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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. Evaluation of the Shielding Characteristics Test around the Reactor Core in the Prototype Fbr Monju

    Science.gov (United States)

    Usami, Shin; Suzuoki, Zenro; Deshimaru, Takehide; Nakashima, Fumiaki; Hikichi, Takuo

    2003-06-01

    In Monju, shielding measurements were made around the reactor core as a part of the system start-up tests in order to evaluate the design margins of the shielding performance, to demonstrate the validity of the shielding analysis method, and to acquire basic data for use in future FBR design. The measured reaction rates have been obtained radially from the core to the in-vessel storage rack and axially to the reactor vessel upper plenum. The measured values (E) were compared with the calculated values (C) obtained with the FBR shielding analysis system on the basis of the nuclear data library JENDL-3.2. Based upon these results, the design margins around the reactor core have been examined.

  3. Neutronic Reactor Shield

    Science.gov (United States)

    Fermi, Enrico; Zinn, Walter H.

    The argument of the present Patent is a radiation shield suitable for protection of personnel from both gamma rays and neutrons. Such a shield from dangerous radiations is achieved to the best by the combined action of a neutron slowing material (a moderator) and a neutron absorbing material. Hydrogen is particularly effective for this shield since it is a good absorber of slow neutrons and a good moderator of fast neutrons. The neutrons slowed down by hydrogen may, then, be absorbed by other materials such as boron, cadmium, gadolinium, samarium or steel. Steel is particularly convenient for the purpose, given its effectiveness in absorbing also the gamma rays from the reactor (both primary gamma rays and secondary ones produced by the moderation of neutrons). In particular, in the present Patent a shield is described, made of alternate layers of steel and Masonite (an hydrolized ligno-cellulose material). The object of the present Patent is not discussed in any other published paper.

  4. Open Rotor Noise Shielding by Blended-Wing-Body Aircraft

    Science.gov (United States)

    Guo, Yueping; Czech, Michael J.; Thomas, Russell H.

    2015-01-01

    This paper presents an analysis of open rotor noise shielding by Blended Wing Body (BWB) aircraft by using model scale test data acquired in the Boeing Low Speed Aeroacoustic Facility (LSAF) with a legacy F7/A7 rotor model and a simplified BWB platform. The objective of the analysis is the understanding of the shielding features of the BWB and the method of application of the shielding data for noise studies of BWB aircraft with open rotor propulsion. By studying the directivity patterns of individual tones, it is shown that though the tonal energy distribution and the spectral content of the wind tunnel test model, and thus its total noise, may differ from those of more advanced rotor designs, the individual tones follow directivity patterns that characterize far field radiations of modern open rotors, ensuring the validity of the use of this shielding data. Thus, open rotor tonal noise shielding should be categorized into front rotor tones, aft rotor tones and interaction tones, not only because of the different directivities of the three groups of tones, but also due to the differences in their source locations and coherence features, which make the respective shielding characteristics of the three groups of tones distinctly different from each other. To reveal the parametric trends of the BWB shielding effects, results are presented with variations in frequency, far field emission angle, rotor operational condition, engine installation geometry, and local airframe features. These results prepare the way for the development of parametric models for the shielding effects in prediction tools.

  5. Results of shielding characteristics tests in Monju

    Energy Technology Data Exchange (ETDEWEB)

    Usami, Shin; Suzuoki, Zenro; Deshimaru, Takehide; Nakashima, Fumiaki [Japan Nuclear Cycle Development Inst., Tsuruga, Fukui (Japan)

    2001-06-01

    In the prototype fast breeder reactor Monju, the shielding characteristics tests were made around the reactor core, the primary heat transport system, and the fuel handling and storage system as a part of the system start-up tests from 0% to 45% of rated power from October 1993 through December 1995. The results of the measurements, analyses and evaluations in these tests validated the FBR shielding analysis methods and demonstrated that there was a safe shielding design margin in Monju. The important basic data for use in future FBR shielding design were successfully acquired. In order to obtain more substantial basic data and to improve the accuracy of the analyses, the next shielding measurements are planned for the period of the system start-up tests at the restart of Monju. (author)

  6. Status of reactor shielding research in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Bartine, D.E.

    1983-01-01

    Shielding research in the United States continues to place emphasis on: (1) the development and refinement of shielding design calculational methods and nuclear data; and (2) the performance of confirmation experiments, both to evaluate specific design concepts and to verify specific calculational techniques and input data. The successful prediction of the radiation levels observed within the now-operating Fast Flux Test Facility (FFTF) has demonstrated the validity of this two-pronged approach, which has since been applied to US fast breeder reactor programs and is now being used to determine radiation levels and possible further shielding needs at operating light water reactors, especially under accident conditions. A similar approach is being applied to the back end of the fission fuel cycle to verify that radiation doses at fuel element storage and transportation facilities and within fuel reprocessing plants are kept at acceptable levels without undue economic penalties.

  7. Fusion reactor blanket/shield design study

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L.; Clemmer, R.G.; Harkness, S.D.

    1979-07-01

    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.

  8. Gravity Scaling of a Power Reactor Water Shield

    Science.gov (United States)

    Reid, Robert S.; Pearson, J. Boise

    2008-01-01

    Water based reactor shielding is being considered as an affordable option for use on initial lunar surface power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxiliary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2007). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa(sup n). These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

  9. Monte Carlo based radial shield design of typical PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2016-11-15

    Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

  10. 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.

  11. 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.

  12. Shielding efficiency of metal hydrides and borohydrides in fusion reactors

    DEFF Research Database (Denmark)

    Singh, Vishvanath P.; Badiger, Nagappa M.; Gerward, Leif

    2016-01-01

    Mass attenuation coefficients, mean free paths and exposure buildup factors have been used to characterize the shielding efficiency of metal hydrides and borohydrides, with high density of hydrogen. Gamma ray exposure buildup factors were computed using five-parameter geometric progression fittin...... combination of low-and high-Z elements. The present work should be useful for the selection and design of blankets and shielding, and for dose evaluation for components in fusion reactors....

  13. Design of shielding-performance testing-device for reactors construction shielding-material based on 252Cf source%基于252Cf 源的反应堆结构屏蔽材料屏蔽性能测试装置设计

    Institute of Scientific and Technical Information of China (English)

    贺丹; 宋英明; 邹树梁; 徐守龙; 王晓冬; 朱志超; 谭桢干; 何志锋; 王新林

    2015-01-01

    Background: New material for the reactor shielding structure is one type of cast-steel impurity doping actinide elements have been adopted. The design parameters of the shielding performance testing device for the various parts of reactor can be obtained by Monte Carlo method. Purpose: This study aims to test the shielding performance of the shielding material of reactor. Methods: First of all, the testing device model was built based on a 252Cf neutron source. Three kinds of models (the thicknesses of the moderation layer, the protective layers of the neutron and gamma ray) are calculated respectively using the MCNP program for the shielding performance testing device. The GEANT4 program was used for the key moderation layer to verify the results obtained by the MCNP program. Results: Through the analysis of simulation results, we obtained optimal shielding materials and the thicknesses of material layer: moderator layer material is paraffin wax, which having a thickness of 8 cm, neutron shielding material is polyethylene, which having a thickness of 38 cm, and γ protective layer material is iron, which having a thickness of 11 cm. Conclusion: Simulation results show that the proposed shielding performance testing device can meet the requirements of neutron moderator, neutron protection and photon protection.%基于252Cf 中子源,构建了反应堆结构屏蔽材料屏蔽性能测试装置设计模型。采用 MCNP 程序建立了测试模型,并逐次模拟计算屏蔽性能测试装置慢化层、中子防护层、γ光子防护层厚度。对于关键的慢化层,采用 Geant4程序进一步验证 MCNP 程序的计算结果。通过分析模拟计算获得了最优屏蔽材料及厚度分别为:慢化层材料为石蜡,厚度为8 cm;中子防护层材料为聚乙烯,厚度为38 cm;γ防护层材料为铁,厚度为11 cm。模拟实验结果表明,所设计屏蔽性能测试装置能够满足中子慢化以及中子、光子防护的需要。

  14. Preliminary Evaluation of Convective Heat Transfer in a Water Shield for a Surface Power Reactor

    Science.gov (United States)

    Pearson J. Boise; Reid, Robert S.

    2007-01-01

    As part of the Vision for Space Exploration, the end of the next decade will bring man back to the surface of the moon. A crucial issue for the establishment of human presence on the moon will be the availability of compact power sources. This presence could require greater than 10's of kWt's in follow on years. Nuclear reactors are well suited to meet the needs for power generation on the lunar or Martian surface. Radiation shielding is a key component of any surface power reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), and boron carbide. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to fix the location of any vapor that could form radiation streaming paths. The water shield concept relies on the predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. NASA Marshall Space Flight Center has developed the experience and facilities necessary to do this evaluation in its Early Flight Fission - Test Facility (EFF-TF).

  15. Shielding efficiency of metal hydrides and borohydrides in fusion reactors

    Directory of Open Access Journals (Sweden)

    Singh Vishvanath P.

    2016-01-01

    Full Text Available Mass attenuation coefficients, mean free paths and exposure buildup factors have been used to characterize the shielding efficiency of metal hydrides and borohydrides, with high density of hydrogen. Gamma ray exposure buildup factors were computed using five-parameter geometric progression fitting at energies 0.015 MeV to15 MeV, and for penetration depths up to 40 mean free paths. Fast-neutron shielding efficiency has been characterized by the effective neutron removal cross-section. It is shown that ZrH2 and VH2 are very good shielding materials for gamma rays and fast neutrons due to their suitable combination of low- and high-Z elements. The present work should be useful for the selection and design of blankets and shielding, and for dose evaluation for components in fusion reactors.

  16. Experimental Evaluation of a Water Shield for a Surface Power Reactor

    Science.gov (United States)

    Pearson, J. B.; Reid, R.; Sadasivan, P.; Stewart, E.

    2007-01-01

    A water based shielding system is being investigated for use on initial lunar surface power systems. The use of water 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. A representative lunar surface reactor design is evaluated at various power levels in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The evaluation compares the experimental data from the WST to CFD models. Performance of a water shield on the lunar surface is predicted by CFD models anchored to test data, and by matching relevant dimensionless parameters.

  17. 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

  18. 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.

  19. 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.

  20. Uranium self-shielding in fast reactor blankets

    Energy Technology Data Exchange (ETDEWEB)

    Kadiroglu, O.K.; Driscoll, M.J.

    1976-03-01

    The effects of heterogeneity on resonance self-shielding are examined with particular emphasis on the blanket region of the fast breeder reactor and on its dominant reaction--capture in /sup 238/U. The results, however, apply equally well to scattering resonances, to other isotopes (fertile, fissile and structural species) and to other environments, so long as the underlying assumptions of narrow resonance theory apply. The heterogeneous resonance integral is first cast into a modified homogeneous form involving the ratio of coolant-to-fuel fluxes. A generalized correlation (useful in its own right in many other applications) is developed for this ratio, using both integral transport and collision probability theory to infer the form of correlation, and then relying upon Monte Carlo calculations to establish absolute values of the correlation coefficients. It is shown that a simple linear prescription can be developed for the flux ratio as a function of only fuel optical thickness and the fraction of the slowing-down source generated by the coolant. This in turn permitted derivation of a new equivalence theorem relating the heterogeneous self-shielding factor to the homogeneous self-shielding factor at a modified value of the background scattering cross section per absorber nucleus. A simple version of this relation is developed and used to show that heterogeneity has a negligible effect on the calculated blanket breeding ratio in fast reactors.

  1. Superconducting shielded core reactor with reduced AC losses

    Energy Technology Data Exchange (ETDEWEB)

    Cha, Yung S.; Hull, John R.

    2006-04-04

    A superconducting shielded core reactor (SSCR) operates as a passive device for limiting excessive AC current in a circuit operating at a high power level under a fault condition such as shorting. The SSCR includes a ferromagnetic core which may be either closed or open (with an air gap) and extends into and through a superconducting tube or superconducting rings arranged in a stacked array. First and second series connected copper coils each disposed about a portion of the iron core are connected to the circuit to be protected and are respectively wound inside and outside of the superconducting tube or rings. A large impedance is inserted into the circuit by the core when the shielding capability of the superconducting arrangement is exceeded by the applied magnetic field generated by the two coils under a fault condition to limit the AC current in the circuit. The proposed SSCR also affords reduced AC loss compared to conventional SSCRs under continuous normal operation.

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

    Science.gov (United States)

    Pearson, J. Boise; Stewart, Eric T.; Reid, Robert S.

    2007-01-01

    A water based shielding system is being investigated for use on initial lunar surface power systems. The use of water 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 representative lunar surface reactor shield design is evaluated at various power levels in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to anchor a CFD model. Performance of a water shield on the lunar surface is then predicted by CFD models anchored to test data. The accompanying viewgraph presentation includes the following topics: 1) Testbed Configuration; 2) Core Heater Placement and Instrumentation; 3) Thermocouple Placement; 4) Core Thermocouple Placement; 5) Outer Tank Thermocouple Placement; 6) Integrated Testbed; 7) Methodology; 8) Experimental Results: Core Temperatures; 9) Experimental Results; Outer Tank Temperatures; 10) CFD Modeling; 11) CFD Model: Anchored to Experimental Results (1-g); 12) CFD MOdel: Prediction for 1/6-g; and 13) CFD Model: Comparison of 1-g to 1/6-g.

  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. 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.

  5. 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.

  6. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ennis, M. G.; Tobin, T. J.; Cha, Y. S.; Hull, J. R.

    2000-06-30

    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. 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.

  9. Reactor Simulator Testing

    Science.gov (United States)

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

    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. Keywords: fission, space power, nuclear, liquid metal, NaK.

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

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    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 {sup 23}Na(n,g){sup 24}Na 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 B{sub 4}C shielding inside the subassembly.

  11. 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.

  12. 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 aircraft....

  13. Shielding efficiency of metal hydrides and borohydrides in fusion reactors

    OpenAIRE

    Singh Vishvanath P.; Badiger Nagappa M.; Gerward Leif

    2016-01-01

    Mass attenuation coefficients, mean free paths and exposure buildup factors have been used to characterize the shielding efficiency of metal hydrides and borohydrides, with high density of hydrogen. Gamma ray exposure buildup factors were computed using five-parameter geometric progression fitting at energies 0.015 MeV to15 MeV, and for penetration depths up to 40 mean free paths. Fast-neutron shielding efficiency has been characterized by the effective neu...

  14. Shielding designs for pressurized water reactors in France

    Energy Technology Data Exchange (ETDEWEB)

    Champion, G.; Forestier, J.; Vergnaud, T.

    1986-07-01

    The efforts made by Electricite de France to reduce exposure from the two-component neutron-gamma radiation fields inside the pressurized water reactor (PWR) building are described. Most of the attention had been focused on the problem of neutron exposure relative to the problem of achieving a highly efficient confinement within the reactor cavity and the state of the art of personnel neutron dosimetry. A description of the general neutron calculation scheme that links the characteristics of the neutron fields escaping from the reactor vessel to the dose equivalent rate cartographies inside the reactor building is provided.

  15. Neutronics optimization study for D-D fusion reactor blanket/shield

    Energy Technology Data Exchange (ETDEWEB)

    Shiba, T.; Kanda, Y.; Nakashima, H.

    1985-12-01

    Position-dependent optimization calculations have been carried out on a D-D fusion reactor blanket/shield to maximize the energy gain in the blanket and to minimize the atomic displacement rate of the copper stabilizer in the superconducting magnet. The results obtained by using the optimization code SWAN indicate the advantage of D/sub 2/O coolant over H/sub 2/O coolant with respect to increasing the energy gain, and the difference in the optimal shield distributions between D-T and D-D neutron sources. The possibility of improving both the energy gain and radiation shielding characteristics is also discussed.

  16. Thermal fatigue tests of a radiative heat shield panel for a hypersonic transport

    Science.gov (United States)

    Webb, Granville L.; Clark, Ronald K.; Sharpe, Ellsworth L.

    1985-01-01

    A pair of corrugation stiffened, beaded skin Rene 41 heat shield panels were exposed to 20,000 thermal cycles between room temperature and 1450 F to evaluate the thermal fatigue response of Rene 41 metallic heat shields for hypersonic cruise aircraft applications. At the conclusion of the tests, the panels retained substantial structural integrity; however, there were cracks and excessive wear in the vicinity of fastener holes and there was an 80-percent loss in ductility of the skin. Shrinkage of the panel which caused the cracks and wear must be considered in design of panels for Thermal Protection Systems (TPS) applications.

  17. Photon spectrum behind biological shielding of the LVR-15 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Klupak, V.; Viererbl, L.; Lahodova, Z.; Marek, M.; Vins, M. [Research Centre Rez Ltd., Husinec-Rez 130 (Czech Republic)

    2011-07-01

    The LVR-15 reactor is a light water research reactor situated at the Research Centre Rez, near Prague. It operates as a multipurpose facility with a maximum thermal power of 10 MW. The reactor core usually contains from 28 to 32 fuel assemblies with a total mass of {sup 235}U of about 5 kg. Emitted radiation from the fuel caused by fission is shielded by moderating water, a steel reactor vessel, and heavy concrete. This paper deals with measurement and analysis of the gamma spectrum near the outer surface of the concrete wall, behind biological shielding, mainly in the 3- to 10-MeV energy range. A portable HPGe detector with a portable multichannel analyzer was used to measure gamma spectra. The origin of energy lines in gamma detector spectra was identified. (authors)

  18. Preliminary Analysis of a Water Shield for a Surface Power Reactor

    Science.gov (United States)

    Pearson, J. Boise

    2006-01-01

    A water based shielding system is being investigated for use on initial lunar surface power systems. The use of water 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. A simple 1-D thermal model indicates the necessity of natural convection to maintain acceptable temperatures and pressures in the water shield. CFD analysis is done to quantify the natural convection in the shield, and predicts sufficient natural convection to transfer heat through the shield with small temperature gradients. A test program will he designed to experimentally verify the thermal hydraulic performance of the shield, and to anchor the CFD models to experimental results.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Junyong; Chang, Yoonsuk [Department of Nuclear Engineering/Kyung Hee Univ., Yongin (Korea, Republic of)

    2013-05-15

    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.

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

    Science.gov (United States)

    Ferrari, A; Pelliccioni, M; Villari, R

    2005-01-01

    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 andambient dose-equivalent rates have been determined inside the aircraft at several locations along the fuselage at a typicaI 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Zagar, Tomaz [Reactor Physics Department, Jozef Stefan Institute, Jamova 39, 1000 Ljubljana (Slovenia)]. E-mail: tomaz.zagar@ijs.si; Bozic, Matjaz [Nuklearna elektrarna Krsko, Vrbina 12, 8270 Krsko (Slovenia); Ravnik, Matjaz [Reactor Physics Department, Jozef Stefan Institute, Jamova 39, 1000 Ljubljana (Slovenia)

    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 ({gamma} emitting) radioactive nuclides in the concrete were found to be {sup 133}Ba, {sup 60}Co and {sup 152}Eu. 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 Jozef Stefan Institute TRIGA reactor. It was observed that the clearance levels of some important long-lived isotopes typical for barite concrete (e.g. {sup 133}Ba, {sup 41}Ca) are not included in the IAEA and EU basic safety standards.

  2. 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.

  3. FASTER Test Reactor Preconceptual Design Report

    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. J. [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, S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-31

    The FASTER test 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.

  4. Application of the kernel method on ET-RR-1 reactor shield

    Science.gov (United States)

    Hathout, A. M.

    1994-07-01

    The kernel method is used to calculate the γ-Dose-Rate (GDR) on the shield surface of the ET-RR-1 reactor. The GDR is obtained in terms of the build-up factor as a function of energy. The build-up factor is calculated for water, cast iron and heavy concrete, as shielding materials, in the energy range 0.5-10.0 MeV. An optimization code was programmed for the main frame VAX to calculate the GDR averaged over the energy range. The results obtained are presented in the tables and discussed.

  5. Application of the kernel method on ET-RR-1 reactor shield

    Energy Technology Data Exchange (ETDEWEB)

    Hathout, A.M. [Atomic Energy Authority, Cairo (Egypt). National Centre of Nuclear Safety and Radiation Control

    1994-07-01

    The kernel method is used to calculate the {gamma}-Dose-Rate (GDR) on the shield surface of the ET-RR-1 reactor. The GDR is obtained in terms of the build-up factor as a function of energy. The build-up factor is calculated for water, cast iron and heavy concrete, as shielding materials, in the energy range 0.5 - 10.0 MeV. An optimization code was programmed for the main frame VAX to calculate the GDR averaged over the energy range. The results obtained are presented in the tables and discussed. (author).

  6. Aircraft Test & Evaluation Facility (Hush House)

    Data.gov (United States)

    Federal Laboratory Consortium — The Aircraft Test and Evaluation Facility (ATEF), or Hush House, is a noise-abated ground test sub-facility. The facility's controlled environment provides 24-hour...

  7. 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.

  8. Radiological characterization of the concrete biological shield of the APSARA reactor

    Directory of Open Access Journals (Sweden)

    Srinivasan Priya

    2013-01-01

    Full Text Available The first Indian research reactor, APSARA, was utilized for various R&D programmes from 1956 until its shutdown in 2009. The biological shield of the reactor developed residual activity due to neutron irradiation during the operation of the reactor. Dose rate mapping and in-situ gamma spectrometry of the concrete structures of the reactor pool were carried out. Representative concrete samples collected from various locations were subjected to high-resolution gamma spectrometry analysis. 60Co and 152Eu were found to be the dominant gamma-emitting radionuclides in most of the locations. 133Ba was also found in some of the concrete structures. The separation of 3H from concrete was achieved using an acid digestion method and beta activity measured using liquid scintillation counting. The depth profile of radionuclide specific activity in the concrete wall of the shielding corner was also studied. Specific activities of the radionuclides were found to decrease exponentially with depth inside the concrete walls. This study would be helpful in bulk waste management during the decommissioning of the reactor.

  9. Radiation distribution through ilmenite-limonite concrete and its application as a reactor biological shield

    Energy Technology Data Exchange (ETDEWEB)

    Makarious, A.S.; El-Kolaly, M.A.; Kansouh, W.A.; Bashter, I.I.

    1989-01-01

    A study of the penetration of primary ..gamma.. rays, secondary ..gamma.. rays and slow neutrons through an ilmenite-limonite concrete shield (heat resistant concrete) and through both ordinary and ilmenite concrete shields has been carried out. A shielding assembly with dimensions of 120 x 120 x 120cm/sup 3/ for each concrete type has been used. Direct, cadmium filtered and B/sub 4/C-filtered reactor beams emitted from one of the horizontal channels of the ET-RR-1 reactor were used. The ..gamma..-ray doses were measured using LiF-7 Teflon disc TLD dosimeters and the slow neutron doses were measured using LiF-6 Teflon disc TLD dosimeters. Ratios of the total ..gamma.. doses, secondary ..gamma.. doses and slow neutron doses for an ilmenite-limonite concrete shield, and for both ordinary and ilmenite concrete, have been obtained. The results show that ilmenite concrete is better than both ordinary and ilmenite-limonite concrete for ..gamma.. ray attenuation, especially at deep penetration. Also it was concluded that ilmenite concrete with a density p = 4.6 g/cm/sup 3/ is better than both ordinary and ilmenite-limonite concrete for slow neutron attenuation.

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

    Energy Technology Data Exchange (ETDEWEB)

    Merz, Stefan [Vienna University of Technology, Atominstitut, Stadionallee 2, 1020 Vienna (Austria); Djuricic, Mile [Vienna University of Technology, Atominstitut, Stadionallee 2, 1020 Vienna (Austria); Nuclear Engineering Seibersdorf, 2444 Seibersdorf (Austria); Villa, Mario; Boeck, Helmuth [Vienna University of Technology, Atominstitut, Stadionallee 2, 1020 Vienna (Austria); Steinhauser, Georg, E-mail: georg.steinhauser@ati.ac.at [Vienna University of Technology, Atominstitut, Stadionallee 2, 1020 Vienna (Austria)

    2011-11-15

    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{sup 9} cm{sup -2} s{sup -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. - Highlights: > Neutron activation is an important process for the waste management of nuclear facilities. > Biological shield of the TRIGA reactor Vienna has been topic of investigation. > Flux values allow a categorization of the concrete concerning radiation protection legislation. > Reactor installations are of great importance as neutron sources into the biological shield. > Every installation shows distinguishable flux profiles.

  11. 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 hybrid...

  12. 压水堆核电站核三级屏蔽电泵试验鉴定程序%Testing of Nuclear Three-stage Shielded Electric Pump for Pressurized Water Reactor Nuclear Power Station

    Institute of Scientific and Technical Information of China (English)

    李作维; 孙悦; 薄大威

    2015-01-01

    This paper analyzes the necessity of obtaining the design/manufacture license of three generation nuclear power plants with two generation of pressurized water reactor nuclear power plant. Describe the function and structure of the three stage pump. Especially introduce the program and test method of the two generation of nuclear power station in three gener-ation pressurized water reactor nuclear power station.%分析了二代加压水堆核电站核三级屏蔽电泵取得民用核安全设备 (屏蔽泵) 设计/制造许可证的必要性, 对核三级屏蔽电泵的功能、 规格及结构进行了详细的阐述, 并重点针对二代加压水堆核电站核三级屏蔽电泵试验鉴定的程序及试验方法进行了介绍, 为核三级屏蔽电泵鉴定提供了依据.

  13. RESONANCE SELF-SHIELDING EFFECT IN UNCERTAINTY QUANTIFICATION OF FISSION REACTOR NEUTRONICS PARAMETERS

    Directory of Open Access Journals (Sweden)

    GO CHIBA

    2014-06-01

    Full Text Available In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

  14. 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.

  15. Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH{sub 2} and B{sub 4}C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor.

  16. 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.

  17. 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.

  18. 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.

  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. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-01

    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.

  2. Bibliography, subject index, and author index of the literature examined by the Radiation Shielding Information Center (Reactor and Weapons Radiation Shielding). [1973--1976

    Energy Technology Data Exchange (ETDEWEB)

    1978-01-01

    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.

  3. Airvolt Aircraft Electric Propulsion Test Stand

    Science.gov (United States)

    Samuel, Aamod; Lin, Yohan

    2015-01-01

    Development of an electric propulsion test stand that collects high-fidelity data of motor, inverter, and battery system efficiencies; thermal dynamics; and acoustics independent of manufacturer reported values will improve understanding of electric propulsion systems to be used in future aircraft. A buildup approach to this development reveals new areas of research and best practices in testing, and attempts to establish a standard for testing these systems.

  4. 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.

  5. Flexible neutron shielding composite material of EPDM rubber with boron trioxide: Mechanical, thermal investigations and neutron shielding tests

    Science.gov (United States)

    Özdemir, T.; Güngör, A.; Reyhancan, İ. A.

    2017-02-01

    In this study, EPDM and boron trioxide composite was produced and mechanical, thermal and neutron shielding tests were performed. EPDM rubber (Ethylene Propylene Diene Monomer) having a considerably high hydrogen content is an effective neutron shielding material. On the other hand, the materials containing boron components have effective thermal neutron absorption crossection. The composite of EPDM and boron trioxide would be an effective solution for both respects of flexibility and effectiveness for developing a neutron shielding material. Flexible nature of EPDM would be a great asset for the shielding purpose in case of intervention action to a radiation accident. The theoretical calculations and experimental neutron absorption tests have shown that the results were in parallel and an effective neutron shielding has been achieved with the use of the developed composite material.

  6. Wake Shield Facility Modal Survey Test in Vibration Acoustic Test Facility

    Science.gov (United States)

    1994-01-01

    Astronaut Ronald M. Sega stands beside the University of Houston's Wake Shield Facility before it undergoes a Modal Survey Test in the Vibration and Acoustic Test Facility Building 49, prior to being flown on space shuttle mission STS-60.

  7. Study of filtration of reactor beam of neutrons with cadmium in a multilayer shield containing boron carbide

    Energy Technology Data Exchange (ETDEWEB)

    Megahid, R.M. (Atomic Energy Establishment, Cairo (Egypt)); El-Kalla, El-Sayed H. (Al-Azhar Univ., Cairo (Egypt)); Esmaiel, M.W. (Ain Shams Univ., Cairo (Egypt))

    1985-08-01

    Experimental measurements have been carried out to study the effect of cadmium on the distribution and attenuation of thermal neutrons emitted from a reactor core and thermal neutrons produced in a heterogeneous shield of water, iron, iron + B/sub 4/C and ordinary concrete. Measurements were performed using a cadmium filtered reactor neutron beam emitted from one of the horizontal channels of ET-RR-1. It was found that the presence of a cadmium sheet at the channel exit causes a marked decrease in the thickness of the shield required for attenuating the thermal neutron flux by a certain factor.

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

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, L.E.; Chou, C.K.; Lo, T.; Schwartz, M.W.

    1988-06-01

    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.

  9. 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.

  10. Testing hot cell shielding in the fuel conditioning facility.

    Science.gov (United States)

    Courtney, J C; Klann, R T

    1997-01-01

    A comprehensive shield test program for a hot cell complex, the Fuel Conditioning Facility at Argonne National Laboratory, has been completed with minimum radiation exposure to participants. The recently modified shielding design for two hot cells and their associated transfer paths for irradiated materials was analyzed and tested for attenuating gamma rays from mixed fission product sources. Testing was accomplished using 0.37 TBq (10 Ci) and 518 TBq (14,000 Ci) 60Co sources. Of specific concern were radiation levels around wall penetrations and the interface between transport casks and the cell floor. Detailed measurements were made for surfaces that bound the hot cells, a transfer tunnel between the two cells, and storage pits that extend below the floors of both cells. In addition to surface measurements, dose equivalent rates in adjacent corridors were determined when the larger source was exposed. Results indicate that with some administrative controls, the facility shields are adequate to meet the design criterion that limits annual dose to less than 10 mSv (1 rem) for facility workers.

  11. Propfan test assessment testbed aircraft flutter model test report

    Science.gov (United States)

    Jenness, C. M. J.

    1987-01-01

    The PropFan Test Assessment (PTA) program includes flight tests of a propfan power plant mounted on the left wind of a modified Gulfstream II testbed aircraft. A static balance boom is mounted on the right wing tip for lateral balance. Flutter analyses indicate that these installations reduce the wing flutter stabilizing speed and that torsional stiffening and the installation of a flutter stabilizing tip boom are required on the left wing for adequate flutter safety margins. Wind tunnel tests of a 1/9th scale high speed flutter model of the testbed aircraft were conducted. The test program included the design, fabrication, and testing of the flutter model and the correlation of the flutter test data with analysis results. Excellent correlations with the test data were achieved in posttest flutter analysis using actual model properties. It was concluded that the flutter analysis method used was capable of accurate flutter predictions for both the (symmetric) twin propfan configuration and the (unsymmetric) single propfan configuration. The flutter analysis also revealed that the differences between the tested model configurations and the current aircraft design caused the (scaled) model flutter speed to be significantly higher than that of the aircraft, at least for the single propfan configuration without a flutter boom. Verification of the aircraft final design should, therefore, be based on flutter predictions made with the test validated analysis methods.

  12. 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.

  13. 14 CFR 135.145 - Aircraft proving and validation tests.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 3 2010-01-01 2010-01-01 false Aircraft proving and validation tests. 135... Aircraft and Equipment § 135.145 Aircraft proving and validation tests. (a) No certificate holder may... safely and in compliance with applicable regulatory standards. Validation tests are required for...

  14. ITER (International Thermonuclear Experimental Reactor) shield and blanket work package report

    Energy Technology Data Exchange (ETDEWEB)

    1988-06-01

    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.

  15. Intercomparison of Different Types of Locally Prepared Concretes and Its Usability for Reactor Neutron Shielding

    Science.gov (United States)

    El-Kolaly, M. A.; Makarious, A. S.; Bashter, I. I.; Kansouh, W. A.

    Measurements have been carried out to study the attenuation of neutron from a horizontal channel of the ET-RR-1 reactor. The assessments of neutron distribution inside three different types of locally prepared concretes have been evaluated.Neutron intensities in ilmenite-limonite concrete shield show an exponential decrease with increasing concrete thickness. Ilmenite concrete is a good attenuator for thermal and intermediate neutrons. However, ordinary and ilmenite-limonite concretes show efficient shielding for fast neutrons.Translated AbstractVergleich verschiedener Zementarten hinsichtlich ihrer Brauchbarkeit zur Neutronenabschirmung von ReaktorenMessungen zur Untersuchung der Neutronenabschwächung in einem horizontalen Kanal eines ET-RR-1-Reaktors wurden durchgeführt. Die Charakteristika der Neutronenverteilung innerhalb dreier unterschiedlich zusammengesetzter Zemente wurden bestimmt. Die Neutronenintensität in einem Schild aus Ilmenite-Limonitezement zeigt einen exponentiellen Abfall mit wachsender Dicke. Ilmenitezement ist ein guter Schild für thermale und mittlere Neutronen. Normaler und Ilmenite-Limonitezement zeigen effektive Abschirmung bei schnellen Neutronen.

  16. 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.

  17. ASDAR (aircraft to satellite data relay) flight test report

    Science.gov (United States)

    Domino, E. J.; Lovell, R. R.; Conroy, M. J.; Culp, D. H.

    1977-01-01

    The aircraft to Satellite Data Relay (ASDAR), an airborne data collection system that gathers meteorological data from existing aircraft instrumentation and relays it to ground user via a geo-synchronous meteorological satellite, is described and the results of the first test flight on a commercial Boeing 747 aircraft are presented. The flight test was successful and verified system performance in the anticipated environment.

  18. 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.

  19. Comparison of the effects of cadmium-shielded and boron carbide-shielded irradiation channel of the Ghana Research Reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Abrefah, R.G., E-mail: adibola2001@yahoo.com [National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O Box LG80, Legon, Accra (Ghana); Sogbadji, R.B.M.; Ampomah-Amoako, E.; Birikorang, S.A. [National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O Box LG80, Legon, Accra (Ghana); Odoi, H.C. [School of Nuclear and Allied Sciences, University of Ghana, P.O Box AE1, Atomic Energy Post, Accra (Ghana); Nyarko, B.J.B. [National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O Box LG80, Legon, Accra (Ghana)

    2011-08-15

    The MCNP model for the Ghana Research Reactor-1 (GHARR-1) was redesigned to incorporate cadmium-shielded irradiation channel as well as boron carbide-shielded channel in one of the outer irradiation channels. Further investigations were made after initial work in the cadmium-shielded channel to consider the boron carbide-shielded channel and both results were compared to determine the best material for the shielded channel. Before arriving at the final design of only one shielded outer irradiation channel extensive investigations were made into several other possible designs; 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 shielded channel is to equip GHARR-1 with the means of performing efficient epithermal neutron activation analysis. The use of epithermal neutron activation analysis can be very useful in many experiments and projects (e.g. it can be used to determine uranium and thorium in sediment samples). After the simulation, a comparison of the results from the boron carbide-shielded channel model for the GHARR-1 and the epicadmium-shielded channel was made. The inner irradiation channels of the two designs recorded peak values of approximately 1.18 x 10{sup 12} {+-} 0.0036 n/cm{sup 2} s, 1.32 x 10{sup 12} {+-} 0.0036 n/cm{sup 2} s and 2.71 x 10{sup 11} {+-} 0.0071 n/cm{sup 2} s for the thermal, epithermal and fast neutron flux, respectively. Likewise the outer irradiation channels of the two designs recorded peak values of approximately 7.36 x 10{sup 11} {+-} 0.0042 n/cm{sup 2} s, 2.53 x 10{sup 11} {+-} 0.0074 n/cm{sup 2} s and 4.73 x 10{sup 10} {+-} 0.0162 n/cm{sup 2} s for the thermal, epithermal and fast neutron flux, respectively. The epicadmium design recorded a peak thermal flux of 7.08 x 10{sup 11} {+-} 0.0033 n/cm{sup 2} s and an epithermal flux of 2.09 x 10{sup 11} {+-} 0.006 n/cm{sup 2} s in the irradiation channel where the shield was

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

    Energy Technology Data Exchange (ETDEWEB)

    Posta, B.A.; Kadar, I. [Bechtel Corp., San Francisco, CA (United States); Rao, A.S. [General Electric Nuclear Engineering, San Jose, CA (United States)

    1996-07-01

    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.

  1. 14 CFR 21.128 - Tests: aircraft engines.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Tests: aircraft engines. 21.128 Section 21.128 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT... engines. (a) Each person manufacturing aircraft engines under a type certificate only shall subject...

  2. 14 CFR 91.1041 - Aircraft proving and validation tests.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 2 2010-01-01 2010-01-01 false Aircraft proving and validation tests. 91... Ownership Operations Program Management § 91.1041 Aircraft proving and validation tests. (a) No program... tests. However, pilot flight training may be conducted during the proving tests. (d) Validation...

  3. Heat shields for aircraft - A new concept to save lives in crash fires.

    Science.gov (United States)

    Neel, C. B.; Parker, J. A.; Fish, R. H.; Henshaw, J.; Newland, J. H.; Tempesta, F. L.

    1971-01-01

    A passenger compartment surrounded by a fire-retardant shell, to protect the occupants long enough for the fire to burn out or for fire-fighting equipment to reach the aircraft and extinguish it, is proposed as a new concept for saving lives in crash fires. This concept is made possible by the recent development of two new fire-retardant materials: a very lightweight foam plastic, called polyisocyanurate foam, and an intumescent paint. Exposed to heat, the intumescent paint expands to many times its original thickness and insulates the surface underneath it. Demonstration tests are illustrated, described and discussed. However, some problems, such as preventing fuselage rupture and protecting windows, must be solved before such a system can be used.

  4. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

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

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J.M., E-mail: Chenjm@swip.ac.cn [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wu, J.H.; Liu, X.; Wang, P.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wang, Z.H.; Li, Z.N. [Ningxia Orient Non-ferrous Metals Group Co. Ltd., P.O. Box 105, Shizuishan (China); Wang, X.S.; Zhang, P.C. [China Academy of Engineering Physics, P.O. Box 919-71, Mianyang 621900 (China); Zhang, N.M.; Fu, H.Y.; Liu, D.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China)

    2011-10-01

    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.

  6. Aircraft

    Science.gov (United States)

    Hibbs, B.D.; Lissaman, P.B.S.; Morgan, W.R.; Radkey, R.L.

    1998-09-22

    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 gases 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. 31 figs.

  7. A Wind Tunnel Captive Aircraft Testing Technique

    Science.gov (United States)

    1976-04-01

    Flight/Wind Tunnel Correlation of Aircraft Longitudinal Motion ....................................... 14 10. Fright/Wind Tunnel Correlation of...I 2 3 4 5 6 T IME, s e c Figure 9. Flight/wind tunnel correla- tion of aircraft longitudinal motion. ’ D A n ~ v i i i | ~ 0 0 - 4 0

  8. Measurement of shielding characteristics in the prototype FBR Monju

    Energy Technology Data Exchange (ETDEWEB)

    Usami, Shin; Sasaki, Kenji; Deshimaru, Takehide; Nakashima, Fumiaki [Japan Nuclear Cycle Development Institute, Tsuruga, Fukui (Japan)

    2000-03-01

    In the prototype fast breeder reactor Monju, shielding measurements were made around the reactor core, the primary heat transport system (PHTS), and the fuel handling and storage system during the system start-up tests at different power levels between 0% and 45%. The objectives of the tests were to evaluate the margins by which the shielding performance exceeds the original design requirements, to demonstrate the validity of the shielding analysis method, and to acquire basic data for use in future FBR design. This paper summarizes the important features of the Monju shielding structures and the shielding measurement. (author)

  9. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG

  10. 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.

  11. 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

  12. Tissue equivalent proportional counter microdosimetry measurements utililzed aboard aircraft and in accelerator based space radiation shielding studies

    Science.gov (United States)

    Gersey, Brad; Wilkins, Richard

    The space radiation environment presents a potential hazard to the humans, electronics and materials that are exposed to it. Particle accelerator facilities such as the NASA Space Ra-diation Laboratory (NSRL) and Loma Linda University Medical Center (LLUMC) provide particle radiation of specie and energy within the range of that found in the space radiation environment. Experiments performed at these facilities determine various endpoints for bio-logical, electronic and materials exposures. A critical factor in the performance of rigorous scientific studies of this type is accurate dosimetric measurements of the exposures. A Tissue Equivalent Proportional Counter (TEPC) is a microdosimeter that may be used to measure absorbed dose, average quality factor (Q) and dose equivalent of the particle beam utilized in these experiments. In this work, results from a variety of space radiation shielding studies where a TEPC was used to perform dosimetry in the particle beam will be presented. These results compare the absorbed dose and dose equivalent measured downstream of equal density thicknesses of stan-dard and multifunctional shielding materials. The standard materials chosen for these shielding studies included High-Density Polyethylene (HDPE) and aluminum alloy, while the multifunc-tional materials included carbon composite infused with single walled carbon nanotubes. High energy particles including proton, silicon and iron nuclei were chosen as the incident radia-tion for these studies. Further, TEPC results from measurements taken during flights aboard ER-2 and KC-135 aircraft will also be discussed. Results from these flight studies include TEPC measurements for shielded and unshielded conditions as well as the effect of vibration and electromagnetic exposures on the TEPC operation. The data selected for presentation will highlight the utility of the TEPC in space radiation studies, and in shielding studies in particular. The lineal energy response function of the

  13. 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.

  14. MCNP Calculations for the Shielding Design of a Beam Tube to Be Installed at the Portuguese Research Reactor

    Science.gov (United States)

    Gonçalves, I. F.; Ramalho, A. G.; Gonçalves, I. C.; Salgado, J.

    The work presented concerns the calculation of the external biological shielding for a neutron beam tube that will be installed at the Portuguese Research Reactor, RPI. This tube will have enough versatility to be used in fields so different as the analysis of the composition of samples or research work in Boron Neutron Capture Therapy, BNCT. The calculation was made by using the MCNP code. This code is a well validated and widely used code, and has therefore become an important tool in the design and optimisation work of experiences related to neutrons and gamma radiation.

  15. Design, fabrication and test of Load Bearing multilayer insulation to support a broad area cooled shield

    Science.gov (United States)

    Dye, S. A.; Johnson, W. L.; Plachta, D. W.; Mills, G. L.; Buchanan, L.; Kopelove, A. B.

    2014-11-01

    Improvements in cryogenic propellant storage are needed to achieve reduced or Zero Boil Off of cryopropellants, critical for long duration missions. Techniques for reducing heat leak into cryotanks include using passive multi-layer insulation (MLI) and vapor cooled or actively cooled thermal shields. Large scale shields cannot be supported by tank structural supports without heat leak through the supports. Traditional MLI also cannot support shield structural loads, and separate shield support mechanisms add significant heat leak. Quest Thermal Group and Ball Aerospace, with NASA SBIR support, have developed a novel Load Bearing multi-layer insulation (LBMLI) capable of self-supporting thermal shields and providing high thermal performance. We report on the development of LBMLI, including design, modeling and analysis, structural testing via vibe and acoustic loading, calorimeter thermal testing, and Reduced Boil-Off (RBO) testing on NASA large scale cryotanks. LBMLI uses the strength of discrete polymer spacers to control interlayer spacing and support the external load of an actively cooled shield and external MLI. Structural testing at NASA Marshall was performed to beyond maximum launch profiles without failure. LBMLI coupons were thermally tested on calorimeters, with superior performance to traditional MLI on a per layer basis. Thermal and structural tests were performed with LBMLI supporting an actively cooled shield, and comparisons are made to the performance of traditional MLI and thermal shield supports. LBMLI provided a 51% reduction in heat leak per layer over a previously tested traditional MLI with tank standoffs, a 38% reduction in mass, and was advanced to TRL5. Active thermal control using LBMLI and a broad area cooled shield offers significant advantages in total system heat flux, mass and structural robustness for future Reduced Boil-Off and Zero Boil-Off cryogenic missions with durations over a few weeks.

