The 'removal cross-section' is redefined and the slowing-down between the multigroup diffusion equations is treated with a complete energy transfer matrix rather than in an age theory approximation. The method, based on the new approach contains an adjustable parameter. Examples of neutron spectra and thermal flux penetrations are given in a number of differing shield configurations and the results compare favorably with experiments and Moments Method calculations

  5. The Evaluation of Lithium Hydride for Use in a Space Nuclear Reactor Shield, Including a Historical Perspective

    Energy Technology Data Exchange (ETDEWEB)

    D. Poeth

    2005-12-09

    LiH was one of the five primary shield materials the NRPCT intended to develop (along with beryllium, boron carbide, tungsten, and water) for potential Prometheus application. It was also anticipated that {sup 10}B metal would be investigated for feasibility at a low level of effort. LiH historically has been selected as a low mass, neutron absorption material for space shields (Systems for Nuclear Auxiliary Power (SNAP), Topaz, SP-100). Initial NRPCT investigations did not produce convincing evidence that LiH was desirable or feasible for a Prometheus mission due to material property issues (primarily swelling and hydrogen cover gas containment), and related thermal design complexity. Furthermore, if mass limits allowed, an option to avoid use of LiH was being contemplated to lower development costs and associated risks. However, LiH remains theoretically the most efficient neutron shield material per unit mass, and, with sufficient testing and development, could be an optimal material choice for future flights.

  6. Models for transient analyses in advanced test reactors

    International Nuclear Information System (INIS)

    Several strategies are developed worldwide to respond to the world's increasing demand for electricity. Modern nuclear facilities are under construction or in the planning phase. In parallel, advanced nuclear reactor concepts are being developed to achieve sustainability, minimize waste, and ensure uranium resources. To optimize the performance of components (fuels and structures) of these systems, significant efforts are under way to design new Material Test Reactors facilities in Europe which employ water as a coolant. Safety provisions and the analyses of severe accidents are key points in the determination of sound designs. In this frame, the SIMMER multiphysics code systems is a very attractive tool as it can simulate transients and phenomena within and beyond the design basis in a tightly coupled way. This thesis is primarily focused upon the extension of the SIMMER multigroup cross-sections processing scheme (based on the Bondarenko method) for a proper heterogeneity treatment in the analyses of water-cooled thermal neutron systems. Since the SIMMER code was originally developed for liquid metal-cooled fast reactors analyses, the effect of heterogeneity had been neglected. As a result, the application of the code to water-cooled systems leads to a significant overestimation of the reactivity feedbacks and in turn to non-conservative results. To treat the heterogeneity, the multigroup cross-sections should be computed by properly taking account of the resonance self-shielding effects and the fine intra-cell flux distribution in space group-wise. In this thesis, significant improvements of the SIMMER cross-section processing scheme are described. A new formulation of the background cross-section, based on the Bell and Wigner correlations, is introduced and pre-calculated reduction factors (Effective Mean Chord Lengths) are used to take proper account of the resonance self-shielding effects of non-fuel isotopes. Moreover, pre-calculated parameters are applied

  7. Models for transient analyses in advanced test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gabrielli, Fabrizio

    2011-02-22

    Several strategies are developed worldwide to respond to the world's increasing demand for electricity. Modern nuclear facilities are under construction or in the planning phase. In parallel, advanced nuclear reactor concepts are being developed to achieve sustainability, minimize waste, and ensure uranium resources. To optimize the performance of components (fuels and structures) of these systems, significant efforts are under way to design new Material Test Reactors facilities in Europe which employ water as a coolant. Safety provisions and the analyses of severe accidents are key points in the determination of sound designs. In this frame, the SIMMER multiphysics code systems is a very attractive tool as it can simulate transients and phenomena within and beyond the design basis in a tightly coupled way. This thesis is primarily focused upon the extension of the SIMMER multigroup cross-sections processing scheme (based on the Bondarenko method) for a proper heterogeneity treatment in the analyses of water-cooled thermal neutron systems. Since the SIMMER code was originally developed for liquid metal-cooled fast reactors analyses, the effect of heterogeneity had been neglected. As a result, the application of the code to water-cooled systems leads to a significant overestimation of the reactivity feedbacks and in turn to non-conservative results. To treat the heterogeneity, the multigroup cross-sections should be computed by properly taking account of the resonance self-shielding effects and the fine intra-cell flux distribution in space group-wise. In this thesis, significant improvements of the SIMMER cross-section processing scheme are described. A new formulation of the background cross-section, based on the Bell and Wigner correlations, is introduced and pre-calculated reduction factors (Effective Mean Chord Lengths) are used to take proper account of the resonance self-shielding effects of non-fuel isotopes. Moreover, pre-calculated parameters are

  8. TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

    International Nuclear Information System (INIS)

    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

  9. Radiation physics and shielding codes and analyses applied to design-assist and safety analyses of CANDUR and ACRTM reactors

    International Nuclear Information System (INIS)

    This paper discusses the radiation physics and shielding codes and analyses applied in the design of CANDU and ACR reactors. The focus is on the types of analyses undertaken rather than the inputs supplied to the engineering disciplines. Nevertheless, the discussion does show how these analyses contribute to the engineering design. Analyses in radiation physics and shielding can be categorized as either design-assist or safety and licensing (accident) analyses. Many of the analyses undertaken are designated 'design-assist' where the analyses are used to generate recommendations that directly influence plant design. These recommendations are directed at mitigating or reducing the radiation hazard of the nuclear power plant with engineered systems and components. Thus the analyses serve a primary safety function by ensuring the plant can be operated with acceptable radiation hazards to the workers and public. In addition to this role of design assist, radiation physics and shielding codes are also deployed in safety and licensing assessments of the consequences of radioactive releases of gaseous and liquid effluents during normal operation and gaseous effluents following accidents. In the latter category, the final consequences of accident sequences, expressed in terms of radiation dose to members of the public, and inputs to accident analysis, e.g., decay heat in fuel following a loss-of-coolant accident, are also calculated. Another role of the analyses is to demonstrate that the design of the plant satisfies the principle of ALARA (as low as reasonably achievable) radiation doses. This principle is applied throughout the design process to minimize worker and public doses. The principle of ALARA is an inherent part of all design-assist recommendations and safety and licensing assessments. The main focus of an ALARA exercise at the design stage is to minimize the radiation hazards at the source. This exploits material selection and impurity specifications and relies

  10. Corrosion of spent Advanced Test Reactor fuel

    International Nuclear Information System (INIS)

    The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented

  11. Proposal of world network on material testing reactors

    International Nuclear Information System (INIS)

    Establishment of an international cooperation system of worldwide testing reactor network (world network) is proposed in order to achieve efficient facility utilization and provide high quality irradiation data by role sharing of irradiation tests with materials testing reactors in the world. As for the first step, mutual understanding among materials testing reactors is thought to be necessary. From this point, an international symposium on materials testing reactors (ISMTR) was held to construct the world network from 2008, and a common understanding of world network has begun to be shared. (author)

  12. Long term testing of materials for tube shielding, stage 2; Laangtidsprovning av tubskyddsmaterial, etapp 2

    Energy Technology Data Exchange (ETDEWEB)

    Norling, Rikard; Hjoernhede, Anders; Mattsson, Mattias

    2012-02-15

    temperature. It varies from a few months to several years. The cost of replacing tube shielding is significant, which calls for improved materials. In a previous study [1] various materials have been tested in the WtE CFB-boiler P14 at Haendeloe in Norrkoeping, Sweden. The study showed that several materials may be cost effective candidates for replacing 253MA, but the test period was too short for safe predictions. This work includes exposures at the same position, but with extended test period up to almost 12000 h to verify the previous results. In addition results from exposures in the WtE CFB-boiler Hoegdalen P6 in Stockholm, Sweden is included. This boiler suffers in particular from severe erosion of the tube shielding requiring it to be exchanged every few months

  13. High Temperature Gas-Cooled Test Reactor Options Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

  14. Post reactor researches of fuel pins, tested under alternating NEMF reactor functioning modes

    International Nuclear Information System (INIS)

    Changing of rod ceramic fuel pins state under their exploitation conditions changing influence at alternating of three-mode nuclear energy-moving facility reactor functioning has been examined. There are presented the results of researches of fuel pins, tested in the reactor IRGIT and RA, firstly under moving mode, then - under energy mode of minor power of NEMF reactor. (author)

  15. The SGR Multipurpose - Generation IV - Transportable Cogeneration Nuclear Reactor with Innovative Shielding

    International Nuclear Information System (INIS)

    Deregulation and liberalization are changing the global energy-markets. At the same time innovative technologies are introduced in the electricity industry; often as a requirement from the upcoming Digital Society. Energy solutions for the future are more seen as a mix of energy-sources for generation-, transmission- and distribution energy-services. The Internet Energy-web based 'Virtual' enterprises are coming up and will gradually change our society. It the fast changing world we have to realize that there will be less time to look for the adequate solutions to anticipate on global developments and the way they will influence our own societies. Global population may reach 9 billion people by 2030; this will put tremendous pressure on energy-, water- and food supply in the global economy. It is time to think about some major issues as described below and come up with the right answers. These are needed on very short term to secure a humane global economic growth and the sustainable global environment. The DOE (Department of Energy - USA) has started the Generation IV initiative for the new generation of nuclear reactors that must lead to much better safety, economics and public acceptance the new reactors. The SGR (Simplified Gas-cooled Reactor) is being proposed as a Generation IV modular nuclear reactor, using graphite pebbles as fuel, whereby an attempt has been made to meet all the DOE requirements, to be used for future nuclear reactors. The focus in this paper is on the changing and emerging global energy-markets and shows some relevant criteria to the nuclear industry and how we can anticipate with improved and new designs towards the coming Digital Society. (author)

  16. Monte Carlo based demonstration of sufficiently dimensioned shielding for a Co-60 testing facility

    International Nuclear Information System (INIS)

    The electrical properties of electronic equipment can be changed in an ionized radiation field. The knowledge of these changes is necessary for applications in space, in air traffic and nuclear medicine. Experimental tests will be performed in Co-60 radiation fields in the irradiation facility (TEC facility) of the Seibersdorf Labor GmbH that is in construction. The contribution deals with a simulation that is aimed to calculate the local dose rate within and outside the building for demonstration of sufficient dimensioning of the shielding in compliance with the legal dose rate limits.

  17. Testing of the sealing arrangements for the Nirex re-usable shielded transport containers

    International Nuclear Information System (INIS)

    As part of its responsibility for the development of a deep repository for intermediate and low level radioactive waste, UK Nirex Ltd is developing a range of Type B re-usable shielded transport containers (RSTCs). A testing programme has been carried out on two alternative concepts for the RSTC sealing arrangements over the temperature range -40oC to 200oC. For each sealing system, a test rig was developed to measure the performance under simulated normal and accident conditions of transport. The elastomer O-rings used for some of the tests had been irradiated to the maximum dose they might receive in normal transport. The performance of both sealing systems was good and it is concluded that either concept would meet the specified leakage criteria over the full temperature range under both normal and accident conditions of transport. However, further testing is required to confirm the performance of Concept N under accident conditions. (author)

  18. Radiation shielding test for hot cells of Irradiated Material Examination Facility(IMEF)

    International Nuclear Information System (INIS)

    Radiation shielding test for IMEF(Irradiated Material Examination Facility) hot cell walls was executed using two Co 60 sources with the activities of 1,600 Ci and 30 Ci respectively. The tested walls are made of heavy concrete or lead, with the maximum thickness of 1,200 mm for concrete cell and 200 mm for lead cell. At first, we measured the dose rates for several standard walls and the result was used as standard reference. We also measured dose rates for hot cell walls by the same method and compared with reference. The number of testing points are 6,000 and we found out defect for several points which are mostly located in boundaries between embedded material and concrete. The defective areas were re tested after repaired and results for the areas were acceptable

  19. Radiation exposure: Cytogenetic tests. Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Forty test subjects who, either during or after the reactor accident of Chernobyl (26th April 1986), stayed at a building site at Shlobin 150 km away, were examined for spontaneously occurring as well as mitomycin C-induced Sister Chromatid Exchanges (SCE). The building site staff, who underwent a whole-body radionuclide count upon their return to Austria (June through September 1986), were used for the cytogenetic tests. The demonstration of the SCE was made from whole-blood cultures by the fluorescence/Giemse technique. At last 20 Metaphases of the 2nd mitotic cycle were evaluated per person. The radiation doses of the test subjects were calculated by adding the external exposure determined on the building site, the estimated thyroid dose through I-131, and the measured incorporation of Cs-134 and Cs-137. The subjects were divided into two groups for statistical analysis: One was a more exposed group (proven stay at Shlobin between 26th April and 31st May 1986, mostly working in the open air) and the other a less exposed group for comparison (staying at Shlobin from 1st Juni 1986 and working mainly indoors). (orig.)

  20. SMART Reactor Flow Distribution Test (SCOP-E-01)

    International Nuclear Information System (INIS)

    A Reactor Flow Distribution Test Facilities for SMART, named SCOP (SMART Core Flow and Pressure Test Facility), were designed in order to simulate the distributions of (1) core flow and (2) reactor sectional flow resistance and flow rates. This report summaries and analyzes the SCOP-E-01 Test which simulated the reactor internal flow distribution under the steady state conditions with the same flow rate at each loop. The primary parameters, which are represented by static/differential pressure, flow rate and temperature were found to satisfy well the requirement of instrumentation and uncertainties. In order to evaluate overall quality of test results, various secondary parameters were selected and analyzed, which shows that the quality of data are good. From the various hydraulic data representing the hydraulics of the SMART reactor, the soundness and performance of the reactor design can be demonstrated. The test data will be utilized as boundary conditions for the thermal margin analysis of SMART reactor

  1. Shielding and criticality safety analyses of a Latin American cask for transportation and interim storage of spent fuel from research reactors

    International Nuclear Information System (INIS)

    Shielding and criticality safety calculations carried out for the Latin American interim storage and transportation cask are presented. Such a dual purpose cask is being designed for the spent fuel elements of research reactors in the region. The Monte Carlo transport code MCNP4B was utilized for the criticality safety analysis part, and SCALE4.4A for shielding. The analyses considered two types of fuel assemblies utilized in the region and the results show that both types can be loaded in the designed cask baskets in compliance with the safety criteria. (author)

  2. Development of fiber optic sensors for advanced aircraft testing and control

    Science.gov (United States)

    Meller, Scott A.; Jones, Mark E.; Wavering, Thomas A.; Kozikowski, Carrie L.; Murphy, Kent A.

    1999-02-01

    Optical fiber sensors, because of the small size, low weight, extremely high information carrying capability, immunity to electromagnetic interference, and large operational temperature range, provide numerous advantages over conventional electrically based sensors. This paper presents preliminary results from optical fiber sensor design for monitoring acceleration on aircraft. Flight testing of the final accelerometer design will be conducted on the F-18 Systems Research Aircraft at NASA Dryden Flight Research Center in Edwards, CA.

  3. Design of a Total Pressure Distortion Generator for Aircraft Engine Testing

    OpenAIRE

    Cramer, Kevin Brendan

    2002-01-01

    Design of Total Pressure Distortion Generator for Aircraft Engine Testing by Kevin B. Cramer Committee Co-Chair: W.F. Oâ Brien Committee Co-Chair: P.S. King Mechanical Engineering (ABSTRACT) A new method and mechanism for generating non-uniform, or distorted, aircraft engine inlet flow is being developed in order to account for dynamic changes during the creation and propagation of the distortion. Total pressure distortions occur in gas turbine engines when the i...

  4. Shielding calculations by using the analytic methods : Application to the radio-isotopes production in the CENM reactor

    International Nuclear Information System (INIS)

    Full text: this work is part of developing an analytical method for solving the neutrons transport equation in improving the treatment of the anisotropy of neutron scattering through heterogeneous shielding. We also develop the tools necessary for the formation of multigroup libraries (cross section) with the best choice of the weighting function. Among the radioprotection problems of radioisotopes production experiments in the research reactor core is mainly the photons gamma generation produced by radiative capture: activation of samples and their capsules. So, in order to review the safety of operating personnel and the public is essential to quantify the neutrons flux and gamma photons produced. In this study a numerical methods is used in two different Fortran program to solve the neutron transport problem and to determine the neutron and photon flux. This program based on the Monte Carlo method: the neutron is born with a unit statistical weight, this corrected after each imposed scattering event during its whole history within the shield. The final neutron statistical weight is used in an appropriate estimator to determine the searched response. The generated gamma rays by neutron capture are calculated of different isotopes, and then the equivalent dose rate is evaluated in biological tissue for different neutron source energies. We have identified and studied the choice of the best weighting function to calculate a library of multigroup cross sections self protected by using the energy weighting function. A Fortran program is used as a mathematical tool to solve the neutron slowing down equation in infinite homogeneous medium for different dilutions. We determined the energetic flux distribution and the effective integrals. The results of both calculations are in a good agreement; the relative error is less than 0.5%.

