A New Equivalence Theory Method for Treating Doubly Heterogeneous Fuel - II. Verifications

DOE PAGES

Choi, Sooyoung; Kong, Chidong; Lee, Deokjung; ...

2015-03-09

A new methodology has been developed recently to treat resonance self-shielding in systems for which the fuel compact region of a reactor lattice consists of small fuel grains dispersed in a graphite matrix. The theoretical development adopts equivalence theory in both micro- and macro-level heterogeneities to provide approximate analytical expressions for the shielded cross sections, which may be interpolated from a table of resonance integrals or Bondarenko factors using a modified background cross section as the interpolation parameter. This paper describes the first implementation of the theoretical equations in a reactor analysis code. In order to reduce discrepancies caused bymore » use of the rational approximation for collision probabilities in the original derivation, a new formulation for a doubly heterogeneous Bell factor is developed in this paper to improve the accuracy of doubly heterogeneous expressions. This methodology is applied to a wide range of pin cell and assembly test problems with varying geometry parameters, material compositions, and temperatures, and the results are compared with continuous-energy Monte Carlo simulations to establish the accuracy and range of applicability of the new approach. It is shown that the new doubly heterogeneous self-shielding method including the Bell factor correction gives good agreement with reference Monte Carlo results.« less

36. DETAILS AND SECTIONS OF SHIELDING TANK, FUEL ELEMENT SUPPORT ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

36. DETAILS AND SECTIONS OF SHIELDING TANK, FUEL ELEMENT SUPPORT FRAME AND SUPPORT PLATFORM, AND SAFETY MECHANISM ASSEMBLY (SPRING-LOADED HINGE). F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-1. INEL INDEX CODE NUMBER: 075 0701 60 851 151975. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

Quantitative Fissile Assay In Used Fuel Using LSDS System

NASA Astrophysics Data System (ADS)

Lee, YongDeok; Jeon, Ju Young; Park, Chang-Je

2017-09-01

A quantitative assay of isotopic fissile materials (U235, Pu239, Pu241) was done at Korea Atomic Energy Research Institute (KAERI), using lead slowing down spectrometer (LSDS). The optimum design of LSDS was performed based on economics, easy maintenance and assay effectiveness. LSDS system consists of spectrometer, neutron source, detection and control. LSDS system induces fissile fission and fast neutrons are collected at fission chamber. The detected signal has a direct relation to the mass of existing fissile isotopes. Many current commercial assay technologies have a limitation in direct application on isotopic fissile assay of spent fuel, except chemical analysis. In the designed system, the fissile assay model was setup and the correction factor for self-shield was obtained. The isotopic fissile content assay was performed by changing the content of Pu239. Based on the fuel rod, the isotopic content was consistent with 2% uncertainty for Pu239. By applying the covering (neutron absorber), the effective shielding was obtained and the activation was calculated on the target. From the assay evaluation, LSDS technique is very powerful and direct to analyze the isotopic fissile content. LSDS is applicable for nuclear fuel cycle and spent fuel management for safety and economics. Additionally, an accurate fissile content will contribute to the international transparency and credibility on spent fuel.

Measurement of carbon distribution in nuclear fuel pin cladding specimens by means of a secondary ion mass spectrometer

NASA Astrophysics Data System (ADS)

Bart, Gerhard; Aerne, Ernst Tino; Burri, Martin; Zwicky, Hans-Urs

1986-11-01

Cladding carburization during irradiation of advanced mixed uranium plutonium carbide fast breeder reactor fuel is possibly a life limiting fuel pin factor. The quantitative assessment of such clad carbon embrittlement is difficult to perform by electron microprobe analysis because of sample surface contamination, and due to the very low energy of the carbon K α X-ray transition. The work presented here describes a method developed at the Swiss Federal Institute for Reactor Research (EIR) to use shielded secondary ion mass spectrometry (SIMS) as an accurate tool to determine radial distribution profiles of carbon in radioactive stainless steel fuel pin cladding. Compared with nuclear microprobe analysis (NMA) [1], which is also an accurate method for carbon analysis, the SIMS method distinguishes itself by its versatility for simultaneous determination of additional impurities.

Post-irradiation-examination of irradiated fuel outside the hot cell

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dawn E. Janney; Adam B. Robinson; Thomas P. O'Holleran

Because of their high radioactivity, irradiated fuels are commonly examined in a hot cell. However, the Idaho National Laboratory (INL) has recently investigated irradiated U-Mo-Al metallic fuel from the Reduced Enrichment for Research and Test Reactors (RERTR) project using a conventional unshielded scanning electron microscope outside a hot cell. This examination was possible because of a two-step sample-preparation approach in which a small volume of fuel was isolated in a hot cell and shielding was introduced during later stages of sample preparation. The resulting sample contained numerous sample-preparation artifacts but allowed analysis of microstructures from selected areas.

NEUTRON REACTOR HAVING A Xe$sup 135$ SHIELD

DOEpatents

Stanton, H.E.

1957-10-29

Shielding for reactors of the type in which the fuel is a chain reacting liquid composition comprised essentially of a slurry of fissionable and fertile material suspended in a liquid moderator is discussed. The neutron reflector comprises a tank containing heavy water surrounding the reactor, a shield tank surrounding the reflector, a gamma ray shield surrounding said shield tank, and a means for conveying gaseous fission products, particularly Xe/sup 135/, from the reactor chamber to the shield tank, thereby serving as a neutron shield by capturing the thermalized neutrons that leak outwardly from the shield tank.

Differential Die-Away Instrument: Report on Fuel Assembly Mock-up Measurements with Neutron Generator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goodsell, Alison Victoria; Swinhoe, Martyn Thomas; Henzl, Vladimir

2014-09-18

Fresh fuel experiments for the differential die-away (DDA) project were performed using a DT neutron generator, a 15x15 PWR fuel assembly, and nine 3He detectors in a water tank inside of a shielded cell at Los Alamos National Laboratory (LANL). Eight different fuel enrichments were created using low enriched (LEU) and depleted uranium (DU) dioxide fuel rods. A list-mode data acquisition system recorded the time-dependent signal and analysis of the DDA signal die-away time was performed. The die-away time depended on the amount of fissile material in the fuel assembly and the position of the detector. These experiments were performedmore » in support of the spent nuclear fuel Next Generation Safeguards Initiative DDA project. Lessons learned from the fresh fuel DDA instrument experiments and simulations will provide useful information to the spent fuel project.« less

630 A MARITIME NUCLEAR STEAM GENERATOR. Progress Report No. 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1962-09-28

A layout of a reduced-height 630A assembly (34 to 23 ft) was prepared and is presently being evaluated for use in a merchant vessel. While shielding studies indicate the need for some rearrangement of the shield materials, the desired radiation constraint can be obtained without an increase in shielding weight. A preliminary stress analysis of the pressure vessel, flow path analysis, and insulation evaluation was completed and showed no major problems. Evaluation of the total containment design indicates a design pressure of 45 psig. The Critical Experiment (CE) mockup is about 80% complete. The CE tank and dolly is aboutmore » 50% complete. The CE hazards report was reviewed and approved. The draft of the test program and procedures document is 75% complete. The LPT control room modifications were made, and the draft of the standard operating procedures completed. The CE fuel was inspected and a significant portion was found to be of no use, about 60% requires recoating. Creep and oxidation test time on some of the fuel sheet has exceeded 3000 hr with no significant oxidation or elongation on any of the samples. The nickel -chromium alloy sheet high- temperature (1750 F) stress and oxidating testing have exceeded 5000 hr with elongations below 0.8% except for one sample of 2.3%. Experimental fuel sheet samples were prepared and comparative property studies were ini-tiated. Fabrication of the ring test assembly, 3-F-1, for test in the MTR is essentially complete. The design of the test ring for seal evaluations was initiated. A detailed schedule for the work in FY 63 was prepared and issued for comments and concurrence. (auth)« less

Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Banerjee, Kaushik; Clarity, Justin B; Cumberland, Riley M

This will be licensed via RSICC. A new, integrated data and analysis system has been designed to simplify and automate the performance of accurate and efficient evaluations for characterizing the input to the overall nuclear waste management system -UNF-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). A relational database within UNF-ST&DARDS provides a standard means by which UNF-ST&DARDS can succinctly store and retrieve modeling and simulation (M&S) parameters for specific spent nuclear fuel analysis. A library of various analysis model templates provides the ability to communicate the various set of M&S parameters to the most appropriate M&S application.more » Interactive visualization capabilities facilitate data analysis and results interpretation. UNF-ST&DARDS current analysis capabilities include (1) assembly-specific depletion and decay, (2) and spent nuclear fuel cask-specific criticality and shielding. Currently, UNF-ST&DARDS uses SCALE nuclear analysis code system for performing nuclear analysis.« less

Nuclear Engine System Simulation (NESS) version 2.0

NASA Technical Reports Server (NTRS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.

1993-01-01

The topics are presented in viewgraph form and include the following; nuclear thermal propulsion (NTP) engine system analysis program development; nuclear thermal propulsion engine analysis capability requirements; team resources used to support NESS development; expanded liquid engine simulations (ELES) computer model; ELES verification examples; NESS program development evolution; past NTP ELES analysis code modifications and verifications; general NTP engine system features modeled by NESS; representative NTP expander, gas generator, and bleed engine system cycles modeled by NESS; NESS program overview; NESS program flow logic; enabler (NERVA type) nuclear thermal rocket engine; prismatic fuel elements and supports; reactor fuel and support element parameters; reactor parameters as a function of thrust level; internal shield sizing; and reactor thermal model.

Two-Dimensional Diffusion Theory Analysis of Reactivity Effects of a Fuel-Plate-Removal Experiment

NASA Technical Reports Server (NTRS)

Gotsky, Edward R.; Cusick, James P.; Bogart, Donald

1959-01-01

Two-dimensional two-group diffusion calculations were performed on the NASA reactor simulator in order to evaluate the reactivity effects of fuel plates removed successively from the center experimental fuel element of a seven- by three-element core loading at the Oak Ridge Bulk Shielding Facility. The reactivity calculations were performed by two methods: In the first, the slowing-down properties of the experimental fuel element were represented by its infinite media parameters; and, in the second, the finite size of the experimental fuel element was recognized, and the slowing-down properties of the surrounding core were attributed to this small region. The latter calculation method agreed very well with the experimented reactivity effects; the former method underestimated the experimental reactivity effects.

SCALE Code System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jessee, Matthew Anderson

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministicmore » and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.SCALE 6.2 provides many new capabilities and significant improvements of existing features.New capabilities include:• ENDF/B-VII.1 nuclear data libraries CE and MG with enhanced group structures,• Neutron covariance data based on ENDF/B-VII.1 and supplemented with ORNL data,• Covariance data for fission product yields and decay constants,• Stochastic uncertainty and correlation quantification for any SCALE sequence with Sampler,• Parallel calculations with KENO,• Problem-dependent temperature corrections for CE calculations,• CE shielding and criticality accident alarm system analysis with MAVRIC,• CE depletion with TRITON (T5-DEPL/T6-DEPL),• CE sensitivity/uncertainty analysis with TSUNAMI-3D,• Simplified and efficient LWR lattice physics with Polaris,• Large scale detailed spent fuel characterization with ORIGAMI and ORIGAMI Automator,• Advanced fission source convergence acceleration capabilities with Sourcerer,• Nuclear data library generation with AMPX, and• Integrated user interface with Fulcrum.Enhanced capabilities include:• Accurate and efficient CE Monte Carlo methods for eigenvalue and fixed source calculations,• Improved MG resonance self-shielding methodologies and data,• Resonance self-shielding with modernized and efficient XSProc integrated into most sequences,• Accelerated calculations with TRITON/NEWT (generally 4x faster than SCALE 6.1),• Spent fuel characterization with 1470 new reactor-specific libraries for ORIGEN,• Modernization of ORIGEN (Chebyshev Rational Approximation Method [CRAM] solver, API for high-performance depletion, new keyword input format)• Extension of the maximum mixture number to values well beyond the previous limit of 2147 to ~2 billion,• Nuclear data formats enabling the use of more than 999 energy groups,• Updated standard composition library to provide more accurate use of natural abundances, andvi• Numerous other enhancements for improved usability and stability.« less

Apparatus for in situ determination of burnup, cooling time and fissile content of an irradiated nuclear fuel assembly in a fuel storage pond

DOEpatents

Phillips, John R.; Halbig, James K.; Menlove, Howard O.; Klosterbuer, Shirley F.

1985-01-01

A detector head for in situ inspection of irradiated nuclear fuel assemblies submerged in a water-filled nuclear fuel storage pond. The detector head includes two parallel arms which extend from a housing and which are spaced apart so as to be positionable on opposite sides of a submerged fuel assembly. Each arm includes an ionization chamber and two fission chambers. One fission chamber in each arm is enclosed in a cadmium shield and the other fission chamber is unshielded. The ratio of the outputs of the shielded and unshielded fission chambers is used to determine the boron content of the pond water. Correcting for the boron content, the neutron flux and gamma ray intensity are then used to verify the declared exposure, cooling time and fissile material content of the irradiated fuel assembly.

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

NASA Technical Reports Server (NTRS)

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 shielding material is not used for dual purpose, and is only used for shielding, then the additional weight per aircraft is estimated to be 428 lb. This weight estimate is based upon a number of assumptions that would need to be revised when applying this concept to an actual airplane design. For example, the weight savings that will result when there is no fan blade containment system, manufacturing limitations which may increase the weight where variable thicknesses was assumed, engine placement on the wing versus aft fuselage, etc.

Horizontal modular dry irradiated fuel storage system

DOEpatents

Fischer, Larry E.; McInnes, Ian D.; Massey, John V.

1988-01-01

A horizontal, modular, dry, irradiated fuel storage system (10) includes a thin-walled canister (12) for containing irradiated fuel assemblies (20), which canister (12) can be positioned in a transfer cask (14) and transported in a horizontal manner from a fuel storage pool (18), to an intermediate-term storage facility. The storage system (10) includes a plurality of dry storage modules (26) which accept the canister (12) from the transfer cask (14) and provide for appropriate shielding about the canister (12). Each module (26) also provides for air cooling of the canister (12) to remove the decay heat of the irradiated fuel assemblies (20). The modules (26) can be interlocked so that each module (26) gains additional shielding from the next adjacent module (26). Hydraulic rams (30) are provided for inserting and removing the canisters (12) from the modules (26).

HAZARDS SUMMARY REPORT FOR A TWO WATT PROMETHIUM-147 FUELED THERMOELECTRIC GENERATOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1959-06-01

Discussions are included of the APU design, vehicle integration, Pm/sup 147/ properties, shielding requirements, hazards design criteria, statistical analysis for impact, and radiation protection. The use of Pm/sup 147/ makes possible the fabrication of an auxiliary power unit which has applications for low power space missions of <10 watts (electrical). (B.O.G.)

Multicanister overpack topical report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lorenz, B.D., Fluor Daniel Hanford

1997-03-25

The Spent Nuclear Fuel MCO is a single-use container that consists of a cylindrical shell, five to six fuel baskets, a shield plug, and features necessary for maintaining the structural integrity of the MCO while providing criticality control and fuel processing capability.