  16. C Reactor overbore test facility review

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, P.A.; Nilson, R.

    1964-04-24

    In 1961, large-size, smooth-bore, Zircaloy process tubes were installed in C-Reactor graphite channels that had been enlarged to 2.275 inches. These tubes were installed to provide a test and demonstration facility for the concept of overboring as a means of securing significant improvement in the production capability of the reactors, After two years of facility operation, it is now appropriate to consider the extent to which original objectives have been achieved, to re-examine the original objectives, and to consider the best future use of this unique facility. This report presents the general results of such a review and re-examination in more detail.

  17. 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.

  18. 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.

  19. 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.

  20. Dielectric Heaters for Testing Spacecraft Nuclear Reactors

    Science.gov (United States)

    Sims, William Herbert; Bitteker, Leo; Godfroy, Thomas

    2006-01-01

    A document proposes the development of radio-frequency-(RF)-driven dielectric heaters for non-nuclear thermal testing of the cores of nuclear-fission reactors for spacecraft. Like the electrical-resistance heaters used heretofore for such testing, the dielectric heaters would be inserted in the reactors in place of nuclear fuel rods. A typical heater according to the proposal would consist of a rod of lossy dielectric material sized and shaped like a fuel rod and containing an electrically conductive rod along its center line. Exploiting the dielectric loss mechanism that is usually considered a nuisance in other applications, an RF signal, typically at a frequency .50 MHz and an amplitude between 2 and 5 kV, would be applied to the central conductor to heat the dielectric material. The main advantage of the proposal is that the wiring needed for the RF dielectric heating would be simpler and easier to fabricate than is the wiring needed for resistance heating. In some applications, it might be possible to eliminate all heater wiring and, instead, beam the RF heating power into the dielectric rods from external antennas.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kostov, M., E-mail: marin.kostov@riskeng.bg [Risk Engineering Ltd., Sofia (Bulgaria); Henkel, F.O. [Woelfel Beratende Ingenieure, Hoechberg (Germany); Andonov, A. [Risk Engineering Ltd., Sofia (Bulgaria)

    2014-04-01

    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. 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.

  3. Test study on the performance of shielding configuration with stuffed layer under hypervelocity impact

    Science.gov (United States)

    Ke, Fa-wei; Huang, Jie; Wen, Xue-zhong; Ma, Zhao-xia; Liu, Sen

    2016-10-01

    In order to study the cracking and intercepting mechanism of stuffed layer configuration on the debris cloud and to develop stuffed layer configuration with better performance, the hypervelocity impact tests on shielding configurations with stuffed layer were carried out. Firstly, the hypervelocity impact tests on the shielding configuration with stuffed layer of 3 layer ceramic fibre and 3 layer aramid fibre were finished, the study results showed that the debris cloud generated by the aluminum sphere impacting bumper at the velocity of about 6.2 km/s would be racked and intercepted by the stuffed layer configuration efficiently when the ceramic fibre layers and aramid fibre layers were jointed together, however, the shielding performance would be declined when the ceramic fibre layers and aramid fibre layers were divided by some distance. The mechanism of stuffed layer racking and intercepting the debris cloud was analyzed according to the above test results. Secondly, based on the mechanism of the stuffed layer cracking and intercepint debirs cloud the hypervelocity impact tests on the following three stuffed layer structures with the equivalent areal density to the 1 mm-thick aluminum plate were also carried out to compare their performance of cracking and intercepting debris cloud. The mechanisms of stuffed layer racking and intercepting the debris cloud were validated by the test result. Thirdly, the influence of the stuffed layer position on the shielding performance was studied by the test, too. The test results would provide reference for the design of better performance shielding configuration with stuffed layer.

  4. 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.

  5. Design and calculation for the main shielding layer of researching reactor%反应堆主屏蔽的设计与计算

    Institute of Scientific and Technical Information of China (English)

    钟文发; 胡永明; 钟兆鹏

    2001-01-01

    为使反应堆处于运行状态时,对辐射源的屏蔽满足辐射安全的要求,以及对堆的各部件和材料满足辐射限制的要求,必须设计堆的主屏蔽层。介绍了主屏蔽的设计与计算方法,以研究堆为设计实例,给出了主屏蔽的主要计算结果表明,以池水和重混凝土作生物屏蔽能满足辐射安全限值的要求,设计的主屏蔽层是适宜的。%To satisfy the national safety criteria of the radiationshielding when the reactor is in operation, We must complete the design of shielding for the main body of reactor. This paper introduces the design and computing methods of the shielding for the main body of reactor, using the researching reactor as an example, and gives the results of the computing for the radiation shielding. The results show that it can satisfy the safety criteria of radiation to use water and heavy concrete as biological layer for radiation, so the design of the main layer for radiation shielding is reliable.

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

    Energy Technology Data Exchange (ETDEWEB)

    Zaritsky, S.; Egorov, A. [National Research Center, Kurchatov Inst., Moscow 123182 (Russian Federation); Osmera, B.; Marik, M.; Rypar, V. [Research Centre Rez Ltd., Rez 25068 (Czech Republic); Cvachovec, F. [Univ. of Defense, Brno 61200 (Czech Republic); Kolros, A. [Czech Technical Univ., Prague 18000 (Czech Republic)

    2011-07-01

    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 {sup 3}He(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 {sup 3}He(n,p)T reaction rate were done with and without an additional moderator-polyethylene insert inside the channel. Second, measurements and calculations of the {sup 3}He(n,p)T reaction rate spatial distribution inside a concrete. The {sup 3}He(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)

  7. Establishment of Testing Device for Shielding Performance of X and Gamma Ray Radiation Protection Materials

    Institute of Scientific and Technical Information of China (English)

    SONG; Ming-zhe; WEI; Ke-xin; HOU; Jin-bing; WANG; Hong-yu; GAO; Fei; NI; Ning

    2015-01-01

    X and gamma ray radiation protective material shielding performance testing device was built based on the international standard IEC61331.1-2014.The device can be used to test attenuation ratio,attenuation equivalent and lead equivalent of radiation protective material in"narrow beam condition","broad beam condition"and"inverse

  8. Prototype test study on mechanical characteristics of segmental lining structure of underwater railway shield tunnel

    Institute of Scientific and Technical Information of China (English)

    He Chuan; Feng Kun; Yan Qixiang

    2014-01-01

    Based on the first underwater railway shield tunnel,the Shiziyang shield tunnel of Guangzhou Zhu-jiang River,the prototype test was carried out against its segmental lining structure by using“multi-function shield tunnel structure test system”. And the mechanical characteristics of segmental lining structure using straight assembling and staggered assembling were studied deeply. The results showed that,the mechanical characteristics of segmental lining structure varied with the water pressures;especially after cracking,the high water pressure played a significant role in slowing down the growing inner force and deformation. It also testi-fied that the failure characteristics varied with straight assembling structure and staggered assembling structure. Shear failure often occurred near longitudinal seam when using straight assembling.

  9. 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.

  10. 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)

  11. Neutron irradiation tests on B4C/epoxy composite for neutron shielding application and the parameters assay

    Science.gov (United States)

    Adeli, Ruhollah; Shirmardi, Seyed Pezhman; Ahmadi, Seyed Javad

    2016-10-01

    In this investigation, epoxy resin with a low viscosity amine-based curing agent was chosen as matrix and additives were added to epoxy resin using low speed stirring with ultrasonic waves approach. The chemical stability of resin during fabrication of composites was studied with Fourier transform infrared spectroscopy (FTIR). The effect of B4C particle size (20 and 150 μm) on neutron shielding was investigated. Besides, in order to develop the high performance composites, the effect of ATH (flame retardant) and WO3 powders (for shielding from against gamma rays) on neutron shielding property is considered. The neutron experiments were based on foil activation analysis in thermal column of Tehran Research Reactor (TRR). According to experimental data, required shield thickness (B4C, 150 μm, 3 wt%) for 80% absorption of neutron fluence was calculated about 9.8 mm. Consequently, data show thermal neutron absorption is dependent also on the size of the boron compound filler and show a significant enhancement in shielding performance when using smaller particle size of B4C filler. Furthermore, data obviously show that the neutron attenuation coefficient of reinforced composites increases to 0.345 cm-1 for B4C (20 μm, 5 wt%)/ Epoxy composite shield. As clearly data indicate, adding WO3 and ATH additive had a significant influence on the thermal neutron attenuation property and hybrid shield shows an enhancement of more than 60% in shielding performance.

  12. 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)

  13. Fusion Reactor and Fusion Reactor Materials:Concept Design of the ITER Test Blanket Modules

    Institute of Scientific and Technical Information of China (English)

    HUANGJinhua; LIZaixing; ZHUYukun; HUGang

    2003-01-01

    Performances required: prospect to be adopted in DEMO. Shielding for V.V. and TFC in ITER. Design principles: the peak temperature and stress should not exceed technical limits. The structure of test blanket modules (TBM) should be simple for easy fabrication, and TBM should be robust for reliability.

  14. 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.

  15. Optimization of a partially non-magnetic primary radiation shielding for the triple-axis spectrometer PANDA at the Munich high-flux reactor FRM-II

    CERN Document Server

    Pyka, N M; Rogov, A

    2002-01-01

    Monte Carlo simulations have been used to optimize the monochromator shielding of the polarized cold-neutron triple-axis spectrometer PANDA at the Munich high-flux reactor FRM-II. By using the Monte Carlo program MCNP-4B, the density of the total spectrum of incoming neutrons and gamma radiation from the beam tube SR-2 has been determined during the three-dimensional diffusion process in different types of heavy concrete and other absorbing material. Special attention has been paid to build a compact and highly efficient shielding, partially non-magnetic, with a total biological radiation dose of less than 10 mu Sv/h at its outsides. Especially considered was the construction of an albedo reducer, which serves to reduce the background in the experiment outside the shielding. (orig.)

  16. 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.,…

  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. 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. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  20. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M. [ed.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG&G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  1. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M. (ed.); Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  2. 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.

  3. 77 FR 65823 - Control of Air Pollution From Aircraft and Aircraft Engines; Emission Standards and Test Procedures

    Science.gov (United States)

    2012-10-31

    ... From the Federal Register Online via the Government Publishing Office ENVIRONMENTAL PROTECTION AGENCY 40 CFR Parts 87 RIN 2060-AO70 Control of Air Pollution From Aircraft and Aircraft Engines; Emission Standards and Test Procedures Correction In rule document 2012-13828 appearing on pages...

  4. 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.

  5. Novel durable bio-photocatalyst purifiers, a non-heterogeneous mechanism: accelerated entrapped dye degradation into structural polysiloxane-shield nano-reactors.

    Science.gov (United States)

    Dastjerdi, Roya; Montazer, Majid; Shahsavan, Shadi; Böttcher, Horst; Moghadam, M B; Sarsour, Jamal

    2013-01-01

    This research has designed innovative Ag/TiO(2) polysiloxane-shield nano-reactors on the PET fabric to develop novel durable bio-photocatalyst purifiers. To create these very fine nano-reactors, oppositely surface charged multiple size nanoparticles have been applied accompanied with a crosslinkable amino-functionalized polysiloxane (XPs) emulsion. Investigation of photocatalytic dye decolorization efficiency revealed a non-heterogeneous mechanism including an accelerated degradation of entrapped dye molecules into the structural polysiloxane-shield nano-reactors. In fact, dye molecules can be adsorbed by both Ag and XPs due to their electrostatic interactions and/or even via forming a complex with them especially with silver NPs. The absorbed dye and active oxygen species generated by TiO(2) were entrapped by polysiloxane shelter and the presence of silver nanoparticles further attract the negative oxygen species closer to the adsorbed dye molecules. In this way, the dye molecules are in close contact with concentrated active oxygen species into the created nano-reactors. This provides an accelerated degradation of dye molecules. This non-heterogeneous mechanism has been detected on the sample containing all of the three components. Increasing the concentration of Ag and XPs accelerated the second step beginning with an enhanced rate. Further, the treated samples also showed an excellent antibacterial activity.

  6. Testing For EM Upsets In Aircraft Control Computers

    Science.gov (United States)

    Belcastro, Celeste M.

    1994-01-01

    Effects of transient electrical signals evaluated in laboratory tests. Method of evaluating nominally fault-tolerant, aircraft-type digital-computer-based control system devised. Provides for evaluation of susceptibility of system to upset and evaluation of integrity of control when system subjected to transient electrical signals like those induced by electromagnetic (EM) source, in this case lightning. Beyond aerospace applications, fault-tolerant control systems becoming more wide-spread in industry; such as in automobiles. Method supports practical, systematic tests for evaluation of designs of fault-tolerant control systems.

  7. Tests of shielding effectiveness of Kevlar and Nextel onboard the International Space Station and the Foton-M3 capsule.

    Science.gov (United States)

    Pugliese, M; Bengin, V; Casolino, M; Roca, V; Zanini, A; Durante, M

    2010-08-01

    Radiation assessment and protection in space is the first step in planning future missions to the Moon and Mars, where mission and number of space travelers will increase and the protection of the geomagnetic shielding against the cosmic radiation will be absent. In this framework, the shielding effectiveness of two flexible materials, Kevlar and Nextel, were tested, which are largely used in the construction of spacecrafts. Accelerator-based tests clearly demonstrated that Kevlar is an excellent shield for heavy ions, close to polyethylene, whereas Nextel shows poor shielding characteristics. Measurements on flight performed onboard of the International Space Station and of the Foton-M3 capsule have been carried out with special attention to the neutron component; shielded and unshielded detectors (thermoluminescence dosemeters, bubble detectors) were exposed to a real radiation environment to test the shielding properties of the materials under study. The results indicate no significant effects of shielding, suggesting that thin shields in low-Earth Orbit have little effect on absorbed dose.

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

    Directory of Open Access Journals (Sweden)

    Košťál Michal

    2016-01-01

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

  9. Measurements of thermal- and slow-neutron dose distributions in ordinary concrete shield using a reactor neutron beam of different energy ranges

    Energy Technology Data Exchange (ETDEWEB)

    Megahid, R.M.; Makarious, A.S.; El-Kolaly, M.A.; Afifi, Y.A.

    1980-01-01

    Experimental studies on the distribution and attenuation of thermal and slow neutron doses in ordinary concrete shield have been carried-out. A collimated beam of reactor neutrons emitted from one of the horizontal channels of the ET-RR-1 reactor was used. Measurements were performed using, a direct beam, cadmium filtered beam and boron carbide filtered beam. The neutron doses were measured using thermolumin-escent Li/sub 2/B/sub 4/O/sub 7/ detectors. The measured data have been analyzed and a group of attenuation curves were given for beams of reactor neutrons of different energy. These curves show that cadmium and boron carbide filters tend to decrease the neutron doses specially at the beginning of penetration. The data were transformed to that which would be obtained using neutron sources of different geometries.

  10. 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.

  11. 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.

  12. Self Diagnostic Accelerometer Testing on the C-17 Aircraft

    Science.gov (United States)

    Tokars, Roger P.; Lekki, John D.

    2013-01-01

    The self diagnostic accelerometer (SDA) developed by the NASA Glenn Research Center was tested for the first time in an aircraft engine environment as part of the Vehicle Integrated Propulsion Research (VIPR) program. The VIPR program includes testing multiple critical flight sensor technologies. One such sensor, the accelerometer, measures vibrations to detect faults in the engine. In order to rely upon the accelerometer, the health of the accelerometer must be ensured. The SDA is a sensor system designed to actively determine the accelerometer structural health and attachment condition, in addition to vibration measurements. The SDA uses a signal conditioning unit that sends an electrical chirp to the accelerometer and recognizes changes in the response due to changes in the accelerometer health and attachment condition. To demonstrate the SDAs flight worthiness and robustness, multiple SDAs were mounted and tested on a C-17 aircraft engine. The engine test conditions varied from engine off, to idle, to maximum power. The SDA attachment conditions were varied from fully tight to loose. The newly developed SDA health algorithm described herein uses cross correlation pattern recognition to discriminate a healthy from a faulty SDA. The VIPR test results demonstrate for the first.

  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. Entrained Flow Reactor Test of Potassium Capture by Kaolin

    DEFF Research Database (Denmark)

    Wang, Guoliang; Jensen, Peter Arendt; Wu, Hao

    2015-01-01

    In the present study a method to simulate the reaction between gaseous KCl and kaolin at suspension fired condition was developed using a pilot-scale entrained flow reactor (EFR). Kaolin was injected into the EFR for primary test of this method. By adding kaolin, KCl can effectively be captured......-bed reactor. The method using the EFR developed in this study will be applied for further systematic investigation of different additives....

  15. The RES Reactor. A test reactor for the French naval propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Pivet, Sylvestre [CEA, Centre de Cadarache, F-13108 Saint Paul lez Durance (France); Minguet, Jean-Luc [AREVA-Technicatome, BP17, 91192 Gif-sur-Yvette (France)

    2006-07-01

    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

  16. High speed wind tunnel tests of the PTA aircraft. [Propfan Test Assessment Program

    Science.gov (United States)

    Aljabri, A. S.; Little, B. H., Jr.

    1986-01-01

    Propfans, advanced highly-loaded propellers, are proposed to power transport aircraft that cruise at high subsonic speeds, giving significant fuel savings over the equivalent turbofan-powered aircraft. NASA is currently sponsoring the Propfan Test Assessment Program (PTA) to provide basic data on the structural integrity and acoustic performance of the propfan. The program involves installation design, wind-tunnel tests, and flight tests of the Hamilton Standard SR-7 propfan in a wing-mount tractor installation on the Gulfstream II aircraft. This paper reports on the high-speed wind-tunnel tests and presents the computational aerodynamic methods that were employed in the analyses, design, and evaluation of the configuration. In spite of the complexity of the configuration, these methods provide aerodynamic predictions which are in excellent agreement with wind-tunnel data.

  17. 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.

  18. 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.

  19. 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

  20. Final safeguards analysis, high temperature lattice test reactor. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Hanthorn, H.E.; Brown, W.W.; Clark, R.G.; Heineman, R.E.; Humes, R.M.

    1966-01-01

    The PMACS `reactor-normal` signal signifies that important process variables do not exceed their set points, that various interlocks are properly set, that functional tests of the computer operation are satisfactory, and that the reactor flux level and period derived from two additional, independent, and dissimilar channels are within set limits. This safety circuit combines the features of redundancy, dissimilar components, and frequent testing which are required for best reliability. The experimental equipment auxiliary to the reactor includes two oscillator mechanisms, one to move the test cell or the adjoining cell into and out of position, the other to move small specimens in the test cell or adjoining cells. They have cooling chambers for the removal of specimens from the test cell without the necessity of cooling the reactor. A neutron chopper and time-of-flight spectrometer are provided; the neutron detectors, at the end of a 25-meter flight tube, are in an adjoining small building. Test cores may be assembled on a core dolly have a load capacity of 14,000 lb. Two wire traverse mechanisms are provided for measurements of flux distribution.

  1. 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.

  2. 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.

  3. Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

    Science.gov (United States)

    Bragg-Sitton, Shannon M.

    2005-01-01

    The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

  4. Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-15

    A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. The various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.

  5. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

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

    OpenAIRE

    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...

  7. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1988-05-01

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

  8. Standard Guide for Benchmark Testing of Light Water Reactor Calculations

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

  9. Neutron shielding effects of spent fuel tank of high temperature reactor%高温堆乏燃料贮罐中子屏蔽性能计算

    Institute of Scientific and Technical Information of China (English)

    李文茜; 李红; 谢锋; 曹建主; 方晟

    2013-01-01

    High temperature gas cooled reactor-pebble bed module (HTR-PM) adopts the coated particle spherical fuel elements, during the reactor's running, the constantly discharged spent fuel spheres will be loaded into the spent fuel tank. The spent fuel tank should use proper materials and thicknesses to shield gammas and neutrons effectively, and guarantee the dose limit not to be exceeded outside the tanks. Both relaxation length method and Monte Carlo simulation method were employed to study the neutrons' shielding capabilities of the spent fuel tank. Iron and borated polyethylene were chosen to be the shielding materials. The shielding capabilities of iron and borated polyethylene with different B4C contents (mass fraction 0, 5%, 10% and 15%) were calculated. The effect of the spent fuel spheres' self-absorption to the dose rate outside the tank was also considered, when the tank was full of the spent fuel spheres. The calculation results of these two methods are in good agreement, and provide important guiding suggestions for the shielding design in the practical engineering.%球床模块式高温气冷堆采用包覆颗粒球形燃料元件,在反应堆运行过程中,不断排出的乏燃料球将被装入乏燃料贮罐.乏燃料贮罐应选取适当的材料和厚度,对光子和中子进行有效屏蔽,使罐外的剂量率满足相应的限值要求.为此,使用张弛长度法和蒙特卡罗模拟法研究乏燃料贮罐的中子屏蔽性能.屏蔽材料为铁和含硼聚乙烯,计算了铁和不同B4C含量聚乙烯的屏蔽性能,并给出了乏燃料贮罐装满乏燃料球后,乏燃料球自吸收对贮罐外剂量率的影响.两种方法计算结果吻合很好,可以为实际工程中的屏蔽设计提供参考意见.

  10. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

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

  11. 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.

  12. Testing piezoelectric sensors in a nuclear reactor environment

    Science.gov (United States)

    Reinhardt, Brian T.; Suprock, Andy; Tittmann, Bernhard

    2017-02-01

    Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE's FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this work piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts Institute of Technology Research reactor to a fast neutron fluence of 8.65×1020 nf/cm2. It is demonstrated that Bismuth Titanate is capable of transduction up to 5 × 1020 nf/cm2, Zinc Oxide is capable of transduction up to at least 6.27 × 1020 nf/cm2, and Aluminum Nitride is capable of transduction up to at least 8.65 × 1020 nf/cm2.

  13. Enhanced in-pile instrumentation at the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T.; Chase, B. M.; Palmer, J.; Condie, K. G.; Davis, K. L. [Idaho National Laboratory, MS 3840, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2011-07-01

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and realtime flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted. (authors)

  14. Shielding evaluation and acceptance testing of a prefabricated, modular, temporary radiation therapy treatment facility.

    Science.gov (United States)

    Ezzell, Gary A

    2004-01-01

    We have recently commissioned a temporary radiation therapy facility that is novel in two aspects: it was constructed using modular components, and the LINAC was installed in one of the modular sections before it was lifted into position. Additional steel and granular fill was added to the modular sections on-site during construction. The building will be disassembled and removed when no longer needed. This paper describes the radiation shielding specifications and survey of the facility, as well as the ramifications for acceptance testing occasioned by the novel installation procedure. The LINAC is a Varian 21EX operating at 6 MV and 18 MV. The radiation levels outside the vault satisfied the design criteria, and no anomalous leakage was detected along the joints of the modular structure. At 18 MV and 600 monitor units (MU) per minute, the radiation level outside the primary barrier walls was 8.5 micro Sv/h of photons; there were no detectable neutrons. Outside the direct-shielded door, the levels were 0.4 micro Sv/h of photons and 3.0 micro Sv/h of neutrons. The isocentricity of the accelerator met the acceptance criteria and was not affected by its preinstallation into an integrated baseframe and subsequent transport to the building site.

  15. 77 FR 14319 - Unmanned Aircraft System Test Sites

    Science.gov (United States)

    2012-03-09

    ... infrastructure and research needs; and (C) Consult with the National Aeronautics and Space Administration and the... Aeronautics and Space Administration (NASA); (4) Address both civil and public unmanned aircraft systems; (5... be flown using a data link to transmit commands to the aircraft. They may perform a variety of...

  16. Laboratory test and acoustic analysis of cabin treatment for propfan test assessment aircraft

    Science.gov (United States)

    Kuntz, H. L.; Gatineau, R. J.

    1991-01-01

    An aircraft cabin acoustic enclosure, built in support of the Propfan Test Assessment (PTA) program, is described. Helmholtz resonators were attached to the cabin trim panels to increase the sidewall transmission loss (TL). Resonators (448) were located between the trim panels and fuselage shell. In addition, 152 resonators were placed between the enclosure and aircraft floors. The 600 resonators were each tuned to a 235 Hz resonance frequency. After flight testing on the PTA aircraft, the enclosure was tested in the Kelly Johnson R and D Center Acoustics Lab. Laboratory noise reduction (NR) test results are discussed. The enclosure was placed in a Gulfstream 2 fuselage section. Broadband (138 dB overall SPL) and tonal (149 dB overall SPL) excitations were used in the lab. Tonal excitation simulated the propfan flight test excitation. The fundamental tone was stepped in 2 Hz intervals from 225 through 245 Hz. The resonators increase the NR of the cabin walls around the resonance frequency of the resonator array. The effects of flanking, sidewall absorption, cabin adsorption, resonator loading of trim panels, and panel vibrations are presented. Increases in NR of up to 11 dB were measured.

  17. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  18. Test Plan for Westinghouse Hanford Company`s Hedgehog Shielded Container, Docket 94-39-7A, Type A Container

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, D.L.

    1995-02-27

    This report documents the US Department of Transportation Specification 7A Type A (DOT-7A) compliance testing to be followed for qualification of the Westinghouse Hanford Company`s Hedgehog Shielded Container for use as a Type A packaging. The packaging configurations being tested are intended for liquids and solids, and for air transportation.

  19. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  20. The Jules Horowitz reactor (JHR), a European material testing reactor (MTR), with extended experimental capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, 13 - Saint-Paul-lez-Durance (France)]|[CEA Saclay Dir. de l' Energie Nucleaire DEN, 91 - Gif sur Yvette (France)

    2003-07-01

    The Jules Horowitz Reactor (JHR) is the European MTR (Material Testing Reactor) designed to provide, after 2010, the necessary knowledge for keeping the existing power plants in operation and to design innovative reactors types with new objectives such as: minimizing the radioactive waste production, taking into account additional safety requirements, preventing risks of nuclear proliferation... To achieve such an ambitious objective. The JHR is designed with a high flexibility in order to satisfy the current demand from European industry, research and to be able to accommodate future requirements. The JHR will offer a wide range of performances and services in gathering, in a single site at Cadarache, all the necessary functionalities and facilities for an effective production of results: e.g. fuel fabrication laboratories, preparation of the instrumented devices, interpretation of the experiments, modelling. The JHR must rely on a top level scientific environment based on experts teams from CEA and EC and local universities. With a thermal flux of 7,4.10{sup 14} ncm{sup -2} s{sup -1} and a fast flux of 6,4.10{sup 14} ncm{sup -2}s{sup -1}, it is possible to carry out irradiation experiments on materials and fuels whatever the reactor type considered. It will also be possible to carry out locally, fast neutron irradiation to achieve damage effect up to 25 dpa/year. (dpa = displacement per atom.) The study of the fuels behavior under accidental conditions, from analytical experiments, on a limited amount of irradiated fuel, is a major objective of the project. These oriented safety tests are possible by taking into account specific requirements in the design of the facility such as the tightness level of the containment building, the addition of an alpha hot cell and a laboratory for on line fission products measurement. (authors)

  1. New results from pulse tests in the CABRI reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schmitz, F.; Papin, J.; Haessler, M. [Institut de Proterction et de Surete Nucleaire, Saint Paul Lez Durance (France)] [and others

    1996-03-01

    At the 21st and 22nd WRSM (1,2), the motivation and objectives of the French program on the behaviour of high burnup PWR fuel under RIA conditions in the CABRI test reactor has been presented. The major results of the three first tests of the test matrix were presented and in particular REP-Na1, which failed at an unexpected low level of fuel enthalpy, was exposed to the community of nuclear safety research. At this time, no final understanding was reached for the origin of the failure. This objective is reached now. Two further tests, REP-Na4 and 5, have been performed in 1995, they demonstrated a satisfactory and safe behaviour by resisting to the early phase of severe loading during the RIA pulse test. Further examination work and analytical testing is in progress and the next tests with MOX fuel are being prepared.

  2. 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.

  3. Radiochemical problems of fusion reactors. 1. Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Crespi, M.B.A.

    1984-02-01

    A list of fusion reactor candidate materials is given, for use in connection with blanket structure, breeding, moderation, neutron multiplication, cooling, magnetic field generation, electrical insulation and radiation shielding. The phenomena being studied for each group of materials are indicated. Suitable irradiation test facilities are discussed under the headings (1) accelerator-based neutron sources, (2) fission reactors, and (3) ion accelerators.

  4. Residual radioactivity guidelines for the heavy water components test reactor at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B. Smith, R.; McNeil, J.

    1997-04-01

    Guidelines were developed for acceptable levels of residual radioactivity in the Heavy Water Components Test Reactor (HWCTR) facility at the conclusion of its decommissioning. Using source terms developed from data generated in a detailed characterization study, the RESRAD and RASRAD-BUILD computer codes were used to calculate derived concentration guideline levels (DCGLs) for the radionuclides that will remain in the facility. The calculated DCGLs, when compared to existing concentrations of radionuclides measured during a 1996 characterization program, indicate that no decontamination of concrete surfaces will be necessary. Also, based on the results of the calculations, activated concrete in the reactor biological shield does not have to be removed, and imbedded radioactive piping in the facility can remain in place. Viewed in another way, the results of the calculations showed that the present inventory of residual radioactivity in the facility (not including that associated with the reactor vessel and steam generators) would produce less than one millirem per year above background to a hypothetical individual on the property. The residual radioactivity is estimated to be approximately 0.04 percent of the total inventory in the facility as of March, 1997. According to the results, the only radionuclides that would produce greater than 0.0.1-millirem per year are Am-241 (0.013 mrem/yr at 300 years), C-14 (0.022 mrem/yr at 1000 years) and U-238 (0.034 mrem/yr at 6000 years). Human exposure would occur only through the groundwater pathways, that is, from water drawn from, a well on the property. The maximum exposure would be approximately one percent of the 4 millirem per year ground water exposure limit established by the U.S. Environmental Protection Agency. 11 refs., 13 figs., 15 tabs.

  5. 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.

  6. Final Physics Report for the Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wolfe, I. B. [Savannah River Site (SRS), Aiken, SC (United States)

    1956-06-25

    This report is a summary of the physics design work performed on the Engineering Test Reactor. The ETR presents computational difficulties not found in other reactors because of the large number of experimental holes in the core. The physics of the ETR depends strongly upon the contents of the in-core experimental facilities. In order to properly evaluate the reactor; taking into account the experiments in the core, multi-region, two-dimensional calculations are required. These calculations require .the use of a large computer such as the Remington Rand Univac and are complex and expensive enough to warrant a five-stage program: 1. In the early stages of design, only preliminary two-dimensional calculations were performed .in order to obtain a rough idea of the general behavior of the reactor and its critical mass with tentative experiments in place. 2. A large amount of work was carried out in which the reactor was approximated as one with a uniform homogeneous core. With this model, detailed studies were carried out to investigate the feasibility and to obtain general design data on such points as the design and properties of the gray and black-control rods, the design of the beryllium reflector, gamma and neutron heating, the use of burnable poisons, etc. In performing these calculations, use was made of the IBM 650 PROD code obtained from KAPL. 3. With stages 1 and 2 carried out, two-dimensional calculations of the core at start-up conditions were performed on the Univac computer. 4. Detailed two-dimensional calculations of the properties of the ETR with a proposed first set of experiments in place were carried out. 5. A series of nuclear tests were performed at the reactivity measurements facility at the MTR site in order to confirm the validity of the analytical techniques in physics analysis. In performing the two-dimensional Univac calculations, the MUG code developed by KAPL and the Cuthill code developed at the David Taylor Model Basin were utilized. In

  7. Advanced Test Reactor National Scientific User Facility Progress

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Todd R. Allen; James I. Cole; Jeff B. Benson; Mary Catherine Thelen

    2012-10-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the world’s premier test reactors for studying the effects of intense neutron radiation on reactor materials and fuels. The ATR began operation in 1967, and has operated continuously since then, averaging approximately 250 operating days per year. The combination of high flux, large test volumes, and multiple experiment configuration options provide unique testing opportunities for nuclear fuels and material researchers. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected highly-enriched uranium fueled, reactor with a maximum operating power of 250 MWth. The ATR peak thermal flux can reach 1.0 x1015 n/cm2-sec, and the core configuration creates five main reactor power lobes (regions) that can be operated at different powers during the same operating cycle. In addition to these nine flux traps there are 68 irradiation positions in the reactor core reflector tank. The test positions range from 0.5” to 5.0” in diameter and are all 48” in length, the active length of the fuel. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation 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. Goals of the ATR NSUF are to define the cutting edge of nuclear technology research in high temperature and radiation environments, contribute to improved industry performance of current and future light water reactors, and stimulate cooperative research between user groups conducting basic and applied research. The ATR NSUF has developed partnerships with other universities and national laboratories to enable ATR NSUF researchers to perform research at these other facilities, when the research objectives

  8. Primary design and optimization of shielding for nuclear medical ship reactor%核医疗船反应堆屏蔽初步设计及优化

    Institute of Scientific and Technical Information of China (English)

    万海霞; 徐治龙; 邵静; 孙征; 李龙; 吴晓春

    2017-01-01

    The program of nuclear medical ship,funded by the cancer healing program of IAEA,was proposed according to the existing Miniature Neutron Source Reactor (MNSR)technology.The nuclear medical ship equipped with Boron Neutron Cap-ture Therapy(BNCT)device was developed,which opened a new scope of nuclear science application.The reactor of nuclear med-ical ship was designed in accordance with In-hospital Neutron Irradiator mark I (IHNI-1).In IHNI-1,heavy concrete was used as shielding material,and the reactor pool was cylindrical.Whereas,the volume and total weight of the reactor were too large to meet the ship’s requirement.After the design and optimization of the reactor’s shielding system by M-C method,stainless steel and B-polyethylene were chosen as shielding materials,and square pool was substituted by compact cylindrical pool.The result shows that on the premise of guaranteeing safety,shielding system’s mass and volume were cut down greatly,i.e.,this design can meet the requirement of nuclear medical ship.%核医疗船的概念是在现有微堆(MNSR)技术的基础上提出的,它基于 IAEA 的癌症关怀项目,开发配备硼中子俘获治疗技术(BNCT)的远洋核医疗船,该项目开辟核科技应用的新领域。核医疗船的反应堆设计参考了已建成的医院中子照射器-1型(IHNI-1)反应堆,该堆采用重混凝土作为主要屏蔽材料,水池为方形,其体积和质量都很大,不能满足船用要求。为了使反应堆能够满足船用要求,使用蒙特卡罗方法对医院中子照射器的反应堆屏蔽系统进行重新设计和优化,通过对多个方案的综合对比,最终确定采用不锈钢、含硼聚乙烯为屏蔽材料,并将水池设计成结构紧凑的圆柱形结构,该屏蔽方案在保证安全的基础上,使屏蔽系统的质量和体积大大降低,满足了船用要求。

  9. Conceptual design study of a scyllac fusion test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thomassen, K.I. (comp.)

    1975-07-01

    The report describes a conceptual design study of a fusion test reactor based on the Scyllac toroidal theta-pinch approach to fusion. It is not the first attempt to describe the physics and technology required for demonstrating scientific feasibility of the approach, but it is the most complete design in the sense that the physics necessary to achieve the device goals is extrapolated from experimentally tested MHD theories of toroidal systems,and it uses technological systems whose engineering performance has been carefully calculated to ensure that they meet the machine requirements.

  10. 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.

  11. Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research

    Energy Technology Data Exchange (ETDEWEB)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials

  12. Interior noise control ground test studies for advanced turboprop aircraft applications

    Science.gov (United States)

    Simpson, Myles A.; Cannon, Mark R.; Burge, Paul L.; Boyd, Robert P.

    1989-01-01

    The measurement and analysis procedures are documented, and the results of interior noise control ground tests conducted on a DC-9 aircraft test section are summarized. The objectives of these tests were to study the fuselage response characteristics of treated and untreated aircraft with aft-mount advanced turboprop engines and to analyze the effectiveness of selected noise control treatments in reducing passenger cabin noise on these aircraft. The results of fuselage structural mode surveys, cabin cavity surveys and sound intensity surveys are presented. The performance of various structural and cabin sidewall treatments is assessed, based on measurements of the resulting interior noise levels under simulated advanced turboprop excitation.

  13. Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

    Energy Technology Data Exchange (ETDEWEB)

    S. Bragg-Sitton; J. Bess; J. Werner; G. Harms; S. Bailey

    2011-09-01

    Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al., 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).

  14. Local stability tests in Dresden 2 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    March-Leuba, J.; Fry, D.N.; Buchanan, M.E.; McNew, C.O.

    1984-04-01

    This report presents the results of a local stability test performed at Dresden Unit 2 in May 1983 to determine the effect of a new fuel element design on local channel stability. This test was performed because the diameter of the new fuel rods increases the heat transfer coefficient, making the reactor more responsive and, thus, more susceptible to instabilities. After four of the new fuel elements with a 9 x 9 array of fuel rods were loaded into Dresden 2, the test was performed by inserting an adjacent control rod all the way in and then withdrawing it to its original position at maximum speed. At the moment of the test, reactor conditions were 52.7% power and 38.9% flow. Both the new 9 x 9 fuel elements and the standard 8 x 8 ones proved to be locally stable when operating at minimum pump speed at the beginning of cycle in Dresden 2, and no significant difference was found between the behavior of the two fuel types. Finally, Dresden 2 showed a high degree of stability during control rod and normal noise type perturbations.

  15. Present status and future perspectives of research and test reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kaneko, Yoshihiko [Atomic Energy Research Laboratory, Musashi Institute of Technology, Kawasaki, Kanagawa (Japan); Kaieda, Keisuke [Department of Research Reactor, Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2000-10-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfill a major role in the study of nuclear energy and fundamental research. At present four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR) are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has recently reached first criticality and now in the power up test. In 1966, the Kyoto University built the Kyoto University Reactor (KUR) and started its operation for joint use program of the Japanese universities. This paper introduces these reactors and describes their present operational status and also efforts for aging management. The recent tendency of utilization and future perspectives is also reported. (author)

  16. High temperature indentation tests on fusion reactor candidate materials

    Energy Technology Data Exchange (ETDEWEB)

    Montanari, R. [Dipartimento di Ingegneria Meccanica, Universita di Roma-Tor Vergata, Via del Politecnico 1, I-00133 Rome (Italy)]. E-mail: roberto.montanari@uniroma2.it; Filacchioni, G. [ENEA CR Casaccia, Via Anguillarese 301, I-00060 S.M. di Galeria, Rome (Italy); Iacovone, B. [Dipartimento di Ingegneria Meccanica, Universita di Roma-Tor Vergata, Via del Politecnico 1, I-00133 Rome (Italy); Plini, P. [Dipartimento di Ingegneria Meccanica, Universita di Roma-Tor Vergata, Via del Politecnico 1, I-00133 Rome (Italy); Riccardi, B. [Associazione EURATOM-ENEA sulla Fusione, P.O. Box 65, I-00044 Frascati, Rome (Italy)

    2007-08-01

    Flat-top cylinder indenter for mechanical characterization (FIMEC) is an indentation technique employing cylindrical punches with diameters ranging from 0.5 to 2 mm. The test gives pressure-penetration curves from which the yield stress can be determined. The FIMEC apparatus was developed to test materials in the temperature range from -180 to +200 {sup o}C. Recently, the heating system of FIMEC apparatus has been modified to operate up to 500 {sup o}C. So, in addition to providing yield stress over a more extended temperature range, it is possible to perform stress-relaxation tests at temperatures of great interest for several nuclear fusion reactor (NFR) alloys. Data on MANET-II, F82H mod., Eurofer-97, EM-10, AISI 316 L, Ti6Al4V and CuCrZr are presented and compared with those obtained by mechanical tests with standard methods.

  17. 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.

  18. Action Memorandum for the Engineering Test Reactor under the Idaho Cleanup Project

    Energy Technology Data Exchange (ETDEWEB)

    A. B. Culp

    2007-01-26

    This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared adn released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessol. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface.

  19. Safety Analyses at the Idaho National Engineering and Environmental Laboratory Test Reactor Area - Past to Present

    Energy Technology Data Exchange (ETDEWEB)

    Ambrosek, Richard Garry; Ingram, Frederick William

    1999-11-01

    Test reactors are unique in that the core configuration may change with each operating interval. The process of safety analyses for test reactors at the Idaho National Engineering and Environmental Test Reactor Area has evolved as the computing capabilities, software, and regulatory requirements have changed. The evaluations for experiments and the reactor have moved from measurements in a set configuration and then application to other configurations with a relatively large error to modeling in three-dimensions and explicit analyses for each experiment and operating interval. This evolution is briefly discussed for the Test Reactor Area.

  20. Analytical and statistical calculation of gamma dose rate for the accident of losing the shield for Tehran Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    In this paper we study the analytical and statistical results of estimating the gamma dose rate at pool access floor in TRR when the core shield accidentally decreases to some non-permitted levels. Due to the risk of experimental techniques, we use the analytical and statistical methods. In normal conditions (no risk),the discrepancies between experiment and two methods are justified and it is found that for such problems we have to normalize these methods to experimental results as follows: the analytical method by factor 0.13 and MCNP by 1.7.

  1. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

  2. EPR/PTFE dosimetry for test reactor environments

    Energy Technology Data Exchange (ETDEWEB)

    Vehar, D.W.; Griffin, P.J.; Quirk, T.J. [Sandia National Laboratories, Albuquerque, NM 87185-1146 (United States)

    2011-07-01

    The use of Electron Paramagnetic Resonance (EPR) spectroscopy with materials such as alanine is well established as a technique for measurement of ionizing radiation absorbed dose in photon and electron fields such as Co-60, high-energy bremsstrahlung and electron-beam fields [1]. In fact, EPR/Alanine dosimetry has become a routine transfer standard for national standards bodies such as NIST and NPL. In 1992 the Radiation Metrology Laboratory (RML) at Sandia National Laboratories implemented EPR/Alanine capabilities for use in routine and calibration activities at its Co-60 and pulsed-power facilities. At that time it also investigated the usefulness of the system for measurement of absorbed dose in the mixed neutron/photon environments of reactors such as the Sandia Pulsed Reactor and the Annular Core Research Reactor used for hardness testing of electronics. The RML concluded that the neutron response of alanine was a sufficiently high fraction of the overall dosimeter response that the resulting uncertainties in the photon dose would be unacceptably large for silicon-device testing. However, it also suggested that non-hydrogenous materials such as polytetrafluoroethylene (PTFE) would exhibit smaller neutron response and might be useful in mixed environments. Preliminary research with PTFE in photon environments indicated considerable promise, but further development was not pursued at that time. Because of renewed interest in absorbed dose measurements that could better define the individual contributions of photon and neutron components to the overall dose delivered to a test object, the RML has re-initiated the development of an EPR/PTFE dosimetry system. This effort consists of three stages: 1) Identification of PTFE materials that may be suitable for dosimetry applications. It was speculated that the inconsistency of EPR signatures in the earlier samples may have been due to variability in PTFE manufacturing processes. 2) Characterization of dosimetry in

  3. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, William Jonathan [Idaho National Laboratory; Barrett, Kristine Eloise [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory

    2015-09-01

    The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

  4. 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) with...