  5. Advanced Gas Cooled Reactor Materials Program. Reducing helium impurity depletion in HTGR materials testing

    International Nuclear Information System (INIS)

    Moisture depletion in HTGR materials testing rigs has been empirically studied in the GE High Temperature Reactor Materials Testing Laboratory (HTRMTL). Tests have shown that increased helium flow rates and reduction in reactive (oxidizable) surface area are effective means of reducing depletion. Further, a portion of the depletion has been shown to be due to the presence of free C released by the dissociation of CH4. This depletion component can be reduced by reducing the helium residence time (increasing the helium flow rate) or by reducing the CH4 concentration in the test gas. Equipment modifications to reduce depletion have been developed, tested, and in most cases implemented in the HTRMTL to date. These include increasing the Helium Loop No. 1 pumping capacity, conversion of metallic retorts and radiation shields to alumina, isolation of thermocouple probes from the test gas by alumina thermowells, and substitution of non-reactive Mo-TZM for reactive metallic structural components

  6. Ames Research Center Shear Tests of SLA-561V Heat Shield Material for Mars-Pathfinder

    Science.gov (United States)

    Tauber, Michael; Tran, Huy; Henline, William; Cartledge, Alan; Hui, Frank; Tran, Duoc; Zimmerman, Norm

    1996-01-01

    This report describes the results of arc-jet testing at Ames Research Center on behalf of Jet Propulsion Laboratory (JPL) for the development of the Mars-Pathfinder heat shield. The current test series evaluated the performance of the ablating SLA-561V heat shield material under shear conditions. In addition, the effectiveness of several methods of repairing damage to the heat shield were evaluated. A total of 26 tests were performed in March 1994 in the 2 in. X 9 in. arc-heated turbulent Duct Facility, including runs to calibrate the facility to obtain the desired shear stress conditions. A total of eleven models were tested. Three different conditions of shear and heating were used. The non-ablating surface shear stresses and the corresponding, approximate, non-ablating surface heating rates were as follows: Condition 1, 170 N/m(exp 2) and 22 W/cm(exp 2); Condition 2, 240 N/m(exp 2) and 40 W/cm(exp 2); Condition 3, 390 N/m(exp 2) and 51 W/cm(exp 2). The peak shear stress encountered in flight is represented approximately by Condition 1; however, the heating rate was much less than the peak flight value. The peak heating rate that was available in the facility (at Condition 3) was about 30 percent less than the maximum value encountered during flight. Seven standard ablation models were tested, of which three models were instrumented with thermocouples to obtain in-depth temperature profiles and temperature contours. An additional four models contained a variety of repair plugs, gaps, and seams. These models were used to evaluated different repair materials and techniques, and the effect of gaps and construction seams. Mass loss and surface recession measurements were made on all models. The models were visually inspected and photographed before and after each test. The SLA-561 V performed well; even at test Condition 3, the char remained intact. Most of the resins used for repairs and gap fillers performed poorly. However, repair plugs made of SLA-561V performed

  7. Rotary Balance Wind Tunnel Testing for the FASER Flight Research Aircraft

    Science.gov (United States)

    Denham, Casey; Owens, D. Bruce

    2016-01-01

    Flight dynamics research was conducted to collect and analyze rotary balance wind tunnel test data in order to improve the aerodynamic simulation and modeling of a low-cost small unmanned aircraft called FASER (Free-flying Aircraft for Sub-scale Experimental Research). The impetus for using FASER was to provide risk and cost reduction for flight testing of more expensive aircraft and assist in the improvement of wind tunnel and flight test techniques, and control laws. The FASER research aircraft has the benefit of allowing wind tunnel and flight tests to be conducted on the same model, improving correlation between wind tunnel, flight, and simulation data. Prior wind tunnel tests include a static force and moment test, including power effects, and a roll and yaw damping forced oscillation test. Rotary balance testing allows for the calculation of aircraft rotary derivatives and the prediction of steady-state spins. The rotary balance wind tunnel test was conducted in the NASA Langley Research Center (LaRC) 20-Foot Vertical Spin Tunnel (VST). Rotary balance testing includes runs for a set of given angular rotation rates at a range of angles of attack and sideslip angles in order to fully characterize the aircraft rotary dynamics. Tests were performed at angles of attack from 0 to 50 degrees, sideslip angles of -5 to 10 degrees, and non-dimensional spin rates from -0.5 to 0.5. The effects of pro-spin elevator and rudder deflection and pro- and anti-spin elevator, rudder, and aileron deflection were examined. The data are presented to illustrate the functional dependence of the forces and moments on angle of attack, sideslip angle, and angular rate for the rotary contributions to the forces and moments. Further investigation is necessary to fully characterize the control effectors. The data were also used with a steady state spin prediction tool that did not predict an equilibrium spin mode.

  8. Reactor design of the SP-100 nuclear assembly test

    International Nuclear Information System (INIS)

    The Nuclear Assembly Test is currently being designed to demonstrate the performance characteristics of a 100-kWe version of the power source for the SP-100 Generic Flight System. Particular emphasis will be placed upon the operation of the prototypical ground test reactor under conditions of high-working temperatures and long life. The key features of the reactor include a small, compact core with component materials consisting of refractory metals and alloys. Because of the unique features of the SP-100 system, extensive use is made of Monte Carlo methods in the design and analysis of the reactor configuration. In addition, detailed testing of the reactor design has been carried out in the Zero Power Physics Reactor facility to provide calibration factors for the principal performance parameters. The key features of the test reactor design are described in this paper

  9. A Review on the Production Methods and Testing of Textiles for Electro Magnetic Interference (EMI) shielding

    OpenAIRE

    Bagavathi M,; Dr.-Ing. Priyadarshini R

    2015-01-01

    The need of the present generation to protect themselves from electromagnetic radiation due the various technological developments has paved way to the birth of EMI shielding of textiles. The shielding effectiveness of the developed fabric will vary depending upon the fabric or the coating constituents. The shielding requirements for different applications vary widely which has resulted in the development of wide variety of shielding mechanisms and materials which can be used in t...

  10. Reactors

    International Nuclear Information System (INIS)

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

  11. Testing new types of personal protective shielding equipment against ionizing radiation

    International Nuclear Information System (INIS)

    X and gamma ray attenuation and dose rate caused by the shielding layers of the various items of personal protective shielding equipment (PPSE) were measured by using different radionuclide point sources. Thermoluminescent detectors were installed in the layers of some pieces of the shielding clothing and their response to the radiation of dispersed radioactive aerosols was measured in different experimental conditions. (orig.)

  12. Alternate shield material feasibility

    International Nuclear Information System (INIS)

    The feasibility and cost/benefit of using materials other than stainless steel for in-vessel neutron shielding in large LMFBRs were investigated. Canned vibratorally compacted B4C powder shields were found to be much more economical than stainless steel (a savings of $1.1M in loop plant designs and $9.4M in pool plant designs). The helium gas pressure buildup in B4C shields placed around LMFBR in-vessel components (direct reactor heat exchangers in a loop reactor and intermediate heat exchangers in a pool reactor) would only be 0.04 atm after 40 y of reactor operation (with 80% dense powder). The irradiation-induced swelling of the B4C would only be 0.002%. No adverse reactor impact would occur if the B4C escaped from the B4C shields

  13. Alternate shield material feasibility

    Energy Technology Data Exchange (ETDEWEB)

    Specht, E.R.; Levitt, L.B.

    1984-04-01

    The feasibility and cost/benefit of using materials other than stainless steel for in-vessel neutron shielding in large LMFBRs were investigated. Canned vibratorally compacted B/sub 4/C powder shields were found to be much more economical than stainless steel (a savings of $1.1M in loop plant designs and $9.4M in pool plant designs). The helium gas pressure buildup in B/sub 4/C shields placed around LMFBR in-vessel components (direct reactor heat exchangers in a loop reactor and intermediate heat exchangers in a pool reactor) would only be 0.04 atm after 40 y of reactor operation (with 80% dense powder). The irradiation-induced swelling of the B/sub 4/C would only be 0.002%. No adverse reactor impact would occur if the B/sub 4/C escaped from the B/sub 4/C shields.

  14. Structural design for aircraft impact loading

    International Nuclear Information System (INIS)

    The distribution of military aircraft and proximity to commercial air routes requires the analysis of aircraft impact effect on nuclear power plant facilities in Europe. The 'hardened-building' approach has led to the consideration of severe shock and vibration caused by the aircraft impact and development of corresponding floor response spectra for component design. The reactor auxiliary system building allows a more defensive alternate in the form of a partially softened design. In this approach the equipment layout is arranged such that equipment performing either safety functions or having the potential for significant release of radioctivity (upon destruction) is located in the central area of the plant and is enclosed in thick concrete walls for shielding and protection purposes. The non-safety class equipment is arranged in the area peripheral to the hardened central area and enclosed in thin concrete walls. Since the kinetic energy of the impacting aircraft is absorbed by the collapsed thin walls and ceilings, the vibrational effect on the safety class equipment is drastically reduced. In order to achieve the objective of absorbing high kinetic energy and yet reduce the shock and vibration effects, the softened exterior walls require low resistance and high ductility. In order not to increase the construction cost, and yet to assure the safety of the plant, some dynamic tests of conventionally reinforced slabs have to be performed all the way to collapse. These calculations have assumptions of achieving the maximum velocity instantaneously after impact, and take into account the kinetic energy in the broken wall. Nonlinear equations of motion are also formulated and solved. The results indicate that the phantom jet would go through the first wall. The second wall would stop the jet, but would sustain some permanent deformation and damage

  15. Full-scale aircraft impact test for evaluatioin of impact force

    International Nuclear Information System (INIS)

    For estimating the global elasto-plastic structural response of critical concrete structures subjected to an aircraft crash, the time dependent impact force of a flat rigid barrier against a normally impacting aircraft was first evaluated and then the response, to the impact force, was calculated. In this approach, a significant problem was to determine the impact force for the aircraft against a rigid target. A review of the method proposed to determine the impact forces showed that all were based on analytical methods. However, in these analytical methods, there were many assumptions and many questions remained to be answered. Because of the uncertainty involved in the analytical prediction of the impact force, a full- scale aircraft impact test was performed and an extensive suite of response measurements was obtained. In this paper, these measurements are analyzed to evaluate the impact force accurately. Also, the results were used to evaluate existing analytical methods for prediction of the impact force. 7 refs., 10 figs

  16. Testing stand for cosmic gas-cooling fast reactor's sample

    International Nuclear Information System (INIS)

    For carrying out of technical decision and nuclear, radiation and technological safety of gas-cooling space nuclear power plants is elaborating gas-cooling fast reactor's testing stand. In the base of its draft is taken conception of the reactor with filling up type reactor core on the base of ball fuel elements and radial coolant flowing. On the testing stand would suggested carrying out testing for study neutron and physical parameters of gas-cooling reactor, its behaviour under accident simulation. In the reactor core will suggest use carbon nitrides fuel elements with tungsten cover, provides under nominal regime relatively low fission products yield to first contour of device. Construction of fuel element was carrying out on reactor and non reactor testing and its calculated on working resource about 3000 hours. Constructive materials of reactor core have lower melting temperature, that provides organized in good time remove fuel element to containers placed under reactor in case connected with hypothetical accident. In the construction of reactor for seen tree-contours system of heat transfer and its provides multistage system of barriers against fission products yield to environment. tabs.1

  17. New Sensors for Irradiation Testing at Materials and Test Reactors

    International Nuclear Information System (INIS)

    Enhanced instrumentation, capable of providing real-time measurements of parameters during fuels and material irradiations, is required to support irradiation testing requested by US nuclear research programs. For example, several research programs funded by the US Department of Energy (US DOE) are emphasizing the use of first principle models to characterize the performance of fuels and materials. To facilitate this approach, high fidelity, real-time data are essential to demonstrate the performance of these new fuels and materials during irradiation testing. Furthermore, sensors that obtain such data in US MTRs, such as the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL), must be miniature, reliable, and able to withstand high fluxes and high temperatures. Depending on program requirements, sensors may need to obtain data in inert gas, pressurized water, or liquid metal environments. To address these needs, INL has developed and deployed several new sensors to support irradiation testing in US DOE programs. The paper identifies the sensors currently available to support higher flux US MTR irradiations. Recent results and products from sensor research and development are highlighted. In particular, progress in deploying enhanced in-pile sensors for detecting temperature, elongation, and thermal conductivity is emphasized. Finally, initial results from research to evaluate the viability of ultrasonic and fiber optic technologies for irradiation testing are summarized. (author)

  18. LOFA [loss of flow accident] and LOCA [loss of coolant accident] in the TIBER-II engineering test reactor: Appendix A-4

    International Nuclear Information System (INIS)

    This paper describes the preliminary analysis of LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor breeding shield. TIBER-II is a compact reactor with a major radius of 3 m and thus requires a thin, high efficiency shield on the inboard side. The use of tungsten in the inboard shield implies a rather high rate of afterheat upon plasma shutdown, which must be dissipated in a controlled manner to avoid the possibility of radioactivity release or threatening the investment. Because the shield is cooled with an aqueous solution, LOFA does not pose a problem as long as natural convection can be established. LOCA, however, has more serious consequences, particularly on the inboard side. Circulation of air by natural convection is proposed as a means for dissipating the inboard shield decay heat. The safety and environmental implications of such a scheme are evaluated. It is shown that the inboard shield temperature never exceeds 5100C following LOCA posing no hazard to reactor personnel and not threatening the investment. 7 refs., 6 figs

  19. Design of high temperature irradiation materials inspection cells. (Spent fuel inspection cells) in the High Temperature Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ino, Hiroichi; Ueta, Shouhei; Suzuki, Hiroshi; Sawa, Kazuhiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tobita, Tsutomu [Nuclear Engineering Company, Ltd., Tokai, Ibaraki (Japan)

    2002-01-01

    This report summarizes design requirements and design results for shields, ventilation system and fuel handling devices for the high temperature irradiation materials inspection cells (spent fuel inspection cells). These cells are small cells to carry out few post-irradiation examinations of spent fuels, specimen, etc., which are irradiated in the High Temperature Engineering Test Reactor, since the cells should be built in limited space in the HTTR reactor building, the cells are designed considering relationship between the cells and the reactor building to utilize the limited space effectively. The cells consist of three partitioned hot cells with wall for neutron and gamma-ray shields, ventilation system including filtering units and fuel handling devices. The post-irradiation examinations of the fuels and materials are planed by using the cells and the Hot Laboratory of the Japan Materials Testing Reactor to establish the technology basis on high temperature gas-cooled reactors (HTGRs). In future, irradiation tests and post-irradiation examinations will be carried out with the cells to upgrade present HTGR technologies and to make the innovative basic research on high-temperature engineering. (author)

  20. Design of high temperature irradiation materials inspection cells. (Spent fuel inspection cells) in the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    This report summarizes design requirements and design results for shields, ventilation system and fuel handling devices for the high temperature irradiation materials inspection cells (spent fuel inspection cells). These cells are small cells to carry out few post-irradiation examinations of spent fuels, specimen, etc., which are irradiated in the High Temperature Engineering Test Reactor, since the cells should be built in limited space in the HTTR reactor building, the cells are designed considering relationship between the cells and the reactor building to utilize the limited space effectively. The cells consist of three partitioned hot cells with wall for neutron and gamma-ray shields, ventilation system including filtering units and fuel handling devices. The post-irradiation examinations of the fuels and materials are planed by using the cells and the Hot Laboratory of the Japan Materials Testing Reactor to establish the technology basis on high temperature gas-cooled reactors (HTGRs). In future, irradiation tests and post-irradiation examinations will be carried out with the cells to upgrade present HTGR technologies and to make the innovative basic research on high-temperature engineering. (author)

  1. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    Energy Technology Data Exchange (ETDEWEB)

    Mourao, Rogerio P.; Leite da Silva, Luiz [Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte (Brazil); Miranda, Carlos A.; Mattar Neto, Miguel [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil); Quintana, Jose F.A.; Saliba, Roberto O. [Comision Nacional de Energia Atomica, Bariloche (Argentina); Novara, Oscar E. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    2013-07-01

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

  2. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    International Nuclear Information System (INIS)

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

  3. Method of testing fuel assemblies for nuclear reactors

    International Nuclear Information System (INIS)

    The stresses occurring in the fuel assemblies are simulated by power excursions. For this purpose the fuel assembly is placed in the neutron field of a test reactor and for a short time can be exposed to the much higher neutron field of a pulsed reactor. One possibility of design provides for the test and the pulsed reactor lying one above the other, separated by a neutron absorber and penetrated by a common irradiation channel. The fuel assembly then is to be moved from the position in the test reactor to the position in the pulsed reactor. The other possibility is to make the irradiation duct pass along the gap between both reactors and, by means of a tube-shaped absorber, open one or the other irradiation field. (DG)

  4. Cross Section Evaluation Group shielding benchmark compilation. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    Rose, P.F.; Roussin, R.W.