Nondestrucive analysis of fuel pins

DOEpatents

Stepan, I.E.; Allard, N.P.; Suter, C.R.

1972-11-03

Disclosure is made of a method and a correspondingly adapted facility for the nondestructive analysis of the concentation of fuel and poison in a nuclear reactor fuel pin. The concentrations of fuel and poison in successive sections along the entire length of the fuel pin are determined by measuring the reactivity of a thermal reactor as each successive small section of the fuel pin is exposed to the neutron flux of the reactor core and comparing the measured reactivity with the reactivities measured for standard fuel pins having various known concentrations. Only a small section of the length of the fuel pin is exposed to the neutron flux at any one time while the remainder of the fuel pin is shielded from the neutron flux. In order to expose only a small section at any one time, a boron-10-lined dry traverse tube is passed through the test region within the core of a low-power thermal nuclear reactor which has a very high fuel sensitivity. A narrow window in the boron-10 lining is positioned at the core center line. The fuel pins are then systematically traversed through the tube past the narrow window such that successive small sections along the length of the fuel pin are exposed to the neutron flux which passes through the narrow window.

Screening of advanced cladding materials and UN-U3Si5 fuel

NASA Astrophysics Data System (ADS)

Brown, Nicholas R.; Todosow, Michael; Cuadra, Arantxa

2015-07-01

In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO2-Zr fuel-cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN-U3Si5 fuels with Kanthal AF or APMT cladding. The objective of the U3Si5 phase in the UN-U3Si5 fuel concept is to shield the nitride phase from water. It was shown that UN-U3Si5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO2-Zr fuel-cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to 14N content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels are not expected to have significantly different relative burn-up distributions at discharge relative to the UO2 reference fuel. However, the overall harder spectrum in the UN ceramic composite fuels increases transuranic build-up, which will increase long-term activity in a once-thru fuel cycle but is expected to be a significant advantage in a fuel cycle with continuous recycling of transuranic material. It is recognized that the fuel and cladding properties assumed in these assessments are preliminary, and that additional data are necessary for these materials, most significantly under irradiation.

ADM. Tanks: from left to right: fuel oil tank, fuel ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ADM. Tanks: from left to right: fuel oil tank, fuel pump house (TAN-611), engine fuel tank, water pump house, water storage tank. Camera facing northwest. Not edge of shielding berm at left of view. Date: November 25, 1953. INEEL negative no. 9217 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

Split-core heat-pipe reactors for out-of-pile thermionic power systems.

NASA Technical Reports Server (NTRS)

Niederauer, G.; Lantz, E.; Breitweiser, R.

1971-01-01

Description of the concept of splitting a heat-pipe reactor for out-of-core thermionics into two identical halves and using the resulting center gap for reactivity control. Short Li-W reactor heat pipes penetrate the axial reflectors and form a heat exchanger with long heat pipes which wind through the shield to the thermionic diodes. With one reactor half anchored to the shield, the other is attached to a long arm with a pivot behind the shield and swings through a small arc for reactivity control. A safety shim prevents large reactivity inputs, and a fueled control arm drive shaft acts as a power stabilizer. Reactors fueled with U-235C and with U-233C have been studied.-

INTERIOR PHOTO OF THE REMOTE ANALYTICAL FACILITY OF SHIELDED GLOVE ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

INTERIOR PHOTO OF THE REMOTE ANALYTICAL FACILITY OF SHIELDED GLOVE BOXES IN OPERATING CORRIDOR (CPP-627). INL PHOTO NUMBER NRTS-55-1524. Unknown Photographer, 1955 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Shield system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Finch, D.R.; Chandler, J.R.; Church, J.P.

1979-01-01

The SHIELD system is a powerful new computational tool for calculation of isotopic inventory, radiation sources, decay heat, and shielding assessment in part of the nuclear fuel cycle. The integrated approach used in this system permitss the communication and management of large fields of numbers efficiently thus permitting the user to address the technical rather than computer aspects of a problem. Emphasis on graphical outputs permits large fields of resulting numbers to be efficiently displayed.

Delayed Gamma-ray Spectroscopy for Safeguards Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mozin, Vladimir

The delayed gamma-ray assay technique utilizes an external neutron source (D-D, D-T, or electron accelerator-driven), and high-resolution gamma-ray spectrometers to perform characterization of SNM materials behind shielding and in complex configurations such as a nuclear fuel assembly. High-energy delayed gamma-rays (2.5 MeV and above) observed following the active interrogation, provide a signature for identification of specific fissionable isotopes in a mixed sample, and determine their relative content. Potential safeguards applications of this method are: 1) characterization of fresh and spent nuclear fuel assemblies in wet or dry storage; 2) analysis of uranium enrichment in shielded or non-characterized containers or inmore » the presence of a strong radioactive background and plutonium contamination; 3) characterization of bulk and waste and product streams at SNM processing plants. Extended applications can include warhead confirmation and warhead dismantlement confirmation in the arms control area, as well as SNM diagnostics for the emergency response needs. In FY16 and prior years, the project has demonstrated the delayed gamma-ray measurement technique as a robust SNM assay concept. A series of empirical and modeling studies were conducted to characterize its response sensitivity, develop analysis methodologies, and analyze applications. Extensive experimental tests involving weapons-grade Pu, HEU and depleted uranium samples were completed at the Idaho Accelerator Center and LLNL Dome facilities for various interrogation time regimes and effects of the neutron source parameters. A dedicated delayed gamma-ray response modeling technique was developed and its elements were benchmarked in representative experimental studies, including highresolution gamma-ray measurements of spent fuel at the CLAB facility in Sweden. The objective of the R&D effort in FY17 is to experimentally demonstrate the feasibility of the delayed gamma-ray interrogation of shielded SNM samples with portable neutron sources suitable for field applications.« less

75 FR 13600 - Nine Mile Point Nuclear Station, LLC, Nine Mile Point Nuclear Station, Unit No. 2; Draft...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-22

... shielding design and the ALARA program would continue in its current form. Offsite Doses at EPU Conditions..., such as fossil fuel or alternative fuel power generation, to provide electric generation capacity to offset future demand. Construction and operation of such a fossil-fueled or alternative-fueled plant may...

PATHFINDER ATOMIC POWER PLANT TECHNICAL PROGRESS REPORT FOR JULY 1, 1959- SEPTEMBER 30, 1959

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1960-10-31

ABS>Fuel Element Research and Development. Dynamic and static corrosion tests on 8001 Al were completed. Annealmmmg of 1100 cladding on 5083 and M400 cladding on X2219 were tested at 500 deg C, and investigation continued on producing X8101 Al alloy cladding in tube plates by extrusion. Boiler fuel element capsule irradiation tests and subassembly tests are described Heat transfer loop studies and fuel fabrication for the critical facility are reported. Boiler fuel element mechanical design and testing progress is desc ribed. and the superheater fuel element temperature evaluating routine is discussed. Low- enrichment superheater fuel element development included design studiesmore » and stainless steel powder and UO/sub 2/ powder fabrication studies Reactor Mechanical Studies. Research is reported on vessel and structure design, fabrication, and testing, recirculation system design, steam separator tests, and control rod studies. Nuclear Analysis. Reactor physics studies are reported on nuclear constants, baffle plate analysis, comparison of core representations, delayed neutron fraction. and shielding analysis of the reactor building. Reactor and system dynamics and critical experiments were also studied. Chemistry. Progress is reported on recombiner. radioactive gas removal and storage, ion exchanger and radiochemical processing. (For preceding period see ACNP-5915.) (T.R.H.)« less

NRC approves spent-fuel cask for general use: Who needs Yucca Mountain?

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simpson, J.

1993-07-01

The Nuclear Regulatory Commission (NRC) on April 7, 1993, added Pacific Sierra Nuclear Associates`s (PSNA`s) VSC-24 spent-fuel container to its list of approved storage casks. Unlike previously approved designs, however, the cask was made available for use by utilities without site-specific approval. The VSC-24 (ventilated storage cask) is a 130-ton, 16-foot high vertical storage container composed of a ventilated concrete cask (VCC) housing a steel multi-assembly sealed basket (MSB). A third component, a transfer cask (MTC), shields, supports, and protects the MSB during fuel loading and VCC loading operations. The VCC is a cylindrical reinforced-concrete cask 29 inches thick, withmore » a 1.75-inch-thick A 36 steel liner. The cask contains eight vents-four on the top and four on the bottom-to provide for MSB (and fuel rod) cooling. Its concrete shell provides protection against shearing and penetration by tornado projectiles, protects the MSB in the event of a drop or tipover, and is designed to withstand internal temperatures of 350 degrees Farenheit. The VCC is closed with a bolted-down cover of 0.75-inch-thick A 36 steel. The MSB, which provides the primary boundary for 24 spent fuel rods, is a cylindrical steel shell with a thick shield plug and steel cover plates welded at each end. The shell and covers are constructed from SA 516 Grade 70 pressure vessel steel. Fuel is housed in a basket fabricated from SA 516 Grade 70 sheet steel. Penetrations in the MSB`s structural and shield lids allow for vacuum drying and backfilling with helium after fuel loading. Although its manufacturer claims a design life of 50 years, the NRC has licensed the VSC-24 cask for 20 years.« less

Spent nuclear fuel dry transfer system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stewart, L.; Agace, S.

The U.S. Department of Energy is currently engaged in a cooperative program with the Electric Power Research Institute (EPRI) to design a spent nuclear fuel dry transfer system (DTS). The system will enable the transfer of individual spent nuclear fuel assemblies between a conventional top loading cask and multi-purpose canister in a shielded overpack, or accommodate spent nuclear fuel transfers between two conventional casks.

Skeletal Mechanism Generation of Surrogate Jet Fuels for Aeropropulsion Modeling

NASA Astrophysics Data System (ADS)

Sung, Chih-Jen; Niemeyer, Kyle E.

2010-05-01

A novel implementation for the skeletal reduction of large detailed reaction mechanisms using the directed relation graph with error propagation and sensitivity analysis (DRGEPSA) is developed and presented with skeletal reductions of two important hydrocarbon components, n-heptane and n-decane, relevant to surrogate jet fuel development. DRGEPSA integrates two previously developed methods, directed relation graph-aided sensitivity analysis (DRGASA) and directed relation graph with error propagation (DRGEP), by first applying DRGEP to efficiently remove many unimportant species prior to sensitivity analysis to further remove unimportant species, producing an optimally small skeletal mechanism for a given error limit. It is illustrated that the combination of the DRGEP and DRGASA methods allows the DRGEPSA approach to overcome the weaknesses of each previous method, specifically that DRGEP cannot identify all unimportant species and that DRGASA shields unimportant species from removal.

LPT. Shield test facility test building interior (TAN646). Camera facing ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

LPT. Shield test facility test building interior (TAN-646). Camera facing south. Distant pool contained EBOR reactor; near pool was intended for fuel rod storage. Other post-1970 activity equipment remains in pool. INEEL negative no. HD-40-9-4 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

Importance of resonance interference effects in multigroup self-shielding calculation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stachowski, R.E.; Protsik, R.

1995-12-31

The impact of the resonance interference method (RIF) on multigroup neutron cross sections is significant for major isotopes in the fuel, indicating the importance of resonance interference in the computation of gadolinia burnout and plutonium buildup. The self-shielding factor method with the RIF method effectively eliminates shortcomings in multigroup resonance calculations.

Nuclear thermal propulsion engine system design analysis code development

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.; Ivanenok, Joseph F.

1992-01-01

A Nuclear Thermal Propulsion (NTP) Engine System Design Analyis Code has recently been developed to characterize key NTP engine system design features. Such a versatile, standalone NTP system performance and engine design code is required to support ongoing and future engine system and vehicle design efforts associated with proposed Space Exploration Initiative (SEI) missions of interest. Key areas of interest in the engine system modeling effort were the reactor, shielding, and inclusion of an engine multi-redundant propellant pump feed system design option. A solid-core nuclear thermal reactor and internal shielding code model was developed to estimate the reactor's thermal-hydraulic and physical parameters based on a prescribed thermal output which was integrated into a state-of-the-art engine system design model. The reactor code module has the capability to model graphite, composite, or carbide fuels. Key output from the model consists of reactor parameters such as thermal power, pressure drop, thermal profile, and heat generation in cooled structures (reflector, shield, and core supports), as well as the engine system parameters such as weight, dimensions, pressures, temperatures, mass flows, and performance. The model's overall analysis methodology and its key assumptions and capabilities are summarized in this paper.

Three Dimensional Forming Simulation of the Shielded Slot Plate for the MCFC Using a Ductile Fracture Criterion

NASA Astrophysics Data System (ADS)

Lee, C. H.; Yang, D. Y.; Lee, S. R.; Chang, I. G.; Lee, T. W.

2011-08-01

The shielded slot plate, which has a sheared corrugated trapezoidal pattern, is a component of the metallic bipolar plate for the molten carbonate fuel cell (MCFC). In order to increase the efficiency of the fuel cell, the unit cell of the shielded slot plate should have a relatively large upper area. Additionally, defects from the forming process should be minimized. In order to simulate the slitting process, whereby sheared corrugated patterns are formed, ductile fracture criteria based on the histories of stress and strain are employed. The user material subroutine VUMAT is employed for implementation of the material and ductile fracture criteria in the commercial FEM software ABAQUS. The variables of the ductile fracture criteria were determined by comparing the simulation results and the experimental results of the tension test and the shearing test. Parametric studies were conducted to determine the critical value of the ductile fracture criterion. Employing these ductile fracture criteria, the three dimensional forming process of the shielded slot plate was numerically simulated. The effects of the slitting process in the forming process of the shielded slot plate were analyzed through a FEM simulation and experimental studies. Finally, experiments involving microscopic and macroscopic observations were conducted to verify the numerical simulations of the 3-step forming process.

Neutron flux measurements on a mock-up of a storage cask for high-level nuclear waste using 2.5 MeV neutrons.

PubMed

Suárez, H Saurí; Becker, F; Klix, A; Pang, B; Döring, T

2018-06-07

To store and dispose spent nuclear fuel, shielding casks are employed to reduce the emitted radiation. To evaluate the exposure of employees handling such casks, Monte Carlo radiation transport codes can be employed. Nevertheless, to assess the reliability of these codes and nuclear data, experimental checks are required. In this study, a neutron generator (NG) producing neutrons of 2.5 MeV was employed to simulate neutrons produced in spent nuclear fuel. Different configurations of shielding layers of steel and polyethylene were positioned between the target of the NG and a NE-213 detector. The results of the measurements of neutron and γ radiation and the corresponding simulations with the code MCNP6 are presented. Details of the experimental set-up as well as neutron and photon flux spectra are provided as reference points for such NG investigations with shielding structures.