  5. Development of a flight test maneuver autopilot for an F-15 aircraft

    Science.gov (United States)

    Alag, G. S.; Duke, E. L.

    1985-01-01

    An autopilot can be used to provide precise control to meet the demanding requirements of flight research maneuvers with high-performance aircraft. This paper presents the development of control laws for a flight test maneuver autopilot for an F-15 aircraft. A linear quadratic regulator approach is used to develop the control laws within the context of flight test maneuver requirements by treating the maneuver as a finite time tracking problem with regulation of state rates. Results are presented to show the effectiveness of the controller in insuring acceptable aircraft performance during a maneuver.

  6. Meso-scale modeling of irradiated concrete in test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giorla, A. [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Vaitová, M. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic); Le Pape, Y., E-mail: lepapeym@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Štemberk, P. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic)

    2015-12-15

    Highlights: • A meso-scale finite element model for irradiated concrete is developed. • Neutron radiation-induced volumetric expansion is a predominant degradation mode. • Confrontation with expansion and damage obtained from experiments is successful. • Effects of paste shrinkage, creep and ductility are discussed. - Abstract: A numerical model accounting for the effects of neutron irradiation on concrete at the mesoscale is detailed in this paper. Irradiation experiments in test reactor (Elleuch et al., 1972), i.e., in accelerated conditions, are simulated. Concrete is considered as a two-phase material made of elastic inclusions (aggregate) subjected to thermal and irradiation-induced swelling and embedded in a cementitious matrix subjected to shrinkage and thermal expansion. The role of the hardened cement paste in the post-peak regime (brittle-ductile transition with decreasing loading rate), and creep effects are investigated. Radiation-induced volumetric expansion (RIVE) of the aggregate cause the development and propagation of damage around the aggregate which further develops in bridging cracks across the hardened cement paste between the individual aggregate particles. The development of damage is aggravated when shrinkage occurs simultaneously with RIVE during the irradiation experiment. The post-irradiation expansion derived from the simulation is well correlated with the experimental data and, the obtained damage levels are fully consistent with previous estimations based on a micromechanical interpretation of the experimental post-irradiation elastic properties (Le Pape et al., 2015). The proposed modeling opens new perspectives for the interpretation of test reactor experiments in regards to the actual operation of light water reactors.

  7. Criticality and shielding calculations for containers in dry of spent fuel of TRIGA Mark III reactor of ININ; Calculos de criticidad y blindaje para contenedores en seco de combustible gastado del reactor Triga Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Barranco R, F.

    2015-07-01

    In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated

  8. 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...

  9. 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...

  10. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  11. Creating a Test Validated Structural Dynamic Finite Element Model of the X-56A Aircraft

    Science.gov (United States)

    Pak, Chan-Gi; Truong, Samson

    2014-01-01

    Small modeling errors in the finite element model will eventually induce errors in the structural flexibility and mass, thus propagating into unpredictable errors in the unsteady aerodynamics and the control law design. One of the primary objectives of the Multi Utility Technology Test-bed, X-56A aircraft, is the flight demonstration of active flutter suppression, and therefore in this study, the identification of the primary and secondary modes for the structural model tuning based on the flutter analysis of the X-56A aircraft. The ground vibration test-validated structural dynamic finite element model of the X-56A aircraft is created in this study. The structural dynamic finite element model of the X-56A aircraft is improved using a model tuning tool. In this study, two different weight configurations of the X-56A aircraft have been improved in a single optimization run. Frequency and the cross-orthogonality (mode shape) matrix were the primary focus for improvement, while other properties such as center of gravity location, total weight, and offdiagonal terms of the mass orthogonality matrix were used as constraints. The end result was a more improved and desirable structural dynamic finite element model configuration for the X-56A aircraft. Improved frequencies and mode shapes in this study increased average flutter speeds of the X-56A aircraft by 7.6% compared to the baseline model.

  12. Development of control laws for a flight test maneuver autopilot for an F-15 aircraft

    Science.gov (United States)

    Alag, G. S.; Duke, E. L.

    1985-01-01

    An autopilot can be used to provide precise control to meet the demanding requirements of flight research maneuvers with high-performance aircraft. The development of control laws within the context of flight test maneuver requirements is discussed. The control laws are developed using eigensystem assignment and command generator tracking. The eigenvalues and eigenvectors are chosen to provide the necessary handling qualities, while the command generator tracking enables the tracking of a specified state during the maneuver. The effectiveness of the control laws is illustrated by their application to an F-15 aircraft to ensure acceptable aircraft performance during a maneuver.

  13. VISTA : thermal-hydraulic integral test facility for SMART reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, K. Y.; Park, H. S.; Cho, S.; Park, C. K.; Lee, S. J.; Song, C. H.; Chung, M. K. [KAERI, Taejon (Korea, Republic of)

    2003-07-01

    Preliminary performance tests were carried out using the thermal-hydraulic integral test facility, VISTA (Experimental Verification by Integral Simulation of Transients and Accidents), which has been constructed to simulate the SMART-P. The VISTA facility is an integral test facility including the primary and secondary systems as well as safety-related Passive Residual Heat Removal (PRHR) systems. Its scaled ratio with respect to the SMART-P is 1/1 in height and 1/96 in volume and heater power. Several steady states and power changing tests have been carried out to verify the overall thermal hydraulic primary and secondary characteristics in the range of 10% to 100% power operation. As for the preliminary results, the steady state conditions were found to coincide with the expected design values of the SMART-P. But the major thermal hydraulic parameters are greatly affected by the initial water level and the nitrogen pressure in the reactor's upper annular cavity. The power step/ramp changing tests are successfully carried out and the system responses are observed. The primary natural circulation operation is achieved, but advanced control logics need to be developed to reach the natural circulation mode without pressure excursion. In the PRHR transient tests, the natural circulation flow rate through the PRHR system was found to be about 10 percent in the early phases of PRHR operation.

  14. Development and flight test of an experimental maneuver autopilot for a highly maneuverable aircraft

    Science.gov (United States)

    Duke, Eugene L.; Jones, Frank P.; Roncoli, Ralph B.

    1986-01-01

    This report presents the development of an experimental flight test maneuver autopilot (FTMAP) for a highly maneuverable aircraft. The essence of this technique is the application of an autopilot to provide precise control during required flight test maneuvers. This newly developed flight test technique is being applied at the Dryden Flight Research Facility of NASA Ames Research Center. The FTMAP is designed to increase the quantity and quality of data obtained in test flight. The technique was developed and demonstrated on the highly maneuverable aircraft technology (HiMAT) vehicle. This report describes the HiMAT vehicle systems, maneuver requirements, FTMAP development process, and flight results.

  15. Development of a flight test maneuver autopilot for a highly maneuverable aircraft

    Science.gov (United States)

    Duke, E. L.; Jones, F. P.; Roncoli, R. B.

    1983-01-01

    This paper details the development of a flight test maneuver autopilot for a highly maneuverable aircraft. This newly developed flight test technique is being applied at the Dryden Flight Research Facility of the NASA Ames Research Center. The flight test maneuver autopilot (FTMAP) is designed to increase the quantity and quality of the data obtained in flight test. The vehicle with which it is being used is the highly maneuverable aircraft technology (HiMAT) vehicle. This paper describes the HiMAT vehicle systems, maneuver requirements, FTMAP development process, and flight results.

  16. Creating a Test-Validated Finite-Element Model of the X-56A Aircraft Structure

    Science.gov (United States)

    Pak, Chan-Gi; Truong, Samson

    2014-01-01

    Small modeling errors in a finite-element model will eventually induce errors in the structural flexibility and mass, thus propagating into unpredictable errors in the unsteady aerodynamics and the control law design. One of the primary objectives of the X-56A Multi-Utility Technology Testbed aircraft is the flight demonstration of active flutter suppression and, therefore, in this study, the identification of the primary and secondary modes for the structural model tuning based on the flutter analysis of the X-56A aircraft. The ground-vibration test-validated structural dynamic finite-element model of the X-56A aircraft is created in this study. The structural dynamic finite-element model of the X-56A aircraft is improved using a model-tuning tool. In this study, two different weight configurations of the X-56A aircraft have been improved in a single optimization run. Frequency and the cross-orthogonality (mode shape) matrix were the primary focus for improvement, whereas other properties such as c.g. location, total weight, and off-diagonal terms of the mass orthogonality matrix were used as constraints. The end result was an improved structural dynamic finite-element model configuration for the X-56A aircraft. Improved frequencies and mode shapes in this study increased average flutter speeds of the X-56A aircraft by 7.6% compared to the baseline model.

  17. 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.

  18. Development and verification test of integral reactor major components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. I.; Kim, Y. W.; Kim, J. H. and others

    1999-03-01

    The conceptual designs for SG, MCP, CEDM to be installed in the integral reactor SMART were developed. Three-dimensional CAD models for the major components were developed to visualize the design concepts. Once-through helical steam generator was conceptually designed for SMART. Canned motor pump was adopted in the conceptual design of MCP. Linear pulse motor type and ballscrew type CEDM, which have fine control capabilities were studied for adoption in SMART. In parallel with the structural design, the electro-magnetic design was performed for the sizing motors and electro-magnet. Prototypes for the CEDM and MCP sub-assemblies were developed and tested to verify the performance. The impeller design procedure and the computer program to analyze the dynamic characteristics of MCP rotor shaft were developed. The design concepts of SG, MCP, CEDM were also invetigated for the fabricability.

  19. Natural radioactive materials at the Arco Reactor Test Site

    Energy Technology Data Exchange (ETDEWEB)

    Singlevich, W; Healy, J W; Paas, H J; Carey, Z E

    1951-05-28

    At the request of the Division of Biology and Medicine of the AEC, the Biophysics Section of the Radiological Sciences Department at Hanford undertook the task of conducting a background survey for naturally occurring radioactive materials in the environs of the Arco Reactor Test Site in Central Idaho. This survey was part of an overall study which included meteorological measurements by the the Air Weather Service, Geological Studies by the USGS, and an ecological survey of plants and animals by members of the Idaho State College at Pocatello. In general, the measurements at Arco followed the pattern established for environmental monitoring at the Hanford Site with some additional measurements made for natural isotopes not normally of concern at Hanford. A number of analysis included materials such as plutonium and I-131 which were carried out for the purpose of establishing analytical backgrounds for the procedures used. 20 refs., 13 figs., 11 tabs.

  20. Recent results on the RIA test in IGR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L. [Nuclear Safety Institute, Moscow (Russian Federation)

    1997-01-01

    At the 23d WRSM meeting the data base characterizing results of VVER high burnup fuel rods tests under reactivity-initiated accident (RIA) conditions was presented. Comparison of PWR and VVER failure thresholds was given also. Additional analysis of the obtained results was being carried out during 1996. The results of analysis show that the two different failure mechanisms were observed for PWR and VVER fuel rods. Some factors which can be as the possible reasons of these differences are presented. First of them is the state of preirradiated cladding. Published test data for PWR high burnup fuel rods demonstrated that the PWR high burnup fuel rods failed at the RIA test are characterized by very high level of oxidation and hydriding for the claddings. Corresponding researches were performed at Institute of Atomic Reactors (RLAR, Dimitrovgrad, Russia) for large set of VVER high burnup fuel rods. Results of these investigations show that preirradiated commercial Zr-1%Nb claddings practically keep their initial levels of oxidation and H{sub 2} concentration. Consequently the VVER preirradiated cladding must keep the high level of mechanical properties. The second reason leading to differences between failure mechanisms for two types of high burnup fuel rods can be the test conditions. Now such kind of analysis have been performed by two methods.

  1. Aircraft Integration and Flight Testing of 4STAR

    Energy Technology Data Exchange (ETDEWEB)

    Flynn, CJ; Kassianov, E; Russell, P; Redemann, J; Dunagan, S; Holben, B

    2012-10-12

    Under funding from the U.S. Dept. of Energy, in conjunction with a funded NASA 2008 ROSES proposal, with internal support from Battelle Pacific Northwest Division (PNWD), and in collaboration with NASA Ames Research Center, we successfully integrated the Spectrometer for Sky-Scanning, Sun-Tracking Atmospheric Research (4STAR-Air) instrument for flight operation aboard Battelle’s G-1 aircraft and conducted a series of airborne and ground-based intensive measurement campaigns (hereafter referred to as “intensives”) for the purpose of maturing the initial 4STAR-Ground prototype to a flight-ready science-ready configuration.

  2. Deterministic Modeling of the High Temperature Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, J.; Cogliati, J. J.; Pope, M. A.; Ferrer, R. M.; Ougouag, A. M.

    2010-06-01

    Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine INL’s current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 column thin annular core, and the fully loaded core critical condition with 30 columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is used in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal z full core solver used in this study and is based on the Green’s Function solution of the transverse integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, provide benchmarking capability for the DRAGON and the nodal diffusion solver codes. The results from this study show a consistent bias of 2–3% for the core multiplication factor. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement stems from the fact that during the experiments the

  3. Testing of an advanced thermochemical conversion reactor system

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective of this project was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions. 6 refs., 59 figs., 29 tabs.

  4. Miniaturized Charpy test for reactor pressure vessel embrittlement characterization

    Energy Technology Data Exchange (ETDEWEB)

    Manahan, M.P. Sr. [MPM Research and Consulting, Lemont, PA (United States)

    1999-10-01

    Modifications were made to a conventional Charpy machine to accommodate the miniaturized Charpy V-Notch (MCVN) specimens which were fabricated from an archived reactor pressure vessel (RPV) steel. Over 100 dynamic MCVN tests were performed and compared to the results from conventional Charpy V-Notch (CVN) tests to demonstrate the efficacy of the miniature specimen test. The optimized sidegrooved MCVN specimens exhibit transitional fracture behavior over essentially the same temperature range as the CVN specimens which indicates that the stress fields in the MCVN specimens reasonably simulate those of the CVN specimens and this fact has been observed in finite element calculations. This result demonstrates a significant breakthrough since it is now possible to measure the ductile-brittle transition temperature (DBTT) using miniature specimens with only small correction factors, and for some materials as in the present study, without the need for any correction factor at all. This development simplifies data interpretation and will facilitate future regulatory acceptance. The non-sidegrooved specimens yield energy-temperature data which is significantly shifted downward in temperature (non-conservative) as a result of the loss of constraint which accompanies size reduction.

  5. High Temperature Gas-Cooled Test Reactor Point Design: Summary 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); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    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 uranium oxycarbide 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 technology readiness level, licensing approach, and costs of the test reactor point design.

  6. High Temperature Gas-Cooled Test Reactor Point Design: Summary 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); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    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 uranium oxycarbide 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 technology readiness level, licensing approach, and costs of the test reactor point design.

  7. Test method for shielding effects of clothing fabrics by shield box method%服装用织物屏蔽效能测试方法—屏蔽箱法的探讨

    Institute of Scientific and Technical Information of China (English)

    刘志才; 何鹏; 徐勤; 王利君; 李妮; 熊杰

    2013-01-01

    基于电磁波的屏蔽原理,采用一种新的测试织物电磁屏蔽效能的方法-屏蔽箱法,对增重率不同的化学镀银聚酰胺6织物进行测试.结果表明:发射源与人的距离可以用发射源与屏蔽材料的距离来模拟;织物屏蔽效能随着镀层增重率的增加而提高,提高到一定程度后屏蔽效能值趋于稳定;材料性能的变化可以用屏蔽箱法测试的数据直接反映;这种测试织物屏蔽效能的方法经济有效,且测试结果具有较好的重复性.%Based on the theory of shielding against electromagnetic radiation,a new device by shield box was designed to test electromagnetic shielding effects of polyamide 6 (PA6) fabrics that had been treated by electroless plating with silver by varying the weighting percentage.The results indicated that the distance between the radiation source and person could be simulated by the distance between the radiation source and the shielding material.The fabric shielding effects increased with the increase of the weighting percentage,but when to a certain extent they tended to be stable,which meant that the test data obtained from the shield box can directly reflect the change of fabric performance.This shield box method is costeffective and has good reproducibility.

  8. CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.

    Energy Technology Data Exchange (ETDEWEB)

    Pytel, K.; Mieleszczenko, W.; Lechniak, J.; Moldysz, A.; Andrzejewski, K.; Kulikowska, T.; Marcinkowska, A.; Garner, P. L.; Hanan, N. A.; Nuclear Engineering Division; Institute of Atomic Energy (Poland)

    2010-03-01

    The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

  9. Design considerations of the irradiation test vehicle for the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    An irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) is being jointly developed by the Lockheed Martin Idaho Technologies Company (LMIT) and the U.S. Fusion Program. The vehicle is intended for neutron irradiation testing of candidate structural materials, including vanadium-based alloys, silicon carbide composites, and low activation steels. It could possibly be used for U.S./Japanese collaboration in the Jupiter Program. The first test train is scheduled to be completed by September 1998. In this report, we present the functional requirements for the vehicle and a preliminary design that satisfies these requirements.

  10. 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.

  11. Aircraft control surface failure detection and isolation using the OSGLR test. [orthogonal series generalized likelihood ratio

    Science.gov (United States)

    Bonnice, W. F.; Motyka, P.; Wagner, E.; Hall, S. R.

    1986-01-01

    The performance of the orthogonal series generalized likelihood ratio (OSGLR) test in detecting and isolating commercial aircraft control surface and actuator failures is evaluated. A modification to incorporate age-weighting which significantly reduces the sensitivity of the algorithm to modeling errors is presented. The steady-state implementation of the algorithm based on a single linear model valid for a cruise flight condition is tested using a nonlinear aircraft simulation. A number of off-nominal no-failure flight conditions including maneuvers, nonzero flap deflections, different turbulence levels and steady winds were tested. Based on the no-failure decision functions produced by off-nominal flight conditions, the failure detection and isolation performance at the nominal flight condition was determined. The extension of the algorithm to a wider flight envelope by scheduling on dynamic pressure and flap deflection is examined. Based on this testing, the OSGLR algorithm should be capable of detecting control surface failures that would affect the safe operation of a commercial aircraft. Isolation may be difficult if there are several surfaces which produce similar effects on the aircraft. Extending the algorithm over the entire operating envelope of a commercial aircraft appears feasible.

  12. 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. EPA actively participated in the United Nation's International Civil Aviation Organization (ICAO... Regulation of Aircraft Engine Emissions E. Brief History of ICAO Regulation of Aircraft Engine Emissions II... under consideration by the United Nation's International Civil Aviation Organization (ICAO)....

  13. Frequency-domain identification of aircraft structural modes from short-duration flight tests

    Science.gov (United States)

    Vayssettes, J.; Mercère, G.; Vacher, P.; De Callafon, R. A.

    2014-07-01

    This article presents identification algorithms dedicated to the modal analysis of civil aircraft structures during in-flight flutter tests. This particular operational framework implies several specifications for the identification procedure. To comply with these requirements, the identification problem is formulated in the frequency domain as an output-error problem. Iterative identification methods based on structured matrix fraction descriptions are used to solve this problem and to identify a continuous-time model. These iterative methods are specifically designed to deal with experiments where short-duration tests with multiple-input excitations are used. These algorithms are first discussed and then evaluated through a simulation example illustrative of the in-flight modal analysis of a civil aircraft. Based on these evaluation results, an efficient iterative algorithm is suggested and applied to real flight-test data measured on board a military aircraft.

  14. NASA Reactor Facility Hazards Summary. Volume 1

    Science.gov (United States)

    1959-01-01

    The Lewis Research Center of the National Aeronautics and Space Administration proposes to build a nuclear research reactor which will be located in the Plum Brook Ordnance Works near Sandusky, Ohio. The purpose of this report is to inform the Advisory Committee on Reactor Safeguards of the U. S. Atomic Energy Commission in regard to the design Lq of the reactor facility, the characteristics of the site, and the hazards of operation at this location. The purpose of this research reactor is to make pumped loop studies of aircraft reactor fuel elements and other reactor components, radiation effects studies on aircraft reactor materials and equipment, shielding studies, and nuclear and solid state physics experiments. The reactor is light water cooled and moderated of the MTR-type with a primary beryllium reflector and a secondary water reflector. The core initially will be a 3 by 9 array of MTR-type fuel elements and is designed for operation up to a power of 60 megawatts. The reactor facility is described in general terms. This is followed by a discussion of the nuclear characteristics and performance of the reactor. Then details of the reactor control system are discussed. A summary of the site characteristics is then presented followed by a discussion of the larger type of experiments which may eventually be operated in this facility. The considerations for normal operation are concluded with a proposed method of handling fuel elements and radioactive wastes. The potential hazards involved with failures or malfunctions of this facility are considered in some detail. These are examined first from the standpoint of preventing them or minimizing their effects and second from the standpoint of what effect they might have on the reactor facility staff and the surrounding population. The most essential feature of the design for location at the proposed site is containment of the maximum credible accident.

  15. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Bradley K. Heath

    2014-03-01

    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

  16. Multi-Purpose Thermal Hydraulic Loop: Advanced Reactor Technology Integral System Test (ARTIST) Facility for Support of Advanced Reactor Technologies

    Energy Technology Data Exchange (ETDEWEB)

    James E. O' Brien; Piyush Sabharwall; SuJong Yoon

    2001-11-01

    Effective and robust high temperature heat transfer systems are fundamental to the successful deployment of advanced reactors for both power generation and non-electric applications. Plant designs often include an intermediate heat transfer loop (IHTL) with heat exchangers at either end to deliver thermal energy to the application while providing isolation of the primary reactor system. In order to address technical feasibility concerns and challenges a new high-temperature multi-fluid, multi-loop test facility “Advanced Reactor Technology Integral System Test facility” (ARTIST) is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water. Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed

  17. A test technique for measuring lightning-induced voltages on aircraft electrical circuits

    Science.gov (United States)

    Walko, L. C.

    1974-01-01

    The development of a test technique used for the measurement of lightning-induced voltages in the electrical circuits of a complete aircraft is described. The resultant technique utilizes a portable device known as a transient analyzer capable of generating unidirectional current impulses similar to lightning current surges, but at a lower current level. A linear relationship between the magnitude of lightning current and the magnitude of induced voltage permitted the scaling up of measured induced values to full threat levels. The test technique was found to be practical when used on a complete aircraft.

  18. Sipping test update device for fuel elements cladding inspections in IPR-r1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, R.R.; Mesquita, A.Z.; Andrade, E.P.D.; Gual, Maritza R., E-mail: rrr@cdtn.br, E-mail: amir@cdtn.br, E-mail: edson@cdtn.br, E-mail: maritzargual@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    It is in progress at the Centro de Desenvolvimento da Tecnologia Nuclear - CDTN (Nuclear Technology Development Center), a research project that aims to investigate possible leaks in the fuel elements of the TRIGA reactor, located in this research center. This paper presents the final form of sipping test device for TRIGA reactor, and results of the first experiments setup. Mechanical support strength tests were made by knotting device on the crane, charged with water from the conventional water supply, and tests outside the reactor pool with the use of new non-irradiated fuel elements encapsulated in stainless steel, and available safe stored in this unit. It is expected that tests with graphite elements from reactor pool are done soon after and also the test experiment with the first fuel elements in service positioned in the B ring (central ring) of the reactor core in the coming months. (author)

  19. 小型屏蔽体屏蔽效能测试系统全向辐射源设计%The design of omnidirectional radiated source for small shield shielding effectiveness tests

    Institute of Scientific and Technical Information of China (English)

    江岩; 钱时祥

    2013-01-01

    针对较小体积的屏蔽体,按照内置辐射法测量屏蔽效能,需将信号源和天线内置到屏蔽体内部.为了将信号源和天线集组合成辐射源,天线的尺寸与信号源的体积是辐射源设计的重要因素.介绍一种基于振荡源激励梳状波发生器获得宽频带的微波信号,结合盘锥天线,实现1~10 GHz频率范围内多频点、高稳定、全方位特性小型辐射源的设计.在屏蔽效能测试系统中,该辐射源作为发射部分,可实现小型屏蔽体的屏蔽效能测量.%For the small size shield, the test of shielding effectiveness in built-in radiation mode needs to set the signal source and the antenna into the shield. In order to combine the source and antenna into the radiation source,the size of the antenna and the volume of the signal source is the important factor in the design of radiant source. This paper introduces a design of small radiant source which based on the oscillator incentive comb generator for the wideband signal and combined with the plate cone antenna,realize such characteristic as the multiple frequency point,high stability and multimode in the frequency range of l~10 GHz. In the shielding effectiveness test system,the radiation source as the emission part, the system can realize the shielding effectiveness test of small shield.

  20. Instrumentation and data acquisition for full-scale aircraft crash testing

    Science.gov (United States)

    Jones, Lisa E.; Fasanella, Edwin L.

    1993-01-01

    The Landing and Impact Dynamics Branch of the NASA Langley Research Center has been conducting full-scale aircraft crash tests since the 1970s. Using a pendulum method, aircraft are suspended by cables from a 240-ft high gantry and swung into the impact surface at various attitudes and velocities. Instrumentation for these tests include on-board high-speed cameras, strain gages, load cells, displacement transducers, and accelerometers. Transducers in the aircraft are hard-wired through a long umbilical cable to the data acquisition room. Up to 96 channels of data can be collected at a typical rate of 4000 samples per second. Data acquisition using an FM multiplexed analog system and a high-speed personal computer based digital system is described.

  1. Development of automatic Ultrasonic testing equipment for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kor R.; Kim, Jae H.; Lee, Jae C.

    1996-06-01

    The selected weld areas of a reactor pressure vessel and adjacent piping are examined by the remote mechanized ultrasonic testing (MUT) equipment. Since the MUT equipment was purchased from southwest Research Institute (SwRI) in April 1985, 15 inservice inspections and 5 preservice inspections are performed with this MUT equipment. However due to the old age of the equipment and frequent movements to plant sites, the reliability of examination was recently decreased rapidly and it is very difficult to keep spare parts. In order to resolve these problems and to meet the strong request from plant sites, we intend to develop a new 3-axis control system including hardware and software. With this control system, we expect more efficient and reliable examination of the nozzle to shell weld areas, which is specified in ASME Code Section XI. The new 3-axis control system hardware and software were designed and development of our own control system, the advanced technologies of computer control mechanism were established and examination reliability of the nozzle to shell weld area was improved. With the development of our 3-axis control system for PaR ISI-2 computer control system, the reliability of nozzle to shell weld area examination has been improved. The established technologies from the development and detailed analysis of existing control system, are expected to be applied to the similar control systems in nuclear power plants. (author). 12 refs., 4 tabs., 33 figs.

  2. Shields-1, A SmallSat Radiation Shielding Technology Demonstration

    Science.gov (United States)

    Thomsen, D. Laurence, III; Kim, Wousik; Cutler, James W.

    2015-01-01

    The NASA Langley Research Center Shields CubeSat initiative is to develop a configurable platform that would allow lower cost access to Space for materials durability experiments, and to foster a pathway for both emerging and commercial-off-the-shelf (COTS) radiation shielding technologies to gain spaceflight heritage in a relevant environment. The Shields-1 will be Langleys' first CubeSat platform to carry out this mission. Radiation shielding tests on Shields-1 are planned for the expected severe radiation environment in a geotransfer orbit (GTO), where advertised commercial rideshare opportunities and CubeSat missions exist, such as Exploration Mission 1 (EM-1). To meet this objective, atomic number (Z) graded radiation shields (Zshields) have been developed. The Z-shield properties have been estimated, using the Space Environment Information System (SPENVIS) radiation shielding computational modeling, to have 30% increased shielding effectiveness of electrons, at half the thickness of a corresponding single layer of aluminum. The Shields-1 research payload will be made with the Z-graded radiation shields of varying thicknesses to create dose-depth curves to be compared with baseline materials. Additionally, Shields-1 demonstrates an engineered Z-grade radiation shielding vault protecting the systems' electronic boards. The radiation shielding materials' performances will be characterized using total ionizing dose sensors. Completion of these experiments is expected to raise the technology readiness levels (TRLs) of the tested atomic number (Z) graded materials. The most significant contribution of the Z-shields for the SmallSat community will be that it enables cost effective shielding for small satellite systems, with significant volume constraints, while increasing the operational lifetime of ionizing radiation sensitive components. These results are anticipated to increase the development of CubeSat hardware design for increased mission lifetimes, and enable

  3. 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

  4. 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.

  5. 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.

  6. Reactor coolant pump testing using motor current signatures analysis

    Energy Technology Data Exchange (ETDEWEB)

    Burstein, N.; Bellamy, J.

    1996-12-01

    This paper describes reactor coolant pump motor testing carried out at Florida Power Corporation`s Crystal River plant using Framatome Technologies` new EMPATH (Electric Motor Performance Analysis and Trending Hardware) system. EMPATH{trademark} uses an improved form of Motor Current Signature Analysis (MCSA), technology, originally developed at Oak Ridge National Laboratories, for detecting deterioration in the rotors of AC induction motors. Motor Current Signature Analysis (MCSA) is a monitoring tool for motor driven equipment that provides a non-intrusive means for detecting the presence of mechanical and electrical abnormalities in the motor and the driven equipment. The base technology was developed at the Oak Ridge National Laboratory as a means for determining the affects of aging and service wear specifically on motor-operated valves used in nuclear power plant safety systems, but it is applicable to a broad range of electric machinery. MCSA is based on the recognition that an electric motor (ac or dc) driving a mechanical load acts as an efficient and permanently available transducer by sensing mechanical load variations, large and small, long-term and rapid, and converting them into variations in the induced current generated in the motor windings. The motor current variations, resulting from changes in load caused by gears, pulleys, friction, bearings, and other conditions that may change over the life of the motor, are carried by the electrical cables powering the motor and are extracted at any convenient location along the motor lead. These variations modulate the 60 Hz carrier frequency and appear as sidebands in the spectral plot.

  7. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1981-08-01

    Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies.

  8. Unique Aspects of Flight Testing Unmanned Aircraft Systems

    Science.gov (United States)

    2010-04-01

    High Altitude Endurance HCI Human Computer Interface INS Inertial Navigation System IR Infrared JITC Joint Interoperability Testing Command...highlighting some of the unique aspects of flight testing unmanned air vehicle systems. It is intended to provide a practical set of guidelines in support of...of unmanned aviation systems, it is especially important that even minor changes to the baseline code be carefully reviewed, and that regression

  9. Adaptation of Crack Growth Detection Techniques to US Material Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis; Joy L. Rempe; Gordon Kohse; Yakov Ostrovsky; David M. Carpenter

    2014-04-01

    A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some materials testing reactors (MTRs) outside the U.S., such as the Halden Boiling Water Reactor (HBWR), have deployed a technique to measure crack growth propagation during irradiation. This technique incorporates a compact loading mechanism to stress the specimen during irradiation. A crack in the specimen is monitored using the Direct Current Potential Drop (DCPD) method. A project is underway to develop and demonstrate the performance of a similar type of test rig for use in U.S. MTRs. The first year of this three year project was devoted to designing, analyzing, fabricating, and bench top testing a mechanism capable of applying a controlled stress to specimens while they are irradiated in a pressurized water loop (simulating PWR reactor conditions). During the second year, the mechanism will be tested in autoclaves containing high pressure, high temperature water with representative water chemistries. In addition, necessary documentation and safety reviews for testing in a reactor environment will be completed. In the third year, the assembly will be tested in the Massachusetts Institute of Technology Reactor (MITR) and Post Irradiation Examinations (PIE) will be performed.

  10. Space Shielding Materials for Prometheus Application

    Energy Technology Data Exchange (ETDEWEB)

    R. Lewis

    2006-01-20

    At the time of Prometheus program restructuring, shield material and design screening efforts had progressed to the point where a down-selection from approximately eighty-eight materials to a set of five ''primary'' materials was in process. The primary materials were beryllium (Be), boron carbide (B{sub 4}C), tungsten (W), lithium hydride (LiH), and water (H{sub 2}O). The primary materials were judged to be sufficient to design a Prometheus shield--excluding structural and insulating materials, that had not been studied in detail. The foremost preconceptual shield concepts included: (1) a Be/B{sub 4}C/W/LiH shield; (2) a Be/B{sub 4}C/W shield; (3) and a Be/B{sub 4}C/H{sub 2}O shield. Since the shield design and materials studies were still preliminary, alternative materials (e.g., {sup nal}B or {sup 10}B metal) were still being screened, but at a low level of effort. Two competing low mass neutron shielding materials are included in the primary materials due to significant materials uncertainties in both. For LiH, irradiation-induced swelling was the key issue, whereas for H{sub 2}O, containment corrosion without active chemistry control was key, Although detailed design studies are required to accurately estimate the mass of shields based on either hydrogenous material, both are expected to be similar in mass, and lower mass than virtually any alternative. Unlike Be, W, and B{sub 4}C, which are not expected to have restrictive temperature limits, shield temperature limits and design accommodations are likely to be needed for either LiH or H{sub 2}O. The NRPCT focused efforts on understanding swelting of LiH, and observed, from approximately fifty prior irradiation tests, that either casting ar thorough out-gassing should reduce swelling. A potential contributor to LiH swelling appears to be LiOH contamination due to exposure to humid air, that can be eliminated by careful processing. To better understand LiH irradiation performance and

  11. Propeller aircraft interior noise model. II - Scale-model and flight-test comparisons

    Science.gov (United States)

    Willis, C. M.; Mayes, W. H.

    1987-01-01

    A program for predicting the sound levels inside propeller driven aircraft arising from sidewall transmission of airborne exterior noise is validated through comparisons of predictions with both scale-model test results and measurements obtained in flight tests on a turboprop aircraft. The program produced unbiased predictions for the case of the scale-model tests, with a standard deviation of errors of about 4 dB. For the case of the flight tests, the predictions revealed a bias of 2.62-4.28 dB (depending upon whether or not the data for the fourth harmonic were included) and the standard deviation of the errors ranged between 2.43 and 4.12 dB. The analytical model is shown to be capable of taking changes in the flight environment into account.

  12. The performance test of anti-scattering x-ray grid with inclined shielding material by MCNP code simulation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Jun Woo; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2016-06-15

    The scattered photons cause reduction of the contrast of radiographic image and it results in the degradation of the quality of the image. In order to acquire better quality image, an anti-scattering x-ray gird should be equipped in radiography system. The X-ray anti-scattering grid of the inclined type based on the hybrid concept for that of parallel and focused type was tested by MCNP code. The MCNPX 2.7.0 was used for the simulation based test. The geometry for the test was based on the IEC 60627 which was an international standard for diagnostic X-ray imaging equipment-Characteristics of general purpose and mammographic anti-scatter grids. The performance of grids with four inclined shielding material types was compared with that of the parallel type. The grid with completely tapered type the best performance where there were little performance difference according to the degree of inclination.

  13. Development of NVG test maneuvers for civilian aircraft

    Science.gov (United States)

    Jennings, Sion; Craig, Greg; Carignan, Stephan; Fischer, Heiko; Brulotte, Michel

    2008-04-01

    This document provides an overview of helicopter flight-test methods used to evaluate night vision goggles at the National Research Council of Canada's Institute for Aerospace Research. These techniques have been used to examine the performance of display systems in actual field conditions. The flight evaluations were based, in large part, on standard flight test maneuvers and rating systems outlined in Aeronautical Design Standard ADS-33. The document describes NVG test maneuvers developed from ADS-33 principles, including a high hover, a mirror C, a vertical descent, a parallel lateral translation, a turn about the tail, a confined area staged landing, a brown-out/white-out simulation and a lit pirouette. The overview also comprises a description of methods for controlling the cueing environment. These methods include an appropriate selection of maneuvers as well as devices for limiting pilot vision such as goggles with filters and apertures, and other devices. The paper concludes with a short discussion on the merits of developing accurate in-flight tests capable of resolving performance differences among displays.

  14. Climatic Test of Aircraft%飞机气候试验

    Institute of Scientific and Technical Information of China (English)

    唐虎; 李喜明

    2012-01-01

    Modern war requires that aircraft should have the ability of all-weather combat. For validating the climatic environment worthiness design and safety of aircraft and attachment, climatic test needs to be done before flight test in complex weather conditions. Recent development in climatic test at home and abroad was introduced. The purpose, importance, test contents, test management, and working process of aircraft climatic test were discussed.%现代战争要求飞机具有全天候作战能力.飞机气候试验是在进行复杂气象条件试飞前需要完成的气候环境试验,用以验证飞机和地面保障设备的气候环境适应性设计,保证飞行安全.结合国内外气候试验发展情况,介绍了飞机气候试验的目的、重要性、试验内容、试验管理及工作程序等.

  15. Testing a combined radiation protection modality: chemical protector and local shielding.

    Science.gov (United States)

    Minkova, M; Baldzhijska, M

    1989-01-01

    The impact of combined radiation protection upon damage to critical organs-spleen, small intestine, and bone marrow was studied in adult rat males 3 days after whole-body exposure to 9.5 Gy gamma-ray dose. Adeturone, the chemical radioprotector used, was administered intraperitoneally at 1/17 of its LD50 dose. Local shielding of the abdomino-lumbal region was accomplished using a lead ring providing on average 28-30% attenuation of radiation exposure. This degree of physical abdomino-lumbal protection combined with adeturone (50 mg/kg) pretreatment resulted in mutual enhancement of the components' action, expansion of the chemical agent's therapeutic range, providing a combination with improved overall antiradiation properties.

  16. IRRADIATION TESTING OF THE RERTR FUEL MINIPLATES WITH BURNABLE ABSORBERS IN THE ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    I. Glagolenko; D. Wachs; N. Woolstenhulme; G. Chang; B. Rabin; C. Clark; T. Wiencek

    2010-10-01

    Based on the results of the reactor physics assessment, conversion of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) can be potentially accomplished in two ways, by either using U-10Mo monolithic or U-7Mo dispersion type plates in the ATR fuel element. Both designs, however, would require incorporation of the burnable absorber in several plates of the fuel element to compensate for the excess reactivity and to flatten the radial power profile. Several different types of burnable absorbers were considered initially, but only borated compounds, such as B4C, ZrB2 and Al-B alloys, were selected for testing primarily due to the length of the ATR fuel cycle and fuel manufacturing constraints. To assess and compare irradiation performance of the U-Mo fuels with different burnable absorbers we have designed and manufactured 28 RERTR miniplates (20 fueled and 8 non-fueled) containing fore-mentioned borated compounds. These miniplates will be tested in the ATR as part of the RERTR-13 experiment, which is described in this paper. Detailed plate design, compositions and irradiations conditions are discussed.

  17. Study of fast reactor safety test facilities. Preliminary report

    Energy Technology Data Exchange (ETDEWEB)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods. (DG)

  18. 2 MW液态钍基熔盐实验堆主屏蔽温度场分析%Temperature field analysis for the main shielding of the 2-MW thorium-based molten salt experimental reactor

    Institute of Scientific and Technical Information of China (English)

    何杰; 夏晓彬; 蔡军; 潘登; 彭玉; 黄建平; 张国庆

    2016-01-01

    Background:Molten salt reactor is a fourth generation advanced reactor. The concrete wall is the key part of this high-temperature reactor shielding, so temperature field analysis is important. Purpose: This study attempts to calculate the temperature field of the TMSR-LF1 (2-MW liquid-fueled molten salt experimental reactor) shielding, and judge if it meets the design requirements. Methods: In accordance with the problem that MCNP (Monte Carlo N Particle Transport Code) results cannot be directly imported into Fluent, a program which converts MCNP results to the spatial distribution of power density, and imports the spatial distribution of power density into the Fluent in the form of User-Defined Function (UDF) was developed by using Python programming language to realize the coupling of the two. According to TMSR-LF1 design parameters, a one-eight physical and thermal model of the whole reactor is established, using code MCNP and Fluent. Reactor radiation shielding thermal analysis adopts the assumptions that the different environment temperatures are 5°C, 18°C, 25°C, 30°C, 35°C and 40°C, respectively.Results: The maximal values of temperature and temperature gradient in the radiation shielding concrete wall are 67.42 °C and 78.40 °C·m?1, which are lower than limit values.Conclusion: The radiation shielding concrete wall can meet the design requirements.%反应堆主屏蔽是核反应堆的重要组成部分,用来有效降低反应堆运行时屏蔽体外的辐射剂量水平,以满足反应堆部件材料对辐射限制的要求.温度是影响反应堆主屏蔽性能的重要因素.针对2 MWth液态熔盐堆(2-MW liquid-fueled molten salt experimental reactor,TMSR-LF1),采用MCNP软件获得功率分布后,利用Fluent软件对主屏蔽进行温度场计算.计算过程中利用Python语言编写了程序(MCNP to Fluent,MTF)来实现将MCNP(Monte Carlo N Particle Transport Code)计算结果转换为功率密度的空间分布,

  19. Research on Power Ramp Testing Method for PWR Fuel Rod at Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In order to develop high performance fuel assembly for domestic nuclear power plant, it is necessary to master some fundamental test technology. So the research on the power ramp testing methods is proposed. A tentative power ramp test for short PWR fuel rod has been conducted at the heavy water research reactor (HWRR) in China Institute of Atomic Energy (CIAE) in May of 2001. The in-pile test rig was placed into the central channel of the reactor . The test rig consists of pressure pipe assembly, thimble, solid neutron absorbing screen and its driving parts, etc.. The test

  20. Evaluation of neutron streaming and future shielding measurement plan in the prototype FBR Monju

    Energy Technology Data Exchange (ETDEWEB)

    Sasaki, Kenji; Usami, Shin; Deshimaru, Takehide; Nakashima, Fumiaki [Japan Nuclear Cycle Development Institute, Tsuruga, Fukui (Japan)

    2000-03-01

    This paper describes the shielding measurements results in the reactor head access area (RHAA) and the primary heat transport system (PHTS) cells performed during the power raising test periods from February through December, 1995 in the prototype fast breeder reactor Monju. We confirmed that the measured values were well below the shielding design requirements in the RHAA and the PHTS cells and that the countermeasures against neutron streaming to the RHAA and the PHTS cells were very effective. Future measurements in order to obtain basic data for use in future FBR design are planned for the next period of operation. (author)

  1. Propfan test assessment testbed aircraft stability and control/performance 1/9-scale wind tunnel tests

    Science.gov (United States)

    Little, B. H., Jr.; Tomlin, K. H.; Aljabri, A. S.; Mason, C. A.

    1988-01-01

    One-ninth scale wind tunnel model tests of the Propfan Test Assessment (PTA) aircraft were performed in three different NASA facilities. Wing and propfan nacelle static pressures, model forces and moments, and flow field at the propfan plane were measured in these tests. Tests started in June 1985 and were completed in January 1987. These data were needed to assure PTA safety of flight, predict PTA performance, and validate analytical codes that will be used to predict flow fields in which the propfan will operate.

  2. An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

    2009-11-01

    This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

  3. Self Diagnostic Accelerometer Ground Testing on a C-17 Aircraft Engine

    Science.gov (United States)

    Tokars, Roger P.; Lekki, John D.

    2013-01-01

    The self diagnostic accelerometer (SDA) developed by the NASA Glenn Research Center was tested for the first time in an aircraft engine environment as part of the Vehicle Integrated Propulsion Research (VIPR) program. The VIPR program includes testing multiple critical flight sensor technologies. One such sensor, the accelerometer, measures vibrations to detect faults in the engine. In order to rely upon the accelerometer, the health of the accelerometer must be ensured. Sensor system malfunction is a significant contributor to propulsion in flight shutdowns (IFSD) which can lead to aircraft accidents when the issue is compounded with an inappropriate crew response. The development of the SDA is important for both reducing the IFSD rate, and hence reducing the rate at which this component failure type can put an aircraft in jeopardy, and also as a critical enabling technology for future automated malfunction diagnostic systems. The SDA is a sensor system designed to actively determine the accelerometer structural health and attachment condition, in addition to making vibration measurements. The SDA uses a signal conditioning unit that sends an electrical chirp to the accelerometer and recognizes changes in the response due to changes in the accelerometer health and attachment condition. In an effort toward demonstrating the SDAs flight worthiness and robustness, multiple SDAs were mounted and tested on a C-17 aircraft engine. The engine test conditions varied from engine off, to idle, to maximum power. The two SDA attachment conditions used were fully tight and loose. The newly developed SDA health algorithm described herein uses cross correlation pattern recognition to discriminate a healthy from a faulty SDA. The VIPR test results demonstrate for the first time the robustness of the SDA in an engine environment characterized by high vibration levels.