    1983-12-01

    At the time of the release of ENDF/B-IV in 1974, the Shielding Subcommittee had identified a series of 12 shielding data testing benchmarks (the SDT series). Most were used in the ENDF/B-IV data testing effort. A new concept and series was begun in the interim, the so-called Shielding Benchmark (SB) series. An effort was made to upgrade the SDT series as far as possible and to add new SB benchmarks. In order to be designated in the SB class, both an experiment and analysis must have been performed. The current recommended benchmark for Shielding Data Testing are listed. Until recently, the philosophy has been to include only citations to published references for shielding benchmarks. It is now our intention to provide adequate information in this volume for proper analysis of any new benchmarks added to the collection. These compilations appear in Section II, with the SB5 Fusion Reactor Shielding Benchmark as the first entry.

  5. Experimental Photogrammetric Techniques Used on Five Full-Scale Aircraft Crash Tests

    Science.gov (United States)

    Littell, Justin D.

    2016-01-01

    Between 2013 and 2015, full-scale crash tests were conducted on five aircraft at the Landing and Impact Research Facility (LandIR) at NASA Langley Research Center (LaRC). Two tests were conducted on CH-46E airframes as part of the Transport Rotorcraft Airframe Crash Testbed (TRACT) project, and three tests were conduced on Cessna 172 aircraft as part of the Emergency Locator Transmitter Survivability and Reliability (ELTSAR) project. Each test served to evaluate a variety of crashworthy systems including: seats, occupants, restraints, composite energy absorbing structures, and Emergency Locator Transmitters. As part of each test, the aircraft were outfitted with a variety of internal and external cameras that were focused on unique aspects of the crash event. A subset of three camera was solely used in the acquisition of photogrammetric test data. Examples of this data range from simple two-dimensional marker tracking for the determination of aircraft impact conditions to entire full-scale airframe deformation to markerless tracking of Anthropomorphic Test Devices (ATDs, a.k.a. crash test dummies) during the crash event. This report describes and discusses the techniques used and implications resulting from the photogrammetric data acquired from each of the five tests.

  6. Dynamic stability testing of aircraft - Needs versus capabilities

    Science.gov (United States)

    Orlik-Rueckemann, K. J.

    1973-01-01

    Highlights of a recent survey of the future needs for dynamic stability information for such aerospace vehicles as the Space Shuttle and advanced high-performance military aircraft, indicating the importance of obtaining this information for high-angle-of-attack high-Reynolds-number conditions. A review of the wind-tunnel capabilities in North America for measuring dynamic stability derivatives reveals an almost total lack of such capabilities for Mach numbers above 0.1 at angles of attack higher than 25 deg. In addition, capabilities to obtain certain new cross-coupling derivatives and information on effects of the coning motion are almost completely lacking. Recommendations are made regarding equipment that should be constructed to remedy this situation.

  7. Study and implementation of the CADIS methodology to research reactor shielding design

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Gregorio S.; Shorto, Julian M.B.; Santos, Adimir dos, E-mail: greguis@gmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    The Consistent Adjoint Driven Importance Sampling (CADIS) is a methodology that basically uses source biasing and a mesh-based importance map. Therefore, to make the best use of an importance map, the map must be consistent with the source biasing. To achieve this consistency, a Sn calculation could be made to improve the importance map and the computational performance. The MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations) code does that and this work intends to study the code options to generate the importance map. A pool type 10 MW research reactor was designed in a simple way just to study the prompt gamma rays penetration in the concrete and therefore study the CADIS methodology applied to point detectors and mesh tallies. By keeping constant the simulation time and the CPU (Central Processing Unit) power a significant improvement was achieved in the relative errors for the point detectors and for the mesh tally. (author)

  8. Status of the irradiation test vehicle for testing fusion materials in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States); Palmer, A.J.; Ingram, F.W. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States); Wiffen, F.W. [Dept. of Energy, Germantown, MD (United States). Office of Fusion Energy

    1998-09-01

    The design of the irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) has been completed. The main application for the ITV is irradiation testing of candidate fusion structural materials, including vanadium-base alloys, silicon carbide composites, and low-activation steels. Construction of the vehicle is underway at the Lockheed Martin Idaho Technology Company (LMITCO). Dummy test trains are being built for system checkout and fine-tuning. Reactor insertion of the ITV with the dummy test trains is scheduled for fall 1998. Barring unexpected difficulties, the ITV will be available for experiments in early 1999.

  9. Measurement and analysis of aircraft and vehicle LRCS in outfield test

    Science.gov (United States)

    Cao, Chang-Qing; Zeng, Xiao-dong; Fan, Zhao-jin; Feng, Zhe-jun; Lai, Zhi

    2015-04-01

    The measurement of aircraft and vehicle Laser Radar Cross Section (LRCS) is of crucial importance for the detection system evaluation and the characteristic research of the laser scattering. A brief introduction of the measuring theory of the laser scattering from the full-scale aircraft and vehicle targets is presented in this paper. By analyzing the measuring condition in outfield test, the laser systems and test steps are designed for full-scale aircraft and vehicle LRCS and verified by the experiment in laboratory. The processing data error 7% below is obtained of the laser radar cross section by using Gaussian compensation and elimination of sky background for original test data. The study of measurement and analysis proves that the proposed method is effective and correct to get laser radar cross section data in outfield test. The objectives of this study were: (1) to develop structural concepts for different LRCS fuselage configurations constructed of conventional materials; (2) to compare these findings with those of aircrafts or vehicles; (3) to assess the application of advanced materials for each configuration; (4) to conduct an analytical investigation of the aerodynamic loads, vertical drag and mission performance of different LRCS configurations; and (5) to compare these findings with those of the aircrafts or vehicles.

  10. Theoretical and experimental substantiation of the fusion reactor divertor and first wall shield concept by means of lithium capillary porous systems

    International Nuclear Information System (INIS)

    Basic properties of the lithium capillary porous systems (LCPS) for protection of divertor and the first wall shield TNR are described. The LCPS behaviour under conditions of impact of high stationary and pulse heat loads are studied. Lithium ions and neutrons migration in the tokamak reactor with liquid lithium is studied numerically and lithium flow in the crossed magnetic field experimentally. The LCPS behaviour is studied also experimentally under the conditions of the tokamak plasma impact

  11. Induced radioactivity and the field of activation products γ-radiation in shielding on investigative reactor IRI-MIFI in problem of decommissioning

    International Nuclear Information System (INIS)

    Problem solution peculiarities for forecasting induced radioactivity and characteristics of activation product γ radiation field in single calculation of the primary and secondary particle transport are considered proceeding from the notion of the activation radiation as the secondary delayed radiation in contrast to the secondary instantaneous captured radiation. Calculation program complex is illustrated by induced activation forecasting problem and 60Co γ-radiation maximum equivalent dose rate in shield of the IRI-MIFI research reactor

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-01-01

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

  14. Measurements and calculation of the activation of the biologic shield of the Lingen BWR power reactor definitively stopped (in view of dismantling)

    International Nuclear Information System (INIS)

    For the dismantling planning of a power reactor, it is important to know among others the depth of activation of the biological shield. A large sampling and measurement program joint to computer calculations, has given data which will allow to avoid in the future high-cost measurement programs. One shows that the calculation of activation induced by neutrons in the median plane of the core, to determine the zone from which concrete is only slightly activated. In the reactor considered, this zone does not reach the external concrete (or first layer of concrete)

  15. Tests of a novel method to assay SNM using polarized photofission and its sensitivity in the presence of shielding

    International Nuclear Information System (INIS)

    A novel method to identify Special Nuclear Material was recently developed (Mueller et al., 2014) [1]. This method relies upon using a linearly polarized γ-ray beam to induce photofission of a sample and then comparing the prompt fission neutron yields in and out of the plane of beam polarization. The present paper will describe experimental tests of this new technique and assess its sensitivity in the presence of shielding. The capability of this technique to measure the enrichment of uranium was tested by using combinations of thin 235U and 238U foils of known enrichments. The sensitivity of this assay to shielding by lead, steel, and polyethylene was experimentally measured and simulated using GEANT4. These tests show that the measured asymmetry can indeed be used to determine the enrichment of materials composed of an admixture of 235U and 238U, and this asymmetry is relatively insensitive to moderate amounts of shielding

  16. Fuels for space nuclear power systems. 3. Innovative Semi-spherical Pb-Hf-Cu Shield for a Fissioning Plasma Core Reactor

    International Nuclear Information System (INIS)

    This study investigated the shielding materials and requirements for a fissioning plasma core reactor (FPCR) with a magnetohydrodynamic (MHD) power conversion system for multimegawatt space power and propulsion applications. The FPCR is a liquid-vapor core reactor concept operating with metallic uranium or uranium tetrafluoride (UF4) vapor as the fissioning fuel and alkali metals or their fluorides as working fluid in a closed Rankine cycle with MHD energy conversion. This concept is under study for its potential to provide space power at a low specific mass of 3 with a length-to-diameter ratio of one was selected. This design based on earlier gas core reactor studies incorporates a 50-cm BeO radial reflector with additional 25-cm-thick BeO disk-shaped reflectors at the top and bottom of the cylindrical core. Liquid hydrogen tanks for propulsion and refrigeration were modeled between the reactor/power generation complex and the payload/habitable regions of the vessel or space station and lying along the boom, which can be from 30 to 60 m in length. Although the liquid hydrogen is not very dense (∼0.1 g/cm3), there is a considerable amount present (50 t is commonly referenced). A model of this system was developed in the MCNP-4C general Monte Carlo code, which was used to calculate the dose rate at various distances from the power-generating system. Sources of both fission neutrons (prompt and delayed) and gamma rays (prompt and decay) from fission were modeled. The neutron sampling distribution was taken as a Watt fission spectrum. The energy distribution of gamma rays from fission was taken from Ref. 1 and consists of a total average of 12.15 gamma rays per fission with ∼70% < 1 MeV and 27% between 1 and 3 MeV. Various shield designs were modeled, and corresponding dose rates were calculated. A criterion of <10 rems/h to the payload module was established for all shield designs. It was assumed that for a manned station or vessel, additional shielding would be

  17. Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP

    International Nuclear Information System (INIS)

    The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245x10-4 s was recorded for the new boron carbide designed model while a value of 1.5571x10-7 s was recorded for the original MCNP design of the GHARR-1.

  18. Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP.

    Science.gov (United States)

    Abrefah, R G; Sogbadji, R B M; Ampomah-Amoako, E; Birikorang, S A; Odoi, H C; Nyarko, B J B

    2011-01-01

    The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245 × 10(-4)s was recorded for the new boron carbide designed model while a value of 1.5571 × 10(-7)s was recorded for the original MCNP design of the GHARR-1. PMID:20637646

  19. Systems for neutronic, thermohydraulic and shielding calculation in personal computers

    International Nuclear Information System (INIS)

    The MTR-PC (Materials Testing Reactors-Personal Computers) system has been developed by the Nuclear Engineering Division of INVAP S.E. with the aim of providing working conditions integrated with personal computers for design and neutronic, thermohydraulic and shielding analysis for reactors employing plate type fuel. (Author)

  20. Proceedings of the international symposium on materials testing reactors

    International Nuclear Information System (INIS)

    This report is the Proceedings of the International Symposium on Materials Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The symposium was held on July 16 to 17, 2008, at the Oarai Research and Development Center of JAEA. This symposium was also held for the 40th anniversary ceremony of Japan Materials Testing Reactor (JMTR) from achieving its first criticality. The objective of the symposium is to exchange the information on current status, future plan and so on among each testing reactors for the purpose of mutual understanding. There were 138 participants from Argentina, Belgium, France, Indonesia, Kazakhstan, Korea, the Russian Federation, Sweden, the United State, Vietnam and Japan. The symposium was divided into four technical sessions and three topical sessions. Technical sessions addressed the general topics of 'status and future plan of materials testing reactors', 'material development for research and testing reactors', irradiation technology (including PIE technology)' and 'utilization with materials testing reactors', and 21 presentations were made. Also the topical sessions addressed 'establishment of strategic partnership', 'management on re-operation work at reactor trouble' and 'basic technology for neutron irradiation tests in MTRs', and panel discussion was made. The 21 of the presented papers are indexed individually. (J.P.N.)

  1. An investigation of scale model testing of VTOL aircraft in hover

    Science.gov (United States)

    Hill, W. G., Jr.; Jenkins, R. C.; Dudley, M. R.

    1982-01-01

    Utilizing the unique opportunity created by full scale hover testing of the twin-jet Grumman Design 698 VTOL aircraft in the NASA-Ames Hover Facility, a series of experiments was conducted to evaluate the effectivness of scale model testing in predicting full scale behavior. Interference forces were found to be sensitive to aircraft lower surface geometry, but when the geometry was modeled accurately the small scale results matched full scale forces guite well. The interference forces were found to be insensitive to core nozzle temperature and fan nozzle pressure ratio. The results clearly demonstrate that small scale models can be reliably utilized for aircraft and technology development when the appropriate sensitivities are recognized.

  2. A low-risk aqueous lithium salt blanket for engineering test reactors

    International Nuclear Information System (INIS)

    A simple blanket concept is proposed based on 1-3 wt.% lithium dissolved as a salt in low temperature (80 degrees C) and low pressure (0.1 MPa) water. This concept can provide, for example, a 0.5 tritium breeding ratio with 60% steel structure and 70% coverage. The use of neutron multipliers, other structural materials (especially zirconium alloys), higher coverage and higher lithium salt concentrations allows tritium breeding ratios over unity if necessary. Other advantages of this concept include the simple shield-like geometry, substantial structural volume for mechanical strength, excellent heat transfer ability of water coolant, efficient neutron and gamma shielding through the combination of high-Z structure and low-Z water, and conventional tritium recovery and control technology. This concept could initially provide the shielding needs for an engineering test reactor and later, by the addition of lithium salt and tritium recovery systems, also provide tritium breeding. This staged operation and liquid breeder/coolant allows control over the tritium inventory in the device without machine disassembly. 14 refs

  3. Development, utilization, and future prospects of materials test reactors

    International Nuclear Information System (INIS)

    Reactor radiation affects the chemical and physical properties of materials. These changes can be very drastic in certain cases. Special test reactors have therefore been built since the 1950's and specific skills were developed to expose materials specimens to the precise irradiation conditions required. Materials testing reactors are those research reactor facilities which are designed and operated predominantly for studies into radiation damage. About a dozen plants in European communities (EC) Member States and in the US can be identified in this category, with 5 to 100 MW fission power and neutron fluxes between 5 x 1013 and 1015 cm-2s-1. The paper elaborates common aspects of development, utilization, and future prospects of US and EC materials testing reactors, and indicates the most significant differences

  4. Safety analysis report for the National Low-Temperature Neutron Irradiation Facility (NLTNIF) at the ORNL Bulk Shielding Reactor (BSR)

    International Nuclear Information System (INIS)

    This report provides information concerning: the experiment facility; experiment assembly; instrumentation and controls; materials; radioactivity; shielding; thermodynamics; estimated or measured reactivity effects; procedures; hazards; and quality assurance

  5. ENFICA-FC: Design of transport aircraft powered by fuel cell & flight test of zero emission 2-seater aircraft powered by fuel cells fueled by hydrogen

    OpenAIRE

    Cestino, Enrico; Borello, Fabio; Romeo, Giulio

    2013-01-01

    Fuel cells could become the main power source for small general aviation aircraft or could replace APU and internal sub-systems on larger aircraft, to obtain all-electric or more-electric air vehicles. There are several potential advantages of using such a power source, that range from environmental and economic issues to performance and operability aspects. A preliminary design is reported. Also, the paper contains a description of testing activities related to experimental flights of an all...

  6. Full-Scale Structural and NDI Validation Tests of Bonded Composite Doublers for Commercial Aircraft Applications

    Energy Technology Data Exchange (ETDEWEB)

    Roach, D.; Walkington, P.