Resonance treatment using pin-based pointwise energy slowing-down method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Choi, Sooyoung, E-mail: csy0321@unist.ac.kr; Lee, Changho, E-mail: clee@anl.gov; Lee, Deokjung, E-mail: deokjung@unist.ac.kr

A new resonance self-shielding method using a pointwise energy solution has been developed to overcome the drawbacks of the equivalence theory. The equivalence theory uses a crude resonance scattering source approximation, and assumes a spatially constant scattering source distribution inside a fuel pellet. These two assumptions cause a significant error, in that they overestimate the multi-group effective cross sections, especially for {sup 238}U. The new resonance self-shielding method solves pointwise energy slowing-down equations with a sub-divided fuel rod. The method adopts a shadowing effect correction factor and fictitious moderator material to model a realistic pointwise energy solution. The slowing-down solutionmore » is used to generate the multi-group cross section. With various light water reactor problems, it was demonstrated that the new resonance self-shielding method significantly improved accuracy in the reactor parameter calculation with no compromise in computation time, compared to the equivalence theory.« less

78 FR 73379 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-06

..., and criticality control. If there is no loss of confinement, shielding, or criticality control, the... would prevent loss of confinement, shielding, and criticality control. If there is no loss of...;Federal Register / Vol. 78, No. 235 / Friday, December 6, 2013 / Rules and Regulations#0;#0; [[Page 73379...

COOLED NEUTRONIC REACTOR

DOEpatents

Binner, C.R.; Wilkie, C.B.

1958-03-18

This patent relates to a design for a reactor of the type in which a fluid coolant is flowed through the active portion of the reactor. This design provides for the cooling of the shielding material as well as the reactor core by the same fluid coolant. The core structure is a solid moderator having coolant channels in which are disposed the fuel elements in rod or slug form. The coolant fluid enters the chamber in the shield, in which the core is located, passes over the inner surface of said chamber, enters the core structure at the center, passes through the coolant channels over the fuel elements and out through exhaust ducts.

Influence of gamma-ray skyshine on nuclear facilities design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ohta, M.; Tsuji, M.; Kimura, Y.

1986-01-01

In safety analysis of nuclear facilities, skyshine dose rate at site boundary is one of the most important shielding design problems. For nuclear power stations in Japan, the skyshine dose rate at the site boundary has been specified not to exceed 5 mR/yr by the authorities, including total dose contribution from all structures on site, and this guide is commonly applied to other nuclear fuel cycle facilities. Therefore the design criterion dose of each structure on site is, considering plot planning, shielding condition, and so on, defined as a value <5 mR/yr. The purpose of this study is to investigatemore » how skyshine dose standards or other factors have an influence on the design of nuclear facilities, in a parametric survey of gamma-ray skyshine.« less

ENGINEERING TEST REACTOR

DOEpatents

De Boisblanc, D.R.; Thomas, M.E.; Jones, R.M.; Hanson, G.H.

1958-10-21

Heterogeneous reactors of the type which is both cooled and moderated by the same fluid, preferably water, and employs highly enriched fuel are reported. In this design, an inner pressure vessel is located within a main outer pressure vessel. The reactor core and its surrounding reflector are disposed in the inner pressure vessel which in turn is surrounded by a thermal shield, Coolant fluid enters the main pressure vessel, fiows downward into the inner vessel where it passes through the core containing tbe fissionable fuel assemblies and control rods, through the reflector, thence out through the bottom of the inner vessel and up past the thermal shield to the discharge port in the main vessel. The fuel assemblles are arranged in the core in the form of a cross having an opening extending therethrough to serve as a high fast flux test facility.

EMMI-Electric solar wind sail facilitated Manned Mars Initiative

NASA Astrophysics Data System (ADS)

Janhunen, Pekka; Merikallio, Sini; Paton, Mark

2015-08-01

The novel propellantless electric solar wind sail concept promises efficient low thrust transportation in the Solar System outside Earth's magnetosphere. Combined with asteroid mining to provide water and synthetic cryogenic rocket fuel in orbits of Earth and Mars, possibilities for affordable continuous manned presence on Mars open up. Orbital fuel and water enable reusable bidirectional Earth-Mars vehicles for continuous manned presence on Mars and allow smaller fuel fraction of spacecraft than what is achievable by traditional means. Water can also be used as radiation shielding of the manned compartment, thus reducing the launch mass further. In addition, the presence of fuel in the orbit of Mars provides the option for an all-propulsive landing, thus potentially eliminating issues of heavy heat shields and augmenting the capability of pinpoint landing. With this E-sail enabled scheme, the recurrent cost of continuous bidirectional traffic between Earth and Mars might ultimately approach the recurrent cost of running the International Space Station, ISS.

IRRADIATION METHOD AND APPARATUS

DOEpatents

Cabell, C.P.

1962-12-18

A method and apparatus are described for changing fuel bodies into a process tube of a reactor. According to this method fresh fuel elements are introduced into one end of the tube forcing used fuel elements out the other end. When sufficient fuel has been discharged, a reel and tape arrangement is employed to pull the column of bodies back into the center of the tube. Due provision is made for providing shielding in the tube. (AEC)

SCALE Code System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rearden, Bradley T.; Jessee, Matthew Anderson

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministicmore » and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.« less

SCALE Code System 6.2.1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rearden, Bradley T.; Jessee, Matthew Anderson

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministicmore » and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.« less

Thermoelectric generator with hinged assembly for fins

DOEpatents

Purdy, David L.; Shapiro, Zalman M.; Hursen, Thomas F.; Maurer, Gerould W.

1976-11-02

A cylindrical casing has a central shielded capsule of radioisotope fuel. A plurality of thermonuclear modules are axially arranged with their hot junctions resiliently pressed toward the shield and with their cold junctions adjacent a transition member having fins radiating heat to the environment. For each module, the assembly of transition member and fins is hinged to the casing for swinging to permit access to and removal of such module. A ceramic plate having gold layers on opposite faces prevents diffusion bonding of the hot junction to the shield.

Reactivity loss validation of high burn-up PWR fuels with pile-oscillation experiments in MINERVE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leconte, P.; Vaglio-Gaudard, C.; Eschbach, R.

2012-07-01

The ALIX experimental program relies on the experimental validation of the spent fuel inventory, by chemical analysis of samples irradiated in a PWR between 5 and 7 cycles, and also on the experimental validation of the spent fuel reactivity loss with bum-up, obtained by pile-oscillation measurements in the MINERVE reactor. These latter experiments provide an overall validation of both the fuel inventory and of the nuclear data responsible for the reactivity loss. This program offers also unique experimental data for fuels with a burn-up reaching 85 GWd/t, as spent fuels in French PWRs never exceeds 70 GWd/t up to now.more » The analysis of these experiments is done in two steps with the APOLLO2/SHEM-MOC/CEA2005v4 package. In the first one, the fuel inventory of each sample is obtained by assembly calculations. The calculation route consists in the self-shielding of cross sections on the 281 energy group SHEM mesh, followed by the flux calculation by the Method Of Characteristics in a 2D-exact heterogeneous geometry of the assembly, and finally a depletion calculation by an iterative resolution of the Bateman equations. In the second step, the fuel inventory is used in the analysis of pile-oscillation experiments in which the reactivity of the ALIX spent fuel samples is compared to the reactivity of fresh fuel samples. The comparison between Experiment and Calculation shows satisfactory results with the JEFF3.1.1 library which predicts the reactivity loss within 2% for burn-up of {approx}75 GWd/t and within 4% for burn-up of {approx}85 GWd/t. (authors)« less

Atmospheric reentry of the in-core thermionic SP-100 reactor system

NASA Technical Reports Server (NTRS)

Stamatelatos, M. G.; Barsell, A. W.; Harris, P. A.; Francisco, J.

1987-01-01

Presumed end-of-life atmospheric reentry of the GA SP-100 system was studied to assess dispersal feasibility and associated hazards. Reentry was studied by sequential use of an orbital trajectory and a heat analysis computer program. Two heating models were used. The first model assumed a thermal equilibrium condition between the stagnation point aerodynamic heating and the radiative cooling of the skin material surface. The second model allowed for infinite conductivity of the skin material. Four reentering configurations were studied representing stages of increased SP-100 breakup: (1) radiator, shield and reactor, (2) shield and reactor, (3) reactor with control drums, and (4) reactor without control drums. Each reentering configuration was started from a circular orbit at 116 km having an inertial velocity near Mach 25. The assumed failing criterion was the attainment of melting temperature of a critical system component. The reentry analysis revealed breakup of the vessel in the neighborhood of 61 km altitude and scattering of the fuel elements. Subsequent breakup of the fuel elements was not predicted. Oxidation of the niobium skin material was calculated to cause an increase in surface temperature of less than ten percent. The concept of thermite analogs for enhancing reactor reentry dispersal was assessed and found to be feasible in principle. A conservative worst-case hazards analysis was performed for radioactive and nonradioactive toxic SP-100 materials assumed to be dispersed during end-of-life reentry. The hazards associated with this phase of the SP-100 mission were calculated to be insignificant.

Integrated Solar Concentrator and Shielded Radiator

NASA Technical Reports Server (NTRS)

Clark, David Larry

2010-01-01

A shielded radiator is integrated within a solar concentrator for applications that require protection from high ambient temperatures with little convective heat transfer. This innovation uses a reflective surface to deflect ambient thermal radiation, shielding the radiator. The interior of the shield is also reflective to provide a view factor to deep space. A key feature of the shield is the parabolic shape that focuses incoming solar radiation to a line above the radiator along the length of the trough. This keeps the solar energy from adding to the radiator load. By placing solar cells along this focal line, the concentration of solar energy reduces the number and mass of required cells. By shielding the radiator, the effective reject temperature is much lower, allowing lower radiator temperatures. This is particularly important for lower-temperature processes, like habitat heat rejection and fuel cell operations where a high radiator temperature is not feasible. Adding the solar cells in the focal line uses the concentrating effect of the shield to advantage to accomplish two processes with a single device. This shield can be a deployable, lightweight Mylar structure for compact transport.

Radiation Templates of Spent Fuel in Casks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vanier, Peter

BNL and INL propose to perform a scoping study, using heavily collimated gamma and fast neutron detectors, to obtain passive radiation templates of dry storage casks containing spent fuel. The goal is to demonstrate sufficient spatial resolution and sensitivity to detect a missing fuel assembly. Such measurements, combined with detailed modeling and decay corrections should provide confidence that the cask contents have not been altered, despite loss of continuity of knowledge (CoK). The concept relies on the leakage of high energy gammas and neutrons through the shielding of the casks. Tests will emphasize organic scintillators with pulse shape discrimination, butmore » baseline comparisons will be made to high purity germanium (HPGe) and collimated moderated 3He detectors deployed in the same locations. Commercial off-the-shelf (COTS) detectors and data acquisition electronics will be used with custom-built collimators and shielding.« less

46 CFR 119.455 - Fuel piping.

Code of Federal Regulations, 2012 CFR

2012-10-01

..., such as seamless steel pipe or tubing, which provide equivalent safety may be used; (ii) Diesel fuel... bowls of other than steel construction must be approved by the Commandant and be protected from mechanical damage. Approval of bowls of other than steel construction will specify if a flame shield is...

46 CFR 119.455 - Fuel piping.

Code of Federal Regulations, 2013 CFR

2013-10-01

..., such as seamless steel pipe or tubing, which provide equivalent safety may be used; (ii) Diesel fuel... bowls of other than steel construction must be approved by the Commandant and be protected from mechanical damage. Approval of bowls of other than steel construction will specify if a flame shield is...

46 CFR 119.455 - Fuel piping.

Code of Federal Regulations, 2011 CFR

2011-10-01

..., such as seamless steel pipe or tubing, which provide equivalent safety may be used; (ii) Diesel fuel... bowls of other than steel construction must be approved by the Commandant and be protected from mechanical damage. Approval of bowls of other than steel construction will specify if a flame shield is...

46 CFR 119.455 - Fuel piping.

Code of Federal Regulations, 2014 CFR

2014-10-01

..., such as seamless steel pipe or tubing, which provide equivalent safety may be used; (ii) Diesel fuel... bowls of other than steel construction must be approved by the Commandant and be protected from mechanical damage. Approval of bowls of other than steel construction will specify if a flame shield is...

MOX fuel assembly design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reese, A.P.; Crowther, R.L. Jr.

1992-02-18

This patent describes improvement in a boiling water reactor core having a plurality of vertically upstanding fuel bundles; each fuel bundle containing longitudinally extending sealed rods with fissile material therein; the improvement comprises the fissile material including a mixture of uranium and recovered plutonium in rods of the fuel bundle at locations other than the corners of the fuel bundle; and, neutron absorbing material being located in rods of the fuel bundle at rod locations adjacent the corners of the fuel bundles whereby the neutron absorbing material has decreased shielding from the plutonium and maximum exposure to thermal neutrons formore » shaping the cold reactivity shutdown zone in the fuel bundle.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Han, Bo-Young; Choi, Daewoong; Park, Se Hwan

Korea Atomic Energy Research Institute (KAERI) have been developing the design and deployment methodology of Laser- Induced Breakdown Spectroscopy (LIBS) instrument for safeguards application within the argon hot cell environment at Advanced spent fuel Conditioning Process Facility (ACPF), where ACPF is a facility being refurbished for the laboratory-scaled demonstration of advanced spent fuel conditioning process. LIBS is an analysis technology used to measure the emission spectra of excited elements in the local plasma of a target material induced by a laser. The spectra measured by LIBS are analyzed to verify the quality and quantity of the specific element in themore » target matrix. Recently LIBS has been recognized as a promising technology for safeguards purposes in terms of several advantages including a simple sample preparation and in-situ analysis capability. In particular, a feasibility study of LIBS to remotely monitor the nuclear material in a high radiation environment has been carried out for supporting the IAEA safeguards implementation. Fiber-Optic LIBS (FO-LIBS) deployment was proposed by Applied Photonics Ltd because the use of fiber optics had benefited applications of LIBS by delivering the laser energy to the target and by collecting the plasma light. The design of FO-LIBS instrument for the measurement of actinides in the spent fuel and high temperature molten salt at ACPF had been developed in cooperation with Applied Photonics Ltd. FO-LIBS has some advantages as followings: the detectable plasma light wavelength range is not limited by the optical properties of the thick lead-glass shield window and the potential risk of laser damage to the lead-glass shield window is not considered. The remote LIBS instrument had been installed at ACPF and then the feasibility study for monitoring actinide elements such as uranium, plutonium, and curium in process materials has been carried out. (authors)« less

Inlet nozzle assembly

DOEpatents

Christiansen, David W.; Karnesky, Richard A.; Precechtel, Donald R.; Smith, Bob G.; Knight, Ronald C.

1987-01-01

An inlet nozzle assembly for directing coolant into the duct tube of a fuel assembly attached thereto. The nozzle assembly includes a shell for housing separable components including an orifice plate assembly, a neutron shield block, a neutron shield plug, and a diffuser block. The orifice plate assembly includes a plurality of stacked plates of differently configurated and sized openings for directing coolant therethrough in a predesigned flow pattern.

Inlet nozzle assembly

DOEpatents

Christiansen, D.W.; Karnesky, R.A.; Knight, R.C.; Precechtel, D.R.; Smith, B.G.

1985-09-09

An inlet nozzle assembly for directing coolant into the duct tube of a fuel assembly attached thereto. The nozzle assembly includes a shell for housing separable components including an orifice plate assembly, a neutron shield block, a neutron shield plug, and a diffuser block. The orifice plate assembly includes a plurality of stacked plates of differently configurated and sized openings for directing coolant therethrough in a predesigned flow pattern.