  4. Reactor

    Science.gov (United States)

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  5. 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.

  6. Design and test of aircraft engine isolators for reduced interior noise

    Science.gov (United States)

    Unruh, J. F.; Scheidt, D. C.

    1982-01-01

    Improved engine vibration isolation was proposed to be the most weight and cost efficient retrofit structure-borne noise control measure for single engine general aviation aircraft. A study was carried out the objectives: (1) to develop an engine isolator design specification for reduced interior noise transmission, (2) select/design candidate isolators to meet a 15 dB noise reduction design goal, and (3) carry out a proof of concept evaluation test. Analytical model of the engine, vibration isolators and engine mount structure were coupled to an empirical model of the fuselage for noise transmission evaluation. The model was used to develop engine isolator dynamic properties design specification for reduced noise transmission. Candidate isolators ere chosen from available product literature and retrofit to a test aircraft. A laboratory based test procedure was then developed to simulate engine induced noise transmission in the aircraft for a proof of concept evaluation test. Three candidate isolator configurations were evaluated for reduced structure-borne noise transmission relative to the original equipment isolators.

  7. Fatigue Test of Domestic Manufactured Reactor Pressure Vessel Steel

    Institute of Scientific and Technical Information of China (English)

    ZHONG; Wei-hua; TONG; Zhen-feng; NING; Guang-sheng; YU; Bin-tao

    2013-01-01

    The CAP1400 will be built by our country,after the self-dependent innovation work on the imported technology of AP1000,which is a 3rd generation NPP.Now,the design of CAP1400 key equipment is ongoing,and the fatigue design of the domestic manufactured key equipment,such as reactor pressure vessel(RPV),is found to be a main problem in the design work,as the fatigue data is lacked.Thus the

  8. Aircraft Engine Noise Research and Testing at the NASA Glenn Research Center

    Science.gov (United States)

    Elliott, Dave

    2015-01-01

    The presentation will begin with a brief introduction to the NASA Glenn Research Center as well as an overview of how aircraft engine noise research fits within the organization. Some of the NASA programs and projects with noise content will be covered along with the associated goals of aircraft noise reduction. Topics covered within the noise research being presented will include noise prediction versus experimental results, along with engine fan, jet, and core noise. Details of the acoustic research conducted at NASA Glenn will include the test facilities available, recent test hardware, and data acquisition and analysis methods. Lastly some of the actual noise reduction methods investigated along with their results will be shown.

  9. Lateral noise attenuation of the advanced propeller of the propfan test assessment aircraft

    Science.gov (United States)

    Chambers, F. W.; Reddy, N. N.; Bartel, H. W.

    1989-01-01

    Lateral noise attenuation characteristics of the advanced propeller are determined using the flight test results of the testbed aircraft, Propfan Test Assessment (PTA), with a single, large-scale propfan. The acoustic data were obtained with an array of ground-mounted microphones positioned at distances up to 2.47 km (8100 feet) to the side of the flight path. The aircraft was flown at a Mach number of 0.31 for a variety of operating conditions. The lateral noise attenuation in a frequency range containing the blade passage frequency of the propeller was found to have positive magnitudes on the propfan side and negative magnitudes on the opposite side. The measured attenuation exhibits a strong dependence upon the elevation angle. The results also display a clear dependence upon the angle at which the propeller and nacelle are mounted on the wing (inflow angle).

  10. 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

  11. 10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

    Energy Technology Data Exchange (ETDEWEB)

    Bohachek, Randolph Charles [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-09-01

    The Advanced Test Reactor (ATR; TRA-670), which is located in the ATR Complex at Idaho National Laboratory, was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. While ATR is safely fulfilling current mission requirements, assessments are continuing. These assessments intend to identify areas to provide defense–in-depth and improve safety for ATR. One of the assessments performed by an independent group of nuclear industry experts recommended that a remote accident management capability be provided. The report stated that: “contemporary practice in commercial power reactors is to provide a remote shutdown station or stations to allow shutdown of the reactor and management of long-term cooling of the reactor (i.e., management of reactivity, inventory, and cooling) should the main control room be disabled (e.g., due to a fire in the control room or affecting the control room).” This project will install remote reactor monitoring and management capabilities for ATR. Remote capabilities will allow for post scram reactor management and monitoring in the event the main Reactor Control Room (RCR) must be evacuated.

  12. Impact Analyses and Tests of Metal Cask Considering Aircraft Engine Crash - 12308

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sanghoon; Choi, Woo-Seok; Kim, Ki-Young; Jeon, Je-Eon; Seo, Ki-Seog [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-07-01

    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)

  13. 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.

  14. Anticipated Effectiveness of Active Noise Control in Propeller Aircraft Interiors as Determined by Sound Quality Tests

    Science.gov (United States)

    Powell, Clemans A.; Sullivan, Brenda M.

    2004-01-01

    Two experiments were conducted, using sound quality engineering practices, to determine the subjective effectiveness of hypothetical active noise control systems in a range of propeller aircraft. The two tests differed by the type of judgments made by the subjects: pair comparisons in the first test and numerical category scaling in the second. Although the results of the two tests were in general agreement that the hypothetical active control measures improved the interior noise environments, the pair comparison method appears to be more sensitive to subtle changes in the characteristics of the sounds which are related to passenger preference.

  15. Liquid Metal Cooled Reactor for Space Power

    Science.gov (United States)

    Weitzberg, Abraham

    2003-01-01

    The conceptual design is for a liquid metal (LM) cooled nuclear reactor that would provide heat to a closed Brayton cycle (CBC) power conversion subsystem to provide electricity for electric propulsion thrusters and spacecraft power. The baseline power level is 100 kWe to the user. For long term power generation, UN pin fuel with Nb1Zr alloy cladding was selected. As part of the SP-100 Program this fuel demonstrated lifetime with greater than six atom percent burnup, at temperatures in the range of 1400-1500 K. The CBC subsystem was selected because of the performance and lifetime database from commercial and aircraft applications and from prior NASA and DOE space programs. The high efficiency of the CBC also allows the reactor to operate at relatively low power levels over its 15-year life, minimizing the long-term power density and temperature of the fuel. The scope of this paper is limited to only the nuclear components that provide heated helium-xenon gas to the CBC subsystem. The principal challenge for the LM reactor concept was to design the reactor core, shield and primary heat transport subsystems to meet mission requirements in a low mass configuration. The LM concept design approach was to assemble components from prior programs and, with minimum change, determine if the system met the objective of the study. All of the components are based on technologies having substantial data bases. Nuclear, thermalhydraulic, stress, and shielding analyses were performed using available computer codes. Neutronics issues included maintaining adequate operating and shutdown reactivities, even under accident conditions. Thermalhydraulic and stress analyses calculated fuel and material temperatures, coolant flows and temperatures, and thermal stresses in the fuel pins, components and structures. Using conservative design assumptions and practices, consistent with the detailed design work performed during the SP-100 Program, the mass of the reactor, shield, primary heat

  16. Moderated heat pipe thermionic reactor (MOHTR) module development and test

    Science.gov (United States)

    Merrigan, Michael A.; Trujillo, Vincent L.

    1992-01-01

    The Moderated Heat Pipe Thermionic Reactor (MOHTR) thermionic space reactor design combines the low risk technology associated with the Thermionic Fuel Element (TFE) Verification Program with the high reliability heat transfer capability of liquid metal heat pipe technology. The resulting design concept, capable of implementation over the power range of 10 to 100 kWe, offers efficiency and reliability with reduced risk of single point failures. The union of TFE and heat pipe technology is achieved by imbedding TFEs and heat pipes in a beryllium matrix to which they are thermally coupled by brazing or by liquid metal (NaK or Na) bonding. The reactor employs an array of TFE modules, each comprising a TFE, a zirconium hydride (ZrH) cylinder for neutron moderation, and heat pipes for transport of heat from the collector surface of the TFE to the waste heat radiator. An advantage of the design is the low temperature drop from the collector surface to the radiating surface. This is a result of the elimination of electrical insulation from the heat transport path through electrical isolation of the modules. The module used in this study consisted of a beryllium core, and electrical cartridge heater simulating the TFE, and three heat pipes to dissipate the waste heat. The investigation was focused on the thermal performance of the assembly, including evaluation of the sodium and braze bonding options for minimizing the thermal resistance between the elements, the temperature distribution in the beryllium matrix, and the heat pipe performance. Continuing subjects of the investigation include performance of the heat pipes through start-up transients, during normal operation, and in a single heat pipe failure mode. Secondary objectives of the investigation include correlation of analytic models for the thermionic element and module including the effects of gap thermal conductances at the modules electrically insulated surfaces.

  17. Shielding Effectiveness of Laminated Shields

    Directory of Open Access Journals (Sweden)

    B. P. Rao

    2008-12-01

    Full Text Available Shielding prevents coupling of undesired radiated electromagnetic energy into equipment otherwise susceptible to it. In view of this, some studies on shielding effectiveness of laminated shields with conductors and conductive polymers using plane-wave theory are carried out in this paper. The plane wave shielding effectiveness of new combination of these materials is evaluated as a function of frequency and thickness of material. Conductivity of the polymers, measured in previous investigations by the cavity perturbation technique, is used to compute the overall reflection and transmission coefficients of single and multiple layers of the polymers. With recent advances in synthesizing stable highly conductive polymers these lightweight mechanically strong materials appear to be viable alternatives to metals for EM1 shielding.

  18. Hexavalent Chrome Free Coatings for Electronics Electromagnetic Interference (EMI) Shielding Effectiveness (SE) Interim Test Report

    Science.gov (United States)

    Kessel, Kurt R.

    2015-01-01

    Test specimen configuration was provided by Parker Chomerics. The EMI gasket used in this project was Cho-Seal 6503E. Black oxide alloy steel socket head bolts were used to hold the plates together. Non-conductive spacers were used to control the amount of compression on the gaskets. The following test fixture specifications were provided by Parker Chomerics. The CHO-TP09 test plate sets selected for this project consist of two aluminum plates manufactured to the specifications detailed in CHO­-TP09. The first plate, referred to as the test frame, is illustrated in Figure 1. The test frame is designed with a cutout in the center and two alternating bolt patterns. One pattern is used to bolt the test frame to the corresponding test cover plate (Figure 2), forming a test plate set. The second pattern accepts the hardware used to mount the fully assembled test plate set to the main adapter plate (Figure 3).

  19. 热管冷却型月球堆的辐射屏蔽设计研究%Radiation Shielding Design for a Heat Pipe Cooled Lunar Surface Reactor

    Institute of Scientific and Technical Information of China (English)

    杜金峰; 宋旺旺; 王三丙; 谢奇林; 范晓强; 梁文峰

    2016-01-01

    针对40 kWe热管型月球堆的辐射屏蔽问题,建立了三维蒙特卡罗计算模型。屏蔽布局方式整体采用地埋式,堆芯上方的轴向屏蔽体则采用影锥构形,选取适用温度范围宽、耐辐照的碳化硼作为中子屏蔽材料,伽玛屏蔽材料则采用钨,模拟计算了空腔及钨层厚度对轴向屏蔽体后端典型位置的辐射影响,同时也考察了百米外人员所受剂量随径向屏蔽体厚度的变化关系。据此得到了热管型月球堆初步的屏蔽体构型,其总质量为2�15 t ,十年运行寿期内紧靠屏蔽体后端的敏感器件最大中子注量(1 MeV等效)为1�86×1014 cm-2、最大伽玛剂量为3�68 Mrad,而人员剂量仅为11�92 mSv/a,在较大的裕量范围内满足辐射安全要求。%Three⁃dimension Monte Carlo calculation model was built to study the radiation shielding of heat pipe cooled lunar surface reactor with 40kWe. The lunar soil buried shielding method was chosen for the overall layout and the cone⁃shadow configuration was adopted for the axis shielding. B4 C was used for neutron shielding owing to its large range of applicable temperature and high radia⁃tion resistant, and W was adopted for gamma shielding. The radiation fluence on the back end of the axis shielding versus the chamber and W’ s thickness were calculated. The dose for an astronaut 100 m away versus the radial shield thickness was also calculated. The primary shielding configuration was provided and the whole mass was 2�15 t. The maximal neutron fluence (1 MeV equivalent) was 1�86 × 1014 cm-2 and the maximal gama dose for the radiation sensitive component was 3. 68Mrad, dose for an astronaut 100 m away was only 11�92 mSv/a. The shield design can satisfy the shielding requirements with a large margin.

  20. The Jules Horowitz Reactor - A new High Performance European Material Testing Reactor open to International Users Present Status and Objectives

    Energy Technology Data Exchange (ETDEWEB)

    Iracane, Daniel; Bignan, Gilles [CEA Atomic Energy Commission Saclay Batiment 121- 91191 Gif Sur Yvette (France); Lindbaeck, Jan-Erik; Blomgren, Jan [VATTENFALL AB Nuclear Power Jaemtlandsgatan 99 SE-16287 Stockholm (Sweden)

    2010-07-01

    The development of sustainable nuclear energy requires R and D on fuel and material behaviour under irradiation with a high level of performance in order to meet the needs and challenges for the benefit of industry, research and public bodies. These stakes require a sustainable and secured access to an up-to-date high performance Material Testing Reactor. Following a broad survey within the European Research Area, the international community agreed that the need for Material Test Reactors in support of nuclear power plant safety and operation will continue in the context of sustainable nuclear energy. The Jules Horowitz Reactor project (JHR) copes with this context. JHR is designed as a user facility addressing the needs of the international community. This means: - flexibility with irradiation loops able to reproduce a large variation in operation conditions of different power reactor technologies, - high flux capacity to address Generations II, III, and IV needs. JHR is designed, built and operated as an international user facility because: - Given the maturity and globalization of the industry, domestic tools have no more the required level of economic and technical efficiency. Meanwhile, countries with nuclear energy need an access to high performance irradiation experimental capabilities to support technical skill and guarantee the competitiveness and safety of nuclear energy. - Many research items related to safety or public policy (waste management, etc.) require international cooperation to share costs and benefits of resulting consensus. JHR design is optimised for offering high performance material and fuel irradiation capability for the coming decades. This project is driven and funded by an international consortium gathering vendors, utilities and public stakeholders. This consortium has been set up in March 2007 when the construction began. The construction is in progress and the start of operation is scheduled for 2014. The JHR is a research

  1. ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis; Gordon Kohse; Yakov Ostrovsky; David M. Carpenter; Joy L. Rempe

    2015-04-01

    A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Current Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in

  2. 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.

  3. Adaptive support for aircraft panel testing: New method and its experimental verification on a beam structure

    Science.gov (United States)

    Sachau, Delf; Baschke, Manuel

    2017-04-01

    Acoustic transmissibility of aircraft panels is measured in full-scale test rigs. The panels are supported at their frames. These boundary conditions do not take into account the dynamic influence of the fuselage, which is significant in the frequency range below 300 Hz. This paper introduces a new adaptive boundary system (ABS). It combines accelerometers and electrodynamic shakers with real-time signal processing. The ABS considers the dynamic effect of the fuselage on the panel. The frames are dominating the dynamic behaviour of a fuselage in the low-frequency range. Therefore, the new method is applied to a beam representing a frame of the aircraft structure. The experimental results are evaluated and the precision of the ABS is discussed. The theoretical apparent mass representing the cut-off part of a frame is calculated and compared with the apparent mass, as provided by the ABS. It is explained how the experimental set-up limits the precision of the ABS.

  4. Production test PTA-002, increased graphite temperature limit -- B, C and D Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Russell, A.

    1965-12-17

    The fundamental objective of the graphite temperature limit is to prevent excessive oxidation of the graphite moderator blocks with carbon dioxide and water vapor in the reactor atmosphere. Laboratory tests have shown that 10% uniform oxidation of graphite results in a loss in strength of approximately 50%. Production Test IP-725 was conducted at F Reactor for a period of six months at graphite temperatures approximately 50 and 100 C higher than the present graphite temperature limit of 650 C. The results from the F Reactor test suggest that an increase in the graphite temperature limit from 650 C to 700 C is technically feasible from the standpoint of oxidation of the graphite moderator with CO{sub 2}. Any significant additional increase was shown to lead to excessively high oxidation rates and is therefore not considered feasible. The objective of this test, therefore, is to extend the higher temperature investigations to B, C, and D Reactors. For the duration of this test, the graphite temperature limit will be increased from 650 C and 700 C, corresponding to an increase in the graphite stringer temperature limit from 735 C to 790 C. The test is expected to last for approximately six months but may be terminated early on any or all the reactors.

  5. THE MECHANICAL AND SHIELDING DESIGN OF A PORTABLE SPECTROMETER AND BEAM DUMP ASSEMBLY AT BNLS ACCELERATOR TEST FACILITY.

    Energy Technology Data Exchange (ETDEWEB)

    HU,J.P.; CASEY,W.R.; HARDER,D.A.; PJEROV,S.; RAKOWSKY,G.; SKARITKA,J.R.

    2002-09-05

    A portable assembly containing a vertical-bend dipole magnet has been designed and installed immediately down-beam of the Compton electron-laser interaction chamber on beamline 1 of the Accelerator Test Facility (ATF) at Brookhaven National Laboratory (BNL). The water-cooled magnet designed with field strength of up to 0.7 Tesla will be used as a spectrometer in the Thompson scattering and vacuum acceleration experiments, where field-dependent electron scattering, beam focusing and energy spread will be analyzed. This magnet will deflect the ATF's 60 MeV electron-beam 90{sup o} downward, as a vertical beam dump for the Compton scattering experiment. The dipole magnet assembly is portable, and can be relocated to other beamlines at the ATF or other accelerator facilities to be used as a spectrometer or a beam dump. The mechanical and shielding calculations are presented in this paper. The structural rigidity and stability of the assembly were studied. A square lead shield surrounding the assembly's Faraday Cup was designed to attenuate the radiation emerging from the 1 inch-copper beam stop. All photons produced were assumed to be sufficiently energetic to generate photoneutrons. A safety evaluation of groundwater tritium contamination due to the thermal neutron capturing by the deuterium in water was performed, using updated Monte Carlo neutron-photon coupled transport code (MCNP). High-energy neutron spallation, which is a potential source to directly generate radioactive tritium and sodium-22 in soil, was conservatively assessed in verifying personal and environmental safety.

  6. Performance and Facility Background Pressure Characterization Tests of NASAs 12.5-kW Hall Effect Rocket with Magnetic Shielding Thruster

    Science.gov (United States)

    Kamhawi, Hani; Huang, Wensheng; Haag, Thomas; Shastry, Rohit; Thomas, Robert; Yim, John; Herman, Daniel; Williams, George; Myers, James; Hofer, Richard; Mikellides, Ioannis; Sekerak, Michael; Polk, James

    2015-01-01

    NASA's Space Technology Mission Directorate (STMD) Solar Electric Propulsion Technology Demonstration Mission (SEP/TDM) project is funding the development of a 12.5-kW Hall thruster system to support future NASA missions. The thruster designated Hall Effect Rocket with Magnetic Shielding (HERMeS) is a 12.5-kW Hall thruster with magnetic shielding incorporating a centrally mounted cathode. HERMeS was designed and modeled by a NASA GRC and JPL team and was fabricated and tested in vacuum facility 5 (VF5) at NASA GRC. Tests at NASA GRC were performed with the Technology Development Unit 1 (TDU1) thruster. TDU1's magnetic shielding topology was confirmed by measurement of anode potential and low electron temperature along the discharge chamber walls. Thermal characterization tests indicated that during full power thruster operation at peak magnetic field strength, the various thruster component temperatures were below prescribed maximum allowable limits. Performance characterization tests demonstrated the thruster's wide throttling range and found that the thruster can achieve a peak thruster efficiency of 63% at 12.5 kW 500 V and can attain a specific impulse of 3,000 s at 12.5 kW and a discharge voltage of 800 V. Facility background pressure variation tests revealed that the performance, operational characteristics, and magnetic shielding effectiveness of the TDU1 design were mostly insensitive to increases in background pressure.

  7. Safety Issues at the DOE Test and Research Reactors. A Report to the U.S. Department of Energy.

    Science.gov (United States)

    National Academy of Sciences - National Research Council, Washington, DC. Commission on Physical Sciences, Mathematics, and Resources.

    This report provides an assessment of safety issues at the Department of Energy (DOE) test and research reactors. Part A identifies six safety issues of the reactors. These issues include the safety design philosophy, the conduct of safety reviews, the performance of probabilistic risk assessments, the reliance on reactor operators, the fragmented…

  8. Limitations of eddy current testing in a fast reactor environment

    Science.gov (United States)

    Wu, Tao; Bowler, John R.

    2016-02-01

    The feasibility of using eddy current probes for detecting flaws in fast nuclear reactor structures has been investigated with the aim of detecting defects immersed in electrically conductive coolant including under liquid sodium during standby. For the inspections to be viable, there is a need to use an encapsulated sensor system that can be move into position with the aid of visualization tools. The initial objective being to locate the surface to be investigated using, for example, a combination of electromagnetic sensors and sonar. Here we focus on one feature of the task in which eddy current probe impedance variations due to interaction with the external surface of a tube are evaluated in order to monitor the probe location and orientation during inspection.

  9. 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.

  10. Design and Testing of a Prototype Pixellated CZT Detector and Shield for Hard X-Ray Astronomy

    CERN Document Server

    Bloser, P F; Narita, T; Jenkins, J A

    1999-01-01

    We report on the design and laboratory testing of a prototype imaging CZT detector intended for balloon flight testing in April 2000. The detector tests several key techniques needed for the construction of large-area CZT arrays, as required for proposed hard X-ray astronomy missions. Two 10 mm x 10 mm x 5 mm CZT detectors, each with a 4 x 4 array of 1.9 mm pixels on a 2.5 mm pitch, will be mounted in a ``flip-chip'' fashion on a printed circuit board carrier card; the detectors will be placed 0.3 mm apart in a tiled configuration such that the pixel pitch is preserved across both crystals. One detector is eV Products high-pressure Bridgman CZT, and the other is IMARAD horizontal Bridgman material. Both detectors are read out by a 32-channel VA-TA ASIC controlled by a PC/104 single-board computer. A passive shield/collimator surrounded by plastic scintillator surrounds the detectors on five sides and provides a ~45 deg field of view. The background spectrum recorded by this instrument will be compared to that...

  11. 77 FR 36341 - Control of Air Pollution From Aircraft and Aircraft Engines; Emission Standards and Test Procedures

    Science.gov (United States)

    2012-06-18

    ... Organization (ICAO) proceedings in which most of these requirements were first developed. These regulatory... E. Brief History of ICAO Regulation of Aircraft Engine Emissions III. Why is EPA taking this action.... This rule will allow us to enforce in the U.S. the emission standards adopted by ICAO, and will...

  12. Safety and Suitability for Service Assessment Testing for Aircraft Launched Munitions

    Science.gov (United States)

    2013-07-01

    aircraft since shock responses can be affected by local influences such as wing and fuselage bending modes, pylon interfaces, and structural damping...and unguided munitions deployed on fixed and rotary wing aircraft (manned or unmanned). The munitions covered by this JOTP include aircraft launched...attached to an aircraft (e.g., missile attached to a launcher on an aircraft wing ). 4. FACILITIES AND INSTRUMENTATION. 4.1 Facilities. All

  13. Sequential probability ratio tests for reactor signal validation and sensor surveillance applications

    Energy Technology Data Exchange (ETDEWEB)

    Humenik, K. (Maryland Univ., Baltimore, MD (USA)); Gross, K.C. (Argonne National Lab., IL (USA))

    1989-11-09

    This paper examines the properties of sequential probability ratio tests (SPRT's) and the application of these tests to nuclear power reactor operation. Recently SPRT's have been applied to delayed-neutron (DN) signal data analysis using actual reactor data from the Experimental Breeder Reactor-II, which is operated by Argonne National Laboratory. The implementation of this research as part of an expert system is described. Mathematical properties of the SPRT are investigated, and theoretical results are validated with tests that use DN-signal data taken from the EBR-II in Idaho. Variations of the basic SPRT and applications to general signal validation are also explored. 16 refs., 3 figs.

  14. Accelerated corrosion test and corrosion failure distribution model of aircraft structural aluminum alloy

    Institute of Scientific and Technical Information of China (English)

    LIU Wen-lin; MU Zhi-tao; JIN Ping

    2006-01-01

    Based on corrosion damage data of 10 years for a type of aircraft aluminum alloy, the statistical analysis was conducted by Gumbel, Normal and two parameters Weibull distribution function. The results show that aluminum alloy structural member has the corrosion history of pitting corrosion-intergranular corrosion-exfoliation corrosion, and the maximum corrosion depth is in conformity to normal distribution. The accelerated corrosion test was carried out with the complied equivalent airport accelerated environment spectrum. The corrosion damage failure modes of aluminum alloy structural member indicate that the period of validity of the former protective coating is about 2.5 to 3 years, and that of the novel protective coating is about 4.0 to 4.5 years. The corrosion kinetics law of aluminum spar flange was established by fitting corrosion damage test data. The law indicates two apparent corrosion stages of high strength aluminum alloy section material: pitting corrosion and intergranular corrosion/exfoliation corrosion.The test results agree with the statistical fit result of corrosion data collected from corrosion member in service. The fractional error is 5.8% at the same calendar year. The accelerated corrosion test validates the corrosion kinetics law of aircraft aluminum alloy in service.

  15. Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

    Energy Technology Data Exchange (ETDEWEB)

    M. Chen; CM Regan; D. Noe

    2006-01-09

    Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

  16. Design of Seismic Test Rig for Control Rod Drive Mechanism of Jordan Research and Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Jongoh; Kim, Gyeongho; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The reactor assembly is submerged in a reactor pool filled with water and its reactivity is controlled by locations of four control absorber rods(CARs) inside the reactor assembly. Each CAR is driven by a stepping motor installed at the top of the reactor pool and they are connected to each other by a tie rod and an electromagnet. The CARs scram the reactor by de-energizing the electromagnet in the event of a safe shutdown earthquake(SSE). Therefore, the safety function of the control rod drive mechanism(CRDM) which consists of a drive assembly, tie rod and CARs is to drop the CAR into the core within an appropriate time in case of the SSE. As well known, the operability for complex equipment such as the CRDM during an earthquake is very hard to be demonstrated by analysis and should be verified through tests. One of them simulates the reactor assembly and the guide tube of the CAR, and the other one does the pool wall where the drive assembly is installed. In this paper, design of the latter test rig and how the test is performed are presented. Initial design of the seismic test rig and excitation table had its first natural frequency at 16.3Hz and could not represent the environment where the CRDM was installed. Therefore, experimental modal analyses were performed and an FE model for the test rig and table was obtained and tuned based on the experimental results. Using the FE model, the design of the test rig and table was modified in order to have higher natural frequency than the cutoff frequency. The goal was achieved by changing its center of gravity and the stiffness of its sliding bearings.

  17. Construction and testing of a pixellated CZT detector and shield for a hard x-ray astronomy balloon flight

    Science.gov (United States)

    Bloser, Peter F.; Narita, Tomohiko; Jenkins, Jonathan A.; Grindlay, Jonathan E.

    2000-12-01

    We report on the construction and laboratory testing of pixellated CZT detectors mounted in a flip-chip, tiled fashion and read out by an ASIC, as required for proposed hard X-ray astronomy missions. Two 10 mm X 10 mm X 5 mm detectors were fabricated, one out of standard eV Products high-pressure Bridgman CZT and one out of IMARAD horizontal Bridgman CZT. Each was fashioned with a 4 X 4 array of gold pixels on 2.5 mm pitch with a surrounding guard ring. The detectors were mounted side by side on a carrier card, such that the pixel pitch was preserved, and read out by a 32-channel VA-TA ASIC from IDE AS Corp. controlled by a PC/104 single-board computer. A passive shield/collimator surrounded by plastic scintillator encloses the detectors on five sides and provides an approximately 40 degree field of view. Thus this experiment tests key techniques required for future hard X-ray survey instruments. The experiment was taken to Ft. Sumner, NM in May 2000 in preparation for a scientific balloon flight aboard the joint Harvard-MSFC EXITE2/HERO payload. Although we did not receive a flight opportunity, and are currently scheduled to fly in September 2000, we present our calibration data in the flight configuration together with data analysis techniques and simulations of the expected flight background spectrum.

  18. Construction and Testing of a Pixellated CZT Detector and Shield for a Hard X-ray Astronomy Balloon Flight

    CERN Document Server

    Bloser, P F; Jenkins, J A; Grindlay, J E

    2000-01-01

    We report on the construction and laboratory testing of pixellated CZT detectors mounted in a flip-chip, tiled fashion and read out by an ASIC, as required for proposed hard X-ray astronomy missions. Two 10 mm x 10 mm x 5 mm detectors were fabricated, one out of standard eV Products high-pressure Bridgman CZT and one out of IMARAD horizontal Bridgman CZT. Each was fashioned with a 4 x 4 array of gold pixels on a 2.5 mm pitch with a surrounding guard ring. The detectors were mounted side by side on a carrier card, such that the pixel pitch was preserved, and read out by a 32-channel VA-TA ASIC from IDE AS Corp. controlled by a PC/104 single-board computer. A passive shield/collimator surrounded by plastic scintillator encloses the detectors on five sides and provides a ~40deg field of view. Thus this experiment tests key techniques required for future hard X-ray survey instruments. The experiment was taken to Ft Sumner, NM in May 2000 in preparation for a scientific balloon flight aboard the joint Harvard-MSFC...

  19. The experimental determination of atmospheric absorption from aircraft acoustic flight tests

    Science.gov (United States)

    Miller, R. L.; Oncley, P. B.

    1971-01-01

    A method for determining atmospheric absorption coefficients from acoustic flight test data is presented. Measurements from five series of acoustic flight tests were included in the study. The number of individual flights totaled 24: six Boeing 707 flights performed in May 1969 in connection with the turbofan nacelle modification program, eight flights from Boeing tests conducted during the same period, and 10 flights of the Boeing 747 airplane. The effects of errors in acoustic, meteorological, and aircraft performance and position measurements are discussed. Tabular data of the estimated sample variance of the data for each test are given for source directivity angles from 75 deg to 120 deg and each 1/3-octave frequency band. Graphic comparisons are made of absorption coefficients derived from ARP 866, using atmospheric profile data, with absorption coefficients determined by the experimental method described in the report.

  20. Novel shielding materials for space and air travel.

    Science.gov (United States)

    Vana, N; Hajek, M; Berger, T; Fugger, M; Hofmann, P

    2006-01-01

    The reduction of dose onboard spacecraft and aircraft by appropriate shielding measures plays an essential role in the future development of space exploration and air travel. The design of novel shielding strategies and materials may involve hydrogenous composites, as it is well known that liquid hydrogen is most effective in attenuating charged particle radiation. As precursor for a later flight experiment, the shielding properties of newly developed hydrogen-rich polymers and rare earth-doped high-density rubber were tested in various ground-based neutron and heavy ion fields and compared with aluminium and polyethylene as reference materials. Absorbed dose, average linear energy transfer and gamma-equivalent neutron absorbed dose were determined by means of LiF:Mg,Ti thermoluminescence dosemeters and CR-39 plastic nuclear track detectors. First results for samples of equal aerial density indicate that selected hydrogen-rich plastics and rare-earth-doped rubber may be more effective in attenuating cosmic rays by up to 10% compared with conventional aluminium shielding. The appropriate adaptation of shielding thicknesses may thus allow reducing the biologically relevant dose. Owing to the lower density of the plastic composites, mass savings shall result in a significant reduction of launch costs. The experiment was flown as part of the European Space Agency's Biopan-5 mission in May 2005.

  1. Radiation shielding for neutron guides

    Science.gov (United States)

    Ersez, T.; Braoudakis, G.; Osborn, J. C.

    2006-11-01

    Models of the neutron guide shielding for the out of bunker guides on the thermal and cold neutron beam lines of the OPAL Reactor (ANSTO) were constructed using the Monte Carlo code MCNP 4B. The neutrons that were not reflected inside the guides but were absorbed by the supermirror (SM) layers were noted to be a significant source of gammas. Gammas also arise from neutrons absorbed by the B, Si, Na and K contained in the glass. The proposed shielding design has produced compact shielding assemblies. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. To verify the design a prototype was assembled consisting of 120 mm thick Pb(96%)Sb(4%) walls resting on a concrete block. There was good agreement between experimental measurements and calculated dose rates for bulk shield regions.

  2. Postirradiation examination of recycle test elements from the Peach Bottom Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tiegs, T.N.; Long, E.L. Jr.

    1978-12-01

    The Recycle Test Elements were a series of tests of High-Temperature Gas-Cooled Reactor fuels irradiated in Core 2 of the Peach Bottom Unit 1 Reactor. They tested a wide variety of fissile and fertile fuel types of prime interest when the tests were designed. The fuel types included UO/sub 2/, UC/sub 2/, (2Th,U)O/sub 2/, (4Th,U)O/sub 2/, ThC/sub 2/, and ThO/sub 2/. The mixed thorium--uranium oxides and the pure thorium oxide were tested as Biso-coated particles only, while the others were tested as both Biso- and Triso-coated particles. The Biso coatings on the fissile kernels contained the fission products inadequately but on the fertile kernels they did so acceptably. The results from accelerated and real-time tests on the particle types agreed well.

  3. Cold Model Study and Commercial Test on Novel Vapor-Liquid Distributor of Hydroprocessing Reactor

    Institute of Scientific and Technical Information of China (English)

    Wang Shaobing; Zhang Zhanzhu; Wu Defei; Guo Qingming

    2007-01-01

    A novel vapor-liquid distributor was developed on the basis of sufficient study on the existing distributors applied in hydroprocessing reactors.The cold model test data showed that the fluid distribution performance of the novel vapor-liquid distributor was evidently better than the traditional one.Commercial tests of the new distributor were carried out in the 300 kt/a gas oil hydrotreating reactor at SINOPEC Changling Branch Company,showing that the new vapor-liquid distributor could improve the fluid distribution,promote the hydrotreating efficiency and lead to better performance than the traditional one.

  4. An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rosenthal, Murray Wilford [ORNL

    2009-08-01

    The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

  5. Measurement and calculation of the fast-neutron and photon spectra from the core boundary to the biological shielding in the WWER-1000 reactor model.

    Science.gov (United States)

    Osmera, B; Cvachovec, F; Kyncl, J; Smutný, V

    2005-01-01

    The fast-neutron and photon space-energy distributions have been measured in an axially (1.25 m active height) and azimuthally (60 degree symmetry sector) shortened model of the WWER-1000 reactor assembled in the LR-0 experimental reactor. The space-energy distributions have been calculated with the stochastic code MCNP and the deterministic three-dimensional code TORT. Selected results are presented and discussed in the paper. This work has been done in the frame of the EU 5th FW project REDOS REDOS, Reactor Dosimetry: Accurate determination and benchmarking of radiation field parameters, relevant for reactor pressure vessel monitoring. EURATOM Programme, Call 2000/C 294/04). All geometry and material composition data of the model as well as the available experimental data were carefully checked and revised.

  6. High Temperature Stress Analysis on 61-pin Test Assembly for Reactor Core Sub-channel Flow Test

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dongwon; Kim, Hyungmo; Lee, Hyeongyeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this study, a high temperature heat transfer and stress analysis of a 61-pin test fuel assembly scaled down from the full scale 217-pin sub-assembly was conducted. The reactor core subchannel flow characteristic test will be conducted to evaluate uncertainties in computer codes used for reactor core thermal hydraulic design. Stress analysis for a 61-pin fuel assembly scaled down from Prototype Generation IV Sodium-cooled Fast Reactor was conducted and structural integrity in terms of load controlled stress limits was conducted. In this study, The evaluations on load-controlled stress limits for a 61-pin test fuel assembly to be used for reactor core subchannel flow distribution tests were conducted assuming that the test assembly is installed in a Prototype Generation IV Sodium-cooled fast reactor core. The 61-pin test assembly has the geometric similarity on P/D and H/D with PGSFR and material of fuel assembly is austenitic stainless steel 316L. The stress analysis results showed that 4.05MPa under primary load occurred at mid part of the test assembly and it was shown that the value of 4.05Mpa was far smaller than the code allowable of 127MPa. , it was shown that the stress intensity due to due to primary load is very small. The stress analysis results under primary and secondary loads showed that maximum stress intensity of 84.08MPa occurred at upper flange tangent to outer casing and the value was well within the code allowable of 268.8MPa. Integrity evaluations based on strain limits and creep-fatigue damage are underway according to the elevated design codes.

  7. Advanced high-pressure bench-scale reactor for testing with hot corrosive gases

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, J.; Bachta, R.P.; Wangerow, J.R. (Inst. of Gas Technology, Chicago, IL (United States)); Mojtahedi, W.; Salo, K. (Enviropower Inc., Espoo (Finland))

    1994-01-01

    A bench-scale, high-pressure/high-temperature fluidized-bed reactor (HPTR) system is described that is capable of operating at a maximum temperature and pressure of 1,000 C and 30 bar in a corrosive atmosphere. The design of the unit is based on a double-shell balanced-pressure system. All the hot parts of the reactor that are wetted by the corrosive (and/or reactive) gases and the entire sampling line are constructed of inert material to prevent corrosion and loss of the reactant gases. The unit has been used for over 200 high-pressure hot coal gas desulfurization tests at 20 bars and up to 750 C without any experimental problem and with excellent sulfur balance, indicating that this reactor system is ideal for testing with reactive and corrosive gases at elevated pressures and temperatures.

  8. Standard Test Method for Stress-Corrosion of Titanium Alloys by Aircraft Engine Cleaning Materials

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2006-01-01

    1.1 This test method establishes a test procedure for determining the propensity of aircraft turbine engine cleaning and maintenance materials for causing stress corrosion cracking of titanium alloy parts. 1.2 The evaluation is conducted on representative titanium alloys by determining the effect of contact with cleaning and maintenance materials on tendency of prestressed titanium alloys to crack when subsequently heated to elevated temperatures. 1.3 Test conditions are based upon manufacturer's maximum recommended operating solution concentration. 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. For specific precautionary statements, see and .

  9. Test evaluation of potential heat shield contamination of an Outer Planet Probe's atmospheric sampling system

    Science.gov (United States)

    Kessler, W. C.; Woeller, F. H.; Wilkins, M. E.

    1975-01-01

    An Outer Planets Probe which retains the charred heatshield during atmospheric descent must deploy a sampling tube through the heatshield to extract atmospheric samples for analysis. Once the sampling tube is deployed, the atmospheric samples ingested must be free of contaminant gases generated by the heatshield. Outgassing products such as methane and water vapor are present in planetary atmospheres and hence, ingestion of such species would result in gas analyzer measurement uncertainties. This paper evaluates the potential for, and design impact of, the extracted atmospheric samples being contaminated by heatshield outgassing products. Flight trajectory data for Jupiter, Saturn and Uranus entries are analyzed to define the conditions resulting in the greatest potential for outgassing products being ingested into the probe's sampling system. An experimental program is defined and described which simulates the key flow field features for a planetary flight in a ground-based test facility. The primary parameters varied in the test include: sampling tube length, injectant mass flow rate and angle of attack. Measured contaminant levels predict the critical sampling tube length for contamination avoidance. Thus, the study demonstrates the compatibility of a retained heatshield concept and high quality atmospheric trace species measurements.

  10. Lithium hydride - A space age shielding material

    Science.gov (United States)

    Welch, F. H.

    1974-01-01

    Men and materials performing in the environment of an operating nuclear reactor require shielding from the escaping neutron particles and gamma rays. For efficient shielding from gamma rays, dense, high atomic number elements such as iron, lead, or tungsten are required, whereas light, low atomic number elements such as hydrogen, lithium, or beryllium are required for efficient neutron shielding. The use of lithium hydride (LiH) as a highly efficient neutron-shielding material is considered. It contains, combined into a single, stable compound, two of the elements most effective in attenuating and absorbing neutrons.

  11. Evaluating and planning the radioactive waste options for dismantling the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rule, K.; Scott, J.; Larson, S. [Princeton Plasma Physics Lab., NJ (United States)] [and others

    1995-12-31

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a kind tritium fusion research reactor, and is planned to be decommissioned within the next several years. This is the largest fusion reactor in the world and as a result of deuterium-tritum reactions is tritium contaminated and activated from 14 Mev neutrons. This presents many unusual challenges when dismantling, packaging and disposing its components and ancillary systems. Special containers are being designed to accommodate the vacuum vessel, neutral beams, and tritium delivery and processing systems. A team of experienced professionals performed a detailed field study to evaluate the requirements and appropriate methods for packaging the radioactive materials. This team focused on several current and innovative methods for waste minimization that provides the oppurtunmost cost effective manner to package and dispose of the waste. This study also produces a functional time-phased schedule which conjoins the waste volume, weight, costs and container requirements with the detailed project activity schedule for the entire project scope. This study and project will be the first demonstration of the decommissioning of a tritium fusion test reactor. The radioactive waste disposal aspects of this project are instrumental in demonstrating the viability of a fusion power reactor with regard to its environmental impact and ultimate success.

  12. Performance tests of a small hydrogen reactor based on Mg-Al pellets

    Energy Technology Data Exchange (ETDEWEB)

    Capurso, Giovanni, E-mail: giovanni.capurso@studenti.unipd.it [Dipartimento di Ingegneria Meccanica, Settore Materiali, Universita di Padova, via Marzolo 9, 35131 Padova (Italy); Agresti, Filippo [Dipartimento di Ingegneria Meccanica, Settore Materiali, Universita di Padova, via Marzolo 9, 35131 Padova (Italy); Russo, Sergio Lo [Dipartimento di Fisica and CNISM, Universita di Padova, via Marzolo 8, 35131 Padova (Italy); Maddalena, Amedeo; Principi, Giovanni [Dipartimento di Ingegneria Meccanica, Settore Materiali, Universita di Padova, via Marzolo 9, 35131 Padova (Italy); Cavallari, Andrea; Guardamagna, Cristina [ERSE s.p.a., via Rubattino 54, 20134 Milano (Italy)

    2011-09-15

    On the basis of a previously acquired experience on scaling up issues concerning the use of magnesium hydride as a base material for solid-state hydrogen storage, a small reactor was designed and tested in different operating conditions. It contains about 10 g of catalyzed magnesium hydride powder mixed with 5 wt.% aluminium powder and pressed in the form of cylindrical pellets and the heat flow is managed by means of an oil circulation system. Carbon paper is used to ensure good heat conductivity between the pellets and the inner wall of the reactor and between one pellet and another. A number of hydrogen absorption and desorption cycles at different temperatures and pressures was carried out to compare the behaviour of the small reactor with the laboratory data obtained on small amounts (fractions of grams) of powdered and pelletized samples. Data acquisition for gas flow, pressure and temperature in different positions of the reactor allow a good understanding of internal dynamics. The results in terms of hydrogen absorption/desorption kinetics and of stability to ongoing cycles are stimulating, so that the tested small reactor can be considered as a basic element for further studies and improvements.