    1999-02-01

    Composite doublers, or repair patches, provide an innovative repair technique which can enhance the way aircraft are maintained. Instead of riveting multiple steel or aluminum plates to facilitate an aircraft repair, it is possible to bond a single Boron-Epoxy composite doubler to the damaged structure. Most of the concerns surrounding composite doubler technology pertain to long-term survivability, especially in the presence of non-optimum installations, and the validation of appropriate inspection procedures. This report focuses on a series of full-scale structural and nondestructive inspection (NDI) tests that were conducted to investigate the performance of Boron-Epoxy composite doublers. Full-scale tests were conducted on fuselage panels cut from retired aircraft. These full-scale tests studied stress reductions, crack mitigation, and load transfer capabilities of composite doublers using simulated flight conditions of cabin pressure and axial stress. Also, structures which modeled key aspects of aircraft structure repairs were subjected to extreme tension, shear and bending loads to examine the composite laminate's resistance to disbond and delamination flaws. Several of the structures were loaded to failure in order to determine doubler design margins. Nondestructive inspections were conducted throughout the test series in order to validate appropriate techniques on actual aircraft structure. The test results showed that a properly designed and installed composite doubler is able to enhance fatigue life, transfer load away from damaged structure, and avoid the introduction of new stress risers (i.e. eliminate global reduction in the fatigue life of the structure). Comparisons with test data obtained prior to the doubler installation revealed that stresses in the parent material can be reduced 30%--60% through the use of the composite doubler. Tests to failure demonstrated that the bondline is able to transfer plastic strains into the doubler and that

  7. NASA rotor system research aircraft flight-test data report: Helicopter and compound configuration

    Science.gov (United States)

    Erickson, R. E.; Kufeld, R. M.; Cross, J. L.; Hodge, R. W.; Ericson, W. F.; Carter, R. D. G.

    1984-01-01

    The flight test activities of the Rotor System Research Aircraft (RSRA), NASA 740, from June 30, 1981 to August 5, 1982 are reported. Tests were conducted in both the helicopter and compound configurations. Compound tests reconfirmed the Sikorsky flight envelope except that main rotor blade bending loads reached endurance at a speed about 10 knots lower than previously. Wing incidence changes were made from 0 to 10 deg.

  8. The photon and fast neutron spectra measurement and calculation in the concrete of the simulator of WWER-1000 reactor biological shielding

    International Nuclear Information System (INIS)

    The measurements have been performed in the WWER-1000 model in experimental reactor LR-0 in N.R.I. (Nuclear Research Institute). The biological shielding simulator consists of serpentinite concrete with stainless steel cover. It is placed behind the reactor pressure vessel (R.P.V.) simulator situated in a concrete hall outside of LR-0 tank. Simulators of reactor internals as well as the driving core are in the LR-0 reactor tank. The fuel assemblies consist of 312 fuel pins (hexagon of WWER-1000 type) with 1.25 m active length. The measurements were performed before concrete shielding and in the channel in the concrete. The photon and neutron spectra have been measured simultaneously with two-parametric spectrometer with extended energy range [1], 0.5 MeV 10 MeV for both parts of radiation field. The results were by means of monitoring system normalized to other ones in the WWER-1000 model. The calculation of the measured spectra has been performed with the deterministic 3D code T.O.R.T. and cross section library B.U.G.L.E. 96. Comparison of calculated and measured results can enable evaluate reliability of calculation results for deep penetration of radiation, as well as the capability for planning of decommissioning issues. (authors)

  9. Global shielding analysis for the three-element core advanced neutron source reactor under normal operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Slater, C.O.; Bucholz, J.A.

    1995-08-01

    Two-dimensional discrete ordinates radiation transport calculations were performed for a model of the three-element core Advanced Neutron Source reactor design under normal operating conditions. The core consists of two concentric upper elements and a lower element radially centered in the annulus between the upper elements. The initial radiation transport calculations were performed with the DORT two-dimensional discrete ordinates radiation transport code using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub 6} quadrature, and a P{sub 1} Legendre polynomial expansion of the cross sections to determine the fission neutron source distribution in the core fuel elements. These calculations were limited to neutron groups only. The final radiation transport calculations, also performed with DORT using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub l0} quadrature, and a P{sub 3} Legendre polynomial expansion of the cross sections, produced neutron and gamma-ray fluxes over the full extent of the geometry model. Responses (or activities) at various locations in the model were then obtained by folding the appropriate response functions with the fluxes at those locations. Some comparisons were made with VENTURE-calculated (diffusion theory) 20-group neutron fluxes that were summed into four broad groups. Tne results were in reasonably good agreement when the effects of photoneutrons were not included, thus verifying the physics model upon which the shielding model was based. Photoneutrons increased the fast-neutron flux levels deep within the D{sub 2}0 several orders of magnitude. Results are presented as tables of activity values for selected radial and axial traverses, plots of the radial and axial traverse data, and activity contours superimposed on the calculational geometry model.

  10. Test of a new gonad shield in radiographic hip joint examinations of sucklings and infants

    International Nuclear Information System (INIS)

    Preparation and application of a shield consisting of lead rubber are described. Using the shield, a considerable decrease of radiation exposure to male and female infants could be achieved. Therefore it is recommended for application in mass examinations of hip joints. (author)

  11. Testing the integrity of packaging radiation shielding by scanning with radiation source and detector

    International Nuclear Information System (INIS)

    This specification deals with the radiological scanning method of inspection for biological shielding (to be used in transport packaging for gamma emitting sources of radiation), of regular thickness, when the sections are to be checked for integrity and homogeneity; it does not establish the adequacy of design. The shielding materials may be lead, iron, steel, heavy alloy (tungsten), and depleted uranium. (author)

  12. Status and future plan of Japan materials testing reactor

    International Nuclear Information System (INIS)

    The Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Agency (JAEA) is a light water cooling tank typed reactor. JMTR has been used for fuel and material irradiation studies for LWRs, HTGR, fusion reactor and RI production. Since the JMTR is connected with hot laboratory through the canal, re-irradiation tests can conduct easily by safety and quick transportation of irradiation samples. First criticality was achieved in March 1968, and operation was stopped from August, 2006 for the refurbishment. The reactor facilities are refurbished during four years from the beginning of FY 2007, and necessary examination and work are carrying out on schedule. The renewed and upgraded JMTR will start from FY 2011 and operate for a period of about 20 years (until around FY 2030). The usability improvement of the JMTR, such as higher reactor available factor, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussing as the preparations for re-operation. (author)

  13. Manufacturing and testing of full scale prototype for ITER blanket shield block

    International Nuclear Information System (INIS)

    Highlights: • 316L(N)-IG forged steel was successfully fabricated and qualified. • Related R&D activities were implemented to resolve the fabrication issues. • SB #8 FSP was successfully manufactured with conventional fabrication techniques. • All of the validation tests were carried out and met the acceptance criteria. - Abstract: Based on the preliminary design of the ITER blanket shield block (SB) #8, the full scale prototype (FSP) has been manufactured and tested in accordance with pre-qualification program, and related R&D was performed to resolve the technical issues of fabrication. The objective of the SB pre-qualification program is to demonstrate the acceptable manufacturing quality by successfully passing the formal test program. 316L(N)-IG stainless steel forging blocks with 1.80L × 1.12W × 0.43t (m) were developed by using an electric arc furnace, and as a result, the material properties were satisfied with technical specification. In the course of applying conventional fabrication techniques such as cutting, milling, drilling and welding of the forged stainless steel block for the manufacturing of the SB #8 FSP, several technical problems have been addressed. And also, the hydraulic connector of cross-forged material re-melted by electro slag or vacuum arc requires the application of advanced joining techniques such as automatic bore TIG and friction welding. Many technical issues – drilling, welding, slitting, non-destructive test and so on – have been raised during manufacturing. Associated R&D including the computational simulation and coupon testing has been done in collaboration with relevant industries in order to resolve these engineering issues. This paper provides technical key issues and their possible resolutions addressed during the manufacture and formal test of the SB #8 FSP, and related R&D

  14. Manufacturing and testing of full scale prototype for ITER blanket shield block

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sa-Woong, E-mail: swkim12@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, Duck-Hoi; Jung, Hun-Chea [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Sung-Ki [WONIL Co., Ltd., Haman (Korea, Republic of); Kang, Sung-Chan [POSCO Specialty Steel Co., Ltd., Changwon (Korea, Republic of); Zhang, Fu; Kim, Byoung-Yoon [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ahn, Hee-Jae; Lee, Hyeon-Gon; Jung, Ki-Jung [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-04-15

    Highlights: • 316L(N)-IG forged steel was successfully fabricated and qualified. • Related R&D activities were implemented to resolve the fabrication issues. • SB #8 FSP was successfully manufactured with conventional fabrication techniques. • All of the validation tests were carried out and met the acceptance criteria. - Abstract: Based on the preliminary design of the ITER blanket shield block (SB) #8, the full scale prototype (FSP) has been manufactured and tested in accordance with pre-qualification program, and related R&D was performed to resolve the technical issues of fabrication. The objective of the SB pre-qualification program is to demonstrate the acceptable manufacturing quality by successfully passing the formal test program. 316L(N)-IG stainless steel forging blocks with 1.80L × 1.12W × 0.43t (m) were developed by using an electric arc furnace, and as a result, the material properties were satisfied with technical specification. In the course of applying conventional fabrication techniques such as cutting, milling, drilling and welding of the forged stainless steel block for the manufacturing of the SB #8 FSP, several technical problems have been addressed. And also, the hydraulic connector of cross-forged material re-melted by electro slag or vacuum arc requires the application of advanced joining techniques such as automatic bore TIG and friction welding. Many technical issues – drilling, welding, slitting, non-destructive test and so on – have been raised during manufacturing. Associated R&D including the computational simulation and coupon testing has been done in collaboration with relevant industries in order to resolve these engineering issues. This paper provides technical key issues and their possible resolutions addressed during the manufacture and formal test of the SB #8 FSP, and related R&D.

  15. Determination of self-shielding factor of activation detectors in neutron flux and spectrum measurements in RSG-Gas reactor

    International Nuclear Information System (INIS)

    Determination of self-shielding factor and cadmium ration of foil and cylindrical probe have been done by measurement and calculation. Self-shielding factor was determined by dividing the activity of detector with its Al-alloy. Theoretically, self-shielding factor can be determined by numerical solution of two-dimensional integral equations in FORTRAN. For gold foil and wire, the calculation result are quite close to the measurement. The relative difference between calculation and measurement of activity, self-shielding factor and cadmium ratio are respectively less than 11%, 9% and 4%. It is therefore, the calculation program can be used for calculation of other kinds of activation detectors. The application in neutron flux measurement gives a better result especially for epithermal flux. For neutron spectrum measurement, self-shielding correction can avoid resonance peaks in epithermal region due to absorption by activation detectors. (author)

  16. Irradiation test of diagnostic components for ITER application in a fission reactor, Japan Materials Testing Reactor

    International Nuclear Information System (INIS)

    Radiation effects on components and materials will be one of the most serious technological issues in fusion systems realizing burning plasmas. Especially, diagnostic components, which should play crucial roles to control plasmas and to understand physics of burning plasmas, will be exposed to high-flux neutrons and gamma-rays. Dynamic radiation effects will affects performance of components substantially from beginning of exposure to radiation environments, and accumulated radiation effects will gradually degrade their functioning abilities in the course of their services. High-power-density fission reactors will be only realistic tools to simulate the irradiation environments expected in burning-plasma fusion machines such as the ITER, at present. Some key diagnostic components, namely magnetic coils, bolometers, and optical fibers, were irradiation-tested in a fission reactor, JMTR, to evaluate their performances under heavy irradiation environments. Results indicate that the ITER-relevant diagnostic components could be developed in time, though there are still some technological problems to overcome. (author)

  17. Irradiation test of diagnostic components for ITER application in a fission reactor, Japan Materials Testing Reactor

    International Nuclear Information System (INIS)

    Radiation effects on components and materials will be one of the most serious technological issues in fusion systems realizing burning plasmas. Especially, diagnostic components, which should play crucial roles to control plasmas and to understand physics of burning plasmas, will be exposed to high-flux neutrons and gamma-rays. Dynamics radiation effects will affects performance of components substantially from beginning of exposure to radiation environments, and accumulated radiation effects will gradually degrade their functioning abilities in the course of their services. High-power-density fission reactors will be only realistic tools to simulate the irradiation environments expected in burning-plasma fusion machines such as the ITER, at present. Some key diagnostic components, namely magnetic coils, bolometers, and optical fibers, were irradiation-tested in a fission reactor, JMTR, to evaluate their performances under heavy irradiation environments. Results indicate that the ITER-relevant diagnostic components could be developed in time, though there are still some technological problems to overcome. (author)

  18. Design and Evaluation of a Wireless Sensor Network Based Aircraft Strength Testing System

    Directory of Open Access Journals (Sweden)

    Yang Wang

    2009-06-01

    Full Text Available The verification of aerospace structures, including full-scale fatigue and static test programs, is essential for structure strength design and evaluation. However, the current overall ground strength testing systems employ a large number of wires for communication among sensors and data acquisition facilities. The centralized data processing makes test programs lack efficiency and intelligence. Wireless sensor network (WSN technology might be expected to address the limitations of cable-based aeronautical ground testing systems. This paper presents a wireless sensor network based aircraft strength testing (AST system design and its evaluation on a real aircraft specimen. In this paper, a miniature, high-precision, and shock-proof wireless sensor node is designed for multi-channel strain gauge signal conditioning and monitoring. A cluster-star network topology protocol and application layer interface are designed in detail. To verify the functionality of the designed wireless sensor network for strength testing capability, a multi-point WSN based AST system is developed for static testing of a real aircraft undercarriage. Based on the designed wireless sensor nodes, the wireless sensor network is deployed to gather, process, and transmit strain gauge signals and monitor results under different static test loads. This paper shows the efficiency of the wireless sensor network based AST system, compared to a conventional AST system.

  19. Aircraft testing of the new Blunt-body Aerosol Sampler (BASE

    Directory of Open Access Journals (Sweden)

    A. Moharreri

    2014-03-01

    Full Text Available There is limited understanding of aerosol role in the formation and modification of clouds partly due to inadequate data on such systems. Aircraft-based aerosol measurements in the presence of cloud particles has proven to be challenging because of the problem of cloud-droplet/ice-particle shatter and the generation of secondary artifact particles that contaminate aerosol samples. Recently, design of a new aircraft inlet, called the blunt-body aerosol sampler (BASE, which enables sampling of interstitial aerosol particles, was introduced. Numerical modeling results and laboratory test data suggested that the BASE inlet should sample interstitial particles with minimal shatter particle contamination. Here, the sampling performance of the inlet is established from aircraft-based measurements. Initial aircraft test results obtained during the PLOWS campaign indicated two problems with the original BASE design: separated flows around the BASE at high altitudes; and a significant shatter problem when sampling in drizzle. The test data was used to improve the accuracy of flow and particle trajectory modeling around the inlet, and the results from the improved flow model informed several design modifications of BASE to overcome the problems identified from its initial deployment. The performance of the modified BASE was tested during the ICE-T campaign and the inlet was seen to provide near shatter-free measurements in a wide range of cloud conditions. The initial aircraft test results, design modifications, and the performance characteristics of BASE relative to another interstitial inlet, the sub-micron aerosol inlet (SMAI, are presented.