Analysis of loss-of-coolant accident for a fast-spectrum lithium-cooled nuclear reactor for space-power applications

NASA Technical Reports Server (NTRS)

Turney, G. E.; Petrik, E. J.; Kieffer, A. W.

1972-01-01

A two-dimensional, transient, heat-transfer analysis was made to determine the temperature response in the core of a conceptual space-power nuclear reactor following a total loss of reactor coolant. With loss of coolant from the reactor, the controlling mode of heat transfer is thermal radiation. In one of the schemes considered for removing decay heat from the core, it was assumed that the 4 pi shield which surrounds the core acts as a constant-temperature sink (temperature, 700 K) for absorption of thermal radiation from the core. Results based on this scheme of heat removal show that melting of fuel in the core is possible only when the emissivity of the heat-radiating surfaces in the core is less than about 0.40. In another scheme for removing the afterheat, the core centerline fuel pin was replaced by a redundant, constant temperature, coolant channel. Based on an emissivity of 0.20 for all material surfaces in the core, the calculated maximum fuel temperature for this scheme of heat removal was 2840 K, or about 90 K less than the melting temperature of the UN fuel.

MEANS FOR SHIELDING AND COOLING REACTORS

DOEpatents

Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

1959-02-10

Reactors of the water-cooled type and a means for shielding such a rcactor to protect operating personnel from harmful radiation are discussed. In this reactor coolant tubes which contain the fissionable material extend vertically through a mass of moderator. Liquid coolant enters through the bottom of the coolant tubes and passes upwardly over the fissionable material. A shield tank is disposed over the top of the reactor and communicates through its bottom with the upper end of the coolant tubes. A hydrocarbon shielding fluid floats on the coolant within the shield tank. With this arrangements the upper face of the reactor can be opened to the atmosphere through the two superimposed liquid layers. A principal feature of the invention is that in the event radioactive fission products enter thc coolant stream. imposed layer of hydrocarbon reduces the intense radioactivity introduced into the layer over the reactors and permits removal of the offending fuel material by personnel shielded by the uncontaminated hydrocarbon layer.

Desert Shield Leader’s Safety Guide

DTIC Science & Technology

1990-12-01

banana oil) vapor is toxic and flammable. Checking the seal of the protective mask should be done in a well-ventilated area away from heat and flames...protective clothing to keep fuel off the skin. (Skin is highly susceptible to drying, cracking, and peeling if it comes in contact with fuel in desert

Casting evaluation of U-Zr alloy system fuel slug for SFR prepared by injection casting method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Song, Hoon; Kim, Jong-Hwan; Kim, Ki-Hwan

2013-07-01

Metal fuel slugs of U-Pu-Zr alloys for Sodium-cooled Fast Reactor (SFR) have conventionally been fabricated by a vacuum injection casting method. Recently, management of minor actinides (MA) became an important issue because direct disposal of the long-lived MA can be a long-term burden for a tentative repository up to several hundreds of thousand years. In order to recycle transuranic elements (TRU) retained in spent nuclear fuel, remote fabrication capability in a shielded hot cell should be prepared. Moreover, generation of long-lived radioactive wastes and loss of volatile species should be minimized during the recycled fuel fabrication step. In order tomore » prevent the evaporation of volatile elements such as Am, alternative fabrication methods of metal fuel slugs have been studied applying gravity casting, and improved injection casting in KAERI, including melting under inert atmosphere. And then, metal fuel slugs were examined with casting soundness, density, chemical analysis, particle size distribution and microstructural characteristics. Based on these results there is a high level of confidence that Am losses will also be effectively controlled by application of a modest amount of overpressure. A surrogate fuel slug was generally soundly cast by improved injection casting method, melted fuel material under inert atmosphere.« less

High-Fidelity Modelng and Simulation for a High Flux Isotope Reactor Low-Enriched Uranium Core Design

DOE PAGES

Betzler, Benjamin R.; Chandler, David; Davidson, Eva E.; ...

2017-05-08

A high-fidelity model of the High Flux Isotope Reactor (HFIR) with a low-enriched uranium (LEU) fuel design and a representative experiment loading has been developed to serve as a new reference model for LEU conversion studies. With the exception of the fuel elements, this HFIR LEU model is completely consistent with the current highly enriched uranium HFIR model. Results obtained with the new LEU model provide a baseline for analysis of alternate LEU fuel designs and further optimization studies. The newly developed HFIR LEU model has an explicit representation of the HFIR-specific involute fuel plate geometry, including the within-plate fuelmore » meat contouring, and a detailed geometry model of the fuel element side plates. Such high-fidelity models are necessary to accurately account for the self-shielding from 238U and the depletion of absorber materials present in the side plates. In addition, a method was developed to account for fuel swelling in the high-density LEU fuel plates during the depletion simulation. In conclusion, calculated time-dependent metrics for the HFIR LEU model include fission rate and cumulative fission density distributions, flux and reaction rates for relevant experiment locations, point kinetics data, and reactivity coefficients.« less

High-Fidelity Modelng and Simulation for a High Flux Isotope Reactor Low-Enriched Uranium Core Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Betzler, Benjamin R.; Chandler, David; Davidson, Eva E.

A high-fidelity model of the High Flux Isotope Reactor (HFIR) with a low-enriched uranium (LEU) fuel design and a representative experiment loading has been developed to serve as a new reference model for LEU conversion studies. With the exception of the fuel elements, this HFIR LEU model is completely consistent with the current highly enriched uranium HFIR model. Results obtained with the new LEU model provide a baseline for analysis of alternate LEU fuel designs and further optimization studies. The newly developed HFIR LEU model has an explicit representation of the HFIR-specific involute fuel plate geometry, including the within-plate fuelmore » meat contouring, and a detailed geometry model of the fuel element side plates. Such high-fidelity models are necessary to accurately account for the self-shielding from 238U and the depletion of absorber materials present in the side plates. In addition, a method was developed to account for fuel swelling in the high-density LEU fuel plates during the depletion simulation. In conclusion, calculated time-dependent metrics for the HFIR LEU model include fission rate and cumulative fission density distributions, flux and reaction rates for relevant experiment locations, point kinetics data, and reactivity coefficients.« less

Demonstrating the Viability and Affordability of Nuclear Surface Power Systems

NASA Technical Reports Server (NTRS)

Vandyke, Melissa K.

2006-01-01

A set of tasks have been identified to help demonstrate the viability, performance, and affordability of surface fission systems. Completion of these tasks will move surface fission systems closer to reality by demonstrating affordability and performance potential. Tasks include fabrication and test of a 19-pin section of a Surface Power Unit Demonstrator (SPUD); design, fabrication, and utilization of thermal simulators optimized for surface fission' applications; design, fabrication, and utilization of GPHS module thermal simulators; design, fabrication, and test of a fission surface power system shield; and work related to potential fission surface power fuel/clad systems. Work on the SPUD will feed directly into joint NASA MSFC/NASA GRC fabrication and test of a surface power plant Engineering Development Unit (EDU). The goal of the EDU will be to perform highly realistic thermal, structural, and electrical testing on an integrated fission surface power system. Fission thermal simulator work will help enable high fidelity non-nuclear testing of pumped NaK surface fission power systems. Radioisotope thermal simulator work will help enable design and development of higher power radioisotope systems (power ultimately limited by Pu-238 availability). Shield work is designed to assess the potential of using a water neutron shield on the surface of the moon. Fuels work is geared toward assessing the current potential of using fuels that have already flown in space.

CFD Analysis of Spray Combustion and Radiation in OMV Thrust Chamber

NASA Technical Reports Server (NTRS)

Giridharan, M. G.; Krishnan, A.; Przekwas, A. J.; Gross, K.

1993-01-01

The Variable Thrust Engine (VTE), developed by TRW, for the Orbit Maneuvering Vehicle (OMV) uses a hypergolic propellant combination of Monomethyl Hydrazine (MMH) and Nitrogen Tetroxide (NTO) as fuel and oxidizer, respectively. The propellants are pressure fed into the combustion chamber through a single pintle injection element. The performance of this engine is dependent on the pintle geometry and a number of complex physical phenomena and their mutual interactions. The most important among these are (1) atomization of the liquid jets into fine droplets; (2) the motion of these droplets in the gas field; (3) vaporization of the droplets (4) turbulent mixing of the fuel and oxidizer; and (5) hypergolic reaction between MMH and NTO. Each of the above phenomena by itself poses a considerable challenge to the technical community. In a reactive flow field of the kind occurring inside the VTE, the mutual interactions between these physical processes tend to further complicate the analysis. The objective of this work is to develop a comprehensive mathematical modeling methodology to analyze the flow field within the VTE. Using this model, the effect of flow parameters on various physical processes such as atomization, spray dynamics, combustion, and radiation is studied. This information can then be used to optimize design parameters and thus improve the performance of the engine. The REFLEQS CFD Code is used for solving the fluid dynamic equations. The spray dynamics is modeled using the Eulerian-Lagrangian approach. The discrete ordinate method with 12 ordinate directions is used to predict the radiative heat transfer in the OMV combustion chamber, nozzle, and the heat shield. The hypergolic reaction between MMH and NTO is predicted using an equilibrium chemistry model with 13 species. The results indicate that mixing and combustion is very sensitive to the droplet size. Smaller droplets evaporate faster than bigger droplets, leading to a well mixed zone in the combustion chamber. The radiative heat flux at combustion chamber and nozzle walls are an order of negligible less than the conductive heat flux. Simulations performed with the heat shield show that a negligible amount of fluid is entrained into the heat shield region. However, the heat shield is shown to be effective in protecting the OMV structure surrounding the engine from the radiated heat.

Commercial D-T FRC Power Plant Systems Analysis

NASA Astrophysics Data System (ADS)

Nguyen, Canh; Santarius, John; Emmert, Gilbert; Steinhauer, Loren; Stubna, Michael

1998-11-01

Results of an engineering issues scoping study of a Field-Reversed Configuration (FRC) burning D-T fuel will be presented. The study primarily focuses on engineering issues, such as tritium-breeding blanket design, radiation shielding, neutron damage, activation, safety, and environment. This presentation will concentrate on plasma physics, current drive, economics, and systems integration, which are important for the overall systems analysis. A systems code serves as the key tool in defining a reference point for detailed physics and engineering calculations plus parametric variations, and typical cases will be presented. Advantages of the cylindrical geometry and high beta (plasma pressure/magnetic-field pressure) are evident.

76 FR 33991 - Airworthiness Directives; The Boeing Company Model 727, 727C, 727-100, 727-100C, 727-200, and 727...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-10

... manufacturer. We are issuing this AD to increase the level of protection from lightning strikes and prevent the... of protection from lightning strikes and prevent the potential of ignition sources inside fuel tanks... existing unshielded fuel quantity indication system (FQIS) wire bundles with double shielded FQIS wire...

5. CONSTRUCTION PROGRESS VIEW OF ASSEMBLY USED TO RAISE AND ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

5. CONSTRUCTION PROGRESS VIEW OF ASSEMBLY USED TO RAISE AND LOWER FUEL ELEMENTS. TAKEN FROM TOP OF SHIELDING TANK WITH CAMERA POINTING TOWARDS BOTTOM OF TANK. SHOWS LADDER, SQUARE LIFTING FRAME, FUEL ELEMENT HOLDERS, AND CABLE CYLINDERS. INEL PHOTO NUMBER 65-5434, TAKEN OCTOBER 20, 1965. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

DOE Office of Scientific and Technical Information (OSTI.GOV)

T.L. Lotz

1997-02-15

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide boiling water reactor (BWR) assembly radiation source term data for use during Waste Package (WP) design. The BWR assembly radiation source terms are to be used for evaluation of radiolysis effects at the WP surface, and for personnel shielding requirements during assembly or WP handling operations. The objectives of this evaluation are to generate BWR assembly radiation source terms that bound selected groupings of BWR assemblies, with regard to assembly average burnup and cooling time, which comprise the anticipated MGDS BWR commercialmore » spent nuclear fuel (SNF) waste stream. The source term data is to be provided in a form which can easily be utilized in subsequent shielding/radiation dose calculations. Since these calculations may also be used for Total System Performance Assessment (TSPA), with appropriate justification provided by TSPA, or radionuclide release rate analysis, the grams of each element and additional cooling times out to 25 years will also be calculated and the data included in the output files.« less

SOLID GAS SUSPENSION NUCLEAR FUEL ASSEMBLY

DOEpatents

Schluderberg, D.C.; Ryon, J.W.

1962-05-01

A fuel assembly is designed for use in a gas-suspension cooled nuclear fuel reactor. The coolant fluid is an inert gas such as nitrogen or helium with particles such as carbon suspended therein. The fuel assembly is contained within an elongated pressure vessel extending down into the reactor. The fuel portion is at the lower end of the vessel and is constructed of cylindrical segments through which the coolant passes. Turbulence promotors within the passageways maintain the particles in agitation to increase its ability to transfer heat away from the outer walls. Shielding sections and alternating passageways above the fueled portion limit the escape of radiation out of the top of the vessel. (AEC)

Summary of Blast Shield and Material Testing for Development of Solid Debris Collection at the National Ignition Facility (NIF)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shaughnessy, D A; Gostic, J M; Moody, K J

2011-11-21

The ability to collect solid debris from the target chamber following a NIF shot has application for both capsule diagnostics, particularly for fuel-ablator mix, and measuring cross sections relevant to the Stockpile Stewardship program and nuclear astrophysics. Simulations have shown that doping the capsule with up to 10{sup 15} atoms of an impurity not otherwise found in the capsule does not affect its performance. The dopant is an element that will undergo nuclear activations during the NIF implosion, forming radioactive species that can be collected and measured after extraction from the target chamber. For diagnostics, deuteron or alpha induced reactionsmore » can be used to probe the fuel-ablator mix. For measuring neutron cross sections, the dopant should be something that is sensitive to the 14 MeV neutrons produced through the fusion of deuterium and tritium. Developing the collector is a challenge due to the extreme environment of the NIF chamber. The collector surface is exposed to a large photon flux from x-rays and unconverted laser light before it is exposed to a debris wind that is formed from vaporized material from the target chamber center. The photons will ablate the collector surface to some extent, possibly impeding the debris from reaching the collector and sticking. In addition, the collector itself must be mechanically strong enough to withstand the large amount of energy it will be exposed to, and it should be something that will be easy to count and chemically process. In order to select the best material for the collector, a variety of different metals have been tested in the NIF chamber. They were exposed to high-energy laser shots in order to evaluate their postshot surface characterization, morphology, degree of melt, and their ability to retain debris from the chamber center. The first set of samples consisted of 1 mm thick pieces of aluminum that had been fielded in the chamber as blast shields protecting the neutron activation diagnostic. Ten of these pieces were fielded at the equator and one was fielded on the pole. The shields were analyzed using a combination of scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), x-ray fluorescence (XRF), neutron activation analysis (NAA) and chemical leaching followed by mass spectrometry. On each shield, gold debris originating from the gold hohlraum was observed, as well as large quantities of debris that were present in the center of the target chamber at the time of the shot (i.e., stainless steel, indium, copper, etc.) Debris was visible in the SEM as large blobs or splats of material that had encountered the surface of the aluminum and stuck. The aluminum itself had obviously melted and condensed, and some of the large debris splats arrived after the surface had already hardened. Melt depth was determined by cross sectioning the pieces and measuring the melted surface layers via SEM. After the SEM analysis was completed, the pieces were sent for NAA at the USGS reactor and were analyzed by U. Greife at the Colorado School of Mines. The NAA showed that the majority of gold mass present on the shields was not in the form of large blobs and splats, but was present as small particulates that had most likely formed as condensed vapor. Further analysis showed that the gold was entrained in the melted aluminum surface layers and did not extend down into the bulk of the aluminum. Once the gold mass was accounted for from the NAA, it was determined that the aluminum fielded at the equator was collecting a fraction of the total gold hohlraum mass equivalent to 120% {+-} 10% of the solid angle subtended by the shield. The attached presentation has more information on the results of the aluminum blast shield analysis. In addition to the information given in the presentation, the surfaces of the shields have been chemically leached and submitted for mass spectrometric analysis. The results from that analysis are expected to arrive after the due date of this report and will be written up at a later time. Based on the results of the aluminum blast shield analysis, it was determined that additional materials needed to be tested as potential collectors in the NIF chamber. 1-2 mm thick pieces of tantalum, niobium, vanadium, silver, titanium, molybdenum, and graphite foil were fielded in the Wedge Range Filter (WRF) mount at a distance of 50 cm from target chamber center during the shock timing campaign. The pieces were subsequently removed and analyzed in a similar fashion to the aluminum shields. As of this writing, the pieces are still under analysis, but initial results indicate that gold debris was collected on the various materials. Currently, the pieces are being cross-sectioned so that the melt depths of each material can be compared. In addition, NAA and/or mass spectrometry will be performed in order to determine the total gold mass that was collected on each surface.« less

Effects of materials and design on the criticality and shielding assessment of canister concepts for the disposal of spent nuclear fuel.