  13. Shielding effect and wakefield pattern of a moving test charge in a non-Maxwellian dusty plasma

    Energy Technology Data Exchange (ETDEWEB)

    Ali, S. [National Centre for Physics (NCP), Quaid-e-Azam University Campus, Shahdra Valley Road, Islamabad 44000 (Pakistan); Khan, S. [National Centre for Physics (NCP), Quaid-e-Azam University Campus, Shahdra Valley Road, Islamabad 44000 (Pakistan); Department of Physics, Gomal University, Dera Ismail Khan 29050 (Pakistan)

    2013-07-15

    By using the Vlasov-Poisson equations, we calculate an expression for the electrostatic potential caused by a test charge in an unmagnetized non-Maxwellian dusty plasma, whose constituents are the superthermal hot-electrons, the mobile cold-electrons with a neutralizing background of cold ions, and charge fluctuating isolated dust grains. The superthermality effects due to hot electrons not only modify the dielectric constant of the electron-acoustic waves but also significantly affect the electrostatic potential. The latter can be decomposed into the Debye-Hückel and oscillatory wake potentials. Analytical and numerical results reveal that the Debye-Hückel and wakefield potentials converge to the Maxwellian case for large values of superthermality parameter. Furthermore, the plasma parameters play a vital role in the formation of shielding and wakefield pattern in a two-electron temperature plasma. The present results should be important for laboratory and space dusty plasmas, where hot-electrons can be assumed to follow the non-Maxwellian distribution function.

  14. Volcanic edifice alignment detection software in MATLAB: Test data and preliminary results for shield fields on Venus

    Science.gov (United States)

    Thomson, Bradley J.; Lang, Nicholas P.

    2016-08-01

    The scarcity of impact craters on Venus make it difficult to infer the relative ages of geologic units. Stratigraphic methods can be used to help infer the relative ordering of surface features, but the relatively coarse resolution of available radar data means ambiguity about the timing of certain features is common. Here we develop a set of statistical tools in MATLAB to help infer the relative timing between clusters of small shield volcanoes and sets of fractures in the surrounding terrain. Specifically, we employed two variants of the two-point azimuth method to detect anisotropy in the distribution of point-like features. The results of these methods are shown to successfully identify anisotropy at two spatial scales: at the whole-field level and at scales smaller than a set fraction of the mean value. Initial results on the test cases presented here are promising, at least for volcanic fields emplaced under uniform conditions. These methods could also be used for detecting anisotropy in other point-like geologic features, such as hydrothermal vents, springs, and earthquake epicenters.

  15. 不同骨料防辐射混凝土屏蔽性能试验%Different concrete aggregate anti-radiation shielding performance test

    Institute of Scientific and Technical Information of China (English)

    郭文强; 石建军; 曾帅; 陈炤瑀; 宋雪

    2016-01-01

    In order to study the shielding performance of different aggregate anti-radiation concrete and the ordinary sand and gravel ag-gregate effect the performance of the radiation protection concrete,selects the magnetite sand and magnetite ore,barite,barite ore,steel ball and ordinary river sand and crushed stone as aggregate,mixing get four groups of the radiation protection concrete,gamma ray shield-ing experiment was carried out.The test results show that the magnetic iron sand,steel ball,magnet ore as the radiation protection concrete aggregate,shielding performance is best;Four groups of the radiation protection concrete contrast each other,can be found among several kinds of aggregate used in the experiment,shielding performance of the steel ball is best,magnetite and barite shielding performance is su-perior to the ordinary sand and gravel aggregate;In the concrete aggregate of common shielding gamma-ray mixed with ordinary river sand and gravel,radiation protection still has a good shielding performance of concrete.%为研究不同骨料防辐射混凝土的屏蔽性能以及普通砂石骨料对防辐射混凝土屏蔽性能的影响,选用磁铁矿砂、磁铁矿石、重晶石、重晶石砂、钢珠以及普通河砂和碎石作为骨料,拌制得到四组防辐射混凝土,进行γ射线屏蔽试验。试验结果表明,以磁铁矿砂、钢珠、磁铁矿石作为骨料的防辐射混凝土,屏蔽性能最好;四组防辐射混凝土相互对比,可以发现试验所用的几种骨料当中,钢珠的屏蔽性能最好,磁铁矿和重晶石骨料的屏蔽性能优于普通砂石;在常用屏蔽γ射线的混凝土骨料当中掺入普通河砂和碎石,防辐射混凝土仍然具有良好的屏蔽性能。

  16. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1983-06-01

    During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

  17. Evaluation of the Start-Up Core Physics Tests at Japan's High Temperature Engineering Test Reactor (Annular Core Loadings)

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-03-01

    The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The Japanese government approved construction of the HTTR in the 1989 fiscal year budget; construction began at the Oarai Research and Development Center in March 1991 and was completed May 1996. Fuel loading began July 1, 1998, from the core periphery. The first criticality was attained with an annular core on November 10, 1998 at 14:18, followed by a series of start-up core physics tests until a fully-loaded core was developed on December 16, 1998. Criticality tests were carried out into January 1999. The first full power operation with an average core outlet temperature of 850ºC was completed on December 7, 2001, and operational licensing of the HTTR was approved on March 6, 2002. The HTTR attained high temperature operation at 950 ºC in April 19, 2004. After a series of safety demonstration tests, it will be used as the heat source in a hydrogen production system by 2015. Hot zero-power critical, rise-to-power, irradiation, and safety demonstration testing , have also been performed with the HTTR, representing additional means for computational validation efforts. Power tests were performed in steps from 0 to 30 MW, with various tests performed at each step to confirm

  18. Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY1999-2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    The HTTR (High Temperature Engineering Test Reactor) with the thermal power of 30 MW and the reactor outlet coolant temperature of 850/950 degC is the first high temperature gas-cooled reactor (HTGR) in Japan, which uses coated fuel particle, graphite for core components, and helium gas for primary coolant. The HTTR, which locates at the south-west area of 50,000 m{sup 2} in the Oarai Research Establishment, had been constructed since 1991 before accomplishing the first criticality on November 10, 1998. Rise to power tests of the HTTR started in September, 1999 and the rated thermal power of 30 MW and the reactor outlet coolant temperature of 850 degC was attained in December 2001. JAERI received the certificate of pre-operation test, that is, the commissioning license for the HTTR in March 2002. This report summarizes operation, tests, maintenance, radiation control, and construction of components and facilities for the HTTR as well as R and Ds on HTGRs from FY1999 to 2001. (author)

  19. Reliability-Based Electronics Shielding Design Tools

    Science.gov (United States)

    Wilson, J. W.; O'Neill, P. J.; Zang, T. A.; Pandolf, J. E.; Tripathi, R. K.; Koontz, Steven L.; Boeder, P.; Reddell, B.; Pankop, C.

    2007-01-01

    Shielding design on large human-rated systems allows minimization of radiation impact on electronic systems. Shielding design tools require adequate methods for evaluation of design layouts, guiding qualification testing, and adequate follow-up on final design evaluation.

  20. Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests

    Energy Technology Data Exchange (ETDEWEB)

    Braley, Jenifer C.; Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2011-06-15

    Scoping tests to evaluate the effects of agitation and pH adjustment on simulant sludge agglomeration and uranium metal oxidation at {approx}95 C were performed under Test Instructions(a,b) and as per sections 5.1 and 5.2 of this Test Plan prepared by AREVA. (c) The thermal testing occurred during the week of October 4-9, 2010. The results are reported here. For this testing, two uranium-containing simulant sludge types were evaluated: (1) a full uranium-containing K West (KW) container sludge simulant consisting of nine predominant sludge components; (2) a 50:50 uranium-mole basis mixture of uraninite [U(IV)] and metaschoepite [U(VI)]. This scoping study was conducted in support of the Sludge Treatment Project (STP) Phase 2 technology evaluation for the treatment and packaging of K-Basin sludge. The STP is managed by CH2M Hill Plateau Remediation Company (CHPRC) for the U.S. Department of Energy. Warm water ({approx}95 C) oxidation of sludge, followed by immobilization, has been proposed by AREVA and is one of the alternative flowsheets being considered to convert uranium metal to UO{sub 2} and eliminate H{sub 2} generation during final sludge disposition. Preliminary assessments of warm water oxidation have been conducted, and several issues have been identified that can best be evaluated through laboratory testing. The scoping evaluation documented here was specifically focused on the issue of the potential formation of high strength sludge agglomerates at the proposed 95 C process operating temperature. Prior hydrothermal tests conducted at 185 C produced significant physiochemical changes to genuine sludge, including the formation of monolithic concretions/agglomerates that exhibited shear strengths in excess of 100 kPa (Delegard et al. 2007).

  1. Digital System Reliability Test for the Evaluation of safety Critical Software of Digital Reactor Protection System

    Directory of Open Access Journals (Sweden)

    Hyun-Kook Shin

    2006-08-01

    Full Text Available A new Digital Reactor Protection System (DRPS based on VME bus Single Board Computer has been developed by KOPEC to prevent software Common Mode Failure(CMF inside digital system. The new DRPS has been proved to be an effective digital safety system to prevent CMF by Defense-in-Depth and Diversity (DID&D analysis. However, for practical use in Nuclear Power Plants, the performance test and the reliability test are essential for the digital system qualification. In this study, a single channel of DRPS prototype has been manufactured for the evaluation of DRPS capabilities. The integrated functional tests are performed and the system reliability is analyzed and tested. The results of reliability test show that the application software of DRPS has a very high reliability compared with the analog reactor protection systems.

  2. Continuous-flow stirred-tank reactor 20-L demonstration test: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D.D.; Collins, J.L.

    2000-02-01

    One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test using the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required.

  3. Analysis of Removal Alternatives for the Heavy Water Components Test Reactor at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1996-08-01

    This engineering study was developed to evaluate different options for decommissioning of the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site. This document will be placed in the DOE-SRS Area reading rooms for a period of 30 days in order to obtain public input to plans for the demolition of HWCTR.

  4. Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lowry, N.J.

    1998-10-21

    This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints.

  5. Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

    2010-01-01

    A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

  6. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-10-01

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition.

  7. Finite Element Simulation of Three Full-Scale Crash Tests for Cessna 172 Aircraft

    Science.gov (United States)

    Mason, Brian H.; Warren, Jerry E., Jr.

    2017-01-01

    The NASA Emergency Locator Transmitter Survivability and Reliability (ELT-SAR) project was initiated in 2013 to assess the crash performance standards for the next generation of emergency locator transmitter (ELT) systems. Three Cessna 172 aircraft were acquired to perform crash testing at NASA Langley Research Center's Landing and Impact Research Facility. Full-scale crash tests were conducted in the summer of 2015 and each test article was subjected to severe, but survivable, impact conditions including a flare-to-stall during emergency landing, and two controlled-flight-into-terrain scenarios. Full-scale finite element analyses were performed using a commercial explicit solver, ABAQUS. The first test simulated impacting a concrete surface represented analytically by a rigid plane. Tests 2 and 3 simulated impacting a dirt surface represented analytically by an Eulerian grid of brick elements using a Mohr-Coulomb material model. The objective of this paper is to summarize the test and analysis results for the three full-scale crash tests. Simulation models of the airframe which correlate well with the tests are needed for future studies of alternate ELT mounting configurations.

  8. Study on Electrically Conductive Elastomer Shielding Efficiency for HIRF Condition of Civil Aircraft Electronics Equipment Bay%民用飞机设备舱舱门导电橡胶屏蔽效能研究

    Institute of Scientific and Technical Information of China (English)

    王永根; 代继刚; 王乐意

    2014-01-01

    通过理论计算和软件仿真相结合的方法,给出了民用飞机电子电气设备舱高强度辐射场( HIRF)条件下的腔体谐振特性,并对腔体内部电磁场进行仿真。通过对比舱门缝隙导电橡胶屏蔽效能的仿真结果,给出了电子电气设备舱不同状态下的电磁场分布结果,为电子电气舱舱门屏蔽设计提供技术支持。%Resonance Characteristic of the civil aircraft electronics equipment bay ( E/E Bay ) on HIRF condition is obtained through combining the theory calculation with software simulation method .Compa-ring the simulation results of the electrically conductive elastomer shield efficiency , the electromagnetic field inside the E/E bay are supplied ,and this simulation results can support the E/E bay door design .

  9. A new method to determine dynamically equivalent finite element models of aircraft structures from modal test data

    Science.gov (United States)

    Karaağaçlı, Taylan; Yıldız, Erdinç N.; Nevzat Özgüven, H.

    2012-08-01

    Flutter analysis is a major requirement to predict safe flight envelops and to decide on flutter testing conditions of newly designed or modified aircraft structures. In order to achieve reliable flutter analysis of an aircraft structure, it is necessary to obtain a good correlation between its finite element (FE) model and experimental modal data. Currently available model updating methods require construction of a detailed initial FE model in order to achieve convergence of the modes obtained from updated FE model to their experimental counterparts. If the updating procedure is not carried out by the original design team of the aircraft structure but a subsidiary company that makes certain modification on it, construction of an appropriate initial FE model from scratch becomes a tedious task requiring considerable amount of engineering work. To overcome the foregoing problem, this paper presents a new method that aims to derive dynamically equivalent FE model of an aircraft structure directly from its experimental modal data. The application of the method is illustrated with two case studies. In the first case study, the performance of the method is tested with the modal test data of a benchmark structure built to simulate dynamic behavior of an airplane, namely GARTEUR SM-AG 19 test bed, and very satisfactory results are obtained: the first 10 elastic FE modes of the test bed closely correlate with experimental data. In the second case study, the method is applied to the modal test data obtained from ground vibration test (GVT) of a real aircraft. In this application, it is observed that only the first 4 modes of the resultant FE model correlate well with experimental data. It is concluded that the method suggested works perfectly well for simple structures like GARTEUR test bed, and it gives quite promising results when applied to real aircraft structures.

  10. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the

  11. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately.

  12. Tests of Lorentz and CPT Violation in the Medium Baseline Reactor Antineutrino Experiment

    CERN Document Server

    Li, Yu-Feng

    2014-01-01

    Tests of Lorentz and CPT violation in the medium baseline reactor antineutrino experiment are presented in the framework of the Standard Model Extension (SME). Both the spectral distortion and sidereal variation are employed to derive the limits of Lorentz violation (LV) coefficients. We do the numerical analysis of the sensitivity of LV coefficients by taking the Jiangmen Underground Neutrino Observatory (JUNO) as an illustration, which can improve the sensitivity by more than two orders of magnitude compared with the current limits from reactor antineutrino experiments.

  13. Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine

    Energy Technology Data Exchange (ETDEWEB)

    Reilly, Raymond W.

    2012-07-30

    This project, Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine was established at the Kharkiv Institute of Physics and Technology (KIPT). The associated CRADA was established with Campbell Applied Physics (CAP) located in El Dorado Hills, California. This project extends an earlier project involving both CAP and KIPT conducted under a separate CRADA. The initial project developed the basic Plasma Chemical Reactor (PCR) for generation of ozone gas. This project built upon the technology developed in the first project, greatly enhancing the output of the PCR while also improving reliability and system control.

  14. Mechanical properties of reactor pressure vessel steels studied by static and dynamic torsion tests

    Science.gov (United States)

    Munier, A.; Maamouri, M.; Schaller, R.; Mercier, O.

    1993-06-01

    Internal friction measurements and torsional plastic deformation tests have been performed in reactor pressure vessel steels (unirradiated, irradiated and irradiated/annealed specimens). The results of these experiments have been interpreted with help of transmission electron microscopy observations (conventional and in situ). It is shown how the interactions between screw dislocations and obstacles (Peierls valleys, impurities and precipitates) could explain the low temperature hardening and the irradiation embrittlement of ferritic steels. In addition, it appears that the nondestructive internal friction technique could be used advantageously to follow the evolution of the material properties under irradiation, as for instance the irradiation embrittlement of the reactor pressure vessel steels.

  15. Pilot-plant testing of magnetic filters for the N-Reactor primary cooling circuit

    Energy Technology Data Exchange (ETDEWEB)

    Emory, B.B.

    1982-01-01

    Data obtained during the laboratory loop test program using the high power HGMF indicates that removal efficiency for /sup 60/Co and subsequently the bulk of the crud, will be greater than 90% at field strength above .1 Tesla for the expanded metal mesh matrix. However, since /sup 54/Mn seems to exhibit paramagnetic behavior and the possibility of quantities of alpha iron forming during reactor shut down from oxygen inleakage, a field strength of .5 to 1 Tesla may be more appropriate for a full scale on-reactor installation. Crud loading of 50 gm per kg of matrix weight are readily obtainable and up to twice that amount has been reached.

  16. Development and testing of the Perseus proof-of-concept aircraft. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Langford, J.S. [Aurora Flight Sciences Corp., Manassas, VA (United States)

    1993-02-26

    Many areas of global climate change research could benefit from a flexible, affordable, and near-term platform that could provide in situ measurements in the upper troposphere and lower stratosphere. To provide such a capability, the Perseus unmanned science research aircraft was proposed in 1989. As a first step toward the development of Perseus, a proof-of-concept (POC) demonstrator was constructed and tested during 1990 and 1991. The POC was a full scale Perseus airframe intended to validate the structural, aerodynamic, and flight control technologies for the Perseus within a total budget of about $1.5 million. Advanced propulsion systems needed for the operational Perseus were not covered in the POC program due to funding limitations. This report documents the design, development, and testing of the Perseus POC.

  17. Dynamic Flight Simulation of aircraft and its comparison to Flight tests

    Directory of Open Access Journals (Sweden)

    Reza Khaki

    2015-09-01

    Full Text Available Nowadays obtaining data for air vehicles researches and analyses is very expensive and risky through the flight tests. Therefore using flight simulation is usually used for the mentioned researches by aerospace science researchers. In this paper, dynamic flight simulation has been performed by airplane nonlinear equations modelling. In these equations, aerodynamic coefficients and stability derivatives have an important role. Therefore, the stability derivatives for typical aircraft are calculated on various flight conditions by analytical and numerical methods. Flight conditions include of Mach number, altitude, angle of attack, control surfaces and CG position variations. The obtained derivatives are used in the form of look up table for dynamic flight simulation and virtual flight. In order to validate the simulation results, the under investigation maneuvres parameters are recorded during many real flights. The obtained data from flight tests are compared with the outputs of flight simulations. The results indicate that less than 13% differences are found in different parts of the maneuvres.

  18. Two-Compartment Photoelectrochemical Reactors Tested under various solar Light Concentration ratios

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez-Ibanez, P.; Malato, S. [Plataforma Solar de Almeria. CIEMAT (Spain)

    1999-07-01

    A new type of photo reactor, made of a cylindrical photo anode placed around an inner comportment, has been adapted to the compound parabolic (CPC's) and to the parabolic trough (PTC, Helioman) solar collectors. The photoelectrochemical performances of such two-compartment photo reactors are noticeably improved with respect to those previously obtained with photo reactors having flat photoanodes and a single compartment. The abatement of model pollutants shows up to threefold higher organic oxidation rates compared to Ti O{sub 2} slurries tested in the same experimental conditions. Clearly, charge separation is much better when an external electrochemical bias is applied to Ti/Ti O{sub 2} photoanodes under irradiation. (Author) 8 refs.

  19. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    Energy Technology Data Exchange (ETDEWEB)

    Agarwal, Vivek [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Human Factors, Controls, and Statistics; Smith, James A. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design; Jewell, James Keith [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design

    2015-02-01

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  20. Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

    Science.gov (United States)

    Garber, Anne E.; Dickens, Ricky E.

    2011-01-01

    The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

  1. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

  2. Proceedings of the 1990 International Meeting on Reduced Enrichment for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The global effort to reduce, and possibly, eliminate the international traffic in highly-enriched uranium caused by its use in research reactors requires extensive cooperation and free exchange of information among all participants. To foster this free exchange of information, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the thirteenth of a series which began in 1978. The common effort brought together, past, a large number of specialists from many countries. On hundred twenty-three participants from 26 countries, including scientists, reactor operators, and personnel from commercial fuel suppliers, research centers, and government organizations, convened in Newport, Rhode Island to discuss their results, their activities, and their plans relative to converting research reactors to low-enriched fuels. As more and more reactors convert to the use of low-enriched uranium, the emphasis of our effort has begun to shift from research and development to tasks more directly related to implementation of the new fuels and technologies that have been developed, and to refinements of those fuels and technologies. It is appropriate, for this reason, that the emphasis of this meeting was placed on safety and on conversion experiences. This individual papers in this report have been cataloged separately.

  3. Creating a Test Validated Structural Dynamic Finite Element Model of the Multi-Utility Technology Test Bed Aircraft

    Science.gov (United States)

    Pak, Chan-Gi; Truong, Samson S.

    2014-01-01

    Small modeling errors in the finite element model will eventually induce errors in the structural flexibility and mass, thus propagating into unpredictable errors in the unsteady aerodynamics and the control law design. One of the primary objectives of Multi Utility Technology Test Bed, X-56A, aircraft is the flight demonstration of active flutter suppression, and therefore in this study, the identification of the primary and secondary modes for the structural model tuning based on the flutter analysis of X-56A. The ground vibration test validated structural dynamic finite element model of the X-56A is created in this study. The structural dynamic finite element model of the X-56A is improved using a model tuning tool. In this study, two different weight configurations of the X-56A have been improved in a single optimization run.

  4. The effect of biological shielding on fast neutron and photon transport in the VVER-1000 mock-up model placed in the LR-0 reactor.

    Science.gov (United States)

    Košťál, Michal; Cvachovec, František; Milčák, Ján; Mravec, Filip

    2013-05-01

    The paper is intended to show the effect of a biological shielding simulator on fast neutron and photon transport in its vicinity. The fast neutron and photon fluxes were measured by means of scintillation spectroscopy using a 45×45 mm(2) and a 10×10 mm(2) cylindrical stilbene detector. The neutron spectrum was measured in the range of 0.6-10 MeV and the photon spectrum in 0.2-9 MeV. The results of the experiment are compared with calculations. The calculations were performed with various nuclear data libraries.

  5. The 5th surveillance testing for Kori unit 2 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-03-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules V, R, P, T and N are 2.837E+18, 1.105E+19, 2.110E+19, 3.705E+19 and 4.831E+19n/cm{sup 2}, respectively. The bias factor, the ratio of measurement/calculation, was 0.918 for the 1st through 5th testing and the calculational uncertainty, 11.6% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.898E+19n/cm{sup 2} based on the end of 15th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 4.203E+19, 5.232E+19, 6.262E+19 and 7.291E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 49 refs., 35 figs., 48 tabs. (Author)

  6. The 4th surveillance testing for Kori unit 3 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-10-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

  7. The RES reactor, a test reactor for naval propulsion; Le reacteur d'essais RES, reacteur d'essais de la propulsion navale

    Energy Technology Data Exchange (ETDEWEB)

    Pivet, S. [CEA Bruyeres-le-Chatel, 91 (France); Minguet, J.L. [AREVA-Technicatome, 13 - Aix en Provence (France)

    2005-07-01

    The RES, the new test reactor for naval propulsion, will replace the RNG that nears the end of its operating life after 30 years in service. The main asset of a land-based installation is to provide an in-core instrumented reactor while the on-board system must stay as simple as possible for robustness reasons. The objective of the RES is fivefold: 1) to foresee and help solving problems likely to happen on on-board reactor, 2) to validate nuclear fuels and reactor systems for naval propulsion, 3) to validate reactor system and equipment for the Barracuda submarine program, 4) to upgrade the on-ground facility located at Cadarache, and 5) to provide the Cea with a new capacity for the storing of spent fuels from naval propulsion systems and from Cea research reactors. The RES facility is made of 2 parts: one that houses the reactor and the other that is dedicated to the handling on spent fuels, their examination through a gamma spectrometry bench and their storing in a pool. The RES facility is scheduled to open in 2009. (A.C.)

  8. Subsonic Ultra Green Aircraft Research: Phase II- Volume III-Truss Braced Wing Aeroelastic Test Report

    Science.gov (United States)

    Bradley, Marty K.; Allen, Timothy J.; Droney, Christopher

    2014-01-01

    This Test Report summarizes the Truss Braced Wing (TBW) Aeroelastic Test (Task 3.1) work accomplished by the Boeing Subsonic Ultra Green Aircraft Research (SUGAR) team, which includes the time period of February 2012 through June 2014. The team consisted of Boeing Research and Technology, Boeing Commercial Airplanes, Virginia Tech, and NextGen Aeronautics. The model was fabricated by NextGen Aeronautics and designed to meet dynamically scaled requirements from the sized full scale TBW FEM. The test of the dynamically scaled SUGAR TBW half model was broken up into open loop testing in December 2013 and closed loop testing from January 2014 to April 2014. Results showed the flutter mechanism to primarily be a coalescence of 2nd bending mode and 1st torsion mode around 10 Hz, as predicted by analysis. Results also showed significant change in flutter speed as angle of attack was varied. This nonlinear behavior can be explained by including preload and large displacement changes to the structural stiffness and mass matrices in the flutter analysis. Control laws derived from both test system ID and FEM19 state space models were successful in suppressing flutter. The control laws were robust and suppressed flutter for a variety of Mach, dynamic pressures, and angle of attacks investigated.

  9. Advanced turboprop testbed systems study. Volume 1: Testbed program objectives and priorities, drive system and aircraft design studies, evaluation and recommendations and wind tunnel test plans

    Science.gov (United States)

    Bradley, E. S.; Little, B. H.; Warnock, W.; Jenness, C. M.; Wilson, J. M.; Powell, C. W.; Shoaf, L.

    1982-01-01

    The establishment of propfan technology readiness was determined and candidate drive systems for propfan application were identified. Candidate testbed aircraft were investigated for testbed aircraft suitability and four aircraft selected as possible propfan testbed vehicles. An evaluation of the four candidates was performed and the Boeing KC-135A and the Gulfstream American Gulfstream II recommended as the most suitable aircraft for test application. Conceptual designs of the two recommended aircraft were performed and cost and schedule data for the entire testbed program were generated. The program total cost was estimated and a wind tunnel program cost and schedule is generated in support of the testbed program.

  10. Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

    2010-06-01

    2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has been restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed

  11. Non destructive testing of irradiated fuel assemblies at the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Jose Eduardo Rosa da; Terremoto, Luis Antonio Albiac; Castanheira, Myrthes; Teodoro, Celso Antonio; Silva, Antonio Teixeira e; Damy, Margaret de Almeida; Lucki, Georgi [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: jersilva@ipen.br; laaterre@ipen.br; myrthes@ipen.br; cteodoro@ipen.br; teixeira@ipen.br; madamy@ipen.br; glucki@ipen.br

    2007-07-01

    Fuel performance and nuclear fuel qualification require a post-irradiation analysis. Non-destructive methods are utilised both in irradiated fuel storage pools and in hot-cells laboratories. As Brazil does not have hot-cells facilities for post-irradiation analysis, a qualification program for the Material Testing Reactor (MTR) fuel elements made at IPEN/CNEN-SP was adopted, based on non-destructive tests. The IPEN Fuel Engineering Group - CENC developed basic facilities for fuels post-irradiated analysis inside the reactor pool, which gives indications of: general state, by visual inspection; the integrity of the irradiated fuel cladding, by sipping tests; thickness measurements of the fuel miniplates during the irradiation time, for swelling evaluation; and, local burn-up evaluation by gamma spectrometry along the active area of the fuel element. This work describes that facilities, equipment and examples of some irradiated fuels analysis performed. (author)

  12. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Quiping [The Ohio State Univ., Columbus, OH (United States); Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Chtistensen, Richard [The Ohio State Univ., Columbus, OH (United States); Blue, Thomas [The Ohio State Univ., Columbus, OH (United States); Yoder, Graydon [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-08

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  13. In-reactor tests of the nuclear light bulb rocket concept

    Science.gov (United States)

    Gauntt, R. O.; Slutz, S. A.; Latham, T. S.; Roman, W. C.; Rogers, R. J.

    1992-07-01

    An overview is given of the closed-cycle Gas Core Nuclear Rocket outlining scenarios for its use in short-duration Mars missions and results of Nuclear Light Bulb (NLB) tests. Isothermal and nonnuclear tests are described which confirmed the fundamental concepts behind the NLB. NLB reference-engine performance characteristics are given for hypothetical engines that could be used for manned Mars missions. Vehicle/propulsion sizing is based on a Mars mission with three trans-Mars impulse burns, capture and escape burns, and a total mission duration of 600 days. The engine would have a specific impulse of 1870 seconds, a 412-kN thrust, and a thrust/weight ratio of 1.3. Reactor tests including small-scale in-reactor tests are shown to be prerequisites for studying: (1) fluid mechanical confinement of the gaseous nuclear fuel; (2) buffer gas separation and circulation; and (3) the minimization of transparent wall-heat loading. The reactor tests are shown to be critical for establishing the feasibility of the NLB concept.

  14. The 5th surveillance testing for Kori unit 1 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-08-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed primarily by Korea Atomic Energy Research Institute and Westinhouse corporation partially involved in testing and calculation data evaluation in order to obtain reliable test result. Fast neutron fluences for capsule V, T, S, R and P were 5.087E+18, 1.115E+19, 1.228E+19, 2.988E+19, and 3.938E+19n/cm2, respectively. The bias factor, the ratio of calculation/measurement, was 0.940 for the 1st through 5th testing and the calculational uncertainty, 7% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.9846E+19n/cm{sup 2} based on the end of 17th fuel cycle and it was predicted that the fluences of vessel inside surface at 24, 32, 40 and 48EFPY would reach 3.0593E+19, 4.0695E+19, 5.0797E+19 and 6.0900E+19n/cm{sup 2} based on the current calculation. PTS analysis for Kori unit 1 showed that 27.93EFPY was the threshold value for 300 deg F requirement. 71 refs., 33 figs., 52 tabs. (Author)

  15. Light Water Reactor Sustainability Program Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith; David Schwieder; Cherie Phelan; Anh Bui; Paul Bayless

    2012-08-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. This report describes the RISMC methodology demonstration where the Advanced Test Reactor (ATR) was used as a test-bed for purposes of determining safety margins. As part of the demonstration, we describe how both the thermal-hydraulics and probabilistic safety calculations are integrated and used to quantify margin management strategies.

  16. Comparison of analysis and flight test data for a drone aircraft with active flutter suppression

    Science.gov (United States)

    Newsom, J. R.; Pototzky, A. S.

    1981-01-01

    A drone aircraft equipped with an active flutter suppression system is considered with emphasis on the comparison of modal dampings and frequencies as a function of Mach number. Results are presented for both symmetric and antisymmetric motion with flutter suppression off. Only symmetric results are given for flutter suppression on. Frequency response functions of the vehicle are presented from both flight test data and analysis. The analysis correlation is improved by using an empirical aerodynamic correction factor which is proportional to the ratio of experimental to analytical steady-state lift curve slope. The mathematical models are included and existing analytical techniques are described as well as an alternative analytical technique for obtaining closed-loop results.

  17. Proving test on the seismic reliability of nuclear power plant: PWR reactor containment vessel

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Hiroshi; Yoshikawa, Teiichi; Ohno, Tokue; Yoshikawa, Eiji.

    1989-01-01

    Seismic reliability proving tests of nuclear power plant facilities are carried out by the Nuclear Power Engineering Test Center, using the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry. In 1982, the seismic reliability proving test of a PWR containment vessel was conducted using a test component of reduced scale 1/3.7. As a result of this test, the test component proved to have structural soundness against earthquakes, and at the same time its stable function was proved by leak tests which were carried out before and after the vibration test. In 1983, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. The seismic analysis and evaluation on the actual containment vessel were then performed using these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed.

  18. Nonlinear analysis of commercial aircraft impact on a reactor building—Comparison between integral and decoupled crash simulation

    Energy Technology Data Exchange (ETDEWEB)

    Siefert, A., E-mail: siefert@woelfel.de; Henkel, F.O.

    2014-04-01

    Since 9/11, the crash of a commercial aeroplane on the reactor building of a nuclear power plant is a realistic design scenario. Before that the structural behaviour under a crash of a military plane was investigated by a procedure using load-time functions (Riera, 1968). Thereby, the computation of the load-time-function was based on a conceptional model considering the main stiffness parts and masses by discrete elements. With respect to the homogeneous structural set-up of a military plane, the application of this model and the derived load-time-function applied as lumped load case seems very feasible. Contrary thereto the structural set-up of a commercial aeroplane, with e.g. the high mass concentration of the turbine or the high stiffness of the wing box compared to other parts, is different. This can be counteracted by using a more detailed finite element (FE) model for the computation of the load-time-function and by dividing the load case for the reactor building in different main load zones. Although this represents a more detailed investigation, the procedure of using a load-time-function still has the disadvantage to separate the real scenario into two steps. Thereby, the direct interaction between the structure and the aeroplane including all softening effects due to material respectively structural compliances is neglected. This leads to the general conclusion that by applying load-time-functions the results are conservative compared to the real behaviour. Due to the increased capabilities of numerical software solutions it is also possible nowadays to carry out integral crash simulations, combining all effects within one simulation. Compared to the procedure of using load-time-functions, the numerical complexity and therefore the amount of work for this integral method are increased. Within this paper both procedures (load-time function by detailed FE-model and the integral method) are exemplarily compared to each other by a crash analysis of an

  19. Criticality Safety Evaluation for the Advanced Test Reactor U-Mo Demonstration Elements

    Energy Technology Data Exchange (ETDEWEB)

    Leland M. Montierth

    2010-12-01

    The Reduced Enrichment Research Test Reactors (RERTR) fuel development program is developing a high uranium density fuel based on a (LEU) uranium-molybdenum alloy. Testing of prototypic RERTR fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. Two RERTR-Full Size Demonstration fuel elements based on the ATR-Reduced YA elements (all but one plate fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). The two fuel elements will be irradiated in alternating cycles such that only one element is loaded in the reactor at a time. Existing criticality analyses have analyzed Standard (HEU) ATR elements (all plates fueled) from which controls have been derived. This criticality safety evaluation (CSE) documents analysis that determines the reactivity of the Demonstration fuel elements relative to HEU ATR elements and shows that the Demonstration elements are bound by the Standard HEU ATR elements and existing HEU ATR element controls are applicable to the Demonstration elements.

  20. Gas-cooled reactor programs: high-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1981

    Energy Technology Data Exchange (ETDEWEB)

    1982-06-01

    Information is presented concerning HTGR chemistry; fueled graphite development; irradiation services for General Atomic Company; prestressed concrete pressure vessel development; HTGR structural materials; graphite development; high-temperature reactor physics studies; shielding studies; component flow test loop studies; core support performance test; and application and project assessments.

  1. Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoder Jr, Graydon L [ORNL; Elkassabgi, Yousri M. [Texas A& M University, Kingsville; De Leon, Gerardo I. [Texas A& M University, Kingsville; Fetterly, Caitlin N. [Texas A& M University, Kingsville; Ramos, Jorge A. [Texas A& M University, Kingsville; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)

    2012-02-01

    Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental

  2. A preliminary neutronic evaluation of high temperature engineering test reactor using the SCALE6 code

    Science.gov (United States)

    Tanure, L. P. A. R.; Sousa, R. V.; Costa, D. F.; Cardoso, F.; Veloso, M. A. F.; Pereira, C.

    2014-02-01

    Neutronic parameters of some fourth generation nuclear reactors have been investigated at the Departamento de Engenharia Nuclear/UFMG. Previous studies show the possibility to increase the transmutation capabilities of these fourth generation systems to achieve significant reduction concerning transuranic elements in spent fuel. To validate the studies, a benchmark on core physics analysis, related to initial testing of the High Temperature Engineering Test Reactor and provided by International Atomic Energy Agency (IAEA) was simulated using the Standardized Computer Analysis for Licensing Evaluation (SCALE). The CSAS6/KENO-VI control sequence and the 44-group ENDF/B-V 0 cross-section neutron library were used to evaluate the keff (effective multiplication factor) and the result presents good agreement with experimental value.

  3. Feasibility of conducting a dynamic helium charging experiment for vanadium alloys in the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.; Strain, R.V.; Smith, D.L. [Argonne National Lab., IL (United States); Matsui, H. [Tohoku Univ. (Japan)

    1996-10-01

    The feasibility of conducting a dynamic helium charging experiment (DHCE) for vanadium alloys in the water-cooled Advanced Test Reactor (ATR) is being investigated as part of the U.S./Monbusho collaboration. Preliminary findings suggest that such an experiment is feasible, with certain constraints. Creating a suitable irradiation position in the ATR, designing an effective thermal neutron filter, incorporating thermocouples for limited specimen temperature monitoring, and handling of tritium during various phases of the assembly and reactor operation all appear to be feasible. An issue that would require special attention, however, is tritium permeation loss through the capsule wall at the higher design temperatures (>{approx}600{degrees}C). If permeation is excessive, the reduced amount of tritium entering the test specimens would limit the helium generation rates in them. At the lower design temperatures (<{approx}425{degrees}C), sodium, instead of lithium, may have to be used as the bond material to overcome the tritium solubility limitation.

  4. Field test and evaluation of the passive neutron coincidence collar for prototype fast reactor fuel subassemblies

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, H.O.; Keddar, A.

    1982-08-01

    The passive neutron Coincidence Collar, which was developed for the verification of plutonium content in fast reactor fuel subassemblies, has been field tested using Prototype Fast Reactor fuel. For passive applications, the system measures the /sup 240/Pu-effective mass from the spontaneous fission rate, and in addition, a self-interrogation technique is used to determine the fissile content in the subassembly. Both the passive and active modes were evaluated at the Windscale Works in the United Kingdom. The results of the tests gave a standard deviation 0.75% for the passive count and 3 to 7% for the active measurement for a 1000-s counting time. The unit will be used in the future for the verification of plutonium in fresh fuel assemblies.

  5. Task Order Number 5TS5702D035P: Testing Alternative Aircraft and Runway/Taxiway Deicers - Phase 2

    Science.gov (United States)

    2004-06-01

    Polymer Matrix Composite Materials 55 10.3 Elastomeric Materials 56 10.4 AircraftWire Insulation "" 58 10.5... Polymer Matrix Composite Shear Stress Test Results 13 Table 8. Polymer Matrix Composite Glass Transition Temperature Test Results """"""’’’’’’’’’’ 14...Table 9. Polymer Matrix Composite Barcol Hardness Test Results 16 Table 10. Polymer Matrix

  6. Highly Perturbed Operational Test Configurations at the WSMR Fast Burst Reactor

    Directory of Open Access Journals (Sweden)

    Flanders T. Michael

    2016-01-01

    Full Text Available The White Sands Missile Range (WSMR MoLLY-G reactor has a long history of producing a well characterized environment for testing electronic systems/devices in fission environments. As an unmoderated, unreflected, bare critical assembly, it provides a slightly degraded fission spectrum with a 1/E tail. For radiation hardness testing of electronics, the neutron fluence is usually reported as the 1-MeV Equivalent Neutron Fluence for Silicon. In this paper, we examine additional neutron environments and characterizations ranging from low intensity neutron fields to more extreme modifications of our normal test environment.

  7. Dynamic parameters test of Haiyang Nuclear Power Engineering in reactor areas, China

    Science.gov (United States)

    Zhou, N.; Zhao, S.; Sun, L.

    2012-12-01

    Haiyang Nuclear Power Project is located in Haiyang city, China. It consists of 6×1000MW AP1000 Nuclear Power generator sets. The dynamic parameters of the rockmass are essential for the design of the nuclear power plant. No.1 and No.2 reactor area are taken as research target in this paper. Sonic logging, single hole and cross-hole wave velocity are carried out respectively on the site. There are four types of rock lithology within the measured depth. They are siltstone, fine sandstone, shale and allgovite. The total depth of sonic logging is 409.8m and 2049 test points. The sound wave velocity of the rocks are respectively 5521 m/s, 5576m/s, 5318 m/s and 5576 m/s. Accroding to the statistic data, among medium weathered fine sandstone, fairly broken is majority, broken and relatively integrity are second, part of integrity. Medium weathered siltstone, relatively integrity is mojority, fairly broken is second. Medium weathered shale, fairly broken is majority, broken and relatively integrity for the next and part of integrity. Slight weathered fine sandstone, siltstone, shale and allgovite, integrity is the mojority, relatively integrity for the next, part of fairly broken.The single hole wave velocity tests are set in two boreholesin No.1 reactor area and No.2 reactor area respectively. The test depths of two holes are 2-24m, and the others are 2-40m. The wave velocity data are calculated at different depth in each holes and dynamic parameters. According to the test statistic data, the wave velocity and the dynamic parameter values of rockmass are distinctly influenced by the weathering degree. The test results are list in table 1. 3 groups of cross hole wave velocity tests are set for No.1 and 2 reactor area, No.1 reactor area: B16, B16-1, B20(Direction:175°, depth: 100m); B10, B10-1, B11(269°, 40m); B21, B21-1, B17(154°, 40m); with HB16, HB10, HB21 as trigger holes; No.2 reactor area: B47, B47-1, HB51(176°, 100m); B40, B40-1, B41(272°, 40m); B42, B42-1, B

  8. After Action Report: Advanced Test Reactor Complex 2015 Evaluated Drill October 6, 2015

    Energy Technology Data Exchange (ETDEWEB)

    Holmes, Forest Howard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-11-01

    The Advanced Test Reactor (ATR) Complex, operated by Battelle Energy Alliance, LLC, at the Idaho National Laboratory (INL) conducted an evaluated drill on October 6, 2015, to allow the ATR Complex emergency response organization (ERO) to demonstrate the ability to respond to and mitigate an emergency by implementing the requirements of DOE O 151.1C, “Comprehensive Emergency Management System.”

  9. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

  10. Design, fabrication, and properties of a continuous carbon-fiber reinforced Sm{sub 2}O{sub 3}/polyimide gamma ray/neutron shielding material

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Peng [Department of Nuclear Science and Engineering, Nanjing University of Aeronautics and Astronautics, Nanjing 210016 (China); Tang, Xiaobin, E-mail: tangxiaobin@nuaa.edu.cn [Department of Nuclear Science and Engineering, Nanjing University of Aeronautics and Astronautics, Nanjing 210016 (China); Jiangsu Key Laboratory of Nuclear Energy Equipment Materials Engineering, Nanjing 210016 (China); Chai, Hao [Department of Nuclear Science and Engineering, Nanjing University of Aeronautics and Astronautics, Nanjing 210016 (China); Chen, Da [Department of Nuclear Science and Engineering, Nanjing University of Aeronautics and Astronautics, Nanjing 210016 (China); Jiangsu Key Laboratory of Nuclear Energy Equipment Materials Engineering, Nanjing 210016 (China); Qiu, Yunlong [ZhongXing Energy Equipment Co., Ltd, Haimen, Nantong 226100 (China)

    2015-12-15

    Highlights: • Sm{sub 2}O{sub 3} is used for neutron absorber instead of B{sub 4}C, and Sm{sub 2}O{sub 3} has a good photon-shielding effect. • Carbon-fiber cloth and polyimide were used to enhance shielding materials’ mechanical behavior and thermal behavior. • Both Monte Carlo method and shielding test were used to evaluate shielding performance of the novel shielding material. - Abstract: The design and fabrication of shielding materials with good heat-resistance and mechanical properties is a major problem in the radiation shielding field. In this paper, based on gamma ray and neutron shielding theory, a continuous carbon-fiber reinforced Sm{sub 2}O{sub 3}/polyimide gamma ray/neutron shielding material was fabricated by hot-pressing method. The material's application behavior was subsequently evaluated using neutron shielding, photon shielding, mechanical tensile, and thermogravimetric analysis–differential scanning calorimetry tests. The results show that the tensile strength of the novel shielding material exceeds 200 MPa, which makes it of similar strength to aluminum alloy. The material does not undergo crosslinking and decomposition reactions at 300 °C and it can be used in such environments for long periods of time. The continuous carbon-fiber reinforced Sm{sub 2}O{sub 3}/polyimide material has a good shielding performance with respect to gamma rays and neutrons. The material thus has good prospects for use in fusion reactor system and nuclear waste disposal applications.