  20. The Monte Carlo method for shielding calculations analysis by MORSE code of a streaming case in the CAORSO BWR power reactor shielding (Italy)

    International Nuclear Information System (INIS)

    In the field of shielding, the requirement of radiation transport calculations in severe conditions, characterized by irreducible three-dimensional geometries has increased the use of the Monte Carlo method. The latter has proved to be the only rigorous and appropriate calculational method in such conditions. However, further efforts at optimization are still necessary to render the technique practically efficient, despite recent improvements in the Monte Carlo codes, the progress made in the field of computers and the availability of accurate nuclear data. Moreover, the personal experience acquired in the field and the control of sophisticated calculation procedures are of the utmost importance. The aim of the work which has been carried out is the gathering of all the necessary elements and features that would lead to an efficient utilization of the Monte Carlo method used in connection with shielding problems. The study of the general aspects of the method and the exploitation techniques of the MORSE code, which has proved to be one of the most comprehensive of the Monte Carlo codes, lead to a successful analysis of an actual case. In fact, the severe conditions and difficulties met have been overcome using such a stochastic simulation code. Finally, a critical comparison between calculated and high-accuracy experimental results has allowed the final confirmation of the methodology used by us

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

    Science.gov (United States)

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

    2016-02-01

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

  2. Improving the proliferation resistance of research and test reactors

    International Nuclear Information System (INIS)

    Elimination, or substantial reduction, of the trade in highly enriched fuel elements for research and test reactors would significantly reduce the proliferation risk associated with the current potential for diversion of these materials. To this end, it is the long-term goal of U.S. policy to fuel all new and existing research and test reactors with uranium of less than 20% enrichment (but substantially greater than natural) excepting, perhaps, only a small number of high-power, high-performance, reactors. The U.S. development program for enrichment reduction in research and test reactor designs currently using 90-93% enriched uranium is based on the practical criterion that enrichment reduction should not cause significant flux performance (flux per unit power) or burnup performance degradation relative to the unmodified reactor design. A program is now beginning in the U.S. to develop the necessary fuel technology, but several years of work will be needed. Accordingly, as an immediate interim step, the U.S. is proposing to convert existing research and test reactors (and new designs) from the use of 90-93% enriched fuel to the use of 30-45% enriched fuel wherever this can be done without unacceptable reactor performance degradation

  3. Optical Autocovariance Wind Lidar (OAWL): aircraft test-flight history and current plans

    Science.gov (United States)

    Tucker, Sara C.; Weimer, Carl; Adkins, Mike; Delker, Tom; Gleeson, David; Kaptchen, Paul; Good, Bill; Kaplan, Mike; Applegate, Jeff; Taudien, Glenn

    2015-09-01

    To address mission risk and cost limitations the US has faced in putting a much needed Doppler wind lidar into space, Ball Aerospace and Technologies Corp, with support from NASA's Earth Science Technology Office (ESTO), has developed the Optical Autocovariance Wind Lidar (OAWL), designed to measure winds from aerosol backscatter at the 355 nm or 532 nm wavelengths. Preliminary proof of concept hardware efforts started at Ball back in 2004. From 2008 to 2012, under an ESTO-funded Instrument Incubator Program, Ball incorporated the Optical Autocovariance (OA) interferometer receiver into a prototype breadboard lidar system by adding a laser, telescope, and COTS-based data system for operation at the 355 nm wavelength. In 2011, the prototype system underwent ground-based validation testing, and three months later, after hardware and software modifications to ensure autonomous operation and aircraft safety, it was flown on the NASA WB-57 aircraft. The history of the 2011 test flights are reviewed, including efforts to get the system qualified for aircraft flights, modifications made during the flight test period, and the final flight data results. We also present lessons learned and plans for the new, robust, two-wavelength, aircraft system with flight demonstrations planned for Spring 2016.

  4. Irradiation testing of miniature fuel plates for the RERTR program. [Reduced Enrichment Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Senn, R.L.; Martin, M.M.

    1981-07-01

    An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates (miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. These tests screen various candidate fuel materials on their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% /sup 235/U in place of highly enriched fuel for these reactors would reduce the potential for /sup 235/U diversion. The irradiation test facility, designated as HFED, is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The miniplates will achieve burnups of up to approx. 2.2 x 10/sup 27/ fissions/m/sup 3/ of fuel.

  5. Utilization of fission reactors for fusion engineering testing

    Energy Technology Data Exchange (ETDEWEB)

    Deis, G.A.; Miller, L.G.

    1985-02-08

    Fission reactors can be used to conduct some of the fusion nuclear engineering tests identified in the FINESSE study. To further define the advantages and disadvantages of fission testing, the technical and programmatic constraints on this type of testing are discussed here. This paper presents and discusses eight key issues affecting fission utilization. Quantitative comparisons with projected fusion operation are made to determine the technical assets and limitations of fission testing. Capabilities of existing fission reactors are summarized and compared with technical needs. Conclusions are then presented on the areas where fission testing can be most useful.

  6. Recent reactor testing and experience with gamma thermometers

    International Nuclear Information System (INIS)

    Recent experience with gamma thermometers for light water reactors has primarily been in the Framatome reactors operated by Electricite de France. Other recent testing has taken place at Oak Ridge National Laboratory and the Otto Hahn ship reactor. Earlier experience with gamma thermometers was in heavy water reactors at Savannah River and Halden. This paper presents recent data from the light water reactor (LWR) programs. The principles of design and operation of the Radcal gamma thermometer were presented in ''Gamma Thermometer Developments for Light Water Reactors'', Leyse and Smith1. Observations from LWRs confirm the earlier experience from heavy water reactors that the gamma thermometer units give signals which are proportional to the power of surrounding fuel rods and virtually independent of exposure, surrounding poison and other conditions which affect signals of neutron sensitive devices. After 200 sensor-years in EdF reactors, there has been no change in the sensitivity of the devices. Nonetheless, the Radcal units can be recalibrated in-reactor by the introduction of electrical heating via a heater cable imbedded in the device. Algorithms and signal processing software have been developed to interpret and display the gamma thermometer signals. The results of this processing are illustrated here

  7. 77 FR 36341 - Control of Air Pollution From Aircraft and Aircraft Engines; Emission Standards and Test Procedures

    Science.gov (United States)

    2012-06-18

    ...EPA is adopting several new aircraft engine emission standards for oxides of nitrogen (NOX), compliance flexibilities, and other regulatory requirements for aircraft turbofan or turbojet engines with rated thrusts greater than 26.7 kilonewtons (kN). We also are adopting certain other requirements for gas turbine engines that are subject to exhaust emission standards as follows.......

  8. Reactor calculation benchmark PCA blind test results

    International Nuclear Information System (INIS)

    Further improvement in calculational procedures or a combination of calculations and measurements is necessary to attain 10 to 15% (1 sigma) accuracy for neutron exposure parameters (flux greater than 0.1 MeV, flux greater than 1.0 MeV, and dpa). The calculational modeling of power reactors should be benchmarked in an actual LWR plant to provide final uncertainty estimates for end-of-life predictions and limitations for plant operations. 26 references, 14 figures, 6 tables

  9. Safety re-assessment of AECL test and research reactors

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor Maple-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX, and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX AND NRU). ZED-2 and PTR are operated by the Reactor Physics Branch; all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis, and the plan for future analysis. In addition, during the last four years safety assessments have been carried out on the new isotope-, heat-, and electricity-producing reactors, as part of the safety design review, commissioning and licensing activities

  10. Dynamic structural aeroelastic stability testing of the XV-15 tilt rotor research aircraft

    Science.gov (United States)

    Schroers, L. G.

    1982-01-01

    For the past 20 years, a significant effort has been made to understand and predict the structural aeroelastic stability characteristics of the tilt rotor concept. Beginning with the rotor-pylon oscillation of the XV-3 aircraft, the problem was identified and then subjected to a series of theoretical studies, plus model and full-scale wind tunnel tests. From this data base, methods were developed to predict the structural aeroelastic stability characteristics of the XV-15 Tilt Rotor Research Aircraft. The predicted aeroelastic characteristics are examined in light of the major parameters effecting rotor-pylon-wing stability. Flight test techniques used to obtain XV-15 aeroelastic stability are described. Flight test results are summarized and compared to the predicted values. Wind tunnel results are compared to flight test results and correlated with predicted values.

  11. In-Research Reactor Tests for SCWR Fuel Verifications

    International Nuclear Information System (INIS)

    The Supercritical water cooled reactors (SCWRs) are essentially light water reactors (LWRs) operating at higher pressure and temperature. The SCWRs achieve high thermal efficiency (i.e., about 45% vs. about 35% efficiency for advanced LWRs) and are simpler plants as the need for many of the traditional LWR components is eliminated. The SCWRs build upon two proven technologies, the LWR and the supercritical coal-fired boiler. The main mission of the SCWR is production of low-cost electricity. Thus the SCWR is also suited for hydrogen generation with electrolysis, and can support the development of the hydrogen economy in the near term. In this paper, the SCWR fuel performance verification tests are reviewed. Based on this review results, in-research reactor verification tests to be performed in a fuel test loop through the international joint program are proposed. In addition, capsule tests and fuel test loop tests to be performed in HANARO are also proposed

  12. Micro-specimen testing techniques for evaluating nuclear reactor materials

    International Nuclear Information System (INIS)

    In the initial construction of nuclear power plant nuclear materials not only have to be high quality in mechanical properties and fracture resistant characteristics, but also considerations have to be given to weakness cause and continued safe operation of power reactor. Recognizing the importance of integrity evaluation test material samples are provided under monitoring program in reactor for evaluation of reactor material property. But because of limited space and necessity of a homogeneous irradiation environment a very limited quantity of micro specimen is provided. The existing test method of toughness property and fracture resistance requires pre-determined size specimen. Therefore, it is very difficult to evaluate those properties by limited micro-specimen provided under monitoring program. In this paper the test technologies of micro-specimen, which can be utilized to evaluate material integrity of reactors in operation, are reviewed. (Hong, J. S.)

  13. Missile and aircraft field test data acquired with the rapid optical beam steering (ROBS) sensor system

    Science.gov (United States)

    MacDonald, Bruce; Dunn, Murray; Herr, David W.; Hyman, Howard; Leslie, Daniel H.; Lovern, Michael G.

    1997-08-01

    The ROBS instrument has recently acquired unique imagery of a missile intercepting an airborne drone target. We present a summary of that mission. We also present imagery of three airborne targets collected while the ROBS instrument simultaneously tracked all three aircraft. The recent test data highlights the capability of the ROBS instrument for autonomous acquisition, tracking, and imaging of multiple targets under field test conditions. We also describe improvements to the optical system currently underway.

  14. SMORN-III benchmark test on reactor noise analysis methods

    International Nuclear Information System (INIS)

    A computational benchmark test was performed in conjunction with the Third Specialists Meeting on Reactor Noise (SMORN-III) which was held in Tokyo, Japan in October 1981. This report summarizes the results of the test as well as the works made for preparation of the test. (author)

  15. HFR irradiation testing of light water reactor (LWR) fuel

    International Nuclear Information System (INIS)

    For the materials testing reactor HFR some characteristic information with emphasis on LWR fuel rod testing capabilities and hot cell investigation is presented. Additionally a summary of LWR fuel irradiation programmes performed and forthcoming programmes are described. Project management information and a list of publications pertaining to LWR fuel rod test programmes is given

  16. Rotating shielded crane system

    Science.gov (United States)

    Commander, John C.

    1988-01-01

    A rotating, radiation shielded crane system for use in a high radiation test cell, comprises a radiation shielding wall, a cylindrical ceiling made of radiation shielding material and a rotatable crane disposed above the ceiling. The ceiling rests on an annular ledge intergrally attached to the inner surface of the shielding wall. Removable plugs in the ceiling provide access for the crane from the top of the ceiling into the test cell. A seal is provided at the interface between the inner surface of the shielding wall and the ceiling.

  17. Crash Testing and Simulation of a Cessna 172 Aircraft: Hard Landing Onto Concrete

    Science.gov (United States)

    Jackson, Karen E.; Fasanella, Edwin L.

    2016-01-01

    A full-scale crash test of a Cessna 172 aircraft was conducted at the Landing and Impact Research Facility at NASA Langley Research Center during the summer of 2015. The purpose of the test was to evaluate the performance of Emergency Locator Transmitters (ELTs) that were mounted at various locations in the aircraft and to generate impact test data for model validation. A finite element model of the aircraft was developed for execution in LSDYNA to simulate the test. Measured impact conditions were 722.4-in/s forward velocity and 276-in/s vertical velocity with a 1.5deg pitch (nose up) attitude. These conditions were intended to represent a survivable hard landing. The impact surface was concrete. During the test, the nose gear tire impacted the concrete, followed closely by impact of the main gear tires. The main landing gear spread outward, as the nose gear stroked vertically. The only fuselage contact with the impact surface was a slight impact of the rearmost portion of the lower tail. Thus, capturing the behavior of the nose and main landing gear was essential to accurately predict the response. This paper describes the model development and presents test-analysis comparisons in three categories: inertial properties, time sequence of events, and acceleration and velocity time-histories.

  18. Loading tests of a wing structure for a hypersonic aircraft

    Science.gov (United States)

    Fields, R. A.; Reardon, L. F.; Siegel, W. H.

    1980-01-01

    Room-temperature loading tests were conducted on a wing structure designed with a beaded panel concept for a Mach 8 hypersonic research airplane. Strain, stress, and deflection data were compared with the results of three finite-element structural analysis computer programs and with design data. The test program data were used to evaluate the structural concept and the methods of analysis used in the design. A force stiffness technique was utilized in conjunction with load conditions which produced various combinations of panel shear and compression loading to determine the failure envelope of the buckling critical beaded panels The force-stiffness data did not result in any predictions of buckling failure. It was, therefore, concluded that the panels were conservatively designed as a result of design constraints and assumptions of panel eccentricities. The analysis programs calculated strains and stresses competently. Comparisons between calculated and measured structural deflections showed good agreement. The test program offered a positive demonstration of the beaded panel concept subjected to room-temperature load conditions.

  19. NASA Langley Distributed Propulsion VTOL Tilt-Wing Aircraft Testing, Modeling, Simulation, Control, and Flight Test Development

    Science.gov (United States)

    Rothhaar, Paul M.; Murphy, Patrick C.; Bacon, Barton J.; Gregory, Irene M.; Grauer, Jared A.; Busan, Ronald C.; Croom, Mark A.

    2014-01-01

    Control of complex Vertical Take-Off and Landing (VTOL) aircraft traversing from hovering to wing born flight mode and back poses notoriously difficult modeling, simulation, control, and flight-testing challenges. This paper provides an overview of the techniques and advances required to develop the GL-10 tilt-wing, tilt-tail, long endurance, VTOL aircraft control system. The GL-10 prototype's unusual and complex configuration requires application of state-of-the-art techniques and some significant advances in wind tunnel infrastructure automation, efficient Design Of Experiments (DOE) tunnel test techniques, modeling, multi-body equations of motion, multi-body actuator models, simulation, control algorithm design, and flight test avionics, testing, and analysis. The following compendium surveys key disciplines required to develop an effective control system for this challenging vehicle in this on-going effort.

  20. Startup of the FFTF sodium cooled reactor. [Acceptance Test Program

    Energy Technology Data Exchange (ETDEWEB)

    Redekopp, R.D.; Umek, A.M.

    1981-03-01

    The Fast Flux Test Facility (FFTF), located on the Department of Energy (DOE) Hanford Reservation near Richland, Washington, is a 3 Loop 400 MW(t) sodium cooled fast reactor with a primary mission to test fuels and materials for development of the Liquid Metal Fast Breeder Reactor (LMFBR). Bringing FFTF to a condition to accomplish this mission is the goal of the Acceptance Test Program (ATP). This program was the mechanism for achieving startup of the FFTF. Highlights of the ATP involving the system inerting, liquid metal and inerted cell testing and initial ascent to full power are discussed.

  1. High-temperature heat and mass transfer in a concrete layer of the biological shield of nuclear reactors at critical heat loads

    International Nuclear Information System (INIS)

    The problem on high-temperature heat and mass transfer in a concrete layer of the biological shield of nuclear reactors at critical heat loads is considered. The processes of the adsorbed moisture evaporation and dehydration reaction are accounted for. It is shown that concrete dehydration process at heating up to 800-900 K leads to formation in the concrete of the high-pressure (up to 5 x 105 Pa) area of evaporation and dehydration gaseous products. The level of stresses originating thereby essentially exceeds the limits of typical concrete strength at the corresponding high temperatures. It is also established that the concrete dehydration process having high-quality and low porosity may lead to higher pressures as compared with low-quality concretes. The results obtained constitute the basis for a more accurate calculation of temperature fields in the biological shield of nuclear reactors and also for calculating the stress-deformed concrete and forecasting the operational reliability of power reactors as a whole

  2. Gamma ray shielding calculation benchmark exercise comparison of Monte Carlo simulation and experimental methods of gamma radiation field mapping in the pool of Apsara reactor

    International Nuclear Information System (INIS)

    Apsara is a swimming pool type research reactor loaded with Highly Enriched Uranium (HEU) fuel. The reactor is designed for a maximum power level of 1 MW and is normally operated up to 400 KW. The pool water serves as coolant, moderator and reflector besides providing shielding. In addition, graphite and beryllium oxide incased in aluminum boxes are used as in-core reflector. The core is mounted on a square grid of aluminum, 56x56x15 cm with 49 holes on a 7 x 7 square lattice (77 mm pitch), containing fuel elements, control elements, reflectors, irradiation tubes, neutron source and fission counter. This study served in validation of the experimental measurements conducted using GM counter based detector and diode based detectors. In addition, the comparison provided a confirmation of the accuracy of the radiation transport simulation techniques used for dose rate evaluation in case of complex source geometries and large shield materials present. The experimental measurements thus served in bench marking the simulation methods adopted for radiation transport used to arrive at reactor physics and radiological safety parameters of interest

  3. Laboratory tests on neutron shields for gamma-ray detectors in space

    CERN Document Server

    Hong, J; Hailey, C J

    2000-01-01

    Shields capable of suppressing neutron-induced background in new classes of gamma-ray detectors such as CdZnTe are becoming important for a variety of reasons. These include a high cross section for neutron interactions in new classes of detector materials as well as the inefficient vetoing of neutron-induced background in conventional active shields. We have previously demonstrated through Monte-Carlo simulations how our new approach, supershields, is superior to the monolithic, bi-atomic neutron shields which have been developed in the past. We report here on the first prototype models for supershields based on boron and hydrogen. We verify the performance of these supershields through laboratory experiments. These experimental results, as well as measurements of conventional monolithic neutron shields, are shown to be consistent with Monte-Carlo simulations. We discuss the implications of this experiment for designs of supershields in general and their application to future hard X-ray/gamma-ray experiments...