PubMed

Gutiérrez, Miguel Morales; Caruso, Stefano; Diomidis, Nikitas

2018-05-19

According to the Swiss disposal concept, the safety of a deep geological repository for spent nuclear fuel (SNF) is based on a multi-barrier system. The disposal canister is an important component of the engineered barrier system, aiming to provide containment of the SNF for thousands of years. This study evaluates the criticality safety and shielding of candidate disposal canister concepts, focusing on the fulfilment of the sub-criticality criterion and on limiting radiolysis processes at the outer surface of the canister which can enhance corrosion mechanisms. The effective neutron multiplication factor (k-eff) and the surface dose rates are calculated for three different canister designs and material combinations for boiling water reactor (BWR) canisters, containing 12 spent fuel assemblies (SFA), and pressurized water reactor (PWR) canisters, with 4 SFAs. For each configuration, individual criticality and shielding calculations were carried out. The results show that k-eff falls below the defined upper safety limit (USL) of 0.95 for all BWR configurations, while staying above USL for the PWR ones. Therefore, the application of a burnup credit methodology for the PWR case is required, being currently under development. Relevant is also the influence of canister material and internal geometry on criticality, enabling the identification of safer fuel arrangements. For a final burnup of 55MWd/kgHM and 30y cooling time, the combined photon-neutron surface dose rate is well below the threshold of 1 Gy/h defined to limit radiation-induced corrosion of the canister in all cases. Copyright © 2018 Elsevier Ltd. All rights reserved.

Apollo 15 mission main parachute failure

NASA Technical Reports Server (NTRS)

1971-01-01

The failure of one of the three main parachutes of the Apollo 15 spacecraft was investigated by studying malfunctions in the forward heat shield, broken riser, and firing the fuel expelled from the command module reaction control system. It is concluded that the most probable cause was the burning of raw fuel being expelled during the latter portion of depletion firing. Recommended corrective actions are included.

Shielding Design for the South Pole nToF Diagnostic at the NIF

NASA Astrophysics Data System (ADS)

Khater, Hesham; Sitaraman, Shiva; Hall, James; Hatarik, Robert; Caggiano, Joseph; Waltz, Cory

2017-09-01

Neutron time of flight (nToF) detectors are fielded at the National Ignition Facility (NIF) to measure neutron yield, ion temperature, and downscattering in the cold fuel for D-T implosions. Anisotropically assembled cold fuel may generate different nToF data when measured by detectors located at the Target Chamber equator and poles. A collimated nToF line of sight has been fielded near the Target Chamber South Pole (SP) to examine any possible anisotropy in the cold fuel. The SP nToF detector is located in the lowest floor level of the NIF's Target Bay and at a distance of 18 m from the Target Chamber Center. The detector utilizes a solid bibenzyl scintillator and four photomultiplier tubes. The line of sight includes a port collimator that is attached to the Target Chamber and a bore hole collimator in the concrete floor above the detector. In addition, a beam line get lost hole is constructed in the Target Bay floor to minimize the backscattered radiation at the detector location. Initial measurements indicated the need for installation of additional shielding to eliminate gamma background during the period before arrival of the 14.1 MeV neutrons to the detector. A set of MCNP Monte Carlo simulations with the full Target Bay model were conducted to provide an estimate of the expected neutron and gamma backgrounds during D-T shots. A new shielding scheme is designed to reduce the gamma background by an order of magnitude.

Depleted uranium hexafluoride: The source material for advanced shielding systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Quapp, W.J.; Lessing, P.A.; Cooley, C.R.

1997-02-01

The U.S. Department of Energy (DOE) has a management challenge and financial liability problem in the form of 50,000 cylinders containing 555,000 metric tons of depleted uranium hexafluoride (UF{sub 6}) that are stored at the gaseous diffusion plants. DOE is evaluating several options for the disposition of this UF{sub 6}, including continued storage, disposal, and recycle into a product. Based on studies conducted to date, the most feasible recycle option for the depleted uranium is shielding in low-level waste, spent nuclear fuel, or vitrified high-level waste containers. Estimates for the cost of disposal, using existing technologies, range between $3.8 andmore » $11.3 billion depending on factors such as the disposal site and the applicability of the Resource Conservation and Recovery Act (RCRA). Advanced technologies can reduce these costs, but UF{sub 6} disposal still represents large future costs. This paper describes an application for depleted uranium in which depleted uranium hexafluoride is converted into an oxide and then into a heavy aggregate. The heavy uranium aggregate is combined with conventional concrete materials to form an ultra high density concrete, DUCRETE, weighing more than 400 lb/ft{sup 3}. DUCRETE can be used as shielding in spent nuclear fuel/high-level waste casks at a cost comparable to the lower of the disposal cost estimates. Consequently, the case can be made that DUCRETE shielded casks are an alternative to disposal. In this case, a beneficial long term solution is attained for much less than the combined cost of independently providing shielded casks and disposing of the depleted uranium. Furthermore, if disposal is avoided, the political problems associated with selection of a disposal location are also avoided. Other studies have also shown cost benefits for low level waste shielded disposal containers.« less

Improved Hybrid Modeling of Spent Fuel Storage Facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bibber, Karl van

This work developed a new computational method for improving the ability to calculate the neutron flux in deep-penetration radiation shielding problems that contain areas with strong streaming. The “gold standard” method for radiation transport is Monte Carlo (MC) as it samples the physics exactly and requires few approximations. Historically, however, MC was not useful for shielding problems because of the computational challenge of following particles through dense shields. Instead, deterministic methods, which are superior in term of computational effort for these problems types but are not as accurate, were used. Hybrid methods, which use deterministic solutions to improve MC calculationsmore » through a process called variance reduction, can make it tractable from a computational time and resource use perspective to use MC for deep-penetration shielding. Perhaps the most widespread and accessible of these methods are the Consistent Adjoint Driven Importance Sampling (CADIS) and Forward-Weighted CADIS (FW-CADIS) methods. For problems containing strong anisotropies, such as power plants with pipes through walls, spent fuel cask arrays, active interrogation, and locations with small air gaps or plates embedded in water or concrete, hybrid methods are still insufficiently accurate. In this work, a new method for generating variance reduction parameters for strongly anisotropic, deep penetration radiation shielding studies was developed. This method generates an alternate form of the adjoint scalar flux quantity, Φ Ω, which is used by both CADIS and FW-CADIS to generate variance reduction parameters for local and global response functions, respectively. The new method, called CADIS-Ω, was implemented in the Denovo/ADVANTG software. Results indicate that the flux generated by CADIS-Ω incorporates localized angular anisotropies in the flux more effectively than standard methods. CADIS-Ω outperformed CADIS in several test problems. This initial work indicates that CADIS- may be highly useful for shielding problems with strong angular anisotropies. This is a benefit to the public by increasing accuracy for lower computational effort for many problems that have energy, security, and economic importance.« less

ADVANCEMENTS IN TIME-SPECTRA ANALYSIS METHODS FOR LEAD SLOWING-DOWN SPECTROSCOPY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Leon E.; Anderson, Kevin K.; Gesh, Christopher J.

2010-08-11

Direct measurement of Pu in spent nuclear fuel remains a key challenge for safeguarding nuclear fuel cycles of today and tomorrow. Lead slowing-down spectroscopy (LSDS) is an active nondestructive assay method that has the potential to provide independent, direct measurement of Pu and U isotopic mass with an uncertainty lower than the approximately 10 percent typical of today’s confirmatory assay methods. Pacific Northwest National Laboratory’s (PNNL) previous work to assess the viability of LSDS for the assay of pressurized water reactor (PWR) assemblies indicated that the method could provide direct assay of Pu-239 and U-235 (and possibly Pu-240 and Pu-241)more » with uncertainties less than a few percent, assuming suitably efficient instrumentation, an intense pulsed neutron source, and improvements in the time-spectra analysis methods used to extract isotopic information from a complex LSDS signal. This previous simulation-based evaluation used relatively simple PWR fuel assembly definitions (e.g. constant burnup across the assembly) and a constant initial enrichment and cooling time. The time-spectra analysis method was founded on a preliminary analytical model of self-shielding intended to correct for assay-signal nonlinearities introduced by attenuation of the interrogating neutron flux within the assembly.« less

Radiation Attenuation and Stability of ClearView Radiation Shielding TM-A Transparent Liquid High Radiation Shield.

PubMed

Bakshi, Jayeesh

2018-04-01

Radiation exposure is a limiting factor to work in sensitive environments seen in nuclear power and test reactors, medical isotope production facilities, spent fuel handling, etc. The established choice for high radiation shielding is lead (Pb), which is toxic, heavy, and abidance by RoHS. Concrete, leaded (Pb) bricks are used as construction materials in nuclear facilities, vaults, and hot cells for radioisotope production. Existing transparent shielding such as leaded glass provides minimal shielding attenuation in radiotherapy procedures, which in some cases is not sufficient. To make working in radioactive environments more practicable while resolving the lead (Pb) issue, a transparent, lightweight, liquid, and lead-free high radiation shield-ClearView Radiation Shielding-(Radium Incorporated, 463 Dinwiddie Ave, Waynesboro, VA). was developed. This paper presents the motivation for developing ClearView, characterization of certain aspects of its use and performance, and its specific attenuation testing. Gamma attenuation testing was done using a 1.11 × 10 Bq Co source and ANSI/HPS-N 13.11 standard. Transparency with increasing thickness, time stability of liquid state, measurements of physical properties, and performance in freezing temperatures are reported. This paper also presents a comparison of ClearView with existing radiation shields. Excerpts from LaSalle nuclear power plant are included, giving additional validation. Results demonstrated and strengthened the expected performance of ClearView as a radiation shield. Due to the proprietary nature of the work, some information is withheld.

A document review to characterize Atomic International SNAP fuels shipped to INEL 1966--1973

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wahnschaffe, S.D.; Lords, R.E.; Kneff, D.W.

1995-09-01

This report provides the results of a document search and review study to obtain information on the spent fuels for the following six Nuclear Auxiliary Power (SNAP) reactor cores now stored at the Idaho National Engineering Laboratory (INEL): SNAP-2 Experimental Reactor, SNAP-2 Development Reactor, SNAP-10A Ground Test Reactor, SNAP-8 Experimental Reactor, SNAP-8 Development Reactor, and Shield Test Reactor. The report also covers documentation on SNAP fuel materials from four in-pile materials tests: NAA-82-1, NAA-115-2, NAA-117-1, and NAA-121. Pieces of these fuel materials are also stored at INEL as part of the SNAP fuel shipments.

Pulverized coal burner

DOEpatents

Sivy, J.L.; Rodgers, L.W.; Koslosy, J.V.; LaRue, A.D.; Kaufman, K.C.; Sarv, H.

1998-11-03

A burner is described having lower emissions and lower unburned fuel losses by implementing a transition zone in a low NO{sub x} burner. The improved burner includes a pulverized fuel transport nozzle surrounded by the transition zone which shields the central oxygen-lean fuel devolatilization zone from the swirling secondary combustion air. The transition zone acts as a buffer between the primary and the secondary air streams to improve the control of near-burner mixing and flame stability by providing limited recirculation regions between primary and secondary air streams. These limited recirculation regions transport evolved NO{sub x} back towards the oxygen-lean fuel pyrolysis zone for reduction to molecular nitrogen. Alternate embodiments include natural gas and fuel oil firing. 8 figs.

Pulverized coal burner

DOEpatents

Sivy, Jennifer L.; Rodgers, Larry W.; Koslosy, John V.; LaRue, Albert D.; Kaufman, Keith C.; Sarv, Hamid

1998-01-01

A burner having lower emissions and lower unburned fuel losses by implementing a transition zone in a low NO.sub.x burner. The improved burner includes a pulverized fuel transport nozzle surrounded by the transition zone which shields the central oxygen-lean fuel devolatilization zone from the swirling secondary combustion air. The transition zone acts as a buffer between the primary and the secondary air streams to improve the control of near-burner mixing and flame stability by providing limited recirculation regions between primary and secondary air streams. These limited recirculation regions transport evolved NO.sub.x back towards the oxygen-lean fuel pyrolysis zone for reduction to molecular nitrogen. Alternate embodiments include natural gas and fuel oil firing.

Laminar flow control leading edge glove flight test article development

NASA Technical Reports Server (NTRS)

Pearce, W. E.; Mcnay, D. E.; Thelander, J. A.

1984-01-01

A laminar flow control (LFC) flight test article was designed and fabricated to fit into the right leading edge of a JetStar aircraft. The article was designed to attach to the front spar and fill in approx. 70 inches of the leading edge that are normally occupied by the large slipper fuel tank. The outer contour of the test article was constrained to align with an external fairing aft of the front spar which provided a surface pressure distribution over the test region representative of an LFC airfoil. LFC is achieved by applying suction through a finely perforated surface, which removes a small fraction of the boundary layer. The LFC test article has a retractable high lift shield to protect the laminar surface from contamination by airborne debris during takeoff and low altitude operation. The shield is designed to intercept insects and other particles that could otherwise impact the leading edge. Because the shield will intercept freezing rain and ice, a oozing glycol ice protection system is installed on the shield leading edge. In addition to the shield, a liquid freezing point depressant can be sprayed on the back of the shield.

PWR upper/lower internals shield

DOE Office of Scientific and Technical Information (OSTI.GOV)

Homyk, W.A.