  11. Flight tests of the total automatic flight control system (Tafcos) concept on a DHC-6 Twin Otter aircraft

    Science.gov (United States)

    Wehrend, W. R., Jr.; Meyer, G.

    1980-01-01

    Flight control systems capable of handling the complex operational requirements of the STOL and VTOL aircraft designs as well as designs using active control concepts are considered. Emphasis is placed on the total automatic flight control system (TACOS) (TAFCOS). Flight test results which verified the performance of the system concept are presented.

  12. Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

  13. Standard Test Method for Effects of Cleaning and Chemical Maintenance Materials on Painted Aircraft Surfaces

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This test method covers determination of the effects of cleaning solutions and liquid cleaner concentrates on painted aircraft surfaces (Note 1). Streaking, discoloration, and blistering may be determined visually. Softening is determined with a series of specially prepared pencils wherein determination of the softest pencil to rupture the paint film is made. Note 1—This test method is applicable to any paint film that is exposed to cleaning materials. MIL-PRF-85285 has been selected as a basic example. When other paint finishes are used, refer to the applicable material specification for panel preparation and system curing prior to testing. 1.2 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard. 1.3 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 ...

  14. Real-time flight test analysis and display techniques for the X-29A aircraft

    Science.gov (United States)

    Hicks, John W.; Petersen, Kevin L.

    1989-01-01

    The X-29A advanced technology demonstrator flight envelope expansion program and the subsequent flight research phase gave impetus to the development of several innovative real-time analysis and display techniques. These new techniques produced significant improvements in flight test productivity, flight research capabilities, and flight safety. These techniques include real-time measurement and display of in-flight structural loads, dynamic structural mode frequency and damping, flight control system dynamic stability and control response, aeroperformance drag polars, and aircraft specific excess power. Several of these analysis techniques also provided for direct comparisons of flight-measured results with analytical predictions. The aeroperformance technique was made possible by the concurrent development of a new simplified in-flight net thrust computation method. To achieve these levels of on-line flight test analysis, integration of ground and airborne systems was required. The capability of NASA Ames Research Center, Dryden Flight Research Facility's Western Aeronautical Test Range was a key factor to enable implementation of these methods.

  15. Results from tests, with van-mounted sensor, of magnetic leader cable for aircraft guidance during roll-out and turnoff

    Science.gov (United States)

    Young, J. C.; Bundick, W. T.; Irwin, S. H.

    1983-01-01

    Tests were conducted with a van mounted experimental magnetic leader cable sensor to evaluate its potential for measuring aircraft displacement and heading with respect to the leader cable during roll out and turnoff. Test results show that the system may be usable in measuring displacement but the heading measurement contains errors introduced by distortion of the magnetic field by the metal van or aircraft.

  16. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  17. Long term out-of-pile thermocouple tests in conditions representative for nuclear gas-cooled high temperature reactors

    OpenAIRE

    LAURIE Mathias; FOURREZ Stéphane; FUETTERER Michael; LAPETITE Jean-Marc; SADLI M.; MORICE Ronan; FAILLEAU G

    2013-01-01

    During irradiation tests at high temperature failure of commercial Inconel 600 sheathed thermocouples is commonly encountered. As instrumentation, in particular thermocouples are considered safety-relevant both for irradiation tests and for commercial reactors, JRC and THERMOCOAX joined forces to solve this issue by performing out-of-pile tests with thermocouples mimicking the environment encountered by high temperature reactor (HTR) in-core instrumentation. The objective was to screen innova...

  18. Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

    Energy Technology Data Exchange (ETDEWEB)

    Palmer, A. J.; Haggard, DC; Herter, J. W.; Swank, W. D.; Knudson, D. L.; Cherry, R. S. [Idaho National Laboratory, P.O. Box 1625, MS 4112, Idaho Falls, ID, (United States); Scervini, M. [University of Cambridge, Department of Material Science and Metallurgy, 27 Charles Babbage Road, CB3 0FS, Cambridge, (United Kingdom)

    2015-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time-dependent change in composition and, as a consequence, a time-dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature applications, previous experiences with Type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly, Type N thermocouples are expected to be only slightly affected by neutron fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly

  19. Summary of Thermocouple Performance During Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor and Out-of-Pile Thermocouple Testing in Support of Such Experiments

    Energy Technology Data Exchange (ETDEWEB)

    A. J. Palmer; DC Haggard; J. W. Herter; M. Scervini; W. D. Swank; D. L. Knudson; R. S. Cherry

    2011-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B); and tungsten-rhenium thermocouples (Types C and W). For lower temperature applications, previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of these Nickel based thermocouples is limited when the temperature exceeds 1000°C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past ten years, three long-term Advanced Gas Reactor (AGR) experiments have been conducted with measured temperatures ranging from 700oC – 1200oC. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out of pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150oC and 1200oC for 2000 hours at each temperature, followed by 200 hours at 1250oC, and 200 hours at 1300oC. The standard Type N design utilizes high purity crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including Haynes 214 alloy sheath, spinel (MgAl2O4) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard fired alumina

  20. NUMERICAL SIMULATION FOR MECHANICAL BEHAVIOR OF U10MO MONOLITHIC MINIPLATES FOR RESEARCH AND TEST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Hakan Ozaltun & Herman Shen

    2011-11-01

    This article presents assessment of the mechanical behavior of U-10wt% Mo (U10Mo) alloy based monolithic fuel plates subject to irradiation. Monolithic, plate-type fuel is a new fuel form being developed for research and test reactors to achieve higher uranium densities within the reactor core to allow the use of low-enriched uranium fuel in high-performance reactors. Identification of the stress/strain characteristics is important for understanding the in-reactor performance of these plate-type fuels. For this work, three distinct cases were considered: (1) fabrication induced residual stresses (2) thermal cycling of fabricated plates; and finally (3) transient mechanical behavior under actual operating conditions. Because the temperatures approach the melting temperature of the cladding during the fabrication and thermal cycling, high temperature material properties were incorporated to improve the accuracy. Once residual stress fields due to fabrication process were identified, solution was used as initial state for the subsequent simulations. For thermal cycling simulation, elasto-plastic material model with thermal creep was constructed and residual stresses caused by the fabrication process were included. For in-service simulation, coupled fluid-thermal-structural interaction was considered. First, temperature field on the plates was calculated and this field was used to compute the thermal stresses. For time dependent mechanical behavior, thermal creep of cladding, volumetric swelling and fission induced creep of the fuel foil were considered. The analysis showed that the stresses evolve very rapidly in the reactor. While swelling of the foil increases the stress of the foil, irradiation induced creep causes stress relaxation.

  1. Electrician, Airplane (aircraft mfg.) 825.281; Aircraft Mechanic, Armament (aircraft mfg.) 801.381 -- Technical Report on Development of USTES Aptitude Test Battery.

    Science.gov (United States)

    Manpower Administration (DOL), Washington, DC. U.S. Training and Employment Service.

    The United States Training and Employment Service General Aptitude Test Battery (GATB), first published in 1947, has been included in a continuing program of research to validate the tests against success in many different occupations. The GATB consists of 12 tests which measure nine aptitudes: General Learning Ability; Verbal Aptitude; Numerical…

  2. A Method for Evaluating Aircraft Stability Parameters from Flight Test Data

    Science.gov (United States)

    1952-06-01

    solving for the aircraft longitudinal moment of inertia. The transfer function A2 is directly proportional to the moment of inertia. All the remaining...a method for evaluating aircraft longitudinal stability derivatives from frequency response data pro- vided the linear dependency existing between the

  3. An Experimental Test Facility to Support Development of the Fluoride Salt Cooled High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoder Jr, Graydon L [ORNL; Aaron, Adam M [ORNL; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK); Fugate, David L [ORNL; Holcomb, David Eugene [ORNL; Kisner, Roger A [ORNL; Peretz, Fred J [ORNL; Robb, Kevin R [ORNL; Wilgen, John B [ORNL; Wilson, Dane F [ORNL

    2014-01-01

    The need for high-temperature (greater than 600 C) energy exchange and delivery systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The Fluoride Salt-Cooled High-Temperature Reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the Fluoride Salt Cooled High Temperature Reactor concept. The facility is capable of operating at up to 700 C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system; trace heating system; and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride salt heat transfer inside a heated pebble bed similar to that used for the core of the pebble bed advanced high-temperature reactor. This document describes the details of the loop design, auxiliary systems used to support the facility, the inductive heating system, and facility capabilities.

  4. Spacecraft Electrostatic Radiation Shielding

    Science.gov (United States)

    2008-01-01

    This project analyzed the feasibility of placing an electrostatic field around a spacecraft to provide a shield against radiation. The concept was originally proposed in the 1960s and tested on a spacecraft by the Soviet Union in the 1970s. Such tests and analyses showed that this concept is not only feasible but operational. The problem though is that most of this work was aimed at protection from 10- to 100-MeV radiation. We now appreciate that the real problem is 1- to 2-GeV radiation. So, the question is one of scaling, in both energy and size. Can electrostatic shielding be made to work at these high energy levels and can it protect an entire vehicle? After significant analysis and consideration, an electrostatic shield configuration was proposed. The selected architecture was a torus, charged to a high negative voltage, surrounding the vehicle, and a set of positively charged spheres. Van de Graaff generators were proposed as the mechanism to move charge from the vehicle to the torus to generate the fields necessary to protect the spacecraft. This design minimized complexity, residual charge, and structural forces and resolved several concerns raised during the internal critical review. But, it still is not clear if such a system is costeffective or feasible, even though several studies have indicated usefulness for radiation protection at energies lower than that of the galactic cosmic rays. Constructing such a system will require power supplies that can generate voltages 10 times that of the state of the art. Of more concern is the difficulty of maintaining the proper net charge on the entire structure and ensuring that its interaction with solar wind will not cause rapid discharge. Yet, if these concerns can be resolved, such a scheme may provide significant radiation shielding to future vehicles, without the excessive weight or complexity of other active shielding techniques.

  5. Safety evaluation for packaging (onsite) for the concrete-shielded RH TRU drum for the 327 Postirradiation Testing Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.J.

    1998-03-31

    This safety evaluation for packaging authorizes onsite transport of Type B quantities of radioactive material in the Concrete Shielded Remote-Handled Transuranic Waste (RH TRU) Drum per HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments. The drum will be used for transport of 327 Building legacy waste from the 300 Area to a solid waste storage facility on the Hanford Site.

  6. Safety evaluation for packaging (onsite) for concrete-shielded RHTRU waste drum for the 327 postirradiation testing laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Adkins, H.E.

    1996-10-29

    This safety evaluation for packaging authorizes onsite transport of Type B quantities of radioactive material in the Concrete- Shielded Remote-Handled Transuranic Waste (RH TRU) Drum per WHC-CM-2-14, Hazardous Material Packaging and Shipping. The drum will be used for transport of 327 Building legacy waste from the 300 Area to the Transuranic Waste Storage and Assay Facility in the 200 West Area and on to a Solid Waste Storage Facility, also in the 200 Area.

  7. The external gamma radiation environment from the Kiwi Phoebus, and Pewee reactors

    Science.gov (United States)

    Malenfant, R. E.

    1972-01-01

    During the past few years, ground tests of high-powered propulsion-prototype reactors have provided several opportunities to observe the external radiation environment. Reactor tests have been conducted in free air and inside of open well shields. Measurements were taken over distances ranging from contact with the pressure vessel out to greater than 5000' both during operation and after shutdown. Some measurements characteristic of each of the systems are presented and compared with results of calculations.

  8. Development of a Technique and Method of Testing Aircraft Models with Turboprop Engine Simulators in a Small-scale Wind Tunnel - Results of Tests

    Directory of Open Access Journals (Sweden)

    A. V. Petrov

    2004-01-01

    Full Text Available This report presents the results of experimental investigations into the interaction between the propellers (Ps and the airframe of a twin-engine, twin-boom light transport aircraft with a Π-shaped tail. An analysis was performed of the forces and moments acting on the aircraft with rotating Ps. The main features of the methodology for windtunnel testing of an aircraft model with running Ps in TsAGI’s T-102 wind tunnel are outlined.The effect of 6-blade Ps slipstreams on the longitudinal and lateral aerodynamic characteristics as well as the effectiveness of the control surfaces was studied on the aircraft model in cruise and takeoff/landing configurations. The tests were conducted at flow velocities of V∞ = 20 to 50 m/s in the ranges of angles of attack α =  -6 to 20 deg, sideslip angles of β = -16 to 16 deg and blade loading coefficient of B 0 to 2.8. For the aircraft of unusual layout studied, an increase in blowing intensity is shown to result in decreasing longitudinal static stability and significant asymmetry of the directional stability characteristics associated with the interaction between the Ps slipstreams of the same (left-hand rotation and the empennage.

  9. Lightning protection guidelines and test data for adhesively bonded aircraft structures

    Science.gov (United States)

    Pryzby, J. E.; Plumer, J. A.

    1984-01-01

    The highly competitive marketplace and increasing cost of energy has motivated manufacturers of general aviation aircraft to utilize composite materials and metal-to-metal bonding in place of conventional fasteners and rivets to reduce weight, obtain smoother outside surfaces and reduce drag. The purpose of this program is protection of these new structures from hazardous lightning effects. The program began with a survey of advance-technology materials and fabrication methods under consideration for future designs. Sub-element specimens were subjected to simulated lightning voltages and currents. Measurements of bond line voltages, electrical sparking, and mechanical strength degradation were made to comprise a data base of electrical properties for new technology materials and basic structural configurations. The second hase of the program involved tests on full scale wing structures which contained integral fuel tanks and which were representative of examples of new technology structures and fuel systems. The purpose of these tests was to provide a comparison between full scale structural measurements and those obtained from the sub-element specimens.

  10. Test and Evaluation Metrics of Crew Decision-Making And Aircraft Attitude and Energy State Awareness

    Science.gov (United States)

    Bailey, Randall E.; Ellis, Kyle K. E.; Stephens, Chad L.

    2013-01-01

    NASA has established a technical challenge, under the Aviation Safety Program, Vehicle Systems Safety Technologies project, to improve crew decision-making and response in complex situations. The specific objective of this challenge is to develop data and technologies which may increase a pilot's (crew's) ability to avoid, detect, and recover from adverse events that could otherwise result in accidents/incidents. Within this technical challenge, a cooperative industry-government research program has been established to develop innovative flight deck-based counter-measures that can improve the crew's ability to avoid, detect, mitigate, and recover from unsafe loss-of-aircraft state awareness - specifically, the loss of attitude awareness (i.e., Spatial Disorientation, SD) or the loss-of-energy state awareness (LESA). A critical component of this research is to develop specific and quantifiable metrics which identify decision-making and the decision-making influences during simulation and flight testing. This paper reviews existing metrics and methods for SD testing and criteria for establishing visual dominance. The development of Crew State Monitoring technologies - eye tracking and other psychophysiological - are also discussed as well as emerging new metrics for identifying channelized attention and excessive pilot workload, both of which have been shown to contribute to SD/LESA accidents or incidents.

  11. Low-temperature neutron irradiation tests of superconducting magnet materials using reactor neutrons at KUR

    Science.gov (United States)

    Yoshida, M.; Nakamoto, T.; Ogitsu, T.; Xu, Q.; Itahashi, T.; Kuno, Y.; Kuriyama, Y.; Mori, Y.; Qin, B.; Sato, A.; Sato, K.; Yoshiie, T.

    2012-06-01

    Radiation resistant superconducting magnets are required for high intensity particle accelerators and associated secondary particle beamlines, such as the LHC upgrade and the COMET experiment at J-PARC. Expected neutron fluence on the superconducting coils reaches 1021 n/m2 or higher, therefore the magnet should be designed taking into account the irradiation effects. Irradiation tests for superconducting magnet materials have been carried out using reactor neutrons at Kyoto Univ. Research Reactor Institute. As a first step of the experiment, aluminum alloy stabilizer for superconducting cable was exposed to the reactor neutrons at low temperature and the resistance has been measured in situ during neutron exposure. After the irradiation at 12 K-15 K, the sample resistance increase was proportional to the integrated neutron fluence, and reached almost double for a fast-neutron fluence of 2.3×1020 n/m2 (>0.1 MeV). It is also confirmed that the induced resistance is fully recovered by thermal cycling to room temperature. Details of the irradiation test and the prospects are described.

  12. Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

    Science.gov (United States)

    Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

    2017-01-01

    The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

  13. CALMOS: Innovative device for the measurement of nuclear heating in material testing reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carcreff, H. [Alternative Energies and Atomic Energy Commission CEA, Saclay Center, DEN/DANS/DRSN/SIREN, Gif Sur Yvette, 91191 (France)

    2011-07-01

    An R and D program has been carried out since 2002 in order to improve gamma heating measurements in the 70 MWth OSIRIS Material Testing Reactor operated by CEA's Nuclear Energy Div. at the Saclay research center. Throughout this program an innovative calorimetric probe associated to a specific handling system has been designed in order to make measurements both along the fissile height and on the upper part of the core, where nuclear heating rates still remain high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for the process validation, while a displacement system has been especially designed to move the probe axially. A final probe has been designed thanks to modeling results and to preliminary measurements obtained with mock-ups irradiated to a heating level of 2W/g, This paper gives an overview of the development, describes the calorimetric probe, and expected advantages such as the possibility to use complementary methods to get the nuclear heating measurement. Results obtained with mock-ups irradiated in ex-core area of the reactor are presented and discussed. (authors)

  14. Assessment of Feasibility of the Beneficial Use of Waste Heat from the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Donna P. Guillen

    2012-07-01

    This report investigates the feasibility of using waste heat from the Advanced Test Reactor (ATR). A proposed glycol waste heat recovery system was assessed for technical and economic feasibility. The system under consideration would use waste heat from the ATR secondary coolant system to preheat air for space heating of TRA-670. A tertiary coolant stream would be extracted from the secondary coolant system loop and pumped to a new plate and frame heat exchanger, where heat would be transferred to a glycol loop for preheating outdoor air in the heating and ventilation system. Historical data from Advanced Test Reactor operations over the past 10 years indicates that heat from the reactor coolant was available (when needed for heating) for 43.5% of the year on average. Potential energy cost savings by using the waste heat to preheat intake air is $242K/yr. Technical, safety, and logistics considerations of the glycol waste heat recovery system are outlined. Other opportunities for using waste heat and reducing water usage at ATR are considered.

  15. Operational Philosophy for the Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    J. Benson; J. Cole; J. Jackson; F. Marshall; D. Ogden; J. Rempe; M. C. Thelen

    2013-02-01

    In 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). At its core, the ATR NSUF Program combines access to a portion of the available ATR radiation capability, the associated required examination and analysis facilities at the Idaho National Laboratory (INL), and INL staff expertise with novel ideas provided by external contributors (universities, laboratories, and industry). These collaborations define the cutting edge of nuclear technology research in high-temperature and radiation environments, contribute to improved industry performance of current and future light-water reactors (LWRs), and stimulate cooperative research between user groups conducting basic and applied research. To make possible the broadest access to key national capability, the ATR NSUF formed a partnership program that also makes available access to critical facilities outside of the INL. Finally, the ATR NSUF has established a sample library that allows access to pre-irradiated samples as needed by national research teams.

  16. The development and the tests of the electrostatic probe for dust particle collection in thermonuclear reactors

    Science.gov (United States)

    Begrambekov, L. B.; Voityuk, A. N.; Zakharov, A. M.

    2016-09-01

    Formation of dust particles in thermonuclear reactors can greatly affect the plasma parameters and lead to accumulation of tritium. The rates of formation and deposition of dust need to be measured, and the parameters of formation of dust particles and clusters need to be studied. A model of a device for collection of fine conductive particles capable of removing them from the reactor chamber for future research is proposed in this paper. The dust collector's operation is based on a principle of applied electrostatic field. The model was tested in different operating conditions: in vacuum, at the atmospheric pressure in the atmosphere of air and dry nitrogen. The experiments were conducted with a stationary system and with the dust collector in motion relative to the dusty surface. It is shown that, during the probe moving relative to the surface, it can remove up to 95% of fine tungsten particles with sizes ranging from 1 to 10 μm.

  17. Testing of Passive Safety System Performance for Higher Power Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    brian G. Woods; Jose Reyes, Jr.; John Woods; John Groome; Richard Wright

    2004-12-31

    This report describes the results of NERI research on the testing of advanced passive safety performance for the Westinghouse AP1000 design. The objectives of this research were: (a) to assess the AP1000 passive safety system core cooling performance under high decay power conditions for a spectrum of breaks located at a variety of locations, (b) to compare advanced thermal hydraulic computer code predictions to the APEX high decay power test data and (c) to develop new passive safety system concepts that could be used for Generation IV higher power reactors.

  18. Supported Pd-Au Membrane Reactor for Hydrogen Production: Membrane Preparation, Characterization and Testing

    OpenAIRE

    Adolfo Iulianelli; Marjan Alavi; Giuseppe Bagnato; Simona Liguori; Jennifer Wilcox; Mohammad Reza Rahimpour; Reza Eslamlouyan; Bryce Anzelmo; Angelo Basile

    2016-01-01

    A supported Pd-Au (Au 7wt%) membrane was produced by electroless plating deposition. Permeation tests were performed with pure gas (H2, H2, N2, CO2, CH4) for long time operation. After around 400 h under testing, the composite Pd-Au membrane achieved steady state condition, with an H2/N2 ideal selectivity of around 500 at 420 °C and 50 kPa as transmembrane pressure, remaining stable up to 1100 h under operation. Afterwards, the membrane was allocated in a membrane reactor module for methane s...

  19. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

  20. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kawakami, Hiroto

    1995-02-07

    A reactor container of the present invention has a structure that the reactor container is entirely at the same temperature as that at the inlet of the reactor and, a hot pool is incorporated therein, and the reactor container has is entirely at the same temperature and has substantially uniform temperature follow-up property transiently. Namely, if the temperature at the inlet of the reactor core changes, the temperature of the entire reactor container changes following this change, but no great temperature gradient is caused in the axial direction and no great heat stresses due to axial temperature distribution is caused. Occurrence of thermal stresses caused by the axial temperature distribution can be suppressed to improve the reliability of the reactor container. In addition, since the laying of the reactor inlet pipelines over the inside of the reactor is eliminated, the reactor container is made compact and the heat shielding structures above the reactor and a protection structure of container walls are simplified. Further, secondary coolants are filled to the outside of the reactor container to simplify the shieldings. The combined effects described above can improve economical property and reliability. (N.H.).

  1. Digital virtual flight testing and evaluation method for flight characteristics airworthiness compliance of civil aircraft based on HQRM

    Institute of Scientific and Technical Information of China (English)

    Liu Fan; Wang Lixin; Tan Xiangsheng

    2015-01-01

    In order to incorporate airworthiness requirements for flight characteristics into the entire development cycle of electronic flight control system (EFCS) equipped civil aircraft, digital virtual flight testing and evaluation method based on handling qualities rating method (HQRM) is proposed. First, according to HQRM, flight characteristics airworthiness requirements of civil aircraft in EFCS failure states are determined. On this basis, digital virtual flight testing model, comprising flight task digitized model, pilot controlling model, aircraft motion and atmospheric tur-bulence model, is used to simulate the realistic process of a pilot controlling an airplane to perform assigned flight tasks. According to the simulation results, flight characteristics airworthiness com-pliance of the airplane can be evaluated relying on the relevant regulations for handling qualities (HQ) rating. Finally, this method is applied to a type of passenger airplane in a typical EFCS failure state, and preliminary conclusions concerning airworthiness compliance are derived quickly. The research results of this manuscript can provide important theoretical reference for EFCS design and actual airworthiness compliance verification of civil aircraft.

  2. Advanced Multifunctional MMOD Shield: Radiation Shielding Assessment

    Science.gov (United States)

    Rojdev, Kristina; Christiansen, Eric

    2013-01-01

    Deep space missions must contend with a harsh radiation environment Impacts to crew and electronics. Need to invest in multifunctionality for spacecraft optimization. MMOD shield. Goals: Increase radiation mitigation potential. Retain overall MMOD shielding performance.

  3. The Tower Shielding Facility: Its glorious past

    Energy Technology Data Exchange (ETDEWEB)

    Muckenthaler, F.J.

    1997-05-07

    The Tower Shielding Facility (TSF) is the only reactor facility in the US that was designed and built for radiation-shielding studies in which both the reactor source and shield samples could be raised into the air to allow measurements to be made without interference from ground scattering or other spurious effects. The TSF proved its usefulness as many different programs were successfully completed. It became active in work for the Defense Atomic Support Agency (DASA) Space Nuclear Auxiliary Power, Defense Nuclear Agency, Liquid Metal Fast Breeder Reactor Program, the Gas-Cooled and High-Temperature Gas-Cooled Reactor programs, and the Japanese-American Shielding Program of Experimental Research, just to mention a few of the more extensive ones. The history of the TSF as presented in this report describes the various experiments that were performed using the different reactors. The experiments are categorized as to the programs which they supported and placed in corresponding chapters. The experiments are described in modest detail, along with their purpose when appropriate. Discussion of the results is minimal, but references are given to more extensive topical reports.

  4. Crack propagation monitoring in a full-scale aircraft fatigue test based on guided wave-Gaussian mixture model

    Science.gov (United States)

    Qiu, Lei; Yuan, Shenfang; Bao, Qiao; Mei, Hanfei; Ren, Yuanqiang

    2016-05-01

    For aerospace application of structural health monitoring (SHM) technology, the problem of reliable damage monitoring under time-varying conditions must be addressed and the SHM technology has to be fully validated on real aircraft structures under realistic load conditions on ground before it can reach the status of flight test. In this paper, the guided wave (GW) based SHM method is applied to a full-scale aircraft fatigue test which is one of the most similar test status to the flight test. To deal with the time-varying problem, a GW-Gaussian mixture model (GW-GMM) is proposed. The probability characteristic of GW features, which is introduced by time-varying conditions is modeled by GW-GMM. The weak cumulative variation trend of the crack propagation, which is mixed in time-varying influence can be tracked by the GW-GMM migration during on-line damage monitoring process. A best match based Kullback-Leibler divergence is proposed to measure the GW-GMM migration degree to reveal the crack propagation. The method is validated in the full-scale aircraft fatigue test. The validation results indicate that the reliable crack propagation monitoring of the left landing gear spar and the right wing panel under realistic load conditions are achieved.

  5. Waste Heat Recovery from the Advanced Test Reactor Secondary Coolant Loop

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen

    2012-11-01

    This study investigated the feasibility of using a waste heat recovery system (WHRS) to recover heat from the Advanced Test Reactor (ATR) secondary coolant system (SCS). This heat would be used to preheat air for space heating of the reactor building, thus reducing energy consumption, carbon footprint, and energy costs. Currently, the waste heat from the reactor is rejected to the atmosphere via a four-cell, induced-draft cooling tower. Potential energy and cost savings are 929 kW and $285K/yr. The WHRS would extract a tertiary coolant stream from the SCS loop and pump it to a new plate and frame heat exchanger, from which the heat would be transferred to a glycol loop for preheating outdoor air supplied to the heating and ventilation system. The use of glycol was proposed to avoid the freezing issues that plagued and ultimately caused the failure of a WHRS installed at the ATR in the 1980s. This study assessed the potential installation of a new WHRS for technical, logistical, and economic feasibility.

  6. Measurement of basic thermal-hydraulic characteristics under the test facility and reactor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Eduard A Boltenko; Victor P Sharov [Elektrogorsk Research and Engineering Center, EREC, Bezimyannaja Street, 6, Elektrogorsk, Moscow Region, 142530 (Russian Federation); Dmitriy E Boltenko [State Scientific Center of Russian Federation IPPE, Bondarenko Square, Obhinsk, Kaluga Region, 249020 (Russian Federation)

    2005-07-01

    Full text of publication follows: The nuclear power of Russia is based on the reactors of two types: water-water - WWER and uranium - graphite channel RBMK. The nuclear power development is possible with performance of the basic condition - level of nuclear power plants (NPP) safety should satisfy the rigid requirements. The calculated proof of NPPs safety made by means of thermal-hydraulic codes of improved estimation, verified on experimental data is the characteristic of this level. The data for code verification can be obtained at the integral facilities simulating a circulation circuit of NPP with the basic units and intended for investigation of circuit behaviour in transient and accident conditions. For verification of mathematical models in transient and accident conditions, development of physically reasonable methods for definition of the various characteristics of two-phase flow the experimental data, as the integrated characteristics of a flow, and data on the local characteristics and structure of a flow is necessary. For safety assurance of NPP it is necessary to monitor and determine the basic thermalhydraulic characteristics of reactor facility (RF). It is possible to refer coolant flow-rate, core input and output water temperature, heat-power. The description of the EREC works in the field completion and adaptation of certain methods with reference to measurements in dynamic modes of test facility conditions and development of methods for measurements of basic thermal-hydraulic characteristics of reactor facilities is presented in the paper. (authors)

  7. Dissolution of Material and Test reactor Fuel in an H-Canyon Dissolver

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-01-26

    In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO3 flowsheet. Since the development of the existing flowsheet, improved experimental methods have been developed to more accurately characterize the offgas composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.

  8. Bench-scale reactor tests of low-temperature, catalytic gasification of wet, industrial wastes

    Energy Technology Data Exchange (ETDEWEB)

    Elliott, D.C.; Neuenschwander, G.G.; Baker, E.G.; Butner, R.S.; Sealock, L.J.

    1990-04-01

    Bench-scale reactor tests are under way at Pacific Northwest Laboratory to develop a low-temperature, catalytic gasification system. The system, licensed under the trade name Thermochemical Environmental Energy System (TEES{reg sign}), is designed for to a wide variety of feedstocks ranging from dilute organics in water to waste sludges from food processing. The current research program is focused on the use of a continuous-feed, tubular reactor. The catalyst is nickel metal on an inert support. Typical results show that feedstocks such as solutions of 2% para-cresol or 5% and 10% lactose in water or cheese whey can be processed to >99% reduction of chemical oxygen demand (COD) at a rate of up to 2 L/hr. The estimated residence time is less than 5 min at 360{degree}C and 3000 psig, not including 1 to 2 min required in the preheating zone of the reactor. The liquid hourly space velocity has been varied from 1.8 to 2.9 L feedstock/L catalyst/hr depending on the feedstock. The product fuel gas contains 40% to 55% methane, 35% to 50% carbon dioxide, and 5% to 10% hydrogen with as much as 2% ethane, but less than 0.1% ethylene or carbon monoxide, and small amounts of higher hydrocarbons. The byproduct water stream carries residual organics amounting to less than 500 mg/L COD. 9 refs., 1 fig., 4 tabs.

  9. Prototype-Technology Evaluator and Research Aircraft (PTERA) Flight Test Assessment Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The Area-I team has developed and fabricated the unmanned Prototype-Technology Evaluation and Research Aircraft or PTERA ("ptera" being Greek for wing, or...

  10. Development of Novel, Optically-Based Instrumentation for Aircraft System Testing and Control Project

    Data.gov (United States)

    National Aeronautics and Space Administration — We propose to develop a compact, robust, optically-based sensor for making temperature and multi-species concentration measurements in aircraft system ground and...

  11. NASA Electric Aircraft Test Bed (NEAT) Development Plan - Design, Fabrication, Installation

    Science.gov (United States)

    Dyson, Rodger W.

    2016-01-01

    As large airline companies compete to reduce emissions, fuel, noise, and maintenance costs, it is expected that more of their aircraft systems will shift from using turbofan propulsion, pneumatic bleed power, and hydraulic actuation, to instead using electrical motor propulsion, generator power, and electrical actuation. This requires new flight-weight and flight-efficient powertrain components, fault tolerant power management, and electromagnetic interference mitigation technologies. Moreover, initial studies indicate some combination of ambient and cryogenic thermal management and relatively high bus voltages when compared to state of practice will be required to achieve a net system benefit. Developing all these powertrain technologies within a realistic aircraft architectural geometry and under realistic operational conditions requires a unique electric aircraft testbed. This report will summarize existing testbed capabilities located in the U.S. and details the development of a unique complementary testbed that industry and government can utilize to further mature electric aircraft technologies.

  12. Design of the VISTA-ITL Test Facility for an Integral Type Reactor of SMART and a Post-Test Simulation of a SBLOCA Test

    OpenAIRE

    2014-01-01

    To validate the performance and safety of an integral type reactor of SMART, a thermal-hydraulic integral effect test facility, VISTA-ITL, is introduced with a discussion of its scientific design characteristics. The VISTA-ITL was used extensively to assess the safety and performance of the SMART design, especially for its passive safety system such as a passive residual heat removal system, and to validate various thermal-hydraulic analysis codes. The VISTA-ITL program includes several tests...

  13. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition as part of a fuel meat thickness optimization effort for reactor performance other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  14. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  15. Fault mechanism analysis and simulation for continuity resistance test of electrical components in aircraft engine

    Science.gov (United States)

    Shi, Xudong; Yin, Yaping; Wang, Jialin; Sun, Zhaorong

    2017-01-01

    A large number of electrical components are used in civil aircraft engines, whose electrical circuits are usually intricate and complicated. Continuity resistance is an important parameter for the operating state of electrical components. Electrical continuity fault has serious impact on the reliability of the aircraft engine. In this paper, mathematical models of electrical components are established, and simulation is made by Simulink to analyze the electrical continuity fault.

  16. Status of the Combined Third and Fourth NGNP Fuel Irradiations In the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti; Michael E. Davenport

    2013-07-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in September 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. Since the purpose of this combined experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is

  17. Two-Dimensional Mapping of the Calculated Fission Power for the Full-Size Fuel Plate Experiment Irradiated in the Advanced Test Reactor

    Science.gov (United States)

    Chang, G. S.; Lillo, M. A.

    2009-08-01

    The National Nuclear Security Administrations (NNSA) Reduced Enrichment for Research and Test Reactors (RERTR) program assigned to the Idaho National Laboratory (INL) the responsibility of developing and demonstrating high uranium density research reactor fuel forms to enable the use of low enriched uranium (LEU) in research and test reactors around the world. A series of full-size fuel plate experiments have been proposed for irradiation testing in the center flux trap (CFT) position of the Advanced Test Reactor (ATR). These full-size fuel plate tests are designated as the AFIP tests. The AFIP nominal fuel zone is rectangular in shape having a designed length of 21.5-in (54.61-cm), width of 1.6-in (4.064-cm), and uniform thickness of 0.014-in (0.03556-cm). This gives a nominal fuel zone volume of 0.482 in3 (7.89 cm3) per fuel plate. The AFIP test assembly has two test positions. Each test position is designed to hold 2 full-size plates, for a total of 4 full-size plates per test assembly. The AFIP test plates will be irradiated at a peak surface heat flux of about 350 W/cm2 and discharged at a peak U-235 burn-up of about 70 at.%. Based on limited irradiation testing of the monolithic (U-10Mo) fuel form, it is desirable to keep the peak fuel temperature below 250°C to achieve this, it will be necessary to keep plate heat fluxes below 500 W/cm2. Due to the heavy U-235 loading and a plate width of 1.6-in (4.064-cm), the neutron self-shielding will increase the local-to-average-ratio (L2AR) fission power near the sides of the fuel plates. To demonstrate that the AFIP experiment will meet the ATR safety requirements, a very detailed 2-dimensional (2D) Y-Z fission power profile was evaluated in order to best predict the fuel plate temperature distribution. The ability to accurately predict fuel plate power and burnup are essential to both the design of the AFIP tests as well as evaluation of the irradiated fuel performance. To support this need, a detailed MCNP Y

  18. Evaluation of dynamic fracture toughness for Yong Gwang unit 5 reactor pressure vessel materials (Baseline Tests)

    Energy Technology Data Exchange (ETDEWEB)

    Chi Se Hwan; Kim, Joo Hag; Hong, Jun Hwa; Kwon, Sun Chil; Lee, Bong Sang [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    The dynamic fracture toughness (K{sub d}) of intermediate shell and its weld in SA 508 CI. 3 Yong Gwang 5 reactor pressure vessel was determined and evaluated. Precracked thirty six Charpy specimens were tested by using an instrumented impact tester. The purpose of present work is to evaluate and confirm the un-irradiated dynamic fracture toughness and to provide pre-irradiation baseline data for future evaluation on dynamic fracture toughness change during operation. 18 refs., 5 figs., 5 tabs. (Author)

  19. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

  20. First Test of Lorentz Violation with a Reactor-based Antineutrino Experiment

    CERN Document Server

    Abe, Y; Anjos, J C dos; Bergevin, M; Bernstein, A; Bezerra, T J C; Bezrukhov, L; Blucher, E; Bowden, N S; Buck, C; Busenitz, J; Cabrera, A; Caden, E; Camilleri, L; Carr, R; Cerrada, M; Chang, P -J; Chimenti, P; Classen, T; Collin, A P; Conover, E; Conrad, J M; Crespo-Anadón, J I; Crum, K; Cucoanes, A; D'Agostino, M V; Damon, E; Dawson, J V; Dazeley, S; Dietrich, D; Djurcic, Z; Dracos, M; Durand, V; Ebert, J; Efremenko, Y; Elnimr, M; Erickson, A; Fallot, M; Fechner, M; von Feilitzsch, F; Felde, J; Fischer, V; Franco, D; Franke, A J; Franke, M; Furuta, H; Gama, R; Gil-Botella, I; Giot, L; Göger-Neff, M; Gonzalez, L F G; Goodman, M C; Goon, J TM; Greiner, D; Haag, N; Habib, S; Hagner, C; Hara, T; Hartmann, F X; Haser, J; Hatzikoutelis, A; Hayakawa, T; Hofmann, M; Horton-Smith, G A; Ishitsuka, M; Jochum, J; Jollet, C; Jones, C L; Kaether, F; Kalousis, L N; Kamyshkov, Y; Kaplan, D M; Katori, T; Kawasaki, T; Keefer, G; Kemp, E; de Kerret, H; Konno, T; Kryn, D; Kuze, M; Lachenmaier, T; Lane, C E; Lasserre, T; Letourneau, A; Lhuillier, D; Lima, H P; Lindner, M; López-Castanõ, J M; LoSecco, J M; Lubsandorzhiev, B K; Lucht, S; McKee, D; Maeda, J; Maesano, C N; Mariani, C; Maricic, J; Martino, J; Matsubara, T; Mention, G; Meregaglia, A; Meyer, M; Miletic, T; Milincic, R; Miyata, H; Mueller, Th A; Nagasaka, Y; Nakajima, K; Novella, P; Obolensky, M; Oberauer, L; Onillon, A; Osborn, A; Ostrovskiy, I; Palomares, C; Pepe, I M; Perasso, S; Perrin, P; Pfahler, P; Porta, A; Potzel, W; Pronost, G; Reichenbacher, J; Reinhold, B; Remoto, A; Röhling, M; Roncin, R; Roth, S; Rybolt, B; Sakamoto, Y; Santorelli, R; Sato, F; Schönert, S; Schoppmann, S; Schwetz, T; Shaevitz, M H; Shrestha, D; Sida, J -L; Sinev, V; Skorokhvatov, M; Smith, E; Spitz, J; Stahl, A; Stancu, I; Stokes, L F F; Strait, M; Stüken, A; Suekane, F; Sukhotin, S; Sumiyoshi, T; Sun, Y; Terao, K; Tonazzo, A; Toups, M; Thi, H H Trinh; Valdiviesso, G; Veyssiere, C; Wagner, S; Watanabe, H; White, B; Wiebusch, C; Winslow, L; Worcester, M; Wurm, M; Yanovitch, E; Yermia, F; Zimmer, V

    2012-01-01

    We present a search for Lorentz violation with 8249 candidate electron antineutrino events taken by the Double Chooz experiment in 227.9 live days of running. This analysis, featuring a search for a sidereal time dependence of the events, is the first test of Lorentz invariance using a reactor-based antineutrino source. No sidereal variation is present in the data and the disappearance results are consistent with sidereal time independent oscillations. Under the Standard-Model Extension (SME), we set the first limits on fourteen Lorentz violating coefficients associated with transitions between electron and tau flavor, and set two competitive limits associated with transitions between electron and muon flavor.

  1. Shielding design for PWR in France

    Energy Technology Data Exchange (ETDEWEB)

    Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.

    1983-05-01

    Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983.

  2. Construction and operation of parallel electric and magnetic field spectrometers for mass/energy resolved multi-ion charge exchange diagnostics on the Tokamak Fusion Test Reactor

    Science.gov (United States)

    Medley, S. S.; Roquemore, A. L.

    1998-07-01

    A novel charge exchange spectrometer using a dee-shaped region of parallel electric and magnetic fields was developed at the Princeton Plasma Physics Laboratory for neutral particle diagnostics on the Tokamak Fusion Test Reactor (TFTR). The E∥B spectrometer has an energy range of 0.5⩽A (amu)E (keV)⩽600 and provides mass-resolved energy spectra of H+, D+, and T+ (or 3He+) ion species simultaneously during a single discharge. The detector plane exhibits parallel rows of analyzed ions, each row containing the energy dispersed ions of a given mass-to-charge ratio. The detector consists of a large area microchannel plate (MCP) which is provided with three rectangular, semicontinuous active area strips, one coinciding with each of the mass rows for detection of H+, D+, and T+ (or 3He+) and each mass row has 75 energy channels. To suppress spurious signals attending operation of the plate in the magnetic fringe field of the spectrometer, the MCP was housed in a double-walled iron shield with a wire mesh ion entrance window. Using an accelerator neutron generator, the MCP neutron detection efficiency was measured to be 1.7×10-3 and 6.4×10-3 counts/neutron/cm2 for 2.5 MeV-DD and 14 MeV-DT neutrons, respectively. The design and calibration of the spectrometer are described in detail, including the effect of MCP exposure to tritium, and results obtained during high performance D-D operation on TFTR are presented to illustrate the performance of the E∥B spectrometer. The spectrometers were not used during D-T plasma operation due to the cost of providing the required radiation shielding.

  3. Completion summary for borehole USGS 136 near the Advanced Test Reactor Complex, Idaho National Laboratory, Idaho

    Science.gov (United States)

    Twining, Brian V.; Bartholomay, Roy C.; Hodges, Mary K.V.