  4. Modeling, Testing and Deploying a Multifunctional Radiation Shielding / Hydrogen Storage Unit Project

    Data.gov (United States)

    National Aeronautics and Space Administration — This project addresses two vital problems for long-term space travel activities: radiation shielding and hydrogen storage for power and propulsion. While both...

  5. Dual-core TRIGA research and materials testing reactor

    International Nuclear Information System (INIS)

    General Atomic Company is under contract from the Romanian Institute for Nuclear Technologies to design, fabricate, and install a research reactor in support of the Romanian National Program for Power Reactor Development. The goal was to select a design concept that provided reasonably high neutron fluxes for long term testing of various fuel-cladding-coolant combinations and also provide high performance pulsing for transient testing of fuel specimens. An effective solution was achieved by the selection of a 14 MW steady-state TRIGA reactor for high flux endurance testing, and an Annular Core Pulsing Reactor (ACPR) for high performance pulsing testing, with both reactors mounted in the same reactor tank and operated independently. The fuel bundles for the steady-state reactor consist of 25 uranium-zirconium hydride rods clad in stainless steel arranged in a square 5 x 5 array. The steady-state core is provided with downflow cooling at a rate of approximately 275 gpm/bundle. Bundle flow tests will be performed with both heated and unheated models. The core will be optimized for peak thermal neutron flux and reactivity lifetime within the constraint of a peak fuel meat temperature of 7500C. The operation of the steady-state reactor at a power level of 14 MW will yield peak unperturbed thermal neutron fluxes in the central experiment position of 2.9 x 1014 n/cm2-sec. The corresponding fast neutron flux (less than 1.125 keV) will be 2.6 x 1014 nv. (U.S.)

  6. Results of the BREST-300 type reactor model fuel elements testing in the IGR reactor

    International Nuclear Information System (INIS)

    Testings of BREST-300 type fast reactor's model fuel elements with nitride fuel in the lead coolant in the central experimental channel of IGR reactor were carried out. In the testing the regime of non-controlled power burst was simulated. In the result of testing the seal failure of fuel elements with 2 % and 10 % 235U enrichment has been occurred, and fragmentation of the part of fuel pellets at interaction with coolant has been taken place. During the reactor testing the measurements and registration of experimental parameters (temperature of fuel, shell, coolant; pressure in fuel elements and testing ampoule; power release in the reactor) were conducted. The physical study of the 'fuel element - ampoule - reactor' was carried out, after-start-up spectrometric and material testing studies, calculated evaluation of temperature fields parameters in the testing ampoule were examined as well. Calculated and experimental values of breaking down specific power releases in the fuel are obtained. The assessment of both fuel fragmentation rate and it character is carried out. Distribution of fuel fragmentation within experimental ampoule volume is studied

  7. Acoustic emission for on-line reactor monitoring: results of intermediate vessel test monitoring and reactor hot functional testing

    International Nuclear Information System (INIS)

    This article discusses a program designed to develop the use of acoustic emission (AE) methods for continuous surveillance to detect and evaluate flaw growth in reactor pressure boundaries. Technology developed in the laboratory for identifying AE from crack growth and for using AE information to estimate flaw severity is now being evaluated on an intermediate vessel test and on a reactor facility. A vessel, designated ZB-1, has been tested under fatigue loadings with simulated reactor conditions at Mannheim, West Germany, in collaboration with the German Materialpruefungsanstalt (MPA), Stuttgart. Fatigue cracking from machined flaws and in a fabrication weld were both detected clearly by AE. AE data were measured on a US nuclear reactor (Watts Bar, Unit 1) during hot functional preservice testing. This demonstrated that coolant flow noise is a manageable problem and that AE can be detected under operational coolant flow and temperature conditions. (author)

  8. Technology Options for a Fast Spectrum Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    D. M. Wachs; R. W. King; I. Y. Glagolenko; Y. Shatilla

    2006-06-01

    Idaho National Laboratory in collaboration with Argonne National Laboratory has evaluated technology options for a new fast spectrum reactor to meet the fast-spectrum irradiation requirements for the USDOE Generation IV (Gen IV) and Advanced Fuel Cycle Initiative (AFCI) programs. The US currently has no capability for irradiation testing of large volumes of fuels or materials in a fast-spectrum reactor required to support the development of Gen IV fast reactor systems or to demonstrate actinide burning, a key element of the AFCI program. The technologies evaluated and the process used to select options for a fast irradiation test reactor (FITR) for further evaluation to support these programmatic objectives are outlined in this paper.

  9. The ''CAMERA'' test facility in the OSIRIS reactor

    International Nuclear Information System (INIS)

    CAMERA is an irradiation installation conceived to measure under neutronic flux and continuously the dimension variations of a fuel pencil of PWR reactors. The device, set in the periphery of the OSIRIS reactor, can receive new, preirradiated or reconstituted pencils. The principles of measurements is explained. Then, a brief description of the installation is given: in-pile part; out-of-pile part; connections. The technical characteristics of the installation are presented. A first qualification test of the installation under flux has been carried out at the end of the first semester 1984 in the OSIRIS reactor

  10. Architecture of the ETR [experimental test reactor] systems code

    International Nuclear Information System (INIS)

    TETRA, a tokamak systems code capable of modeling experimental test reactors (ETRs), was developed in a joint effort by participants of the fusion community. The first version of this code was constructed to model devices similar to the Tokamak Ignition/Burn Engineering Reactor (TIBER) in configuration and design. A major feature of this code is its ability to perform optimization studies. Future work will include broadening the scope of the code, particularly in the area of materials selection, to more accurately simulate tokamak configurations such as the Next European Torus (NET) and the Fusion Engineering Reactor (FER). 18 refs., 2 figs., 4 tabs

  11. Closing down and dismantling of research - material testing - and teaching reactors. Stillegung und Beseitigung von Forschungs-, Materialpruef- und Unterrichtsreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Petrasch, P.; Seidler, M.; Stasch, W.

    1983-02-01

    This study is subdivided in six topics: - determination of mass and radioactivity of the parts to be dismantled, - identification of future tasks of research and development, - estimation of radiation exposure for workers charged with dismantling, - determination of cost for closing-down and dismantling of research reactors. In total, 22 research-, materials testing- and traning reactors are taken into consideration here. Only those component that belong directly to the reactor proper plus the auxiliary - and service plants are dismantled. The reactor buildings will only be dismantled if the are a direct reactor component serving for example as a biological shield. The waste quantity created by closing-down and dismantling of all research reactors comes up to about 25200 Mg out of that 720 Mg are radioactive wastes. Planning and carrying out of closing-down and dismantling of all research reactors need about 4870 man-months the total cost will be about 86,4 Mio DM. There are vast differences between the individual research reactors. On 10 Mg will have to be disposed in the case of the Siemens training-reactor 100 (SUR-100) of which a very small share consists of radioactive waste; in the case of the research reactor Neuherberg (FRN) there are about 3500 Mg, about 94 Mg out of it is radioactive waste. The work needed per reactor varies between 26 man-months (SUR-100) and about 740 man-months (FRN). Costs for dosing-down and dismantling range between 0,4 Mio DM (SUR-100) and about 13 Mio DM (FRN).

  12. The reactor core analysis code CITATION-1000VP for High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Reactor core analysis with full core model has been necessary for the High Temperature Engineering Test Reactor (HTTR) design. The CITATION-1000VP code has been developed to enable reactor core analysis of HTTR with full core model through extending the number of zones and meshes, and enhancing the calculation speed of CITATION code. This report describes the program changes for extending the number of zones and meshes, and for vectorization. The maximum number of zones and meshes becomes 999 and 500, respectively. The calculation speed is enhanced up to 21 times. (author)

  13. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  14. Integral test of JENDL-3.3 for fast reactors

    International Nuclear Information System (INIS)

    An integral test of JENDL-3.3 was performed for fast reactors. Various types of fast reactors were analyzed. Calculation values of the nuclear characteristics were greatly especially affected by the revisions of the cross sections of U-235 capture and elastic scattering reactions. The C/E values were improved for ZPPR cross where plutonium is mainly fueled, but not for BFS cores where uranium is mainly fueled. (author)

  15. Models for transient analyses in advanced test reactors

    OpenAIRE

    Gabrielli, Fabrizio

    2011-01-01

    Several strategies are developed worldwide to respond to the world’s increasing demand for electricity. Modern nuclear facilities are under construction or in the planning phase. In parallel, advanced nuclear reactor concepts are being developed to achieve sustainability, minimize waste, and ensure uranium resources. To optimize the performance of components (fuels and structures) of these systems, significant efforts are under way to design new Material Test Reactors facilities in Europe whi...

  16. Conversion of hydrocarbon fuel in thermal protection reactors of hypersonic aircraft

    Science.gov (United States)

    Kuranov, A. L.; Mikhaylov, A. M.; Korabelnikov, A. V.

    2016-07-01

    Thermal protection of heat-stressed surfaces of a high-speed vehicle flying in dense layers of atmosphere is one of the topical issues. Not of a less importance is also the problem of hydrocarbon fuel combustion in a supersonic air flow. In the concept under development, it is supposed that in the most high-stressed parts of airframe and engine, catalytic thermochemical reactors will be installed, wherein highly endothermic processes of steam conversion of hydrocarbon fuel take place. Simultaneously with heat absorption, hydrogen generation will occur in the reactors. This paper presents the results of a study of conversion of hydrocarbon fuel in a slit reactor.

  17. Design, fabrication and testing of a liquid hydrogen fuel tank for a long duration aircraft

    Science.gov (United States)

    Mills, Gary L.; Buchholtz, Brian; Olsen, Al

    2012-06-01

    Liquid hydrogen has distinct advantages as an aircraft fuel. These include a specific heat of combustion 2.8 times greater than gasoline or jet fuel and zero carbon emissions. It can be utilized by fuel cells, turbine engines and internal combustion engines. The high heat of combustion is particularly important in the design of long endurance aircraft with liquid hydrogen enabling cruise endurance of several days. However, the mass advantage of the liquid hydrogen fuel will result in a mass advantage for the fuel system only if the liquid hydrogen tank and insulation mass is a small fraction of the hydrogen mass. The challenge is producing a tank that meets the mass requirement while insulating the cryogenic liquid hydrogen well enough to prevent excessive heat leak and boil off. In this paper, we report on the design, fabrication and testing of a liquid hydrogen fuel tank for a prototype high altitude long endurance (HALE) demonstration aircraft. Design options on tank geometry, tank wall material and insulation systems are discussed. The final design is an aluminum sphere insulated with spray on foam insulation (SOFI). Several steps and organizations were involved in the tank fabrication and test. The tank was cold shocked, helium leak checked and proof pressure tested. The overall thermal performance was verified with a boil off test using liquid hydrogen.

  18. Testing of rubber O-rings for R-5 reactor

    International Nuclear Information System (INIS)

    This paper summarises the results of various experiments and tests conducted for the selection of suitable O-rings to be used in R-5 reactor. O-rings of various elastomeric compositions obtained from different manufacturers were tested for heat aging, fluid aging, radiation stability and specific gravity. They were irradiated to various dose levels at Apsara Reactor. The changes in axial thickness and hardness, after each test, were measured and results were correlated. The tests reveal that O-rings made of ethylene-propylene rubber are the best suited for the use in R-5 reactor. The O-rings made of nitrile rubber are also good. Neoprene rubber O-rings were found unsuitable mainly because of their low radiation resistance. (author)

  19. Monitoring test piece for a reactor pressure vessel

    International Nuclear Information System (INIS)

    Purpose: To obtain a test piece capable of measurement for neutron exposure ranging 0.1 -- 2 MeV in a reactor pressure vessel Constitution: Fissionable materials causing nuclear fission by fast neutrons are contained within a sealed container in addition to a test piece for monitoring the change in the mechanical properties and a monitor wire for measuring the neutron dose. If uranium 238 and thorium 232 are selected as the fissionable materials, for instance, they cause nuclear fission by the reaction with neutrons of higher than about 2 MeV and about 0.2 MeV respectively. Then, after the stop of the reactor operation, the monitoring test piece is taken out from the reactor pressure vessel to determine the radioactivity, whereby the neutron dose within the energy range of 0.1 - 2 MeV applied to the fissionable materials of the test piece can be estimated with ease. (Horiuchi, T.)

  20. Flight Test Evaluation of Mission Computer Algorithms for a Modern Trainer Aircraft

    Directory of Open Access Journals (Sweden)

    Gargi Meharu

    2013-03-01

    Full Text Available A low cost integrated avionics system has been realized on a modern trainer aircraft. Without using an expensive inertial navigation system onboard, acceptable level of accuracy for navigation, guidance, and weapon aiming is achieved by extensive data fusion within mission computer. The flight test evaluation of mission computer is carried out by assessing the overall performance under various navigation and guidance modes. In flight simulation is carried out for weapon aiming modes. The mission computer interfaces with various subsystems and implements the functional requirements for flight management and mission management. The aim of this paper is to discuss the algorithms of a data fusion intensive mission computer and flight test evaluation of these algorithms, for a typical modern trainer aircraft. The challenges and innovations involved in the work are also discussed.Defence Science Journal, 2013, 63(2, pp.164-173, DOI:http://dx.doi.org/10.14429/dsj.63.4259

  1. Reactor fault simulation at the closure of the Windscale advanced gas-cooled reactor: analysis of reactor transient tests

    International Nuclear Information System (INIS)

    The testing of fault transient analysis methods by direct simulation of fault sequences on a commercial reactor is clearly excluded on safety and economic grounds. The closure of the Windscale prototype advanced gas-cooled reactor (WAGR) therefore offered a unique opportunity to test fault study methods under extreme conditions relatively unfettered by economic constraints, although subject to appropriate safety regulations. One aspect of these important experiments was a series of reactor transient tests. The objective of these reactor transients was to increase confidence in the fault study computer models used for commercial AGR safety assessment by extending their range of validation to cover large amplitude and fast transients in temperature, power and flow, relevant to CAGR faults, and well beyond the conditions achievable experimentally on commercial reactors. A large number of tests have now been simulated with the fault study code KINAGRAX. Agreement with measurement is very good and sensitivity studies show that such discrepancies as exist may be due largely to input data errors. It is concluded that KINAGRAX is able to predict steady state conditions and transient amplitudes in both power and temperature to within a few percent. (author)

  2. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

  3. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris oe National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately

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

    OpenAIRE

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

    1987-01-01

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

  5. Impact Analyses and Tests of Metal Cask Considering Aircraft Engine Crash - 12308

    International Nuclear Information System (INIS)

    The structural integrity of a dual purpose metal cask currently under development by the Korea Radioactive Waste Management Cooperation (KRMC) is evaluated through analyses and tests under a high-speed missile impact considering the targeted aircraft crash conditions. The impact conditions were carefully chosen through a survey on accident cases and recommendations from the literature. The missile impact velocity was set at 150 m/s, and two impact orientations were considered. A simplified missile simulating a commercial aircraft engine is designed from an impact load history curve provided in the literature. In the analyses, the focus is on the evaluation of the containment boundary integrity of the metal cask. The analyses results are compared with the results of tests using a 1/3 scale model. The results show very good agreements, and the procedure and methodology adopted in the structural analyses are validated. While the integrity of the cask is maintained in one evaluation where the missile impacts the top side of the free standing cask, the containment boundary is breached in another case in which the missile impacts the center of the cask lid in a perpendicular orientation. A safety assessment using a numerical simulation of an aircraft engine crash into spent nuclear fuel storage systems is performed. A commercially available explicit finite element code is utilized for the dynamic simulation, and the strain rate effect is included in the modeling of the materials used in the target system and missile. The simulation results show very good agreement with the test results. It is noted that this is the first test considering an aircraft crash in Korea. (authors)

  6. Aircraft Design Automation and Subscale Testing : With Special Reference to Micro Air Vehicles

    OpenAIRE

    Lundström, David

    2012-01-01

    This dissertation concerns how design automation as well as rapid prototyping and testing of subscale prototypes can support aircraft design. A framework for design automation has been developed and is applied specifically to Micro Air Vehicles (MAV). MAVs are an interesting area for design automation as they are an application where the entire design, from requirements to manufacturing, can indeed be automated. From a complexity point of view it can be considered to be similar to conceptual ...