1995-03-01

During refueling of a nuclear power plant, the reactor upper internals must be removed from the reactor vessel to permit transfer of the fuel. The upper internals are stored in the flooded reactor cavity. Refueling personnel working in containment at a number of nuclear stations typically receive radiation exposure from a portion of the highly contaminated upper intervals package which extends above the normal water level of the refueling pool. This same issue exists with reactor lower internals withdrawn for inservice inspection activities. One solution to this problem is to provide adequate shielding of the unimmersed portion. The use ofmore » lead sheets or blankets for shielding of the protruding components would be time consuming and require more effort for installation since the shielding mass would need to be transported to a support structure over the refueling pool. A preferable approach is to use the existing shielding mass of the refueling pool water. A method of shielding was devised which would use a vacuum pump to draw refueling pool water into an inverted canister suspended over the upper internals to provide shielding from the normally exposed components. During the Spring 1993 refueling of Indian Point 2 (IP2), a prototype shield device was demonstrated. This shield consists of a cylindrical tank open at the bottom that is suspended over the refueling pool with I-beams. The lower lip of the tank is two feet below normal pool level. After installation, the air width of the natural shielding provided by the existing pool water. This paper describes the design, development, testing and demonstration of the prototype device.« less

10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

Code of Federal Regulations, 2013 CFR

2013-01-01

... maintained in a subcritical condition under credible conditions. (d) Radiation shielding and confinement... confinement of radioactive material under normal, off-normal, and credible accident conditions. (m) To the...

10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

Code of Federal Regulations, 2012 CFR

2012-01-01

... maintained in a subcritical condition under credible conditions. (d) Radiation shielding and confinement... confinement of radioactive material under normal, off-normal, and credible accident conditions. (m) To the...

10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

Code of Federal Regulations, 2010 CFR

2010-01-01

... maintained in a subcritical condition under credible conditions. (d) Radiation shielding and confinement... of radioactive material under normal, off-normal, and credible accident conditions. (m) To the extent...

10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

Code of Federal Regulations, 2011 CFR

2011-01-01

... maintained in a subcritical condition under credible conditions. (d) Radiation shielding and confinement... of radioactive material under normal, off-normal, and credible accident conditions. (m) To the extent...

10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

Code of Federal Regulations, 2014 CFR

2014-01-01

... maintained in a subcritical condition under credible conditions. (d) Radiation shielding and confinement... confinement of radioactive material under normal, off-normal, and credible accident conditions. (m) To the...

Evaluation of the concrete shield compositions from the 2010 criticality accident alarm system benchmark experiments at the CEA Valduc SILENE facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miller, Thomas Martin; Celik, Cihangir; Dunn, Michael E

In October 2010, a series of benchmark experiments were conducted at the French Commissariat a l'Energie Atomique et aux Energies Alternatives (CEA) Valduc SILENE facility. These experiments were a joint effort between the United States Department of Energy Nuclear Criticality Safety Program and the CEA. The purpose of these experiments was to create three benchmarks for the verification and validation of radiation transport codes and evaluated nuclear data used in the analysis of criticality accident alarm systems. This series of experiments consisted of three single-pulsed experiments with the SILENE reactor. For the first experiment, the reactor was bare (unshielded), whereasmore » in the second and third experiments, it was shielded by lead and polyethylene, respectively. The polyethylene shield of the third experiment had a cadmium liner on its internal and external surfaces, which vertically was located near the fuel region of SILENE. During each experiment, several neutron activation foils and thermoluminescent dosimeters (TLDs) were placed around the reactor. Nearly half of the foils and TLDs had additional high-density magnetite concrete, high-density barite concrete, standard concrete, and/or BoroBond shields. CEA Saclay provided all the concrete, and the US Y-12 National Security Complex provided the BoroBond. Measurement data from the experiments were published at the 2011 International Conference on Nuclear Criticality (ICNC 2011) and the 2013 Nuclear Criticality Safety Division (NCSD 2013) topical meeting. Preliminary computational results for the first experiment were presented in the ICNC 2011 paper, which showed poor agreement between the computational results and the measured values of the foils shielded by concrete. Recently the hydrogen content, boron content, and density of these concrete shields were further investigated within the constraints of the previously available data. New computational results for the first experiment are now available that show much better agreement with the measured values.« less

Methods and systems to thermally protect fuel nozzles in combustion systems

DOEpatents

Helmick, David Andrew; Johnson, Thomas Edward; York, William David; Lacy, Benjamin Paul

2013-12-17

A method of assembling a gas turbine engine is provided. The method includes coupling a combustor in flow communication with a compressor such that the combustor receives at least some of the air discharged by the compressor. A fuel nozzle assembly is coupled to the combustor and includes at least one fuel nozzle that includes a plurality of interior surfaces, wherein a thermal barrier coating is applied across at least one of the plurality of interior surfaces to facilitate shielding the interior surfaces from combustion gases.

Development, Testing and Validation of a Waste Assay System for the Measurement and Characterisation of Active Spent Fuel Element Debris From UK Magnox Reactors - 12533

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mason, John A.; Burke, Kevin J.; Looman, Marc R.

2012-07-01

This paper describes the development, testing and validation of a waste measurement instrument for characterising active remote handled radioactive waste arising from the operation of Magnox reactors in the United Kingdom. Following operation in UK Magnox gas cooled reactors and a subsequent period of cooling, parts of the magnesium-aluminium alloy cladding were removed from spent fuel and the uranium fuel rods with the remaining cladding were removed to Sellafield for treatment. The resultant Magnox based spent fuel element debris (FED), which constitutes active intermediate level waste (ILW) has been stored in concrete vaults at the reactor sites. As part ofmore » the decommissioning of the FED vaults the FED must be removed, measured and characterised and placed in intermediate storage containers. The present system was developed for use at the Trawsfynydd nuclear power station (NPS), which is in the decommissioning phase, but the approach is potentially applicable to FED characterisation at all of the Magnox reactors. The measurement system consists of a heavily shielded and collimated high purity Germanium (HPGe) detector with electromechanical cooling and a high count-rate preamplifier and digital multichannel pulse height analyser. The HPGe based detector system is controlled by a software code, which stores the measurement result and allows a comprehensive analysis of the measured FED data. Fuel element debris is removed from the vault and placed on a tray to a uniform depth of typically 10 cm for measurement. The tray is positioned approximately 1.2 meters above the detector which views the FED through a tungsten collimator with an inverted pyramid shape. At other Magnox sites the positions may be reversed with the shielded and collimated HPGe detector located above the tray on which the FED is measured. A comprehensive Monte Carlo modelling and analysis of the measurement process has been performed in order to optimise the measurement geometry and eliminate interferences from radioactive sources and FED in the immediate vicinity of the measurement position. The detector system has been calibrated and high activity radioactive sources of Cs-137, Co-60 and Na-22 have been used to validate the measurement process. The data acquisition and analysis software code has been tested and validated in keeping with the software quality assurance requirements of both ISO:9001-2008 - TICK-IT in the UK and NQA-1. The measurement and analysis system has been comprehensively tested with high activity sources, is flexible and may be applicable to a wide range of remote handled radioactive waste measurement applications. It is due to be installed at Trawsfynydd NPS later this year. This paper describes the Waste Tray Assay System (WTAS) that has been developed for the measurement of Magnox FED waste. The WTAS has been tested with a range of radioactive sources and its operation has been simulated with benchmarked MCNP Monte Carlo calculations. The measurement software has been validated as has the operation of the system for a range of strong radioactive sources. A system based on the design is due for installation and operation in 2012. The system has application to the measurement of Magnox Fuel Element Debris (FED) waste at other Magnox reactor sites. The major design objective of the WTAS that has been achieved is the ability of the assay system to determine the content of Cs-137, and in turn to enable the fissile burden to be assessed using a radionuclide fingerprint, in the presence of higher and highly variable quantities of Co-60, typically from nimonic springs. The approach can be used in other Magnox FED waste configurations where the detector is located above the FED waste sorting tray and where the collimation is fixed below the detector and at a distance above the tray. In this case, which has also been investigated, there are different shielding problems and mechanical support issues. The extensive use of MCNP Monte Carlo modelling to simulate the geometry of the sorting cell and the distribution of radioactive sources has helped to ensure that all of the detector shielding requirements are addressed and suitable Cs-137 and Co-60 discrimination can be achieved. The WTAS in its present form or in other configurations has relevance to the measurement of other active ILW and highly active RH waste. Examples include high activity RH LLW and RH TRU (Transuranic) waste as defined in the United States arising from both commercial nuclear and Department of Energy (DOE) operations. The analysis is able to analyse a range of radionuclides beyong those expected in the Magnox FED cases. (authors)« less

APPARATUS FOR LOADING AND UNLOADING A MACHINE

DOEpatents

Payne, J.H. Jr.

1962-07-17

An arrangement for loading and unloading a nuclear reactor is described. Depleted fuel elements are removed from the reactor through one of a small number of holes in a shielding plug that is rotatably mounted in an eccentric annular plug rotatably mounted in the top of the reactor. The fuel elements removed are stored in a plurality of openings in a rotatable magazine or storage means rotatably mounted over the plugs. (AEC)

Firing of pulverized solvent refined coal

DOEpatents

Lennon, Dennis R.; Snedden, Richard B.; Foster, Edward P.; Bellas, George T.

1990-05-15

A burner for the firing of pulverized solvent refined coal is constructed and operated such that the solvent refined coal can be fired successfully without any performance limitations and without the coking of the solvent refined coal on the burner components. The burner is provided with a tangential inlet of primary air and pulverized fuel, a vaned diffusion swirler for the mixture of primary air and fuel, a center water-cooled conical diffuser shielding the incoming fuel from the heat radiation from the flame and deflecting the primary air and fuel steam into the secondary air, and a watercooled annulus located between the primary air and secondary air flows.

Measurement of the 235U Induced Fission Gamma-ray Spectrum as an Active Non-destructive Assay of Fresh Nucleear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sarnoski, Sarah E.; Fast, James E.; Fulsom, Bryan G.

2017-07-17

Non-destructive assay is a powerful tool the International Atomic Energy Agency (IAEA) employs to verify adherence to safeguards agreements. Current IAEA veri- cation techniques for fresh nuclear fuel include passive gamma-ray spectroscopy to determine fuel enrichment. This technique suers from self-shielding and lakes the percision to detect diversion of central fuel rods. The aim of this research is to develop a new, more capable non-destructive analysis technique using active neutron interroga- tion of fuel assemblies and determining the yields of short-lived ssion products from high-resolution gamma-ray spectroscopy using high-purity germanium (HPGe). This paper reports results from irradiation of a onemore » meter tall mock fresh fuel assembly with low enriched uranium (LEU) or depleted uranium (DU) rods using a down-scattered deuterium-tritium (D-T) neutron source. Both prompt and delayed gamma-ray spec- tra were collected as time-stamped list-mode data in a coax detector and without list mode data in a planar strip detector. No dierentiating signatures were observed in the prompt spectra in either detector; however, both detectors observed several short-lived ssion product signatures in LEU and not DU fuel, indicating that this technique has potential for determination of enrichment of fresh fuel assemblies. There were eight unique ssion products observed in the LEU spectra with the coax detector spectra, and three ssion products were observed in the LEU spectra with the strip detector.« less

Radioactive waste disposal via electric propulsion

NASA Technical Reports Server (NTRS)

Burns, R. E.

1975-01-01

It is shown that space transportation is a feasible method of removal of radioactive wastes from the biosphere. The high decay heat of the isotopes powers a thermionic generator which provides electrical power for ion thrust engines. The massive shields (used to protect ground and flight personnel) are removed in orbit for subsequent reuse; the metallic fuel provides a shield for the avionics that guides the orbital stage to solar system escape. Performance calculations indicate that 4000 kg. of actinides may be removed per Shuttle flight. Subsidiary problems - such as cooling during ascent - are discussed.

A Glucose Fuel Cell for Implantable Brain–Machine Interfaces

PubMed Central

Rapoport, Benjamin I.; Kedzierski, Jakub T.; Sarpeshkar, Rahul

2012-01-01

We have developed an implantable fuel cell that generates power through glucose oxidation, producing steady-state power and up to peak power. The fuel cell is manufactured using a novel approach, employing semiconductor fabrication techniques, and is therefore well suited for manufacture together with integrated circuits on a single silicon wafer. Thus, it can help enable implantable microelectronic systems with long-lifetime power sources that harvest energy from their surrounds. The fuel reactions are mediated by robust, solid state catalysts. Glucose is oxidized at the nanostructured surface of an activated platinum anode. Oxygen is reduced to water at the surface of a self-assembled network of single-walled carbon nanotubes, embedded in a Nafion film that forms the cathode and is exposed to the biological environment. The catalytic electrodes are separated by a Nafion membrane. The availability of fuel cell reactants, oxygen and glucose, only as a mixture in the physiologic environment, has traditionally posed a design challenge: Net current production requires oxidation and reduction to occur separately and selectively at the anode and cathode, respectively, to prevent electrochemical short circuits. Our fuel cell is configured in a half-open geometry that shields the anode while exposing the cathode, resulting in an oxygen gradient that strongly favors oxygen reduction at the cathode. Glucose reaches the shielded anode by diffusing through the nanotube mesh, which does not catalyze glucose oxidation, and the Nafion layers, which are permeable to small neutral and cationic species. We demonstrate computationally that the natural recirculation of cerebrospinal fluid around the human brain theoretically permits glucose energy harvesting at a rate on the order of at least 1 mW with no adverse physiologic effects. Low-power brain–machine interfaces can thus potentially benefit from having their implanted units powered or recharged by glucose fuel cells. PMID:22719888

DOE Office of Scientific and Technical Information (OSTI.GOV)

Venkataraman, M.; Natarajan, R.; Raj, Baldev

The reprocessing of spent fuel from Fast Breeder Test Reactor (FBTR) has been successfully demonstrated in the pilot plant, CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell). Since commissioning in 2003, spent mixed carbide fuel from FBTR of different burnups and varying cooling period, have been reprocessed in this facility. Reprocessing of the spent fuel with a maximum burnup of 100 GWd/t has been successfully carried out so far. The feed backs from these campaigns with progressively increasing specific activities, have been useful in establishing a viable process flowsheet for reprocessing the Prototype Fast Breeder Reactor (PFBR)more » spent fuel. Also, the design of various equipments and processes for the future plants, which are either under design for construction, namely, the Demonstration Fast Reactor Fuel Reprocessing Plant (DFRP) and the Fast reactor fuel Reprocessing Plant (FRP) could be finalized. (authors)« less

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

NASA Astrophysics Data System (ADS)

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.

Tower Shielding Reactor II design and operation report: Vol. 2. Safety Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holland, L. B.; Kolb, J. O.

1970-01-01

Information on the Tower Shielding Reactor II is contained in the TSR-II Design and Operation Report and in the Tower Shielding Facility Manual. The TSR-II Design and Operating Report consists of three volumes. Volume 1 is Descriptions of the Tower Shielding Reactor II and Facility; Volume 2 is Safety analysis of the Tower Shielding Reactor II; and Volume 3 is the Assembly and Testing of the Tower Shielding Reactor II Control Mechanism Housing.