    2012-01-01

    In 2011, the U.S. Geological Survey, in cooperation with the U.S. Department of Energy, cored and completed borehole USGS 136 for stratigraphic framework analyses and long-term groundwater monitoring of the eastern Snake River Plain aquifer at the Idaho National Laboratory. The borehole was initially cored to a depth of 1,048 feet (ft) below land surface (BLS) to collect core, open-borehole water samples, and geophysical data. After these data were collected, borehole USGS 136 was cemented and backfilled between 560 and 1,048 ft BLS. The final construction of borehole USGS 136 required that the borehole be reamed to allow for installation of 6-inch (in.) diameter carbon-steel casing and 5-in. diameter stainless-steel screen; the screened monitoring interval was completed between 500 and 551 ft BLS. A dedicated pump and water-level access line were placed to allow for aquifer testing, for collecting periodic water samples, and for measuring water levels. Geophysical and borehole video logs were collected after coring and after the completion of the monitor well. Geophysical logs were examined in conjunction with the borehole core to describe borehole lithology and to identify primary flow paths for groundwater, which occur in intervals of fractured and vesicular basalt. A single-well aquifer test was used to define hydraulic characteristics for borehole USGS 136 in the eastern Snake River Plain aquifer. Specific-capacity, transmissivity, and hydraulic conductivity from the aquifer test were at least 975 gallons per minute per foot, 1.4 × 105 feet squared per day (ft2/d), and 254 feet per day, respectively. The amount of measureable drawdown during the aquifer test was about 0.02 ft. The transmissivity for borehole USGS 136 was in the range of values determined from previous aquifer tests conducted in other wells near the Advanced Test Reactor Complex: 9.5 × 103 to 1.9 × 105 ft2/d. Water samples were analyzed for cations, anions, metals, nutrients, total organic

  4. Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

    Energy Technology Data Exchange (ETDEWEB)

    MH Lane

    2006-02-15

    This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

  5. Aircraft electromagnetic compatibility

    Science.gov (United States)

    Clarke, Clifton A.; Larsen, William E.

    1987-06-01

    Illustrated are aircraft architecture, electromagnetic interference environments, electromagnetic compatibility protection techniques, program specifications, tasks, and verification and validation procedures. The environment of 400 Hz power, electrical transients, and radio frequency fields are portrayed and related to thresholds of avionics electronics. Five layers of protection for avionics are defined. Recognition is given to some present day electromagnetic compatibility weaknesses and issues which serve to reemphasize the importance of EMC verification of equipment and parts, and their ultimate EMC validation on the aircraft. Proven standards of grounding, bonding, shielding, wiring, and packaging are laid out to help provide a foundation for a comprehensive approach to successful future aircraft design and an understanding of cost effective EMC in an aircraft setting.

  6. Non-Nuclear Testing of Compact Reactor Technologies at NASA MSFC

    Science.gov (United States)

    Houts, Michael G.; Pearson, J. Boise; Godfroy, Thomas J.

    2011-01-01

    Safe, reliable, compact, autonomous, long-life fission systems have numerous potential applications, both terrestrially and in space. Technologies and facilities developed in support of these systems could be useful to a variety of concepts. At moderate power levels, fission systems can be designed to operate for decades without the need for refueling. In addition, fast neutron damage to cladding and structural materials can be maintained at an acceptable level. Nuclear design codes have advanced to the stage where high confidence in the behavior and performance of a system can be achieved prior to initial testing. To help ensure reactor affordability, an optimal strategy must be devised for development and qualification. That strategy typically involves a combination of non-nuclear and nuclear testing. Non-nuclear testing is particularly useful for concepts in which nuclear operating characteristics are well understood and nuclear effects such as burnup and radiation damage are not likely to be significant. To be mass efficient, a SFPS must operate at higher coolant temperatures and use different types of power conversion than typical terrestrial reactors. The primary reason is the difficulty in rejecting excess heat to space. Although many options exist, NASA s current reference SFPS uses a fast spectrum, pumped-NaK cooled reactor coupled to a Stirling power conversion subsystem. The reference system uses technology with significant terrestrial heritage while still providing excellent performance. In addition, technologies from the SFPS system could be applicable to compact terrestrial systems. Recent non-nuclear testing at NASA s Early Flight Fission Test Facility (EFF-TF) has helped assess the viability of the reference SFPS and evaluate methods for system integration. In July, 2011 an Annular Linear Induction Pump (ALIP) provided by Idaho National Laboratory was tested at the EFF-TF to assess performance and verify suitability for use in a10 kWe technology

  7. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; R. Schley; J. Palmer; K. Condie

    2014-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support the growth of nuclear science and technology in the United States (US). By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort at the Idaho National Laboratory (INL) is to design, develop, and deploy 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 was completed. Based on this initial review, recommendations were made with respect to what instrumentation is needed at the ATR, and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. Since 2009, annual reports have been issued to provide updates on the program strategy and the progress made on implementing the strategy. This report provides an update reflecting progress as of January 2014.

  8. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; J. E. Daw

    2011-03-01

    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 NSUF facilitates 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 was completed. Based on this review, recommendations were made with respect to what instrumentation is needed at the ATR; and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. In 2009, a report was issued documenting this instrumentation development strategy and initial progress toward accomplishing instrumentation development program objectives. This document reports progress toward implementing this strategy in 2010.

  9. STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

    Energy Technology Data Exchange (ETDEWEB)

    Anthony L. Alberti; Todd S. Palmer; Javier Ortensi; Mark D. DeHart

    2016-05-01

    With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.

  10. Verifying seismic design of nuclear reactors by testing. Volume 1: test plan

    Energy Technology Data Exchange (ETDEWEB)

    1979-07-20

    This document sets forth recommendations for a verification program to test the ability of operational nuclear power plants to achieve safe shutdown immediately following a safe-shutdown earthquake. The purpose of the study is to develop a program plan to provide assurance by physical demonstration that nuclear power plants are earthquake resistant and to allow nuclear power plant operators to (1) decide whether tests should be conducted on their facilities, (2) specify the tests that should be performed, and (3) estimate the cost of the effort to complete the recommended test program.

  11. Standard Practice for Total Immersion Corrosion Test for Aircraft Maintenance Chemicals

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This practice covers the determination of the corrosiveness of aircraft maintenance chemicals on aircraft metals with time under conditions of total immersion by a combination of weight change measurements and visual qualitative determination of change. 1.2 The values stated in SI units are to be regarded as standard. The values in parentheses are for information only. 1.3 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.

  12. Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters

    Energy Technology Data Exchange (ETDEWEB)

    Darmann, Frank [Zenergy Power, Inc., Burlingame, CA (United States); Lombaerde, Robert [Zenergy Power, Inc., Burlingame, CA (United States); Moriconi, Franco [Zenergy Power, Inc., Burlingame, CA (United States); Nelson, Albert [Zenergy Power, Inc., Burlingame, CA (United States)

    2012-03-01

    Zenergy Power has successfully designed, built, tested, and installed in the US electrical grid a saturable reactor Fault Current Limiter. Beginning in 2007, first as SC Power Systems and from 2008 as Zenergy Power, Inc., ZP used DOE matching grant and ARRA funds to help refine the design of the saturated reactor fault current limiter. ZP ultimately perfected the design of the saturated reactor FCL to the point that ZP could reliably design a suitable FCL for most utility applications. Beginning with a very basic FCL design using 1G HTS for a coil housed in a LN2 cryostat for the DC bias magnet, the technology progressed to a commercial system that was offered for sale internationally. Substantial progress was made in two areas. First, the cryogenics cooling system progressed from a sub-cooled liquid nitrogen container housing the HTS coils to cryostats utilizing dry conduction cooling and reaching temperatures down to less than 20 degrees K. Large, round cryostats with warm bore diameters of 1.7 meters enabled the design of large tanks to hold the AC components. Second, the design of the AC part of the FCL was refined from a six legged spider design to a more compact and lighter design with better fault current limiting capability. Further refinement of the flux path and core shape led to an efficient saturated reactor design requiring less Ampere-turns to saturate the core. In conclusion, the development of the saturable reactor FCL led to a more efficient design not requiring HTS magnets and their associated peripheral equipment, which yielded a more economical product in line with the electric utility industry expectations. The original goal for the DOE funding of the ZP project Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters was to stimulate the HTS wire industry with, first 1G, then 2G, HTS wire applications. Over the approximately 5 years of ZP's product development program, the amount of HTS

  13. Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance, E706 (IIIC)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This test method describes the concept and use of helium accumulation for neutron fluence dosimetry for reactor vessel surveillance. Although this test method is directed toward applications in vessel surveillance, the concepts and techniques are equally applicable to the general field of neutron dosimetry. The various applications of this test method for reactor vessel surveillance are as follows: 1.1.1 Helium accumulation fluence monitor (HAFM) capsules, 1.1.2 Unencapsulated, or cadmium or gadolinium covered, radiometric monitors (RM) and HAFM wires for helium analysis, 1.1.3 Charpy test block samples for helium accumulation, and 1.1.4 Reactor vessel (RV) wall samples for helium accumulation. 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.

  14. U.S. aerospace industry opinion of the effect of computer-aided prediction-design technology on future wind-tunnel test requirements for aircraft development programs

    Science.gov (United States)

    Treon, S. L.

    1979-01-01

    A survey of the U.S. aerospace industry in late 1977 suggests that there will be an increasing use of computer-aided prediction-design technology (CPD Tech) in the aircraft development process but that, overall, only a modest reduction in wind-tunnel test requirements from the current level is expected in the period through 1995. Opinions were received from key spokesmen in 23 of the 26 solicited major companies or corporate divisions involved in the design and manufacture of nonrotary wing aircraft. Development programs for nine types of aircraft related to test phases and wind-tunnel size and speed range were considered.

  15. Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.

    1981-09-01

    Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies. (DLC)

  16. Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

    Science.gov (United States)

    Springborn, R. H.

    1971-01-01

    Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

  17. Test of Different Air Distribution Concepts for a Single-Aisle Aircraft Cabin

    DEFF Research Database (Denmark)

    Nielsen, Peter V.; Damsgaard, Charlotte; Liu, Li;

    2013-01-01

    Traditionally, air is supplied to the aircraft cabin either by individual nozzles or by supply slots. The air is expected to be fully mixed in the cabin, and the system is considered to be a mixing ventilation system. This paper will describe different air distribution systems known from other...

  18. Testing a theory of aircraft noise annoyance: A structural equation analysis

    NARCIS (Netherlands)

    Kroesen, M.; Molin, E.J.E.; Van Wee, B.

    2008-01-01

    Previous research has stressed the relevance of nonacoustical factors in the perception of aircraft noise. However, it is largely empirically driven and lacks a sound theoretical basis. In this paper, a theoretical model which explains noise annoyance based on the psychological stress theory is empi

  19. 14 CFR 61.45 - Practical tests: Required aircraft and equipment.

    Science.gov (United States)

    2010-01-01

    ... limited, primary, or light-sport category. (2) At the discretion of the examiner who administers the..., or light-sport category, but that otherwise meets the requirements of paragraph (a)(1) of this... each person to operate the aircraft safely; and (iv) Cockpit and outside visibility adequate...

  20. TCV software test and validation tools and technique. [Terminal Configured Vehicle program for commercial transport aircraft operation

    Science.gov (United States)

    Straeter, T. A.; Williams, J. R.

    1976-01-01

    The paper describes techniques for testing and validating software for the TCV (Terminal Configured Vehicle) program which is intended to solve problems associated with operating a commercial transport aircraft in the terminal area. The TCV research test bed is a Boeing 737 specially configured with digital computer systems to carry out automatic navigation, guidance, flight controls, and electronic displays research. The techniques developed for time and cost reduction include automatic documentation aids, an automatic software configuration, and an all software generation and validation system.

  1. Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pope, M. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, M. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Morrell, S. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jamison, R. K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nef, E. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nigg, D. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses, a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.

  2. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2010

    Energy Technology Data Exchange (ETDEWEB)

    Rahmat Aryaeinejad; Douglas S. Crawford; Mark D. DeHart; George W. Griffith; D. Scott Lucas; Joseph W. Nielsen; David W. Nigg; James R. Parry; Jorge Navarro

    2010-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or “Core Modeling Update”) Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  3. Measurement of photon flux with a miniature gas ionization chamber in a Material Testing Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fourmentel, D., E-mail: damien.fourmentel@cea.fr [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Filliatre, P.; Villard, J.F.; Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Reynard-Carette, C. [Aix-Marseille Université, LISA EA 4672, cedex 20, Marseille 13397 (France); Carcreff, H. [CEA, DEN, DRSN, Saclay, F-91191 Gif-sur-Yvette (France)

    2013-10-01

    Nuclear heating measurements in Material Testing Reactors (MTR) are crucial for the design of the experimental devices and the prediction of the temperature of the hosted samples. Nuclear heating in MTR materials (except fuel) is mainly due to the energy deposition by the photon flux. Therefore, the photon flux is a key input parameter for the computer codes which simulate nuclear heating and temperature reached by samples/devices under irradiation. In the Jules Horowitz MTR under construction at the CEA Cadarache, the maximal expected nuclear heating levels will be about 15 to 18 W g{sup −1} and it will be necessary to assess this parameter with the best accuracy. An experiment was performed at the OSIRIS reactor to combine neutron flux, photon flux and nuclear heating measurements to improve the knowledge of the nuclear heating in MTR. There are few appropriate sensors for selective measurement of the photon flux in MTR even if studies and developments are ongoing. An experiment, called CARMEN-1, was conducted at the OSIRIS MTR and we used in particular a gas ionization chamber based on miniature fission chamber design to measure the photon flux. In this paper, we detail Monte-Carlo simulations to analyze the photon fluxes with ionization chamber measurements and we compare the photon flux calculations to the nuclear heating measurements. These results show a good accordance between photon flux measurements and nuclear heating measurement and allow improving the knowledge of these parameters.

  4. Test of a prototype neutron spectrometer based on diamond detectors in a fast reactor

    CERN Document Server

    Osipenko, M; Ripani, M; Pillon, M; Ricco, G; Caiffi, B; Cardarelli, R; Verona-Rinati, G; Argiro, S

    2015-01-01

    A prototype of neutron spectrometer based on diamond detectors has been developed. This prototype consists of a $^6$Li neutron converter sandwiched between two CVD diamond crystals. The radiation hardness of the diamond crystals makes it suitable for applications in low power research reactors, while a low sensitivity to gamma rays and low leakage current of the detector permit to reach good energy resolution. A fast coincidence between two crystals is used to reject background. The detector was read out using two different electronic chains connected to it by a few meters of cable. The first chain was based on conventional charge-sensitive amplifiers, the other used a custom fast charge amplifier developed for this purpose. The prototype has been tested at various neutron sources and showed its practicability. In particular, the detector was calibrated in a TRIGA thermal reactor (LENA laboratory, University of Pavia) with neutron fluxes of $10^8$ n/cm$^2$s and at the 3 MeV D-D monochromatic neutron source na...

  5. Monte Carlo Simulation Study of a Differential Calorimeter Measuring the Nuclear Heating in Material Testing Reactors

    Science.gov (United States)

    Amharrak, H.; Reynard-Carette, C.; Lyoussi, A.; Carette, M.; Brun, J.; De Vita, C.; Fourmentel, D.; Villard, J.-F.; Guimbal, P.

    2016-02-01

    The nuclear heating measurements in Material Testing Reactors (MTRs) are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. Then these measurements are used for other materials, other geometries, or other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. This paper will present new simulations with MCNP Monte-Carlo transport code to determine the gamma heating profile inside the calorimeter. The whole complex geometry of the sensor has been considered. We use as an input source in the model, the photon spectra calculated in various positions of CARMEN-1 irradiation program in OSIRIS reactor. After a description of the differential calorimeter device, the MCNP modeling used for the calculations of radial profile of nuclear heating inside the calorimeter elements will be introduced. The obtained results of different simulations will be detailed and discussed in this paper. The charged particle equilibrium inside the calorimeter elements will be studied. Then we will focus on parametric studies of the various components of the calorimeter. The influence of source type will be also took into account. Moreover the influence of the material used for the sample will be described.

  6. Monte Carlo Simulation Study of a Differential Calorimeter Measuring the Nuclear Heating in Material Testing Reactors

    Directory of Open Access Journals (Sweden)

    Amharrak H.

    2016-01-01

    Full Text Available The nuclear heating measurements in Material Testing Reactors (MTRs are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. Then these measurements are used for other materials, other geometries, or other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. This paper will present new simulations with MCNP Monte-Carlo transport code to determine the gamma heating profile inside the calorimeter. The whole complex geometry of the sensor has been considered. We use as an input source in the model, the photon spectra calculated in various positions of CARMEN-1 irradiation program in OSIRIS reactor. After a description of the differential calorimeter device, the MCNP modeling used for the calculations of radial profile of nuclear heating inside the calorimeter elements will be introduced. The obtained results of different simulations will be detailed and discussed in this paper. The charged particle equilibrium inside the calorimeter elements will be studied. Then we will focus on parametric studies of the various components of the calorimeter. The influence of source type will be also took into account. Moreover the influence of the material used for the sample will be described.

  7. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2013

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg

    2013-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  8. Preliminary Feasibility, Design, and Hazard Analysis of a Boiling Water Test Loop Within the Idaho National Laboratory Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Douglas M. Gerstner

    2009-05-01

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The ATR and its support facilities are located at the Idaho National Laboratory (INL). A Boiling Water Test Loop (BWTL) is being designed for one of the irradiation test positions within the. The objective of the new loop will be to simulate boiling water reactor (BWR) conditions to support clad corrosion and related reactor material testing. Further it will accommodate power ramping tests of candidate high burn-up fuels and fuel pins/rods for the commercial BWR utilities. The BWTL will be much like the pressurized water loops already in service in 5 of the 9 “flux traps” (region of enhanced neutron flux) in the ATR. The loop coolant will be isolated from the primary coolant system so that the loop’s temperature, pressure, flow rate, and water chemistry can be independently controlled. This paper presents the proposed general design of the in-core and auxiliary BWTL systems; the preliminary results of the neutronics and thermal hydraulics analyses; and the preliminary hazard analysis for safe normal and transient BWTL and ATR operation.

  9. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors AGENCY... Commission (NRC) is issuing a new regulatory guide (RG), 1.79.1, ``Initial Test Program of Emergency Core... System (ADAMS): You may access publicly available documents online in the NRC Library at...

  10. Guide to verification and validation of the SCALE-4 radiation shielding software

    Energy Technology Data Exchange (ETDEWEB)

    Broadhead, B.L.; Emmett, M.B.; Tang, J.S.

    1996-12-01

    Whenever a decision is made to newly install the SCALE radiation shielding software on a computer system, the user should run a set of verification and validation (V&V) test cases to demonstrate that the software is properly installed and functioning correctly. This report is intended to serve as a guide for this V&V in that it specifies test cases to run and gives expected results. The report describes the V&V that has been performed for the radiation shielding software in a version of SCALE-4. This report provides documentation of sample problems which are recommended for use in the V&V of the SCALE-4 system for all releases. The results reported in this document are from the SCALE-4.2P version which was run on an IBM RS/6000 work-station. These results verify that the SCALE-4 radiation shielding software has been correctly installed and is functioning properly. A set of problems for use by other shielding codes (e.g., MCNP, TWOTRAN, MORSE) performing similar V&V are discussed. A validation has been performed for XSDRNPM and MORSE-SGC6 utilizing SASI and SAS4 shielding sequences and the SCALE 27-18 group (27N-18COUPLE) cross-section library for typical nuclear reactor spent fuel sources and a variety of transport package geometries. The experimental models used for the validation were taken from two previous applications of the SASI and SAS4 methods.

  11. Design and operation of the pellet charge exchange diagnostic for measurement of energetic confined α particles and tritons on the Tokamak Fusion Test Reactor

    Science.gov (United States)

    Medley, S. S.; Mansfield, D. K.; Roquemore, A. L.; Fisher, R. K.; Duong, H. H.; McChesney, J. M.; Parks, P. B.; Petrov, M. P.; Khudoleev, A. V.; Gorelenkov, N. N.

    1996-09-01

    Radially resolved energy and density distributions of the confined α particles in D-T experiments on the Tokamak Fusion Test Reactor (TFTR) are being measured with the pellet charge exchange (PCX) diagnostic. Other energetic ion species can be detected as well, such as tritons produced in D-D plasmas and H, He3, or tritium rf-driven minority ion tails. The ablation cloud formed by injected low-Z impurity pellets provides the neutralization target for this active charge exchange technique. Because the cloud neutralization efficiency is uncertain, the PCX diagnostic is not absolutely calibrated so only relative density profiles are obtained. A mass and energy resolving E∥B neutral particle analyzer (NPA) is used which has eight energy channels covering the energy range of 0.3-3.7 MeV for α particles with energy resolution ranging from 5.8% to 11.3% and a spatial resolution of ˜5 cm. The PCX diagnostic views deeply trapped ions in a narrow pitch angle range around a mean value of v∥/v=-0.048±10-3. For D-T operation, the NPA was shielded by a polyethylene-lead enclosure providing 100× attenuation of ambient γ radiation and 14 MeV neutrons. The PCX diagnostic technique and its application on TFTR are described in detail.

  12. Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

    Energy Technology Data Exchange (ETDEWEB)

    Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-11-30

    The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To support this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.

  13. Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

    Energy Technology Data Exchange (ETDEWEB)

    1979-10-01

    A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection.

  14. Mechanical testing and modelling of carbon-carbon composites for aircraft disc brakes

    Science.gov (United States)

    Bradley, Luke R.

    The objective of this study is to improve the understanding of the stress distributions and failure mechanisms experienced by carbon-carbon composite aircraft brake discs using finite element (FE) analyses. The project has been carried out in association with Dunlop Aerospace as an EPSRC CASE studentship. It therefore focuses on the carbon-carbon composite brake disc material produced by Dunlop Aerospace, although it is envisaged that the approach will have broader applications for modelling and mechanical testing of carbon-carbon composites in general. The disc brake material is a laminated carbon-carbon composite comprised of poly(acrylonitrile) (PAN) derived carbon fibres in a chemical vapour infiltration (CVI) deposited matrix, in which the reinforcement is present in both continuous fibre and chopped fibre forms. To pave the way for the finite element analysis, a comprehensive study of the mechanical properties of the carbon-carbon composite material was carried out. This focused largely, but not entirely, on model composite materials formulated using structural elements of the disc brake material. The strengths and moduli of these materials were measured in tension, compression and shear in several orientations. It was found that the stress-strain behaviour of the materials were linear in directions where there was some continuous fibre reinforcement, but non-linear when this was not the case. In all orientations, some degree of non-linearity was observed in the shear stress-strain response of the materials. However, this non-linearity was generally not large enough to pose a problem for the estimation of elastic moduli. Evidence was found for negative Poisson's ratio behaviour in some orientations of the material in tension. Additionally, the through-thickness properties of the composite, including interlaminar shear strength, were shown to be positively related to bulk density. The in-plane properties were mostly unrelated to bulk density over the range of

  15. Predictions for Radiation Shielding Materials

    Science.gov (United States)

    Kiefer, Richard L.

    2002-01-01

    Radiation from galactic cosmic rays (GCR) and solar particle events (SPE) is a serious hazard to humans and electronic instruments during space travel, particularly on prolonged missions outside the Earth s magnetic fields. Galactic cosmic radiation (GCR) is composed of approx. 98% nucleons and approx. 2% electrons and positrons. Although cosmic ray heavy ions are 1-2% of the fluence, these energetic heavy nuclei (HZE) contribute 50% of the long-term dose. These unusually high specific ionizations pose a significant health hazard acting as carcinogens and also causing microelectronics damage inside spacecraft and high-flying aircraft. These HZE ions are of concern for radiation protection and radiation shielding technology, because gross rearrangements and mutations and deletions in DNA are expected. Calculations have shown that HZE particles have a strong preference for interaction with light nuclei. The best shield for this radiation would be liquid hydrogen, which is totally impractical. For this reason, hydrogen-containing polymers make the most effective practical shields. Shielding is required during missions in Earth orbit and possibly for frequent flying at high altitude because of the broad GCR spectrum and during a passage into deep space and LunarMars habitation because of the protracted exposure encountered on a long space mission. An additional hazard comes from solar particle events (SPEs) which are mostly energetic protons that can produce heavy ion secondaries as well as neutrons in materials. These events occur at unpredictable times and can deliver a potentially lethal dose within several hours to an unshielded human. Radiation protection for humans requires safety in short-term missions and maintaining career exposure limits within acceptable levels on future long-term exploration missions. The selection of shield materials can alter the protection of humans by an order of magnitude. If improperly selected, shielding materials can actually

  16. Creep-fatigue interaction in aircraft gas turbine components by simulation and testing at scaled temperatures

    Science.gov (United States)

    Sabour, Mohammad Hossein

    Advanced gas turbine engines, which use hot section airfoil cooling, present a wide range of design problems. The frequencies of applied loads and the natural frequencies of the blade also are important since they have significant effects on failure of the component due to fatigue phenomenon. Due to high temperature environment the thermal creep and fatigue are quite severe. One-dimensional creep model, using ANSYS has been formulated in order to predict the creep life of a gas turbine engine blade. Innovative mathematical models for the prediction of the operating life of aircraft components, specifically gas turbine blades, which are subjected to creep-fatigue at high temperatures, are proposed. The components are modeled by FEM, mathematically, and using similitude principles. Three models have been suggested and evaluated numerically and experimentally. Using FEM method for natural frequencies causes phenomena such as curve veering which is studied in more detail. The simulation studies on the life-limiting modes of failure, as well as estimating the expected lifetime of the blade, using the proposed models have been carried out. Although the scale model approach has been used for quite some time, the thermal scaling has been used in this study for the first time. The only thermal studies in literature using scaling for structures is by NASA in which materials of both the prototype and the model are the same, but in the present study materials also are different. The finite element method is employed to model the structure. Because of stress redistribution due to the creep process, it is necessary to include a full inelastic creep step in the finite element formulation. Otherwise over-conservative creep life predictions will be estimated if only the initial elastic stresses are considered. The experimental investigations are carried out in order to validate the models. The main contributions in the thesis are: (1) Using similitude theory for life prediction of

  17. A miniaturized test method for the mechanical characterization of structural materials for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gondi, P. [Rome-2 Univ. (Italy). Mech. Eng. Dept.; Donato, A. [ENEA CRE, Fusion Sector, Frascati, Rome (Italy); Montanari, R. [Rome-2 Univ. (Italy). Mech. Eng. Dept.; Sili, A. [Rome-2 Univ. (Italy). Mech. Eng. Dept.

    1996-10-01

    This work deals with a non-destructive method for mechanical tests which is based on the indentation of materials at a constant rate by means of a cylinder with a small radius and penetrating flat surface. The load versus penetration depth curves obtained using this method have shown correspondences with those of tensile tests and have given indications about the mechanical properties on a reduced scale. In this work penetration tests have been carried out on various kinds of Cr martensitic steels (MANET-2, BATMAN and modified F82H) which are of interest for first wall and structural applications in future fusion reactors. The load versus penetration depth curves have been examined with reference to data obtained in tensile tests and to microhardness measurements. Penetration tests have been performed at various temperature (from -180 to 100 C). Conclusions, which can be drawn for the ductile to brittle transition, are discussed for MANET-2 steel. Preliminary results obtained on BATMAN and modified F82H steels are reported. The characteristics of the indenter imprints have been studied by scanning electron microscopy. (orig.).

  18. Modeling, Simulation, and Flight Test for Automatic Flight Control of the Condor Hybrid-Electric Remote Piloted Aircraft

    Science.gov (United States)

    2012-03-01

    Aircraft Longitudinal Definitions (Nelson, 1998 p.123) ................................... 15 Table 4: Aircraft Lateral/Directional Derivatives...plane. Also known as the Roll angle. 15 Table 3: Aircraft Longitudinal Definitions (Nelson, 1998 p.123) Table 4...22 with tuning the longitudinal parameters arise when shifts in the center of gravity due to cargo loading change the aircraft longitudinal moment

  19. Model for collisional fast ion diffusion into Tokamak Fusion Test Reactor loss cone

    Energy Technology Data Exchange (ETDEWEB)

    Chang, C.S. [New York Univ., NY (United States). Courant Inst. of Mathematical Sciences]|[Korea Advanced Inst. of Science and Technology, Seoul (Korea, Republic of); Zweben, S.J.; Schivell, J.; Budny, R.; Scott, S. [Princeton Univ., NJ (United States). Plasma Physics Lab.

    1994-08-01

    An analytic model is developed to estimate the classical pitch angle scattering loss of energetic fusion product ions into prompt loss orbits in a tokamak geometry. The result is applied to alpha particles produced by deutrium-tritium fusion reactions in a plasma condition relevant to Tokamak Fusion Test Reactor (TFTR). A poloidal angular distribution of collisional fast ion loss at the first wall is obtained and the numerical result from the TRANSP code is discussed. The present model includes the effect that the prompt loss boundary moves away from the slowing-down path due to reduction in banana thickness, which enables us to understand, for the first time. the dependence of the collisional loss rate on Z{sub eff}.

  20. Safety Evaluation for Packaging for onsite Transfer of plutonium recycle test reactor ion exchange columns

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.J.

    1995-09-11

    The purpose of this Safety Evaluation for Packaging (SEP) is to authorize the use of three U.S. Department of Transportation (DOT) 7A, Type A metal boxes (Capital Industries Part No. S 0600-0600-1080- 0104) to package 12 Plutonium Recycle Test Reactor (PRTR) ion exchange columns as low-level waste (LLW). The packages will be transferred from the 309 Building in the 300 Area to low level waste burial in the 200 West Area. Revision 1 of WHC-SD-TP-SEP-035 (per ECN No. 621467) documents that the boxes containing ion exchange columns and grout will maintain the payload under normal conditions of transport if transferred without the box lids

  1. Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask

    Energy Technology Data Exchange (ETDEWEB)

    Romano, T.

    1997-09-29

    This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

  2. Crack growth tests on a ferritic reactor pressure vessel steel under the simultaneous influence of simulated BWR coolant and irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, H. [VGB PowerTech e.V., Essen (Germany); Huettner, F. [Hamburgische Electricitaets-Werke AG, Hamburg (Germany); Ilg, U. [EnBW Kraftwerke AG, Philippsburg (Germany); Wachter, O. [E.ON Kernkraft GmbH, Hannover(Germany); Widera, M. [RWE Power AG, Essen (Germany); Brozova, A.; Ernestova, M.; Kysela, J.; Vsolak, R. [Nuclear Research Institute Rez plc (Czech Republic)

    2004-07-01

    Crack growth tests under constant load with initial in-situ cycling were performed on the low alloy reactor pressure vessel (RPV) steel 22 NiMoCr 3 7 (A 508 Cl. 2) with the goal to determine crack growth rates of irradiated and non-irradiated steel under the simultaneous influence of simulated BWR coolant and irradiation. The tests were performed under conditions as near as possible to operational conditions in a commercial BWR reactor. The research results are summarized and are compared with international data. (orig.)

  3. Follow on Research for Multi-Utility Technology Test Bed Aircraft at NASA Dryden Flight Research Center (FY13 Progress Report)

    Science.gov (United States)

    Pak, Chan-Gi

    2013-01-01

    Modern aircraft employ a significant fraction of their weight in composite materials to reduce weight and improve performance. Aircraft aeroservoelastic models are typically characterized by significant levels of model parameter uncertainty due to the composite manufacturing process. Small modeling errors in the finite element model will eventually induce errors in the structural flexibility and mass, thus propagating into unpredictable errors in the unsteady aerodynamics and the control law design. One of the primary objectives of Multi Utility Technology Test-bed (MUTT) aircraft is the flight demonstration of active flutter suppression, and therefore in this study, the identification of the primary and secondary modes for the structural model tuning based on the flutter analysis of MUTT aircraft. The ground vibration test-validated structural dynamic finite element model of the MUTT aircraft is created in this study. The structural dynamic finite element model of MUTT aircraft is improved using the in-house Multi-disciplinary Design, Analysis, and Optimization tool. In this study, two different weight configurations of MUTT aircraft have been improved simultaneously in a single model tuning procedure.

  4. A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

    Energy Technology Data Exchange (ETDEWEB)

    John Darrell Bess

    2008-06-01

    A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for

  5. Estimation of Stability & Control Derivatives from Flight Test Data of Fighter Aircraft

    Directory of Open Access Journals (Sweden)

    L. Saraswathi

    2000-04-01

    Full Text Available Longitudinal stability and tontrol derivatives of a fightlfr aircraft are estimated by output error method for different types of input excitation. The uncertainties in the parameters are computed by cortfcting Cramer-Ra(j bounds using fudge factor. In general, the step input response data is not usedfor estimating the derivatives. Therefore, step response time history trajectories were cross-validated using tIle estimated derivatives for standard inputs like doublet and 3211. This proves that the model parameters are estimated with high confidence. By appropriately choosing the mathematical modeland using the corrected flight data for bias and scale factor errors by compatability check for parameter  estimation proves beyond doubt that such a procedure can be adopted for estimating stability and control derivatives of any aircraft.

  6. Analysis of volatile organic compounds emitted from aircraft carpets:comparison using headspace and dynamic chamber tests

    Institute of Scientific and Technical Information of China (English)

    WANG Chao; YANG Xu-Dong; GAO Peng

    2014-01-01

    Volatile organic compounds (VOCs) emitted from three types of carpets used in aircrafts were compared by using headspace and dynamic chamber tests. The headspace samples contained many compounds that were not detected in the dynamic chamber test;in addition, the dominant VOCs found by these two methods were different. The findings indicate that for highly sorptive materials such as carpets, headspace analysis may give inaccurate indication of actual VOC emissions, and it is necessary to conduct dynamic chamber tests over a certain period of time in order to identify the true emission characteristics. From the dynamic chamber tests, 2-ethyl-1-hexanol was the main VOC emitted from all three carpets. The study also examined the emission characteristics of aircraft carpets. In all experiments, total VOC (TVOC) concentration peaked within a few hours after the start of the experiment and was followed by rapid decay. The emission parameters of TVOC emitted by all three carpets were calculated and the simulated data matched the measured data well.

  7. James Webb Space Telescope (JWST) Integrated Science Instruments Module (ISIM) Electronics Compartment (IEC) Conformal Shields Composite Bond Structure Qualification Test Method

    Science.gov (United States)

    Yew, Calinda; Stephens, Matt

    2015-01-01

    The JWST IEC conformal shields are mounted onto a composite frame structure that must undergo qualification testing to satisfy mission assurance requirements. The composite frame segments are bonded together at the joints using epoxy, EA 9394. The development of a test method to verify the integrity of the bonded structure at its operating environment introduces challenges in terms of requirements definition and the attainment of success criteria. Even though protoflight thermal requirements were not achieved, the first attempt in exposing the structure to cryogenic operating conditions in a thermal vacuum environment resulted in approximately 1 bonded joints failure during mechanical pull tests performed at 1.25 times the flight loads. Failure analysis concluded that the failure mode was due to adhesive cracks that formed and propagated along stress concentrated fillets as a result of poor bond squeeze-out control during fabrication. Bond repairs were made and the structures successfully re-tested with an improved LN2 immersion test method to achieve protoflight thermal requirements.

  8. Summary of experimental tests of elastomeric seismic isolation bearings for use in nuclear reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    Seidensticker, R.W.; Chang, Y.W.; Kulak, R.F.

    1992-05-01

    This paper describes an experimental test program for isolator bearings which was developed to help establish the viability of using laminated elastomer bearings for base isolation of nuclear reactor plants. The goal of the test program is to determine the performance characteristics of laminated seismic isolation bearings under a wide range of loadings. Tests were performed on scale-size laminated seismic isolators both within the design shear strain range to determine the response of the bearing under expected earthquake loading conditions, and beyond the design range to determine failure modes and to establish safety margins. Three types of bearings, each produced from a different manufacturer, have been tested: (1) high shape factor-high damping-high shear modulus bearings; (2) medium shape factor-high damping-high shear modulus bearings; and (3) medium shape factor-high damping-low shear modulus bearings. All of these tests described in this report were performed at the Earthquake Engineering Research Center at the University of California, Berkeley, with technical assistance from ANL. The tests performed on the three types of bearings have confirmed the high performance characteristics of the high damping-high and low shear modulus elastomeric bearings. The bearings have shown that they are capable of having extremely large shear strains before failure occurs. The most common failure mechanism was the debonding of the top steel plate from the isolators. This failure mechanism can be virtually eliminated by improved manufacturing quality control. The most important result of the failure test of the isolators is the fact that bearings can sustain large horizontal displacement, several times larger than the design value, with failure. Their performance in moderate and strong earthquakes will be far superior to conventional structures.

  9. Summary of experimental tests of elastomeric seismic isolation bearings for use in nuclear reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    Seidensticker, R.W.; Chang, Y.W.; Kulak, R.F.

    1992-01-01

    This paper describes an experimental test program for isolator bearings which was developed to help establish the viability of using laminated elastomer bearings for base isolation of nuclear reactor plants. The goal of the test program is to determine the performance characteristics of laminated seismic isolation bearings under a wide range of loadings. Tests were performed on scale-size laminated seismic isolators both within the design shear strain range to determine the response of the bearing under expected earthquake loading conditions, and beyond the design range to determine failure modes and to establish safety margins. Three types of bearings, each produced from a different manufacturer, have been tested: (1) high shape factor-high damping-high shear modulus bearings; (2) medium shape factor-high damping-high shear modulus bearings; and (3) medium shape factor-high damping-low shear modulus bearings. All of these tests described in this report were performed at the Earthquake Engineering Research Center at the University of California, Berkeley, with technical assistance from ANL. The tests performed on the three types of bearings have confirmed the high performance characteristics of the high damping-high and low shear modulus elastomeric bearings. The bearings have shown that they are capable of having extremely large shear strains before failure occurs. The most common failure mechanism was the debonding of the top steel plate from the isolators. This failure mechanism can be virtually eliminated by improved manufacturing quality control. The most important result of the failure test of the isolators is the fact that bearings can sustain large horizontal displacement, several times larger than the design value, with failure. Their performance in moderate and strong earthquakes will be far superior to conventional structures.

  10. In-situ Creep Testing Capability Development for Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    B. G. Kim; J. L. Rempe; D. L. Knudson; K. G. Condie; B. H. Sencer

    2010-08-01

    Creep is the slow, time-dependent strain that occurs in a material under a constant strees (or load) at high temperature. High temperature is a relative term, dependent on the materials being evaluated. A typical creep curve is shown in Figure 1-1. In a creep test, a constant load is applied to a tensile specimen maintained at a constant temperature. Strain is then measured over a period of time. The slope of the curve, identified in the figure below, is the strain rate of the test during Stage II or the creep rate of the material. Primary creep, Stage I, is a period of decreasing creep rate due to work hardening of the material. Primary creep is a period of primarily transient creep. During this period, deformation takes place and the resistance to creep increases until Stage II, Secondary creep. Stage II creep is a period with a roughly constant creep rate. Stage II is referred to as steady-state creep because a balance is achieved between the work hardening and annealing (thermal softening) processes. Tertiary creep, Stage III, occurs when there is a reduction in cross sectional area due to necking or effective reduction in area due to internal void formation; that is, the creep rate increases due to necking of the specimen and the associated increase in local stress.

  11. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for FHRs

    Science.gov (United States)

    Lu, Qiuping

    Direct Reactor Auxiliary Cooling System (DRACS) is a passive decay heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines coated particle fuel and a graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three coupled natural circulation/convection loops, relying completely on buoyancy as the driving force. These loops are coupled through two heat exchangers, namely, the DRACS Heat Exchanger (DHX) and the Natural Draft Heat Exchanger (NDHX). In addition, a fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during normal operation of the reactor, but to keep the DRACS ready for activation, if needed, during accidents. While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for FHRs built or tested in the literature. The primary goal of the present research is to design, test, and model the DRACS for FHR applications. Previously, a detailed modular design of the DRACS for a 20-MWth FHR was developed. As a starting point, the DRACS was designed to remove 1% of the reactor nominal power, i.e., 200 kW decay power. In addition, a detailed scaling analysis has been performed to develop the key non-dimensional numbers that characterize the DRACS system. Based on the previous work on the prototypic DRACS design and scaling analysis, two scaled-down test facilities have been designed and constructed, namely, Low-temperature DRACS Test Facility (LTDF) and High-temperature DRACS Test Facility (HTDF). The LTDF has a nominal power capacity of 6 kW. It uses 1.0-MPa water as the primary coolant, 0.1-MPa water as the secondary coolant, and ambient air as the ultimate heat sink. The main purpose of the LTDF is to examine the couplings among the three natural circulation/convection loops in the DRACS, as well as to provide design and operation experience for the HTDF. An extensive test matrix has

  12. UO2 and PuO2 utilization in high temperature engineering test reactor with helium coolant

    Science.gov (United States)

    Waris, Abdul; Aji, Indarta K.; Novitrian, Pramuditya, Syeilendra; Su'ud, Zaki

    2016-03-01

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO2 fuel. In this study, we have evaluated the use of UO2 and PuO2 in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of 235U in loaded fuel is 18.0% or above.

  13. Force Reconstruction from Ejection Tests of Stores from Aircraft Used for Model Predictions and Missing/Bad Gages

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Michael; Cap, Jerome S.; Starr, Michael J.; Urbina, Angel; Brink, Adam Ray

    2015-12-01

    One of the more severe environments for a store on an aircraft is during the ejection of the store. During this environment it is not possible to instrument all component responses, and it is also likely that some instruments may fail during the environment testing. This work provides a method for developing these responses from failed gages and uninstrumented locations. First, the forces observed by the store during the environment are reconstructed. A simple sampling method is used to reconstruct these forces given various parameters. Then, these forces are applied to a model to generate the component responses. Validation is performed on this methodology.

  14. Long Term Out-of-pile Thermocouple Tests in Conditions Representative for Nuclear Gas-cooled High Temperature Reactors

    OpenAIRE

    LAURIE Mathias; FOURREZ Stephane; FUETTERER Michael; LAPETITE Jean-Marc

    2011-01-01

    During irradiation tests at high temperature, failure of commercial Inconel 600 sheathed thermocouples is commonly encountered. To understand and remediate this problem, out-ofpile tests were performed with thermocouples in carburizing atmospheres which can be assumed to be at least locally representative for High Temperature Reactors. The objective was to screen those thermocouples which would consecutively be used under irradiation. Two such screening tests have been performed with a set of...

  15. Supported Pd-Au Membrane Reactor for Hydrogen Production: Membrane Preparation, Characterization and Testing

    Directory of Open Access Journals (Sweden)

    Adolfo Iulianelli

    2016-05-01

    Full Text Available A supported Pd-Au (Au 7wt% membrane was produced by electroless plating deposition. Permeation tests were performed with pure gas (H2, H2, N2, CO2, CH4 for long time operation. After around 400 h under testing, the composite Pd-Au membrane achieved steady state condition, with an H2/N2 ideal selectivity of around 500 at 420 °C and 50 kPa as transmembrane pressure, remaining stable up to 1100 h under operation. Afterwards, the membrane was allocated in a membrane reactor module for methane steam reforming reaction tests. As a preliminary application, at 420 °C, 300 kPa of reaction pressure, space velocity of 4100 h−1, 40% methane conversion and 35% hydrogen recovery were reached using a commercial Ni/Al2O3 catalyst. Unfortunately, a severe coke deposition affected irreversibly the composite membrane, determining the loss of the hydrogen permeation characteristics of the supported Pd-Au membrane.

  16. Recirculation pump discharge line break tests at ROSA-III for a boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, M.; Anoda, Y.; Kumamaru, H.; Nakamura, H.; Shiba, M.; Tasaka, K.

    1985-08-01

    Three loss-of-coolant accident (LOCA) tests were conducted at the Rig of Safety Assessment (ROSA)-III test facility, which simulates boiling water reactor (BWR)/6-251 with a volumetric scaling factor of 1/424. The fundamental features of the recirculation pump discharge line break LOCA and the effects of break areas on the features are investigated. It has been confirmed experimentally that the LOCA phenomena in the discharge line break are analogous to those in the suction line break with the same effective choking flow area, which is a sum of the least choking flow areas along the break flow paths and controls the system pressure responses. In general, the maximum effective choking flow area is (A /SUB j/ + A /SUB p/ ) for discharge line breaks and (A /SUB j/ + A /SUB o/ ) for suction line breaks, where A /SUB j/ , A /SUB p/ , and A /SUB o/ are the flow areas of the jet pump drive nozzles, the main recirculation pump discharge nozzle, and the break, respectively. The similarity between the ROSA-III test and a BWR LOCA has been confirmed in the key phenomena by the analyses using the RELAP5/MOD1 code. An atypical behavior is observed in the fuel rod surface temperature transient in the early phase of blowdown due to the limitation of the ROSA-III initial core power.