  7. SPECIFICATIONS FOR HIGH TEMPERATURE LATTICE TEST REACTOR BUILDING 318 PROJECT CAH-100

    Energy Technology Data Exchange (ETDEWEB)

    Vitro Engineering Company

    1964-07-15

    This is the specifications for the High Temperature Lattice Test Reactor Building 318 and it is divided into the following 21 divisions or chapters: (1) Excavating, Backfill & Grading; (2) Reinforced Concrete; (3) Masonry; (4) Structural Steel & Miscellaneous Metal Items, Contents - Division V; (5) Plumbing, Process & Service Piping; (6) Welding; (7) Insulated Metal Siding; (8) Roof Decks & Roofing; (9) Plaster Partitions & Ceiling; (10) Standard Doors, Windows & Hardware; (11) Shielding Doors; (12) Sprinkler System & Fire Extinguishers, Contents - Division XIII; (13) Heating, Ventilating & Air Conditioning; (14) Painting, Protective Coating & Floor Covering, Contents - Division XV; (15) Electrical; (16) Communications & Alarm Systems; (17) Special Equipment & Furnishings; (18) Overhead Bridge Crane; (19) Prefabricated Steel Building; (20) Paved Drive; and (21) Landscaping & Irrigation Sprinklers.

  8. DT and DHe3 tokamak test reactor concepts using advanced, high field superconductors

    International Nuclear Information System (INIS)

    If practical high temperature superconducting ceramic magnets can be developed, there could be a significant impact on reactor design. Potential advantages include a simpler, more robust magnet design, the possibility of demountable superconducting toroidal field coils and reduced shielding requirements. The high temperature superconductors can also have very high critical fields and could provide super high field operation. This could substantially increase eta tau/sub E/ values, reduce β requirements, and improve prospects for ohmic heating to ignition. The combination of moderately high β and super high field could make DHe3 operation possible in a JET size tokamak. In this paper we discuss possibilities for test reactor designs using high temperature high field superconductors. An illustrative design has a field at the plasma of 15 T. This reduces the required β to less than 2% for DT operation. The required plasma current is 5 MA. For a reactor size of R0 = 3.4m and a = 0.6m, the neutron wall loading is 3.3 MW/m2 at β = 1.5% for DT operation and an equal amount of fusion power is produced at β = 10% for DHe3 operation. One possible mode of operation is to use ohmic heating to ignition in a DT plasma followed by thermal runaway to DHe3 temperatures. 7 refs., 1 fig., 2 tabs

  9. A personal sampler for aircraft engine cold start particles: laboratory development and testing.

    Science.gov (United States)

    Armendariz, Alfredo; Leith, David

    2003-01-01

    Industrial hygienists in the U.S. Air Force are concerned about exposure of their personnel to jet fuel. One potential source of exposure for flightline ground crews is the plume emitted during the start of aircraft engines in extremely cold weather. The purpose of this study was to investigate a personal sampler, a small tube-and-wire electrostatic precipitator (ESP), for assessing exposure to aircraft engine cold start particles. Tests were performed in the laboratory to characterize the sampler's collection efficiency and to determine the magnitude of adsorption and evaporation artifacts. A low-temperature chamber was developed for the artifact experiments so tests could be performed at temperatures similar to actual field conditions. The ESP collected particles from 0.5 to 20 micro m diameter with greater than 98% efficiency at particle concentrations up to 100 mg/m(3). Adsorption artifacts were less than 5 micro g/m(3) when sampling a high concentration vapor stream. Evaporation artifacts were significantly lower for the ESP than for PVC membrane filters across a range of sampling times and incoming vapor concentrations. These tests indicate that the ESP provides more accurate exposure assessment results than traditional filter-based particle samplers when sampling cold start particles produced by an aircraft engine. PMID:14674798

  10. Inductive testing of reactor pressure vessels

    International Nuclear Information System (INIS)

    In Service Inspection of Reactor Pressure Vessels is mostly done with ultrasonics. Using special 2 crystal-probes good detectability is achieved for near surface defects. The problem is to detect closely spaced cracks, to decide if the defects are surface braking and, if not, to decide the remaining ligament. The purpose of this study is to investigate to what extent Eddy Current can solve these problems. Detecting surfacebreaking cracks and fields of cracks can be done using conventional Eddy Current techniques. Mapping of closely spaced cracks requires a small probe and a high frequency. Measurement of depths a larger probe, a lower frequency and knowledge of the crackfield since 2 closely spaced shallow cracks might be mistaken for one deep crack. Depths of singel cracks can be measured down to 7-8 mm. In closely spaced crackfields the depths can not be measured. The measurement is mostly based on amplitude. For not surface breaking defects the problem is to decide the ligament, i.e. the distance between surface and cracktip. To achieve good penetration a large probe, low frequency and high energy or pulsed energy is used. Ligament up to 4 mm can be measured with good accuracy. The measurements is mostly based on phase. Noise, which originates from rough surface, varied material structure and lift off, can be reduced using multi frequency mix, probe design and scanning pattern. (author)

  11. Preliminary design studies on the Broad Application Test Reactor

    International Nuclear Information System (INIS)

    This report describes progress made at the Idaho National Engineering Laboratory during the first three quarters of Fiscal Year (FY) 1992 on the Laboratory-Directed Research and Development (LDRD) project to perform preliminary design studies on the Broad Application Test Reactor (BATR). This work builds on the FY-92 BATR studies, which identified anticipated mission and safety requirements for BATR and assessed a variety of reactor concepts for their potential capability to meet those requirements. The main accomplishment of the FY-92 BATR program is the development of baseline reactor configurations for the two conventional conceptual test reactors recommended in the FY-91 report. Much of the present report consists of descriptions and neutronics and thermohydraulics analyses of these baseline configurations. In addition, we considered reactor safety issues, compared the consequences of steam explosions for alternative conventional fuel types, explored a Molten Chloride Fast Reactor concept as an alternate BATR design, and examined strategies for the reduction of operating costs. Work planned for the last quarter of FY-92 is discussed, and recommendations for future work are also presented

  12. Validation test for carbon-14 migration and accumulation in a Canadian shield lake

    International Nuclear Information System (INIS)

    This particular BIOMOVS II Technical Report is concerned with modelling the transfer of C-14 through the aquatic food chain following release to a Canadian shield lake. Model performance has been tested against field data supplied by Atomic Energy of Canada Limited. Carbon-14 was added in 1978 to the epilimnion of a small Canadian Shield lake to investigate primary production and carbon dynamics. Data from this experiment were used within BIOMOVS II to provide a validation test, which involved modelling the fate of the C-14 added to the lake. The nature of the spike and the subsequent monitoring allowed the investigation of both short-term processes relevant to evaluation of the impacts of accidental releases as well as longer-term processes relevant to routine release and to solid waste disposal. Four models participated in the scenario: 1) a simple mass balance model of a lake (AECL, Whiteshell Laboratories, Canada); 2) a relatively complex deterministic dynamic compartment model (QuantiSci Ltd.,UK); 3) a complex deterministic model (Studsvik Model A) and a more complex probabilistic model (Studsvik Model B; Studsvik Eco and Safety AB, Sweden). Endpoints were C-14 concentrations in water, sediment and whitefish over a thirteen year period. Each model produced reasonable predictions when compared to the observed data and when uncertainty is taken into consideration. About 0.2 to 0.4% of the initial C-14 inventory to the lakes remained in the water at the end of the study, because of internal recycling of C-14 from sediments. The simple AECL model did not account for this internal recycling of C-14 and, in this respect, its predictions were not as realistic as those of the QuantiSci and Studsvik models for concentrations in water. However, the AECL model predictions for the C-14 inventory remaining in lake sediment were closest to the observed values. Overall, Studsvik Model B was the most accurate in simulating C-14 concentrations in water and in whitefish, but

  13. Lightning effects on aircraft

    Science.gov (United States)

    1977-01-01

    Direct and indirect effects of lightning on aircraft were examined in relation to aircraft design. Specific trends in design leading to more frequent lightning strikes were individually investigated. These trends included the increasing use of miniaturized, solid state components in aircraft electronics and electric power systems. A second trend studied was the increasing use of reinforced plastics and other nonconducting materials in place of aluminum skins, a practice that reduces the electromagnetic shielding furnished by a conductive skin.

  14. Vibration tests on some models of PEC reactor core elements

    International Nuclear Information System (INIS)

    This paper describes the aims of the experimental tests carried out at ISMES, within an agreement with the Department of Fast Reactors of ENEA, on some models of the elements of PEC Fast Nuclear Reactor Core in the frame of the activities for the seismic verification of the PEC core. The seismic verification is briefly described with particular attention to the problems arising from the shocks among the various elements during an earthquake, as well as the computer code used, the purpose and the techniques used to perform tests, some results and the first comparison between the theory and the experimental data

  15. EMERIS: an advanced information system for a materials testing reactor

    International Nuclear Information System (INIS)

    The basic features of the Materials Testing Reactor of IAE, Moscow (MR) Information System (EMERIS) are outlined. The purpose of the system is to support reactor and experimental test loop operators by a flexible, fully computerized and user-friendly tool for the aquisition, analysis, archivation and presentation of data obtained during operation of the experimental facility. High availability of EMERIS services is ensured by redundant hardware and software components, and by automatic configuration procedure. A novel software feature of the system is the automatic Disturbance Analysis package, which is aimed to discover primary causes of irregularities occurred in the technology. (author) 2 refs.; 2 figs

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

    International Nuclear Information System (INIS)

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

  17. Determination and Fabrication of New Shield Super Alloys Materials for Nuclear Reactor Safety by Experiments and Cern-Fluka Monte Carlo Simulation Code, Geant4 and WinXCom

    Science.gov (United States)

    Aygun, Bünyamin; Korkut, Turgay; Karabulut, Abdulhalik

    2016-05-01

    Despite the possibility of depletion of fossil fuels increasing energy needs the use of radiation tends to increase. Recently the security-focused debate about planned nuclear power plants still continues. The objective of this thesis is to prevent the radiation spread from nuclear reactors into the environment. In order to do this, we produced higher performanced of new shielding materials which are high radiation holders in reactors operation. Some additives used in new shielding materials; some of iron (Fe), rhenium (Re), nickel (Ni), chromium (Cr), boron (B), copper (Cu), tungsten (W), tantalum (Ta), boron carbide (B4C). The results of this experiments indicated that these materials are good shields against gamma and neutrons. The powder metallurgy technique was used to produce new shielding materials. CERN - FLUKA Geant4 Monte Carlo simulation code and WinXCom were used for determination of the percentages of high temperature resistant and high-level fast neutron and gamma shielding materials participated components. Super alloys was produced and then the experimental fast neutron dose equivalent measurements and gamma radiation absorpsion of the new shielding materials were carried out. The produced products to be used safely reactors not only in nuclear medicine, in the treatment room, for the storage of nuclear waste, nuclear research laboratories, against cosmic radiation in space vehicles and has the qualities.

  18. An Integrated Architecture for Aircraft Engine Performance Monitoring and Fault Diagnostics: Engine Test Results

    Science.gov (United States)

    Rinehart, Aidan W.; Simon, Donald L.

    2015-01-01

    This paper presents a model-based architecture for performance trend monitoring and gas path fault diagnostics designed for analyzing streaming transient aircraft engine measurement data. The technique analyzes residuals between sensed engine outputs and model predicted outputs for fault detection and isolation purposes. Diagnostic results from the application of the approach to test data acquired from an aircraft turbofan engine are presented. The approach is found to avoid false alarms when presented nominal fault-free data. Additionally, the approach is found to successfully detect and isolate gas path seeded-faults under steady-state operating scenarios although some fault misclassifications are noted during engine transients. Recommendations for follow-on maturation and evaluation of the technique are also presented.

  19. Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2005-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

  20. Full-scale testing, production and cost analysis data for the advanced composite stabilizer for Boeing 737 aircraft, volume 2

    Science.gov (United States)

    Aniversario, R. B.; Harvey, S. T.; Mccarty, J. E.; Parson, J. T.; Peterson, D. C.; Pritchett, L. D.; Wilson, D. R.; Wogulis, E. R.

    1982-01-01

    The development, testing, production activities, and associated costs that were required to produce five-and-one-half advanced-composite stabilizer shipsets for Boeing 737 aircraft are defined and discussed.

  1. Production and benchmark tests of fast reactor group constant set JFS-3-J2

    International Nuclear Information System (INIS)

    The JFS-3-J2 set is an advanced set for JENDL-2B-70, which was produced as the group constants of the JAERI-Fast set type. The JFS-3-J2 has been produced with the use of a processing code system TIMS-PGG. The characteristics of JFS-3-J2 are as follows. The group averaged cross sections are calculated by weighting with the collision density spectrum for core composition in a typical fast reactor, to correct the overestimate of elastic removal cross sections caused by using ''1/E-spectrum''. The weighting method of the collision density spectrum is called ''REMO-correction'', and produces harder neutron spectra than those calculated from ''1/E-spectrum'' method, especially for low energy range below about 1 keV. Hence, the nuclear characteristics such as Doppler reactivity coefficients sensitive to the flux shape in the low energy region are considerably affected by the harder spectra. The temperature dependent self-shielding factors for structural materials, Fe, Cr and Ni are calculated. The Doppler effect of structural materials is 0.24% δk/k for temperature change from 0 to 3000K in the ZPR-6 assembly 7. The scheme of self-shielding factor tables is corrected to obtain high accuracy of interpolation for f-tables by using cubic apline function. The self-shielding effect of inelastic scattering cross sections is considered. The neutron group to group transfer materices of elastic scattering are expanded from P0 to P3. In this report, the effects of ''REMO-correction'' considered in the generation of JFS-3-J2 on the nuclear characteristics are studied. The benchmark tests of JFS-3-J2 are performed and the results are discussed by being compared with those calculated with the JENDL-2B-70 set. (author)

  2. The RES Reactor. A test reactor for the French naval propulsion

    International Nuclear Information System (INIS)

    In the Cadarache nuclear research centre the French Atomic Energy Commission (CEA) operates, with the support of TECHNICATOME as nuclear operator, the experimental facilities which are necessary for the French naval propulsion program. Since the sixties these facilities have brought a large contribution to the development and to the technical support for the nuclear propulsion; they have been used also to train the French Navy operators. The last experimental reactor, the RNG, is now at the end of its life cycle after thirty years of a profitable operation. A replacement reactor is needed to sustain any evolution of the naval propulsion reactors as well as to guarantee a safe operation and a high level of availability of the existing onboard reactors. The aim of the RES program is namely to build such a test facility. Its construction program started in 2003. By the year 2009 the RES reactor will take over the mission of the RNG. We present hereafter: - A brief history of the French experimental reactors built in support to the naval propulsion, - The needs of the naval propulsion and the related objectives of the RES program, - The corresponding architecture and main characteristics of the RES facility, - The current status of the RES construction. The contents of the paper is as follows: 1. Introduction; 2. History of the French nuclear propulsion experimental reactors; 3. Needs of the naval propulsion and related objectives of the RES reactor; 4. RES architecture and main characteristics; 4.1. The pool module; 4.2. The reactor module; 4.3. The RES reactor, an innovative concept; 5. Realisation status; 6. Conclusion. To summarize, from the year 2009 the RES will be an efficient facility available for irradiation and qualification programs. Its large experimental capabilities will allow relevant fuel and core irradiations. This will give access to a real progress in the knowledge of fuel and core physics as well as in the related simulation tools. This reactor

  3. Fast Breeder Test Reactor: 15 years of operating experience

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled, loop type, mixed carbide-fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 1985 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 1993 and power was raised to 10.5 MWt in Dec 1993. Turbine generator was synchronized to the grid in Jul 1997. The indigenously developed mixed carbide fuel has achieved a peak burn up of 88,000 MWd/t till now at a linear heat rating of 320 W/cm and reactor power of 13.4 MWt without any fuel-clad failure. The paper presents operating and decontamination experience, performance of fuel, steam generator and sodium circuits, certain unusual occurrences encountered by the plant and various improvements carried out in reactor systems to enhance plant availability. (author)

  4. REX 2000 core : a new material testing reactor project

    International Nuclear Information System (INIS)