Evaluation of a Method for Remote Detection of Fuel Relocation Outside the Original Core Volumes of Fukushima Reactor Units 1-3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Douglas W. Akers; Edwin A. Harvego

2012-08-01

This paper presents the results of a study to evaluate the feasibility of remotely detecting and quantifying fuel relocation from the core to the lower head, and to regions outside the reactor vessel primary containment of the Fukushima 1-3 reactors. The goals of this study were to determine measurement conditions and requirements, and to perform initial radiation transport sensitivity analyses for several potential measurement locations inside the reactor building. The radiation transport sensitivity analyses were performed based on reactor design information for boiling water reactors (BWRs) similar to the Fukushima reactors, ORIGEN2 analyses of 3-cycle BWR fuel inventories, and datamore » on previously molten fuel characteristics from TMI- 2. A 100 kg mass of previously molten fuel material located on the lower head of the reactor vessel was chosen as a fuel interrogation sensitivity target. Two measurement locations were chosen for the transport analyses, one inside the drywell and one outside the concrete biological shield surrounding the drywell. Results of these initial radiation transport analyses indicate that the 100 kg of previously molten fuel material may be detectable at the measurement location inside the drywell, but that it is highly unlikely that any amount of fuel material inside the RPV will be detectable from a location outside the concrete biological shield surrounding the drywell. Three additional fuel relocation scenarios were also analyzed to assess detection sensitivity for varying amount of relocated material in the lower head of the reactor vessel, in the control rods perpendicular to the detector system, and on the lower head of the drywell. Results of these analyses along with an assessment of background radiation effects and a discussion of measurement issues, such as the detector/collimator design, are included in the paper.« less

Experiments and Monte Carlo modeling of a higher resolution Cadmium Zinc Telluride detector for safeguards applications

NASA Astrophysics Data System (ADS)

Borella, Alessandro

2016-09-01

The Belgian Nuclear Research Centre is engaged in R&D activity in the field of Non Destructive Analysis on nuclear materials, with focus on spent fuel characterization. A 500 mm3 Cadmium Zinc Telluride (CZT) with enhanced resolution was recently purchased. With a full width at half maximum of 1.3% at 662 keV, the detector is very promising in view of its use for applications such as determination of uranium enrichment and plutonium isotopic composition, as well as measurement on spent fuel. In this paper, I report about the work done with such a detector in terms of its characterization. The detector energy calibration, peak shape and efficiency were determined from experimental data. The data included measurements with calibrated sources, both in a bare and in a shielded environment. In addition, Monte Carlo calculations with the MCNPX code were carried out and benchmarked with experiments.

SIMPLIFIED SODIUM GRAPHITE REACTOR SYSTEM

DOEpatents

Dickinson, R.W.

1963-03-01

This patent relates to a nuclear power reactor comprising a reactor vessel, shielding means positioned at the top of said vessel, means sealing said reactor vessel to said shielding means, said vessel containing a quantity of sodium, a core tank, unclad graphite moderator disposed in said tank, means including a plurality of process tubes traversing said tank for isolating said graphite from said sodium, fuel elements positioned in said process tubes, said core tank being supported in spaced relation to the walls and bottom of said reactor vessel and below the level of said sodium, neutron shielding means positioned adjacent said core tank between said core tank and the walls of said vessel, said neutron shielding means defining an annuiar volume adjacent the inside wall of said reactor vessel, inlet plenum means below said core tank for providing a passage between said annular volume and said process tubes, heat exchanger means removably supported from the first-named shielding means and positioned in said annular volume, and means for circulating said sodium over said neutron shielding means down through said heat exchanger, across said inlet plenum and upward through said process tubes, said last-named means including electromagnetic pumps located outside said vessel and supported on said vessel wall between said heat exchanger means and said inlet plenum means. (AEC)

Natural chemical shielding analysis of nuclear magnetic resonance shielding tensors from gauge-including atomic orbital calculations

NASA Astrophysics Data System (ADS)

Bohmann, Jonathan A.; Weinhold, Frank; Farrar, Thomas C.

1997-07-01

Nuclear magnetic shielding tensors computed by the gauge including atomic orbital (GIAO) method in the Hartree-Fock self-consistent-field (HF-SCF) framework are partitioned into magnetic contributions from chemical bonds and lone pairs by means of natural chemical shielding (NCS) analysis, an extension of natural bond orbital (NBO) analysis. NCS analysis complements the description provided by alternative localized orbital methods by directly calculating chemical shieldings due to delocalized features in the electronic structure, such as bond conjugation and hyperconjugation. Examples of NCS tensor decomposition are reported for CH4, CO, and H2CO, for which a graphical mnemonic due to Cornwell is used to illustrate the effect of hyperconjugative delocalization on the carbon shielding.

Fission Product Appearance Rate Coefficients in Design Basis Source Term Determinations - Past and Present

NASA Astrophysics Data System (ADS)

Perez, Pedro B.; Hamawi, John N.

2017-09-01

Nuclear power plant radiation protection design features are based on radionuclide source terms derived from conservative assumptions that envelope expected operating experience. Two parameters that significantly affect the radionuclide concentrations in the source term are failed fuel fraction and effective fission product appearance rate coefficients. Failed fuel fraction may be a regulatory based assumption such as in the U.S. Appearance rate coefficients are not specified in regulatory requirements, but have been referenced to experimental data that is over 50 years old. No doubt the source terms are conservative as demonstrated by operating experience that has included failed fuel, but it may be too conservative leading to over-designed shielding for normal operations as an example. Design basis source term methodologies for normal operations had not advanced until EPRI published in 2015 an updated ANSI/ANS 18.1 source term basis document. Our paper revisits the fission product appearance rate coefficients as applied in the derivation source terms following the original U.S. NRC NUREG-0017 methodology. New coefficients have been calculated based on recent EPRI results which demonstrate the conservatism in nuclear power plant shielding design.

Development of Impregnated Agglomerate Pelletization (IAP) process for fabrication of (Th,U)O 2 mixed oxide pellets

NASA Astrophysics Data System (ADS)

Khot, P. M.; Nehete, Y. G.; Fulzele, A. K.; Baghra, Chetan; Mishra, A. K.; Afzal, Mohd.; Panakkal, J. P.; Kamath, H. S.

2012-01-01

Impregnated Agglomerate Pelletization (IAP) technique has been developed at Advanced Fuel Fabrication Facility (AFFF), BARC, Tarapur, for manufacturing (Th, 233U)O 2 mixed oxide fuel pellets, which are remotely fabricated in hot cell or shielded glove box facilities to reduce man-rem problem associated with 232U daughter radionuclides. This technique is being investigated to fabricate the fuel for Indian Advanced Heavy Water Reactor (AHWR). In the IAP process, ThO 2 is converted to free flowing spheroids by powder extrusion route in an unshielded facility which are then coated with uranyl nitrate solution in a shielded facility. The dried coated agglomerate is finally compacted and then sintered in oxidizing/reducing atmosphere to obtain high density (Th,U)O 2 pellets. In this study, fabrication of (Th,U)O 2 mixed oxide pellets containing 3-5 wt.% UO 2 was carried out by IAP process. The pellets obtained were characterized using optical microscopy, XRD and alpha autoradiography. The results obtained were compared with the results for the pellets fabricated by other routes such as Coated Agglomerate Pelletization (CAP) and Powder Oxide Pelletization (POP) route.

Neutron radiography of irradiated nuclear fuel at Idaho National Laboratory

DOE PAGES

Craft, Aaron E.; Wachs, Daniel M.; Okuniewski, Maria A.; ...

2015-09-10

Neutron radiography of irradiated nuclear fuel provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Idaho National Laboratory (INL) has multiple nuclear fuels research and development programs that routinely evaluate irradiated fuels using neutron radiography. The Neutron Radiography reactor (NRAD) sits beneath a shielded hot cell facility where neutron radiography and other evaluation techniques are performed on these highly radioactive objects. The NRAD currently uses the foil-film transfer technique for imaging fuel that is time consuming but provides high spatial resolution. This study describes the NRAD and hot cell facilities,more » the current neutron radiography capabilities available at INL, planned upgrades to the neutron imaging systems, and new facilities being brought online at INL related to neutron imaging.« less

Isomer Energy Source for Space Propulsion Systems

DTIC Science & Technology

2004-03-01

1,590 Engine F/W (no shield) 3.4 5.0 20.0 A similar core design replacing the fission fuel with the isomer 178Hfm2 is the starting point for this...particles interact and collide with other atoms in the fuel material, reactor core , or coolant, their energy can be transferred to thermal energy...thrust (44). The program produced several reactors that made it all the way through the testing stages of development . The reactors used uranium-235

LOADING MACHINE FOR REACTORS

DOEpatents

Simon, S.L.

1959-07-01

An apparatus is described for loading or charging slugs of fissionable material into a nuclear reactor. The apparatus of the invention is a "muzzle loading" type comprising a delivery tube or muzzle designed to be brought into alignment with any one of a plurality of fuel channels. The delivery tube is located within the pressure shell and it is also disposed within shielding barriers while the fuel cantridges or slugs are forced through the delivery tube by an externally driven flexible ram.

Concrete Shield Performance of the VSC-17 Spent Nuclear Fuel Cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sheryl L. Morton; Philip L. Winston; Toshiari Saegusa

2006-04-01

In 2003, representatives from the Central Research Institute of Electric Power Industry (CRIEPI) requested development of a project with the objective of determining the performance of a concrete spent nuclear fuel storage cask. Radiation and environmental effects may cause chemical alteration of the concrete that could result in excessive cracking, spalling, and loss of compressive strength. The Idaho National Laboratory (INL) project team and CRIEPI representatives identified the Ventilated Storage Cask (VSC-17) spent nuclear fuel storage cask as a candidate to study cask performance, because it had been used to store fuel as part of a dry cask storage demonstrationmore » project for more than 15 years. The project involved investigating the properties of the concrete shield. INL performed a survey of the cask in the summers of 2003 and 2004. Preliminary cask evaluations performed in 2003 indicated that the cask has no visual degradation. However, a 4-5 mrem/hr step-change in the radiation levels about halfway up the cask and a localized hot spot beneath an upper air vent indicate that there may be variability in the density of the concrete or localized cracking. In 2005, INL and CRIEPI scientists performed additional surveys on the VSC-17 cask. This document summarizes the methods used on the VSC-17 to evaluate the cask for compressive strength, concrete cracking, concrete thickness, and temperature distribution.« less

High pressure, energy, and impulse loading of the wall in a 1-GJ Laboratory Microfusion Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harrach, R.J.

1989-07-24

A proposed Laboratory Microfusion Facility (LMF) must be able to withstand repeated, low-repetition-rate fusion explosions at the 1-GJ (one-quarter ton) yield level. The energy release will occur at the center of a chamber only a few meters in radius, subjecting the interior or first wall to severe levels of temperature, pressure, and impulse. We show by theory and computation that the wall loading can be ameliorated by interposing a spherical shell of low-Z material between the fuel and the wall. This sacrificial shield converts the source energy components that are most damaging to the wall (soft x-rays and fast ions)more » to more benign plasma kinetic energy from the vaporized shield, and stretches the time duration over which this energy is delivered to the wall from nanoseconds to microseconds. Numerical calculations emphasize thin, volleyball-sized plastic shields, and much thicker ones of frozen nitrogen. Wall shielding criteria of small (or no) amount of surface ablation, low impulse and pressure loading, minimal shrapnel danger, small expense, and convenience in handling all favor the thin plastic shields. 7 refs., 4 figs.« less

Mars Exploration Rover Heat Shield Recontact Analysis

NASA Technical Reports Server (NTRS)

Raiszadeh, Behzad; Desai, Prasun N.; Michelltree, Robert

2011-01-01

The twin Mars Exploration Rover missions landed successfully on Mars surface in January of 2004. Both missions used a parachute system to slow the rover s descent rate from supersonic to subsonic speeds. Shortly after parachute deployment, the heat shield, which protected the rover during the hypersonic entry phase of the mission, was jettisoned using push-off springs. Mission designers were concerned about the heat shield recontacting the lander after separation, so a separation analysis was conducted to quantify risks. This analysis was used to choose a proper heat shield ballast mass to ensure successful separation with low probability of recontact. This paper presents the details of such an analysis, its assumptions, and the results. During both landings, the radar was able to lock on to the heat shield, measuring its distance, as it descended away from the lander. This data is presented and is used to validate the heat shield separation/recontact analysis.

Optimization of armored spherical tanks for storage on the lunar surface

NASA Technical Reports Server (NTRS)

Bents, D. J.; Knight, D. A.

1992-01-01

A redundancy strategy for reducing micrometeroid armoring mass is investigated, with application to cryogenic reactant storage for a regenerative fuel cell (RFC) on the lunar surface. In that micrometeoroid environment, the cryogenic fuel must be protected from loss due to tank puncture. The tankage must have a sufficiently high probability of survival over the length of the mission so that the probability of system failure due to tank puncture is low compared to the other mission risk factors. Assuming that a single meteoroid penetration can cause a storage tank to lose its contents, two means are available to raise the probability of surviving micrometeoroid attack to the desired level. One can armor the tanks to a thickness sufficient to reduce probability of penetration of any tank to the desired level or add extra capacity in the form of spare tanks that results in survival of a given number out of the ensemble at the desired level. A combination of these strategies (armoring and redundancy) is investigated. The objective is to find the optimum combination which yields the lowest shielding mass per cubic meter of surviving fuel out of the original ensemble. The investigation found that, for the volumes of fuel associated with multikilowatt class cryo storage RFC's, and the armoring methodology and meteoroid models used, storage should be fragmented into small individual tanks. Larger installations (more fuel) pay less of a shielding penalty than small installations. For the same survival probability over the same time period, larger volumes will require less armoring mass per unit volume protected.

Evaluation of Rutter Sigma S6 Ice Navigation Radar on USCGC Healy during Arctic Shield 2014

DTIC Science & Technology

2015-03-01

useful in making decisions about the pressure ridges ahead of time instead of making an immediate decision. Figure 33. CG radar display of... use the radar to help chart an efficient path through an ice field to reduce transit time and fuel expenses. This includes a clear picture of the ice...a ship would be able to use the radar to help chart an efficient path through an ice field to reduce transit time and fuel expenses. This includes

Decommissioning of the Northrop TRIGA reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cozens, George B.; Woo, Harry; Benveniste, Jack

1986-07-01

An overview of the administrative and operational aspects of decommissioning and dismantling the Northrop Mark F TRIGA Reactor, including: planning and preparation, personnel requirements, government interfacing, costs, contractor negotiations, fuel shipments, demolition, disposal of low level waste, final survey and disposition of the concrete biological shielding. (author)

Goals of thermionic program for space power

NASA Technical Reports Server (NTRS)

English, R. E.

1981-01-01

The thermionic and Brayton reactor concepts were compared for application to space power. For a turbine inlet temperature of 15000 K the Brayton powerplant weighted 5 to 40% less than the thermionic concept. The out of core concept separates the thermionic converters from their reactor. Technical risks are diminished by: (1) moving the insolator out of the reactor; (2) allowing a higher thermal flux for the thermionic converters than is required of the reactor fuel; and (3) eliminating fuel swelling's threat against lifetime of the thermionic converters. Overall performance can be improved by including power processing in system optimization for design and technology on more efficient, higher temperature power processors. The thermionic reactors will be larger than those for competitive systems with higher conversion efficiency and lower reactor operating temperatures. It is concluded that although the effect of reactor size on shield weight will be modest for unmanned spacecraft, the penalty in shield weight will be large for manned or man-tended spacecraft.