  17. Flight Services and Aircraft Access: Active Flow Control Vertical Tail and Insect Accretion and Mitigation Flight Test

    Science.gov (United States)

    Whalen, Edward A.

    2016-01-01

    This document serves as the final report for the Flight Services and Aircraft Access task order NNL14AA57T as part of NASA Environmentally Responsible Aviation (ERA) Project ITD12A+. It includes descriptions of flight test preparations and execution for the Active Flow Control (AFC) Vertical Tail and Insect Accretion and Mitigation (IAM) experiments conducted on the 757 ecoDemonstrator. For the AFC Vertical Tail, this is the culmination of efforts under two task orders. The task order was managed by Boeing Research & Technology and executed by an enterprise-wide Boeing team that included Boeing Research & Technology, Boeing Commercial Airplanes, Boeing Defense and Space and Boeing Test and Evaluation. Boeing BR&T in St. Louis was responsible for overall Boeing project management and coordination with NASA. The 757 flight test asset was provided and managed by the BCA ecoDemonstrator Program, in partnership with Stifel Aircraft Leasing and the TUI Group. With this report, all of the required deliverables related to management of this task order have been met and delivered to NASA as summarized in Table 1. In addition, this task order is part of a broader collaboration between NASA and Boeing.

  18. Testing and Analysis of a Composite Non-Cylindrical Aircraft Fuselage Structure. Part 1; Ultimate Design Loads

    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 finite element analysis and 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 I of the paper considers the five most critical load conditions, which are internal pressure only and positive and negative g-loads with and without internal pressure. Analysis results are compared with measurements acquired during testing. Performance of the test article is found to be closely aligned with predictions and, consequently, able to support the hybrid wing body design loads in pristine and barely visible impact damage conditions.

  19. Fabrication and Comparison of Fuels for Advanced Gas Reactor Irradiation Tests

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey Phillips; Charles Barnes; John Hunn

    2010-10-01

    As part of the program to demonstrate TRISO-coated fuel for the Next Generation Nuclear Plant, a series of irradiation tests of Advanced Gas Reactor (AGR) fuel are being performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. In the first test, called “AGR-1,” graphite compacts containing approximately 300,000 coated particles were irradiated from December 2006 until November 2009. Development of AGR-1 fuel sought to replicate the properties of German TRISO-coated particles. No particle failures were seen in the nearly 3-year irradiation to a burn up of 19%. The AGR-1 particles were coated in a two-inch diameter coater. Following fabrication of AGR-1 fuel, process improvements and changes were made in each of the fabrication processes. Changes in the kernel fabrication process included replacing the carbon black powder feed with a surface-modified carbon slurry and shortening the sintering schedule. AGR-2 TRISO particles were produced in a six-inch diameter coater using a change size about twenty-one times that of the two-inch diameter coater used to coat AGR-1 particles. Changes were also made in the compacting process, including increasing the temperature and pressure of pressing and using a different type of press. Irradiation of AGR-2 fuel began in late spring 2010. Properties of AGR-2 fuel compare favorably with AGR-1 and historic German fuel. Kernels are more homogeneous in shape, chemistry and density. TRISO-particle sphericity, layer thickness standard deviations, and defect fractions are also comparable. In a sample of 317,000 particles from deconsolidated AGR-2 compacts, 3 exposed kernels were found in a leach test. No SiC defects were found in a sample of 250,000 deconsolidated particles, and no IPyC defects in a sample of 64,000 particles. The primary difference in properties between AGR-1 and AGR-2 compacts is that AGR-2 compacts have a higher matrix density, 1.6 g/cm3 compared to about 1.3 g/cm3 for AGR-1 compacts. Based on

  20. Counter-measure to prevent temperature rise of stand pipe and primary upper shielding in HTTR

    Energy Technology Data Exchange (ETDEWEB)

    Kunitomi, Kazuhiko; Tachibana, Yukio; Hontani, Kohji [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    1997-09-01

    When primary coolant temperature reached approximately 110degC during a preliminary non-nuclear heat up test in the High Temperature Engineering Test Reactor (HTTR), temperatures of stand pipes and a primary upper shielding increased more than expected. The cause of the temperature rise was investigated by tests and analyses, and we determined a counter-measure. We also confirmed that a modified structure due to this counter-measure does not affect flow distribution in the core, and is not in contact with a control rod wire. This paper describes the cause of the temperature rise, the modified structure and evaluation of effect of the modified structure. (author)

  1. THE COMPONENT TEST FACILITY – A NATIONAL USER FACILITY FOR TESTING OF HIGH TEMPERATURE GAS-COOLED REACTOR (HTGR) COMPONENTS AND SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    David S. Duncan; Vondell J. Balls; Stephanie L. Austad

    2008-09-01

    The Next Generation Nuclear Plant (NGNP) and other High-Temperature Gas-cooled Reactor (HTGR) Projects require research, development, design, construction, and operation of a nuclear plant intended for both high-efficiency electricity production and high-temperature industrial applications, including hydrogen production. During the life cycle stages of an HTGR, plant systems, structures and components (SSCs) will be developed to support this reactor technology. To mitigate technical, schedule, and project risk associated with development of these SSCs, a large-scale test facility is required to support design verification and qualification prior to operational implementation. As a full-scale helium test facility, the Component Test facility (CTF) will provide prototype testing and qualification of heat transfer system components (e.g., Intermediate Heat Exchanger, valves, hot gas ducts), reactor internals, and hydrogen generation processing. It will perform confirmation tests for large-scale effects, validate component performance requirements, perform transient effects tests, and provide production demonstration of hydrogen and other high-temperature applications. Sponsored wholly or in part by the U.S. Department of Energy, the CTF will support NGNP and will also act as a National User Facility to support worldwide development of High-Temperature Gas-cooled Reactor technologies.

  2. Appropriate concrete for nuclear reactor shielding.

    Science.gov (United States)

    Alhajali, S; Yousef, S; Naoum, B

    2016-01-01

    A survey for the neutron activation significant trace-minerals in limestones from different sites in Syria has been carried out. The specific saturation activation radioactivity was calculated and compared to the clearance levels. The results show that the limestone composition varies significantly from one site to another. For example Al-Tamoura site has ∑(C(i)/CL(i))≤1 at thermal neutron flux less than 10(7) (n cm(-2) s(-1)), while the limestone from other the three sites has ∑(C(i)/CL(i))≤1 at a flux less than 3×10(6) (n cm(-2) s(-1)).

  3. Hydrothermal Testing of K Basin Sludge and N Reactor Fuel at Sludge Treatment Project Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, Calvin H.; Schmidt, Andrew J.; Thornton, Brenda M.

    2007-03-30

    The Sludge Treatment Project (STP), managed for the U. S. DOE by Fluor Hanford (FH), was created to design and operate a process to eliminate uranium metal from K Basin sludge prior to packaging for Waste Isolation Pilot Plant (WIPP). The STP process uses high temperature liquid water to accelerate the reaction, produce uranium dioxide from the uranium metal, and safely discharge the hydrogen. Under nominal process conditions, the sludge will be heated in pressurized water at 185°C for as long as 72 hours to assure the complete reaction (corrosion) of up to 0.25-inch diameter uranium metal pieces. Under contract to FH, the Pacific Northwest National Laboratory (PNNL) conducted bench-scale testing of the STP hydrothermal process in November and December 2006. Five tests (~50 ml each) were conducted in sealed, un-agitated reaction vessels under the hydrothermal conditions (e.g., 7 to 72 h at 185°C) of the STP corrosion process using radioactive sludge samples collected from the K East Basin and particles/coupons of N Reactor fuel also taken from the K Basins. The tests were designed to evaluate and understand the chemical changes that may be occurring and the effects that any changes would have on sludge rheological properties. The tests were not designed to evaluate engineering aspects of the process. The hydrothermal treatment affected the chemical and physical properties of the sludge. In each test, significant uranium compound phase changes were identified, resulting from dehydration and chemical reduction reactions. Physical properties of the sludge were significantly altered from their initial, as-settled sludge values, including, shear strength, settled density, weight percent water, and gas retention.

  4. The Active Muon Shield

    CERN Document Server

    Bezshyiko, Iaroslava

    2016-01-01

    In the SHiP beam-dump of the order of 1011 muons will be produced per second. An active muon-shield is used to magnetically deflect these muons out of the acceptance of the spectrom- eter. This note describes how this shield is modelled and optimized. The SHiP spectrometer is being re-optimized using a conical decay-vessel, and utilizing the possibility to magnetize part of the beam-dump shielding iron. A shield adapted to these new conditions is presented which is significantly shorter and lighter than the shield used in the Technical Proposal (TP), while showing a similar performance.

  5. The research Of Multilayer Thermal Insulation With Mechanical Properties Based On Model Analysis Test

    Science.gov (United States)

    Lianhua, Yin

    The heat shield of aircraft is made of the major thrusts structure with multilayer thermal insulation part. For protecting against thermo-radiation from larger thrusting force engine,the heat shield is installed around this engine nearby.The multilayer thermal insulation part with multilayer radiation/reflection structure is made of reflection layer and interval layer.At vacuum condition,these materials is higher heat insulation capability than other material,is applied for lots of pats on aircraft extensively.But because of these material is made of metal and nonmetal,it is impossible to receive it's mechanical properties of materials from mechanical tests.These paper describes a new measure of mechanical properties of materials in the heat shield based on model analysis test.At the requirement for the first order lateral frequency,these measure provide for the FEM analysis foundation on the optimization structure of the heat shield.

  6. Hyper-velocity impact test and simulation of a double-wall shield concept for the Wide Field Monitor aboard LOFT

    CERN Document Server

    Perinati, E; Santangelo, A; Suchy, S; Tenzer, C; Del Monte, E; Herder, J -W den; Diebold, S; Feroci, M; Rachevski, A; Vacchi, A; Zampa, G; Zampa, N

    2014-01-01

    The space mission LOFT (Large Observatory For X-ray Timing) was selected in 2011 by ESA as one of the candidates for the M3 launch opportunity. LOFT is equipped with two instruments, the Large Area Detector (LAD) and the Wide Field Monitor (WFM), based on Silicon Drift Detectors (SDDs). In orbit, they would be exposed to hyper-velocity impacts by environmental dust particles, which might alter the surface properties of the SDDs. In order to assess the risk posed by these events, we performed simulations in ESABASE2 and laboratory tests. Tests on SDD prototypes aimed at verifying to what extent the structural damages produced by impacts affect the SDD functionality have been performed at the Van de Graaff dust accelerator at the Max Planck Institute for Nuclear Physics (MPIK) in Heidelberg. For the WFM, where we expect a rate of risky impacts notably higher than for the LAD, we designed, simulated and successfully tested at the plasma accelerator at the Technical University in Munich (TUM) a double-wall shield...

  7. Fabrication of U-10 wt.%Zr Metallic Fuel Rodlets for Irradiation Test in BOR-60 Fast Reactor

    OpenAIRE

    Ki-Hwan Kim; Jong-Hwan Kim; Seok-Jin Oh; Jung-Won Lee; Ho-Jin Lee; Chan-Bock Lee

    2016-01-01

    The fabrication technology for metallic fuel has been developed to produce the driver fuel in a PGSFR in Korea since 2007. In order to evaluate the irradiation integrity and validate the in-reactor of the starting metallic fuel with FMS cladding for the loading of the metallic fuel, U-10 wt.%Zr fuel rodlets were fabricated and evaluated for a verification of the starting driver fuel through an irradiation test in the BOR-60 fast reactor. The injection casting method was applied to U-10 wt.%Zr...

  8. Heat-pipe thermionic reactor concept

    DEFF Research Database (Denmark)

    Storm Pedersen, E.

    1967-01-01

    Main components are reactor core, heat pipe, thermionic converter, secondary cooling system, and waste heat radiator; thermal power generated in reactor core is transported by heat pipes to thermionic converters located outside reactor core behind radiation shield; thermionic emitters are in direct...

  9. Application of automatic inspection system to nondestructive test of heat transfer tubes of primary pressurized water cooler in the high temperature engineering test reactor. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Furusawa, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Miyamoto, Satoshi [Japan Atomic Power Company, Tokyo (Japan)

    2001-07-01

    Heat transfer tubes of a primary pressurized water cooled (PPWC) in the high temperature engineering test reactor (HTTR) form the reactor pressure boundary of the primary coolant, therefore are important from the viewpoint of safety. To establish inspection techniques for the heat transfer tubes of the PPWC, an automatic inspection system was developed. The system employs a bobbin coil probe, a rotating probe for eddy current testing (ECT) and a rotating probe for ultrasonic testing (UT). Nondestructive test of a half of the heat transfer tubes of the PPWC was carried out by the automatic inspection system during reactor shutdown period of the HTTR (about 55% in the maximum reactor power in this paper). The nondestructive test results showed that the maximum signal-to-noise ratio was 1.8 in ECT. Pattern and phase of Lissajous wave, which were obtained for the heat transfer tube of the PPWC, were different from those obtained for the artificially defected tube. In UT echo amplitude of the PPWC tubes inspected was lower than 20% of distance-amplitude calibration curve. Thus, it was confirmed that there was no defect in depth, which was more than the detecting standard of the probes, on the outer surface of the heat transfer tubes of the PPWC inspected. (author)

  10. A test study on treatment of high-strength polyester wastewater with anaerobic reactor

    Institute of Scientific and Technical Information of China (English)

    韩洪军; 陈秀荣; 徐春艳

    2002-01-01

    The treatment of polyester wastewater using Up-flow activated sludge bed anaerobic filer ( UASB-AF), demonstrated that UASB-AF reactors has a high efficiency, its volume loading is 10 ~ 12 kgCOD/( m3 @d) ,HRT is 22 ~24 h, and the removal of COD is about 80%. The reactor has advantage of fast starting andenduring pulse loading.

  11. Parametric experimental tests of steam gasification of pine wood in a fluidized bed reactor

    Directory of Open Access Journals (Sweden)

    L. Vecchione

    2013-09-01

    Full Text Available Among Renewable Energy Sources (RES, biomass represent one of the most common and suitable solution in order to contribute to the global energy supply and to reduce greenhouse gases (GHG emissions. The disposal of some residual biomass, as pruning from pine trees, represent a problem for agricultural and agro-industrial sectors. But if the residual biomass are used for energy production can become a resource. The most suitable energy conversion technology for the above-mentioned biomass is gasification process because the high C/N ratio and the low moisture content, obtained from the analysis. In this work a small-pilot bubbling-bed gasification plant has been designed, constructed and used in order to obtain, from the pine trees pruning, a syngas with low tar and char contents and high hydrogen content. The activities showed here are part of the activities carried out in the European 7FP UNIfHY project. In particular the aim of this work is to develop experimental test on a bench scale steam blown fluidized bed biomass gasifier. These tests will be utilized in future works for the simulations of a pilot scale steam fluidized bed gasifier (100 kWth fed with different biomass feedstock. The results of the tests include produced gas and tar composition as well gas, tar and char yield. Tests on a bench scale reactor (8 cm I.D. were carried out varying steam to biomass ratio from 0.5, 0.7 and 1 to 830°C.

  12. Laboratory test reactor for the investigation of liquid reducing agents in the selective catalytic reduction of NOx

    Science.gov (United States)

    Peitz, D.; Bernhard, A.; Elsener, M.; Kröcher, O.

    2011-08-01

    A test reactor was designed and built for investigating liquid reducing agents in the selective catalytic reduction (SCR) process in the laboratory. The design of the experimental setup is described in detail and its performance was evaluated. Using a glass nebulizer, liquid reducing agents were sprayed directly onto a catalyst positioned in a heated glass reactor with a length of 250 mm and an internal diameter of 20.4 mm or 40 mm. Model exhaust gases were mixed from individual gas components and were heated up to 450 °C in a heat exchanger before entering the reactor. The off-gas was analyzed using two complimentary techniques, a multi-component online FTIR gas analysis and a liquid quench gas absorption setup, to detect higher molecular compounds and aerosols. Due to the versatility of construction, processes not related to SCR, but involving three-phase reactions with gases, liquids and a catalyst, can also be investigated.

  13. Design and Testing of a Prototype Pixellated CZT Detector and Shield for Hard X-Ray Astronomy

    OpenAIRE

    Bloser, P. F.; Grindlay, J.E.; Narita, T; Jenkins, J. A.

    1999-01-01

    We report on the design and laboratory testing of a prototype imaging CZT detector intended for balloon flight testing in April 2000. The detector tests several key techniques needed for the construction of large-area CZT arrays, as required for proposed hard X-ray astronomy missions. Two 10 mm x 10 mm x 5 mm CZT detectors, each with a 4 x 4 array of 1.9 mm pixels on a 2.5 mm pitch, will be mounted in a ``flip-chip'' fashion on a printed circuit board carrier card; the detectors will be place...

  14. TEMPERATURE MONITORING OPTIONS AVAILABLE AT THE IDAHO NATIONAL LABORATORY ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    J.E. Daw; J.L. Rempe; D.L. Knudson; T. Unruh; B.M. Chase; K.L Davis

    2012-03-01

    As part of the Advanced Test Reactor National Scientific User Facility (ATR NSUF) program, the Idaho National Laboratory (INL) has developed in-house capabilities to fabricate, test, and qualify new and enhanced sensors for irradiation testing. To meet recent customer requests, an array of temperature monitoring options is now available to ATR users. The method selected is determined by test requirements and budget. Melt wires are the simplest and least expensive option for monitoring temperature. INL has recently verified the melting temperature of a collection of materials with melt temperatures ranging from 100 to 1000 C with a differential scanning calorimeter installed at INL’s High Temperature Test Laboratory (HTTL). INL encapsulates these melt wires in quartz or metal tubes. In the case of quartz tubes, multiple wires can be encapsulated in a single 1.6 mm diameter tube. The second option available to ATR users is a silicon carbide temperature monitor. The benefit of this option is that a single small monitor (typically 1 mm x 1 mm x 10 mm or 1 mm diameter x 10 mm length) can be used to detect peak irradiation temperatures ranging from 200 to 800 C. Equipment has been installed at INL’s HTTL to complete post-irradiation resistivity measurements on SiC monitors, a technique that has been found to yield the most accurate temperatures from these monitors. For instrumented tests, thermocouples may be used. In addition to Type-K and Type-N thermocouples, a High Temperature Irradiation Resistant ThermoCouple (HTIR-TC) was developed at the HTTL that contains commercially-available doped molybdenum paired with a niobium alloy thermoelements. Long duration high temperature tests, in furnaces and in the ATR and other MTRs, demonstrate that the HTIR-TC is accurate up to 1800 C and insensitive to thermal neutron interactions. Thus, degradation observed at temperatures above 1100 C with Type K and N thermocouples and decalibration due to transmutation with tungsten

  15. Lewis icing research tunnel test of the aerodynamic effects of aircraft ground deicing/anti-icing fluids

    Science.gov (United States)

    Runyan, L. James; Zierten, Thomas A.; Hill, Eugene G.; Addy, Harold E., Jr.

    1992-01-01

    A wind tunnel investigation of the effect of aircraft ground deicing/anti-icing fluids on the aerodynamic characteristics of a Boeing 737-200ADV airplane was conducted. The test was carried out in the NASA Lewis Icing Research Tunnel. Fluids tested include a Newtonian deicing fluid, three non-Newtonian anti-icing fluids commercially available during or before 1988, and eight new experimental non-Newtonian fluids developed by four fluid manufacturers. The results show that fluids remain on the wind after liftoff and cause a measurable lift loss and drag increase. These effects are dependent on the high-lift configuration and on the temperature. For a configuration with a high-lift leading-edge device, the fluid effect is largest at the maximum lift condition. The fluid aerodynamic effects are related to the magnitude of the fluid surface roughness, particularly in the first 30 percent chord. The experimental fluids show a significant reduction in aerodynamic effects.

  16. 飞机除冰液稳定性研究%Study on stability test of aircraft deicing fluids

    Institute of Scientific and Technical Information of China (English)

    彭华乔; 王强; 曾萍; 吴海涛; 张亚博; 张帆; 赵芯; 苏正良; 夏祖西

    2015-01-01

    在SAE AMS 1424和SAE AMS 1428的基础上,分析了I型、II型、III型和IV型飞机除冰液稳定性试验要求和步骤。建议我国积极开展飞机除冰液稳定性检测和研究工作,为除冰液的适航审定提供技术支持。%Based on SAE AMS 1424 and SAE AMS 1428 ,the requirements and procedures of stability test for aircraft deicing fluids type I,II,III and IV are introduced. In order to provide the technical support for deicing fluids airworthiness certification,the corresponding test and research shall be improved.

  17. Propulsion Control and Health Management (PCHM) Technology for Flight Test on the C-17 T-1 Aircraft

    Science.gov (United States)

    Simon, Donald L.; Garg, Sanjay; Venti, Michael

    2004-01-01

    The C-I 7 T-l Globemaster III is an Air Force flight research vehicle located at Edwards Air Force Base. NASA Dryden and the C-17 System Program Office have entered into a Memorandum of Agreement to permit NASA the use of the C-I 7 T-I to conduct flight research on a mutually coordinated schedule. The C-17 Propulsion Control and Health Management (PCHM) Working Group was formed in order to foster discussion and coordinate planning amongst the various government agencies conducting PCHM research with a potential need for flight testing, and to communicate to the PCHM community the capabilities of the C-17 T-l aircraft to support such flight testing. This paper documents the output of this Working Group, including a summary of the candidate PCHM technologies identified and their associated benefits relative to NASA goals and objectives.

  18. Thermal analysis for a spent reactor fuel storage test in granite

    Energy Technology Data Exchange (ETDEWEB)

    Montan, D.N.

    1980-09-01

    A test is conducted in which spent fuel assemblies from an operating commercial nuclear power reactor are emplaced in the Climax granite at the US Department of Energy`s Nevada Test Site. In this generic test, 11 canisters of spent PWR fuel are emplaced vertically along with 6 electrical simulator canisters on 3 m centers, 4 m below the floor of a storage drift which is 420 m below the surface. Two adjacent parallel drifts contain electrical heaters, operated to simulate (in the vicinity of the storage drift) the temperature fields of a large repository. This test, planned for up to five years duration, uses fairly young fuel (2.5 years out of core) so that the thermal peak will occur during the time frame of the test and will not exceed the peak that would not occur until about 40 years of storage had older fuel (5 to 15 years out of core) been used. This paper describes the calculational techniques and summarizes the results of a large number of thermal calculations used in the concept, basic design and final design of the spent fuel test. The results of the preliminary calculations show the effects of spacing and spent fuel age. Either radiation or convection is sufficient to make the drifts much better thermal conductors than the rock that was removed to create them. The combination of radiation and convection causes the drift surfaces to be nearly isothermal even though the heat source is below the floor. With a nominal ventilation rate of 2 m{sup 3}/s and an ambient rock temperature of 23{sup 0}C, the maximum calculated rock temperature (near the center of the heat source) is about 100{sup 0}C while the maximum air temperature in the drift is around 40{sup 0}C. This ventilation (1 m{sup 3}/s through the main drift and 1/2 m{sup 3}/s through each of the side drifts) will remove about 1/3 of the heat generated during the first five years of storage.

  19. Scaling Studies for High Temperature Test Facility and Modular High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schult; Paul D. Bayless; Richard W. Johnson; James R. Wolf; Brian Woods

    2012-02-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5-year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC).

  20. Anomalous fast ion losses at high β on the tokamak fusion test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fredrickson, E. D.; Bell, M. G.; Budny, R. V.; Darrow, D. S.; White, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

    2015-03-15

    This paper describes experiments carried out on the Tokamak Fusion Test Reactor (TFTR) [R. J. Hawryluk et al., Plasma Phys. Controlled Fusion 33, 1509 (1991)] to investigate the dependence of β-limiting disruption characteristics on toroidal field strength. The hard disruptions found at the β-limit in high field plasmas were not found at low field, even for β's 50% higher than the empirical β-limit of β{sub n} ≈ 2 at high field. Comparisons of experimentally measured β's to TRANSP simulations suggest anomalous loss of up to half of the beam fast ions in the highest β, low field shots. The anomalous transport responsible for the fast ion losses may at the same time broaden the pressure profile. Toroidal Alfvén eigenmodes, fishbone instabilities, and Geodesic Acoustic Modes are investigated as possible causes of the enhanced losses. Here, we present the first observations of high frequency fishbones [F. Zonca et al., Nucl. Fusion 49, 085009 (2009)] on TFTR. The interpretation of Axi-symmetric Beam-driven Modes as Geodesic Acoustic Modes and their possible correlation with transport barrier formation are also presented.

  1. Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report October 2014

    Energy Technology Data Exchange (ETDEWEB)

    Ogden, Dan [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

    Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report October 2014 Highlights • Rory Kennedy, Dan Ogden and Brenden Heidrich traveled to Germantown October 6-7, for a review of the Infrastructure Management mission with Shane Johnson, Mike Worley, Bradley Williams and Alison Hahn from NE-4 and Mary McCune from NE-3. Heidrich briefed the group on the project progress from July to October 2014 as well as the planned path forward for FY15. • Jim Cole gave two invited university seminars at Ohio State University and University of Florida, providing an overview of NSUF including available capabilities and the process for accessing facilities through the peer reviewed proposal process. • Jim Cole and Rory Kennedy co-chaired the NuMat meeting with Todd Allen. The meeting, sponsored by Elsevier publishing, was held in Clearwater, Florida, and is considered one of the premier nuclear fuels and materials conferences. Over 340 delegates attended with 160 oral and over 200 posters presented over 4 days. • Thirty-one pre-applications were submitted for NSUF access through the NE-4 Combined Innovative Nuclear Research Funding Opportunity Announcement. • Fourteen proposals were received for the NSUF Rapid Turnaround Experiment Summer 2014 call. Proposal evaluations are underway. • John Jackson and Rory Kennedy attended the Nuclear Fuels Industry Research meeting. Jackson presented an overview of ongoing NSUF industry research.

  2. Long term out-of-pile thermocouple tests in conditions representative for nuclear gas-cooled high temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Laurie, M., E-mail: mathias.laurie@ec.europa.eu [European Commission, Joint Research Centre, Institut für Transurane, Hermann-von-Helmholtz-Platz 1, 76344, Eggenstein-Leopoldshafen (Germany); Fourrez, S. [THERMOCOAX SAS, Rue du Pré Neuf, 61100 Saint Georges des Groseillers (France); Fütterer, M.A.; Lapetite, J.M. [European Commission, Joint Research Centre, Institut für Transurane, Hermann-von-Helmholtz-Platz 1, 76344, Eggenstein-Leopoldshafen (Germany); Sadli, M.; Morice, R.; Failleau, G. [Laboratoire commun de métrologie LNE-Cnam, 61 rue du Landy, F-92310 Saint-Denis (France)

    2014-05-01

    During irradiation tests at high temperature failure of commercial Inconel 600 sheathed thermocouples is commonly encountered. As instrumentation, in particular thermocouples are considered safety-relevant both for irradiation tests and for commercial reactors, JRC and THERMOCOAX joined forces to solve this issue by performing out-of-pile tests with thermocouples mimicking the environment encountered by high temperature reactor (HTR) in-core instrumentation. The objective was to screen innovative sheathed thermocouples which would consecutively be tested under irradiation. Two such screening tests have been performed in high temperature environment (i.e. temperature in the range 1100–1150 °C) with purposely contaminated helium atmosphere (mainly CH{sub 4}, CO, CO{sub 2}, O{sub 2} impurities) representative for high temperature reactor carburizing atmospheres. The first set of thermocouples embedded in graphite (mainly conventional N type thermocouples and thermocouples with innovative sheaths) was tested in a dedicated furnace at THERMOCOAX lab with helium flushing. The second out-of-pile test at JRC with a partly different set of thermocouples replicated the original test for comparison. Performance indicators such as thermal drift, insulation resistance and loop resistance were monitored. Through these long-term screening tests the effect of several parameters were investigated: niobium sleeves, bending, diameter, sheath composition as well as the chemical environment. SEM examinations were performed to analyze local damage (bending zone, sheath). The present paper describes the two tests, sums up data collected during these tests in terms of thermocouple behavior and describes further instrumentation testing work with fixed point mini cells for qualification under irradiation.

  3. Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014

    Energy Technology Data Exchange (ETDEWEB)

    Soelberg, Renae [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-11-01

    Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014 Highlights Rory Kennedy and Sarah Robertson attended the American Nuclear Society Winter Meeting and Nuclear Technology Expo in Anaheim, California, Nov. 10-13. ATR NSUF exhibited at the technology expo where hundreds of meeting participants had an opportunity to learn more about ATR NSUF. Dr. Kennedy briefed the Nuclear Engineering Department Heads Organization (NEDHO) on the workings of the ATR NSUF. • Rory Kennedy, James Cole and Dan Ogden participated in a reactor instrumentation discussion with Jean-Francois Villard and Christopher Destouches of CEA and several members of the INL staff. • ATR NSUF received approval from the NE-20 office to start planning the annual Users Meeting. The meeting will be held at INL, June 22-25. • Mike Worley, director of the Office of Innovative Nuclear Research (NE-42), visited INL Nov. 4-5. Milestones Completed • Recommendations for the Summer Rapid Turnaround Experiment awards were submitted to DOE-HQ Nov. 12 (Level 2 milestone due Nov. 30). Major Accomplishments/Activities • The University of California, Santa Barbara 2 experiment was unloaded from the GE-2000 at HFEF. The experiment specimen packs will be removed and shipped to ORNL for PIE. • The Terrani experiment, one of three FY 2014 new awards, was completed utilizing the Advanced Photon Source MRCAT beamline. The experiment investigated the chemical state of Ag and Pd in SiC shell of irradiated TRISO particles via X-ray Absorption Fine Structure (XAFS) spectroscopy. Upcoming Meetings/Events • The ATR NSUF program review meeting will be held Dec. 9-10 at L’Enfant Plaza. In addition to NSUF staff and users, NE-4, NE-5 and NE-7 representatives will attend the meeting. Awarded Research Projects Boise State University Rapid Turnaround Experiments (14-485 and 14-486) Nanoindentation and TEM work on the T91, HT9, HCM12A and 9Cr ODS specimens has been completed at

  4. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2012

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg, Principal Investigator; Kevin A. Steuhm, Project Manager

    2012-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to properly verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the next anticipated ATR Core Internals Changeout (CIC) in the 2014-2015 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its third full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL under various licensing arrangements. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core

  5. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2011

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Devin A. Steuhm

    2011-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or 'Core Modeling Update') Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the anticipated ATR Core Internals Changeout (CIC) in the 2014 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its first full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (SCALE, KENO-6, HELIOS, NEWT, and ATTILA) have been installed at the INL under various permanent sitewide license agreements and corresponding baseline models of the ATR and ATRC are now operational, demonstrating the basic feasibility of these code packages for their intended purpose. Furthermore

  6. Fabrication of U-10 wt.%Zr Metallic Fuel Rodlets for Irradiation Test in BOR-60 Fast Reactor

    Directory of Open Access Journals (Sweden)

    Ki-Hwan Kim

    2016-01-01

    Full Text Available The fabrication technology for metallic fuel has been developed to produce the driver fuel in a PGSFR in Korea since 2007. In order to evaluate the irradiation integrity and validate the in-reactor of the starting metallic fuel with FMS cladding for the loading of the metallic fuel, U-10 wt.%Zr fuel rodlets were fabricated and evaluated for a verification of the starting driver fuel through an irradiation test in the BOR-60 fast reactor. The injection casting method was applied to U-10 wt.%Zr fuel slugs with a diameter of 5.5 mm. Consequently, fuel slugs per melting batch without casting defects were fabricated through the development of advanced casting technology and evaluation tests. The optimal GTAW welding conditions were also established through a number of experiments. In addition, a qualification test was carried out to prove the weld quality of the end plug welding of the metallic fuel rodlets. The wire wrapping of metallic fuel rodlets was successfully accomplished for the irradiation test. Thus, PGSFR fuel rodlets have been soundly fabricated for the irradiation test in a BOR-60 fast reactor.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-03-01

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

  8. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950`s are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  9. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950's are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  10. CO2 Absorption in a Lab-Scale Fixed Solid Bed Reactor: Modelling and Experimental Tests

    Directory of Open Access Journals (Sweden)

    Roberto Gabbrielli

    2004-09-01

    Full Text Available The CO2 absorption in a lab-scale fixed solid bed reactor filled with different solid sorbents has been studied under different operative conditions regarding temperature (20-200°C and input gas composition (N2, O2, CO2, H2O at 1bar pressure. The gas leaving the reactor has been analysed to measure the CO2 and O2 concentrations and, consequently, to evaluate the overall CO2 removal efficiency. In order to study the influence of solid sorbent type (i.e. CaO, coal bottom ash, limestone and blast furnace slag and of mass and heat transfer processes on CO2 removal efficiency, a one-dimensional time dependent mathematical model of the reactor, which may be considered a Plug Flow Reactor, has been developed. The quality of the model has been confirmed using the experimental results.

  11. Small Punch Test on Before and Post Irradiated Domestic Reactor Pressure Steel

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    Problems may be caused when applying the standard specimen to study the properties of irradiated reactor materials, because of its big dimension, e.g.: The inner temperature gradient of the specimen is high when irradiated, the radiation

  12. Turbulent jet erosion of a stably stratified gas layer in a nuclear reactor test containment

    Energy Technology Data Exchange (ETDEWEB)

    Ishay, Liel [Department of Mechanical Engineering, Ben-Gurion University of the Negev, Beer-Sheva 84105 (Israel); Bieder, Ulrich [Commissariat à l’énergie atomique et aux énergies alternatives, Centre de SACLAY DEN/SAC/DANS/DM2S/STMF/LMSF, F-91191 Gif-sur-Yvette (France); Ziskind, Gennady [Department of Mechanical Engineering, Ben-Gurion University of the Negev, Beer-Sheva 84105 (Israel); Rashkovan, Alex, E-mail: rashbgu@gmail.com [Physics Department, Nuclear Research Center Negev (NRCN), PO Box 9001, Beer-Sheva 84190 (Israel)

    2015-10-15

    Highlights: • We model stably stratified layer erosion by vertical turbulent round jet. • Separate effect studies are performed as a platform for choosing modeling approach. • A test performed in MISTRA facility, CEA, Saclay is modeled using Fluent and Trio-U codes. • The proposed modeling approach showed good agreement with the MISTRA facility LOWMA-3 test. - Abstract: A number of integral and separate effect experiments were performed in the last two decades for validation of containment computational tools. The main goal of these benchmark experiments was to assess the ability of turbulence models and computational fluid dynamics codes to predict hydrogen concentration distribution and steam condensation rate in a nuclear reactor containment in the course of severe accidents. It appears from the published literature that the predictive capability of the existing computational tools still needs to be improved. This work examines numerically the temporal evolution of helium concentration in the experiment called LOWMA-3, performed in the MISTRA facility of CEA-Saclay, France. In the experiment, helium is used to mimic hydrogen of a real-case accident. The aim of this separate effect experiment, where steam condensation was not involved, is to predict helium concentration field. The conditions of the experiment are such that both the momentum transport and molecular diffusion contributions to the mixing process are of the same order of magnitude (Fr ∼ 1). A commercial CFD code, Fluent, and a CEA in-house code, Trio-U, are used for flow and helium concentration fields temporal evolution prediction in the present study. The preliminary separate effect studies provide guidance to an optimal modeling approach for the LOWMA-3 experiment. Temporal evolution of helium concentration in the stratification layer is shown, and a comparison to the experiment is discussed. It is shown that correct modeling of the round jet flowfield is essential for a reliable

  13. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  14. The design, development, and flight test results of the Boeing 737 aircraft antennas for the ICAO demonstration of the TRSB microwave landing system

    Science.gov (United States)

    Campbell, T. G.; White, W. E.; Gilreath, M. C.

    1976-01-01

    The Research Support Flight System, a modified Boeing 737, was used to evaluate the performance of several aircraft antennas and locations for the Time Reference Scanning Beam (TRSB) Microwave Landing System (MLS). These tests were conducted at the National Aviation Facilities Experimental Center (NAFEC), Atlantic City, New Jersey on December 18, 1975. The flight tests measured the signal strength and all pertinent MLS data during a straight-in approach, a racetrack approach, and ICAO approach profiles using the independent antenna-receiver combinations simultaneously on the aircraft. Signal drop-outs were experienced during the various approaches but only a small percentage could be attributed to antenna pattern effects.

  15. Construction and Testing of a Pixellated CZT Detector and Shield for a Hard X-ray Astronomy Balloon Flight

    OpenAIRE

    Bloser, P. F.; Narita, T; Jenkins, J. A.; Grindlay, J.E.

    2000-01-01

    We report on the construction and laboratory testing of pixellated CZT detectors mounted in a flip-chip, tiled fashion and read out by an ASIC, as required for proposed hard X-ray astronomy missions. Two 10 mm x 10 mm x 5 mm detectors were fabricated, one out of standard eV Products high-pressure Bridgman CZT and one out of IMARAD horizontal Bridgman CZT. Each was fashioned with a 4 x 4 array of gold pixels on a 2.5 mm pitch with a surrounding guard ring. The detectors were mounted side by si...

  16. Post test calculations of a severe accident experiment for VVER-440 reactors by the ATHLET code

    Energy Technology Data Exchange (ETDEWEB)

    Gyoergy, Hunor [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques (BME NTI); Trosztel, Istvan [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research (MTA EK)

    2013-09-15

    Severe accident - if no mitigation action is taken - leads to core melt. An effective severe accident management strategy can be the external reactor pressure vessel cooling for corium localization and stabilization. For some time discussion was going on, whether the in-vessel retention can be applied for the VVER-440 type reactors. It had to be demonstrated that the available space between the reactor vessel and biological protection allows sufficient cooling to keep the melted core in the vessel, without the reactor pressure vessel losing its integrity. In order to demonstrate the feasibility of the concept an experimental facility was realized in Hungary. The facility called Cooling Effectiveness on the Reactor External Surface (CERES) is modeling the vessel external surface and the biological protection of Paks NPP. A model of the CERES facility for the ATHLET TH system code was developed. The results of the ATHLET calculation agree well with the measurements showing that the vessel cooling can be insured for a long time in a VVER-440 reactor. (orig.)

  17. Aerothermodynamic Testing of Protuberances and Penetrations on the NASA Crew Exploration Vehicle Heat Shield in the NASA Langley 20-Inch Mach 6 Air Tunnel

    Science.gov (United States)

    Liechty, Derek S.

    2008-01-01

    An experimental wind tunnel program is being conducted in support of an Agency wide effort to develop a replacement for the Space Shuttle and to support the NASA s long-term objective of returning to the moon and then on to Mars. This paper documents experimental measurements made on several scaled ceramic heat transfer models of the proposed Crew Exploration Vehicle. Global heat transfer images and heat transfer distributions obtained using phosphor thermography were used to infer interference heating on the Crew Exploration Vehicle Cycle 1 heat shield from local protuberances and penetrations for both laminar and turbulent heating conditions. Test parametrics included free stream Reynolds numbers of 1.0x10(exp 6)/ft to 7.25x10(exp 6)/ft in Mach 6 air at a fixed angle-of-attack. Single arrays of discrete boundary layer trips were used to trip the boundary layer approaching the protuberances/penetrations to a turbulent state. Also, the effects of three compression pad diameters, two radial locations of compression pad/tension tie location, compression pad geometry, and rotational position of compression pad/tension tie were examined. The experimental data highlighted in this paper are to be used to validate CFD tools that will be used to generate the flight aerothermodynamic database. Heat transfer measurements will also assist in the determination of the most appropriate engineering methods that will be used to assess local flight environments associated with protuberances/penetrations of the CEV thermal protection system.

  18. Modelling of turbulent hydrocarbon combustion. Test of different reactor concepts for describing the interactions between turbulence and chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, C.; Kremer, H. [Ruhr-Universitaet Bochum, Lehrstuhl fuer Energieanlagentechnik, Bochum (Germany); Kilpinen, P.; Hupa, M. [Aabo Akademi, Turku (Finland). Combustion Chemistry Research Group

    1997-12-31

    The detailed modelling of turbulent reactive flows with CFD-codes is a major challenge in combustion science. One method of combining highly developed turbulence models and detailed chemistry in CFD-codes is the application of reactor based turbulence chemistry interaction models. In this work the influence of different reactor concepts on methane and NO{sub x} chemistry in turbulent reactive flows was investigated. Besides the classical reactor approaches, a plug flow reactor (PFR) and a perfectly stirred reactor (PSR), the Eddy-Dissipation Combustion Model (EDX) and the Eddy Dissipation Concept (EDC) were included. Based on a detailed reaction scheme and a simplified 2-step mechanism studies were performed in a simplified computational grid consisting of 5 cells. The investigations cover a temperature range from 1273 K to 1673 K and consider fuel-rich and fuel-lean gas mixtures as well as turbulent and highly turbulent flow conditions. All test cases investigated in this study showed a strong influence of the reactor residence time on the species conversion processes. Due to this characteristic strong deviations were found for the species trends resulting from the different reactor approaches. However, this influence was only concentrated on the `near burner region` and after 4-5 cells hardly any deviation and residence time dependence could be found. The importance of the residence time dependence increased when the species conversion was accelerated as it is the case for overstoichiometric combustion conditions and increased temperatures. The study focused furthermore on the fine structure in the EDC. Unlike the classical approach this part of the cell was modelled as a PFR instead of a PSR. For high temperature conditions there was hardly any difference between both reactor types. However, decreasing the temperature led to obvious deviations. Finally, the effect of the selective species transport between the cells on the conversion process was investigated

  19. SOC-DS computer code provides tool for design evaluation of homogeneous two-material nuclear shield

    Science.gov (United States)

    Disney, R. K.; Ricks, L. O.

    1967-01-01

    SOC-DS Code /Shield Optimization Code-Direc Search/, selects a nuclear shield material of optimum volume, weight, or cost to meet the requirments of a given radiation dose rate or energy transmission constraint. It is applicable to evaluating neutron and gamma ray shields for all nuclear reactors.

  20. Application of a Virtual Reactivity Feedback Control Loop in Non-Nuclear Testing of a Fast Spectrum Reactor

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Forsbacka, Matthew

    2004-01-01

    For a compact, fast-spectrum reactor, reactivity feedback is dominated by core deformation at elevated temperature. Given the use of accurate deformation measurement techniques, it is possible to simulate nuclear feedback in non-nuclear electrically heated reactor tests. Implementation of simulated reactivity feedback in response to measured deflection is being tested at the NASA Marshall Space Flight Center Early Flight Fission Test Facility (EFF-TF). During tests of the SAFE-100 reactor prototype, core deflection was monitored using a high resolution camera. "virtual" reactivity feedback was accomplished by applying the results of Monte Carlo calculations (MCNPX) to core deflection measurements; the computational analysis was used to establish the reactivity worth of van'ous core deformations. The power delivered to the SAFE-100 prototype was then dusted accordingly via kinetics calculations, The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kilowatt(sub t), to 10 kilowatts(sub t), held approximately constant at 10 kilowatts (sub t), and then allowed to decrease based on the negative thermal reactivity coefficient.