    REX 2000 is a new research reactor project entirely dedicated to technological irradiations, which should be located on the CEA site of CADARACHE. It will be aimed at satisfying the future needs for the validation of new concepts of nuclear materials and fuels, and will take over and replace the present experimental reactors, which are 30 to 40 years old. The fundamental studies started by the CEA in 1993, on future irradiation needs expected in 2005, lead to the design of a reactor which will essentially meet the needs of PWRs, without forgetting the other fields such as FBRs, fusion... The current reactor project is based on a light water open pool concept, with a thermal power of 100 MW, in about 150 l, and characterized by an in-core-central hole. It reaches neutronic flux levels twice those of present French reactor fluxes. It allows many irradiations in the central loop under high fast neutron flux, in order to accelerate the aging of materials and analyze their behaviour. It also enables the achievement of power transient tests under high thermal neutron flux gradients. These performances are obtained with high forced flow rates and upward flow in the core, in order to preserve the operating flexibility of the reactor. This leads to the design of a specific assembly design. (author)

  5. Improving the proliferation resistance of research and test reactors

    International Nuclear Information System (INIS)

    Elimination, or substantial reduction, of the trade in unirradiated highly-enriched fuel elements for research and test reactors would significantly reduce the proliferation risk associated with the current potential for diversion of these materials. To this end, it is the long-term goal of U.S. policy to fuel all new and existing research and test reactors with uranium of less-than-20% enrichment (but substantially greater than natural) excepting, perhaps, only a small number of high-power, high-performance, reactors. The U.S. development program for enrichment reduction in research and test reactor designs currently using 90-93% enriched uranium is based on the practical criterion that enrichment reduction should not cause significant flux performance (flux per unit power) or burnup performance degradation relative to the unmodified reactor design. To first order, this implies the requirement that the 235U loading in the reduced-enrichment fuel elements be the same as the 235U loading in the 90-93% enriched fuel elements. This can be accomplished by substitution of higher uranium density fuel technology for currently-used fuel technology in the fuel meat volume of the current fuel element design and/or by increasing the usable fuel meat volume. For research and test reactors of power greater than 5-10 megawatts, fuel technology does not currently exist that would permit enrichment reductions to below 20% utilizing this criterion. A program is now beginning in the U.S. to develop the necessary fuel technology. Currently-proven fuel technology is capable, however, of accommodating enrichment reductions to the 30-45% range (from 90-93%) for many reactors in the 5-50MW range. Accordingly the U.S. is proposing to convert existing reactors (and new designs) in the 5-50MW range from the use of highly-enriched fuel to the use of 30-45% enriched fuel, and reactors of less that about 5MW to less-than-20% enrichment, wherever this can be done without significant performance

  6. 76 FR 78096 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-12-16

    ... are specifically designed to ensure that the reactor can be shutdown and decay heat can be removed..., 2009 (74 FR 62829). On June 12, 2009 (74 FR 28112), the NRC amended its regulations to require... proposed rule in the Federal Register on January 20, 2011 (76 FR 3540). The public comment period for...

  7. The Philosophy which underlies the structural tests of a supersonic transport aircraft with particular attention to the thermal cycle

    Science.gov (United States)

    Ripley, E. L.

    1972-01-01

    The information presented is based on data obtained from the Concorde. Much of this data also applies to other supersonic transport aircraft. The design and development of the Concorde is a joint effort of the British and French, and the structural test program is shared, as are all the other activities. Vast numbers of small specimens have been tested to determine the behavior of the materials used in the aircraft. Major components of the aircraft structure, totalling almost a complete aircraft, have been made and are being tested to help the constructors in each country in the design and development of the structure. Tests on two complete airframes will give information for the certification of the aircraft. A static test was conducted in France and a fatigue test in the United Kingdom. Fail-safe tests are being made to demonstrate the crack-propagation characteristics of the structure and its residual strength. Aspects of the structural test program are described in some detail, dealing particularly with the problems associated with the thermal cycle. The biggest of these problems is the setting up of the fatigue test on the complete airframe; therefore, this is covered more extensively with a discussion about how the test time can be shortened and with a description of the practical aspects of the test.

  8. Hypervelocity Impact Testing of International Space Station Meteoroid/Orbital Debris Shielding Using an Inhibited Shaped Charge Launcher

    Science.gov (United States)

    Kerr, Justin H.; Grosch, Donald

    2001-01-01

    Engineers at the NASA Johnson Space Center have conducted hypervelocity impact (HVI) performance evaluations of spacecraft meteoroid and orbital debris (M/OD) shields at velocities in excess of 7 km/s. The inhibited shaped charge launcher (ISCL), developed by the Southwest Research Institute, launches hollow, circular, cylindrical jet tips to approximately 11 km/s. Since traditional M/OD shield ballistic limit performance is defined as the diameter of sphere required to just perforate or spall a spacecraft pressure wall, engineers must decide how to compare ISCL derived data with those of the spherical impactor data set. Knowing the mass of the ISCL impactor, an equivalent sphere diameter may be calculated. This approach is conservative since ISCL jet tips are more damaging than equal mass spheres. A total of 12 tests were recently conducted at the Southwest Research Institute (SWRI) on International Space Station M/OD shields. Results of these tests are presented and compared to existing ballistic limit equations. Modification of these equations is suggested based on the results.

  9. The technology development for surveillance test of reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Sun Phil; Park, Day Young; Choi, Kwen Jai

    1997-12-01

    Benchmark test was performed in accordance with the requirement of US NRC Reg. Guide DG-1053 for Kori unit-1 in order to determine best-estimated fast neutron fluence irradiated into reactor vessel. Since the uncertainty of radiation analysis comes from the calculation error due to neutron cross-section data, reactor core geometrical dimension, core source, mesh density, angular expansion and convergence criteria, evaluation of calculational uncertainty due to analytical method was performed in accordance with the regulatory guide and the proof was performed for entire analysis by comparing the measurement value obtained by neutron dosimetry located in surveillance capsule. Best-estimated neutron fluence in reactor vessel was calculated by bias factor, neutron flux measurement value/calculational value, from reanalysis result from previous 1st through 4th surveillance testing and finally fluence prediction was performed for the end of reactor life and the entire period of plant life extension. Pressurized thermal shock analysis was performed in accordance with 10 CFR 50.61 using the result of neutron fluence analysis in order to predict the life of reactor vessel material and the criteria of safe operation for Kori unit 1 was reestablished. (author). 55 refs., 55 figs.

  10. Reactor protection system with automatic self-testing and diagnostic

    International Nuclear Information System (INIS)

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically ''identical'' values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic. 16 figs

  11. The technology development for surveillance test of reactor vessel materials

    International Nuclear Information System (INIS)

    Benchmark test was performed in accordance with the requirement of US NRC Reg. Guide DG-1053 for Kori unit-1 in order to determine best-estimated fast neutron fluence irradiated into reactor vessel. Since the uncertainty of radiation analysis comes from the calculation error due to neutron cross-section data, reactor core geometrical dimension, core source, mesh density, angular expansion and convergence criteria, evaluation of calculational uncertainty due to analytical method was performed in accordance with the regulatory guide and the proof was performed for entire analysis by comparing the measurement value obtained by neutron dosimetry located in surveillance capsule. Best-estimated neutron fluence in reactor vessel was calculated by bias factor, neutron flux measurement value/calculational value, from reanalysis result from previous 1st through 4th surveillance testing and finally fluence prediction was performed for the end of reactor life and the entire period of plant life extension. Pressurized thermal shock analysis was performed in accordance with 10 CFR 50.61 using the result of neutron fluence analysis in order to predict the life of reactor vessel material and the criteria of safe operation for Kori unit 1 was reestablished. (author). 55 refs., 55 figs

  12. Evaluation of earth pressure on triple multi-face shield tunnel based on centrifuge model tests; Enshin mokei jikken ni yoru yoko sanren shield tunnel no doatsu hyoka

    Energy Technology Data Exchange (ETDEWEB)

    Sugihara, Y.; Igarashi, H.; Fujisaki, K. [Kajima Corp., Tokyo (Japan)

    1995-12-20

    Recently, various needs for the utilization of underground space in urban areas, i.e., those for utilizing underground spaces with complicated cross-section shapes, have been encouraging the research and development of new shield methods. There have been considerations of building many tunnels with various cross-section shapes, such as horizontal double multi-face, horizontal triple multi-face, vertical double multi-face, MMST (multi-micro shield tunnel), and large-section tunnels with combinations of those shapes. The horizontal triple multi-face shield method is a new technique applied to underground railway stations with respect to construction period and safety. Since the horizontal triple multi-face shield tunnel has a laterally flatter shape than the former one circle shield tunnel, it is necessary to verify the validity of design using a usual method of evaluating earth pressure mainly for circle cross-sections. Therefore, the force on segments and deformation of surrounding soil were measured by using a centrifuge model equipment. Consequently, it was found that the earth pressure on horizontal triple multi-face shield segments in sandy ground is between earth pressure as evaluated by Terzaghi`s method and full earth pressure. 5 refs., 14 figs., 3 tabs.

  13. In situ tests on the PEC fast reactor building

    International Nuclear Information System (INIS)

    This paper describes forced excitation tests carried out at the PEC reactor building, to determine seismic motion amplifications produced in the building itself. Experimental results are used to gauge numerical methodologies capable of assessing the margins existing in the design analysis. (orig./HP)

  14. Natural convection test in Phenix reactor and associated CATHARE calculation

    International Nuclear Information System (INIS)

    The Phenix sodium cooled fast reactor started operation in 1973 and was stopped in 2009. Before the reactor was definitively stopped, final tests were performed, including a natural convection test in the primary circuit. One objective of this natural convection test in Phenix reactor is the qualification of plant dynamic codes as CATHARE code for future safety studies. The paper firstly describes the Phenix reactor primary circuit. The initial test conditions and the detailed transient scenario are presented. Then, the CATHARE modelling of the Phenix primary circuit is described. The whole transient scenario is calculated, including the nominal state, the steam generators dry out, the scram, the onset of natural convection in the primary circuit and the natural convection phases. The CATHARE calculations are compared to the Phenix measurements. A particular attention is paid to the significant decrease of the core power before the scram. Then, the evolution of main components inlet and outlet temperatures is compared. The need of coupling a system code with a CFD code to model the 3D behaviour of large pools is pointed out. This work is in progress. (author)

  15. Testing and Analysis of a Composite Non-Cylindrical Aircraft Fuselage Structure . Part II; Severe Damage

    Science.gov (United States)

    Przekop, Adam; Jegley, Dawn C.; Lovejoy, Andrew E.; Rouse, Marshall; Wu, Hsi-Yung T.

    2016-01-01

    The Environmentally Responsible Aviation Project aimed to develop aircraft technologies enabling significant fuel burn and community noise reductions. Small incremental changes to the conventional metallic alloy-based 'tube and wing' configuration were not sufficient to achieve the desired metrics. One airframe concept identified by the project as having the potential to dramatically improve aircraft performance was a composite-based hybrid wing body configuration. Such a concept, however, presented inherent challenges stemming from, among other factors, the necessity to transfer wing loads through the entire center fuselage section which accommodates a pressurized cabin confined by flat or nearly flat panels. This paper discusses a finite element analysis and the testing of a large-scale hybrid wing body center section structure developed and constructed to demonstrate that the Pultruded Rod Stitched Efficient Unitized Structure concept can meet these challenging demands of the next generation airframes. Part II of the paper considers the final test to failure of the test article in the presence of an intentionally inflicted severe discrete source damage under the wing up-bending loading condition. Finite element analysis results are compared with measurements acquired during the test and demonstrate that the hybrid wing body test article was able to redistribute and support the required design loads in a severely damaged condition.

  16. Research and Test Reactor Fuel Elements (RTRFE)

    International Nuclear Information System (INIS)

    BWX Technologies Inc. (BWXT) has experienced several production improvements over the past year. The homogeneity yields in 4.8 gU/cc U3Si2 plates have increased over last year's already high yields. Through teamwork and innovative manufacturing techniques, maintaining high quality surface finishes on plates and elements is becoming easier and less expensive. Currently, BWXT is designing a fabrication development plan to reach a fuel loading of 9 gU/cc within 2 - 4 years. This development will involve a step approach requested by ANL to produce plates using U-8Mo at a loading of 6 gU/cc first and qualify the fuel at those levels. In achieving the goal of a very high-density fuel loading of 9 gU/cc, BWXT is considering employing several new, state of the art, ultrasonic testing techniques for fuel core evaluation. (author)

  17. Sipping test of fuel assemblies in LVR-15 reactor

    International Nuclear Information System (INIS)

    The LVR-15 reactor is a light water research type which is situated at NRI in Rez near Prague. The poster describes the procedure and methodology used for sipping test of the fuel assemblies. These tests are designed to evaluate the leakage of fuel and fission products from the tested fuel assembly. From 1995 to 2003 there have been performed about 200 tests. Examples of results of sipping water activity measurements are presented. The values of activities of 137Cs and 134Cs are used for decision if the fuel assembly can be used in reactor core, transported to storage pool or if it is necessary to put the fuel assembly into the special protective can. The used limits of activities are discussed. (author)

  18. Radiation protection/shield design

    International Nuclear Information System (INIS)

    Radiation protection/shielding design of a nuclear facility requires a coordinated effort of many engineering disciplines to meet the requirements imposed by regulations. In the following discussion, the system approach to Clinch River Breeder Reactor Plant (CRBRP) radiation protection will be described, and the program developed to implement this approach will be defined. In addition, the principal shielding design problems of LMFBR nuclear reactor systems will be discussed in realtion to LWR nuclear reactor system shielding designs. The methodology used to analyze these problems in the U.S. LMFBR program, the resultant design solutions, and the experimental verification of these designs and/or methods will be discussed. (orig.)

  19. Testing of research reactor fuel in the high flux reactor (Petten)

    International Nuclear Information System (INIS)

    The two types of fuel most frequently used by the main research reactors are metallic: highly enriched uranium (>90%) and silicide low enriched uranium (3. However, a need exists for research on new reactor fuel. This would permit some plants to convert without losses in flux or in cycle length and would allow new reactor projects to achieve higher possibilities especially in fluxes. In these cases research is made either on silicide with higher density, or on other types of fuel (UMo, etc.). In all cases when new fuel is proposed, there is a need, for safety reasons, to test it, especially regarding the mechanical evolution due to burn-up (swelling, etc.). Initially, such tests are often made with separate plates, but lately, using entire elements. Destructive examinations are often necessary. For this type of test, the High Flux Reactor, located in Petten (The Netherlands) has many specific advantages: a large core, providing a variety of interesting positions with high fluence rate; a downward coolant flow simplifies the engineering of the device; there exists easy access with all handling possibilities to the hot-cells; the high number of operating days (>280 days/year), together with the high flux, gives a possibility to reach quickly the high burn-up needs; an experienced engineering department capable of translating specific requirements to tailor-made experimental devices; a well equipped hot-cell laboratory on site to perform all necessary measurements (swelling, γ-scanning, profilometry) and all destructive examinations. In conclusion, the HFR reactor readily permits experimental research on specific fuels used for research reactors with all the necessary facilities on the Petten site. (author)

  20. Analysis of a loss of forced cooling test using the High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    The High Temperature Engineering Test Reactor (HTTR) is the first High Temperature Gas-cooled Reactor (HTGR) built at the Oarai Research and Development Center of JAEA, with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950degC (Saito, 1994). Test researches are being conducted using the HTTR to improve HTGR technologies and to collaborate with domestic industries to contribute to foreign projects for acceleration of HTGR development worldwide. To improve HTGR technologies, advanced analysis techniques are being developed using data obtained with the HTTR, which include reactor kinetics, thermal-hydraulics, safety evaluation, and fuel performance evaluation data (including the behavior of fission products). A three-gas-circulators trip test and a vessel-cooling-system stop test were planned as a loss-of-forced-cooling test and demonstrate the inherent safety features of HTGR. The vessel-cooling-system stop test consists of stopping the vessel-cooling-system located outside the reactor pressure vessel (RPV), to remove the residual heat of the reactor core as soon as the three-gas-circulators are tripped. All three-gas-circulators is tripped at 9 MW. The primary coolant flow rate is reduced from the rated 45 t/h to 0 t/h. The control rods are not inserted into the core and the reactor power control system does not operated. A core dynamics analysis of the loss-of-forced-cooling test of the HTTR is performed. Analytical results for the reactor transient during the test are presented in this report. It is determined that the reactor power immediately decreases to the decay heat level due to the negative reactivity feedback effect of the core, even though the reactor shutdown system is not operational, and that the temperature distribution in the core changes slowly because of the high heat capacity due to the large amount of core graphite. Furthermore, the relation between the reactivities (namely, the Doppler, moderator temperature, and