Cosmic Ray Muon Imaging of Spent Nuclear Fuel in Dry Storage Casks

DOE PAGES

Durham, J. Matthew; Guardincerri, Elena; Morris, Christopher L.; ...

2016-04-29

In this paper, cosmic ray muon radiography has been used to identify the absence of spent nuclear fuel bundles inside a sealed dry storage cask. The large amounts of shielding that dry storage casks use to contain radiation from the highly radioactive contents impedes typical imaging methods, but the penetrating nature of cosmic ray muons allows them to be used as an effective radiographic probe. This technique was able to successfully identify missing fuel bundles inside a sealed Westinghouse MC-10 cask. This method of fuel cask verification may prove useful for international nuclear safeguards inspectors. Finally, muon radiography may findmore » other safety and security or safeguards applications, such as arms control verification.« less

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

NASA Astrophysics Data System (ADS)

Durham, J. M.; Poulson, D.; Bacon, J.; Chichester, D. L.; Guardincerri, E.; Morris, C. L.; Plaud-Ramos, K.; Schwendiman, W.; Tolman, J. D.; Winston, P.

2018-04-01

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. Here we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. This application of technology and methods commonly used in high-energy particle physics provides a potential solution to this long-standing problem in international nuclear safeguards.

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

DOE PAGES

Sharp, Nicholas; Ticknor, Brian W.; Bronikowski, Michael; ...

2016-11-17

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

NEUTRONIC REACTOR

DOEpatents

Wigner, E.P.

1957-09-17

A reactor of the type having coolant liquid circulated through clad fuel elements geometrically arranged in a solid moderator, such as graphite, is described. The core is enclosed in a pressure vessel and suitable shielding, wherein means is provided for circulating vapor through the core to superheat the same. This is accomplished by drawing off the liquid which has been heated in the core due to the fission of the fuel, passing it to a nozzle within a chamber where it flashes into a vapor, and then passing the vapor through separate tubes extending through the moderator to pick up more heat developed in the core due to the fission of the fuel, thereby producing superheated vapor.

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sharp, Nicholas; Ticknor, Brian W.; Bronikowski, Michael

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

What can nuclear energy do for society.

NASA Technical Reports Server (NTRS)

Rom, F. E.

1971-01-01

Nuclear fuel is a compact and abundant source of energy. Its cost per unit of energy is less than that of fossil fuel. Disadvantages of nuclear fuel are connected with the high cost of capital equipment required for releasing nuclear energy and the heavy weight of the necessary shielding. In the case of commercial electric power production and marine propulsion the advantages have outweighed the disadvantages. It is pointed out that nuclear commercial submarines have certain advantages compared to surface ships. Nuclear powerplants might make air-cushion vehicles for transoceanic ranges feasible. The problems and advantages of a nuclear aircraft are discussed together with nuclear propulsion for interplanetary space voyages.

GARLIC, A SHIELDING PROGRAM FOR GAMMA RADIATION FROM LINE- AND CYLINDER- SOURCES

DOE Office of Scientific and Technical Information (OSTI.GOV)

Roos, M.

1959-06-01

GARLlC is a program for computing the gamma ray flux or dose rate at a shielded isotropic point detector, due to a line source or the line equivalent of a cylindrical source. The source strength distribution along the line must be either uniform or an arbitrary part of the positive half-cycle of a cosine function The line source can be orierted arbitrarily with respect to the main shield and the detector, except that the detector must not be located on the line source or on its extensionThe main source is a homogeneous plane slab in which scattered radiation is accountedmore » for by multiplying each point element of the line source by a point source buildup factor inside the integral over the point elements. Between the main shield and the line source additional shields can be introduced, which are either plane slabs, parallel to the main shield, or cylindrical rings, coaxial with the line source. Scattered radiation in the additional shields can only be accounted for by constant build-up factors outside the integral. GARLlC-xyz is an extended version particularly suited for the frequently met problem of shielding a room containing a large number of line sources in diHerent positions. The program computes the angles and linear dimensions of a problem for GARLIC when the positions of the detector point and the end points of the line source are given as points in an arbitrary rectangular coordinate system. As an example the isodose curves in water are presented for a monoenergetic cosine-distributed line source at several source energies and for an operating fuel element of the Swedish reactor R3, (auth)« less

Preliminary Analysis of a Water Shield for a Surface Power Reactor

NASA Technical Reports Server (NTRS)

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.

Depleted uranium dioxide melting in cold crucible melter and production of granules from the melt for use in casks for spent nuclear fuel and radioactive wastes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gotovchikov, V.T.; Seredenko, V.A.; Shatalov, V.V.

2007-07-01

This paper describes the results of a joint research program between the Russian Research Institute of Chemical Technology and Oak Ridge National Laboratory in the United States to develop new radiation shielding materials for use in the construction of casks for spent nuclear fuel (SNF) and radioactive wastes. Research and development is underway to develop SNF storage, transport, and disposal casks using shielding made with two new depleted uranium dioxide (DUO{sub 2}) materials: a DUO{sub 2}-steel cermet, and, DUCRETE with DUAGG (DUO{sub 2} aggregate). Melting the DUO{sub 2} and allowing it to freeze will produce a near 100% theoretical densitymore » product and assures that the product produces no volatile materials upon subsequent heating. Induction cold-crucible melters (ICCM) are being developed for this specific application. An ICCM is, potentially, a high throughput low-cost process. Schematics of a pilot facility were developed for the production of molten DUO{sub 2} from DU{sub 3}O{sub 8} to produce granules <1 mm in diameter in a continuous mode of operation. Thermodynamic analysis was conducted for uranium-oxygen system in the temperature range from 300 to 4000 K in various gas mediums. Temperature limits of stability for various uranium oxides were determined. Experiments on melting DUO{sub 2} were carried out in a high frequency ICCM in a cold crucible with a 120 mm in diameter. The microstructure of molten DUO{sub 2} was studied and lattice parameters were determined. It was experimentally proved, and validated by X-ray analysis, that an opportunity exists to produce molten DUO{sub 2} from mixed oxides (primarily DU{sub 3}O{sub 8}) by reduction melting in ICCM. This will allow using DU{sub 3}O{sub 8} directly to make DUO{sub 2}-a separate unit operation to produce UO{sub 2} feed material is not needed. Experiments were conducted concerning the addition of alloying components, gadolinium et al. oxides, into the DUO{sub 2} melt while in the crucible. These additives improve neutron and gamma radiation shielding and operation properties of the final solids. Cermet samples of 50 wt % DUO{sub 2} were produced. (authors)« less

INTERIOR PHOTO OF HOT PILOT PLANT SECOND FLOOR DEPICTING DETAIL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

INTERIOR PHOTO OF HOT PILOT PLANT SECOND FLOOR DEPICTING DETAIL OF SHIELDED CAVE (CPP-640) LOOKING SOUTHWEST. PHOTO TAKEN FROM NORTH. INL PHOTO NUMBER HD-54-40-2. Mike Crane, Photographer, 7/2006 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR PHOTO OF HOT PILOT PLANT SECOND FLOOR WITH SOUTH ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

INTERIOR PHOTO OF HOT PILOT PLANT SECOND FLOOR WITH SOUTH SECTION OF SHIELDED CAVE IN FOREGROUND (CPP-640) LOOKING NORTHWEST. INL PHOTO NUMBER HD-54-40-1. Mike Crane, Photographer, 7/2006 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

75 FR 53020 - Proposed Collection: Comment Request for Regulation Project

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-30

... record for the tax imposed on the entry of taxable fuel into the U.S. and revises definition of ``enterer.... ADDRESSES: Direct all written comments to Gerald Shields, Internal Revenue Service, Room 6129, 1111... the administration of any internal revenue law. Generally, tax returns and tax return information are...

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

DOE PAGES

Durham, J. M.; Poulson, D.; Bacon, J.; ...

2018-04-10

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. In this paper, we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. Finally, this application of technology and methods commonly used in high-energy particle physics providesmore » a potential solution to this long-standing problem in international nuclear safeguards.« less

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durham, J. M.; Poulson, D.; Bacon, J.

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. In this paper, we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. Finally, this application of technology and methods commonly used in high-energy particle physics providesmore » a potential solution to this long-standing problem in international nuclear safeguards.« less

Dismantlement of the TSF-SNAP Reactor Assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peretz, Fred J

2009-01-01

This paper describes the dismantlement of the Tower Shielding Facility (TSF)?Systems for Nuclear Auxiliary Power (SNAP) reactor, a SNAP-10A reactor used to validate radiation source terms and shield performance models at Oak Ridge National Laboratory (ORNL) from 1967 through 1973. After shutdown, it was placed in storage at the Y-12 National Security Complex (Y-12), eventually falling under the auspices of the Highly Enriched Uranium (HEU) Disposition Program. To facilitate downblending of the HEU present in the fuel elements, the TSF-SNAP was moved to ORNL on June 24, 2006. The reactor assembly was removed from its packaging, inspected, and the sodium-potassiummore » (NaK) coolant was drained. A superheated steam process was used to chemically react the residual NaK inside the reactor assembly. The heat exchanger assembly was removed from the top of the reactor vessel, and the criticality safety sleeve was exchanged for a new safety sleeve that allowed for the removal of the vessel lid. A chain-mounted tubing cutter was used to separate the lid from the vessel, and the 36 fuel elements were removed and packaged in four U.S. Department of Transportation 2R/6M containers. The fuel elements were returned to Y-12 on July 13, 2006. The return of the fuel elements and disposal of all other reactor materials accomplished the formal objectives of the dismantlement project. In addition, a project model was established for the handling of a fully fueled liquid-metal?cooled reactor assembly. Current criticality safety codes have been benchmarked against experiments performed by Atomics International in the 1950s and 1960s. Execution of this project provides valuable experience applicable to future projects addressing space and liquid-metal-cooled reactors.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Silva, R.A.; Cron, J.

This design analysis has shown that, on a conceptual level, the emplacement of drip shields is feasible with current technology and equipment. A plan for drip shield emplacement was presented using a Drip Shield Transporter, a Drip Shield Emplacement Gantry, a locomotive, and a Drip Shield Gantry Carrier. The use of a Drip Shield Emplacement Gantry as an emplacement concept results in a system that is simple, reliable, and interfaces with the numerous other exising repository systems. Using the Waste Emplacement/Retrieval System design as a basis for the drip shield emplacement concept proved to simplify the system by using existingmore » equipment, such as the gantry carrier, locomotive, Electrical and Control systems, and many other systems, structures, and components. Restricted working envelopes for the Drip Shield Emplacement System require further consideration and must be addressed to show that the emplacement operations can be performed as the repository design evolves. Section 6.1 describes how the Drip Shield Emplacement System may use existing equipment. Depending on the length of time between the conclusion of waste emplacement and the commencement of drip shield emplacement, this equipment could include the locomotives, the gantry carrier, and the electrical, control, and rail systems. If the exisiting equipment is selected for use in the Drip Shield Emplacement System, then the length of time after the final stages of waste emplacement and start of drip shield emplacement may pose a concern for the life cycle of the system (e.g., reliability, maintainability, availability, etc.). Further investigation should be performed to consider the use of existing equipment for drip shield emplacement operations. Further investigation will also be needed regarding the interfaces and heat transfer and thermal effects aspects. The conceptual design also requires further design development. Although the findings of this analysis are accurate for the assumptions made, further refinements of this analysis are needed as the project parameters change. The designs of the drip shield, the Emplacement Drift, and the other drip shield emplacement equipment all have a direct effect on the overall design feasibility.« less

Hollow-Wall Heat Shield for Fuel Injector Component

NASA Technical Reports Server (NTRS)

Hanson, Russell B. (Inventor)

2018-01-01

A fuel injector component includes a body, an elongate void and a plurality of bores. The body has a first surface and a second surface. The elongate void is enclosed by the body and is integrally formed between portions of the body defining the first surface and the second surface. The plurality of bores extends into the second surface to intersect the elongate void. A process for making a fuel injector component includes building an injector component body having a void and a plurality of ports connected to the void using an additive manufacturing process that utilizes a powdered building material, and removing residual powdered building material from void through the plurality of ports.

SHIPPING CONTAINER FOR RADIOACTIVE MATERIAL

DOEpatents

Nachbar, H.D.; Biggs, B.B.; Tariello, P.J.; George, K.O.

1963-01-15

A shipping container is described for transponting a large number of radioactive nuclear fuel element modules which produce a substantial amount of heat. The container comprises a primary pressure vessel and shield, and a rotatable head having an access port that can be indexed with module holders in the container. In order to remove heat generated in the fuel eleme nts, a heat exchanger is arranged within the container and in contact with a heat exchange fluid therein. The heat exchanger communicates with additional external heat exchangers, which dissipate heat to the atmosphere. (AEC)

Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.

2013-07-01

Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. Themore » initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage requires integration with current facility operations, and selection of equipment that will allow safe operation within the constraints of existing facility conditions. Examples of such constraints that are evaluated and addressed by the dry storage program include limited basin depth, varying fuel lengths up to 4 m, (13 ft), fissile loading limits, canister closure design, post-load drying and closure of the canisters, instrument selection and installation, and movement of the canisters to storage casks. The initial pilot phase restricts the fuels to shorter length fuels that can be loaded to the canister directly underwater; subsequent phases will require use of a shielded transfer system. Removal of the canister from the basin, followed by drying, inerting, closure of the canister, and transfer of the canister to the storage cask are completed with remotely operated equipment and appropriate shielding to reduce personnel radiation exposure. (authors)« less

Space crew radiation exposure analysis system based on a commercial stand-alone CAD system

NASA Technical Reports Server (NTRS)

Appleby, Matthew H.; Golightly, Michael J.; Hardy, Alva C.

1992-01-01

Major improvements have recently been completed in the approach to spacecraft shielding analysis. A Computer-Aided Design (CAD)-based system has been developed for determining the shielding provided to any point within or external to the spacecraft. Shielding analysis is performed using a commercially available stand-alone CAD system and a customized ray-tracing subroutine contained within a standard engineering modeling software package. This improved shielding analysis technique has been used in several vehicle design projects such as a Mars transfer habitat, pressurized lunar rover, and the redesigned Space Station. Results of these analyses are provided to demonstrate the applicability and versatility of the system.

Design and analysis of a nuclear reactor core for innovative small light water reactors

NASA Astrophysics Data System (ADS)

Soldatov, Alexey I.

In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.

Dynamic Impact Analyses and Tests of Concrete Overpacks - 13638

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Sanghoon; Cho, Sang-Soon; Kim, Ki-Young

Concrete cask is an option for spent nuclear fuel interim storage which is prevailingly used in US. A concrete cask usually consists of metallic canister which confines the spent nuclear fuel and concrete overpack. When the overpack undergoes a severe missile impact which might be caused by a tornado or an aircraft crash, it should sustain acceptable level of structural integrity so that its radiation shielding capability and the retrievability of canister are maintained. Missile impact against a concrete overpack involves two damage modes, local damage and global damage. Local damage of concrete is usually evaluated by empirical formulas whilemore » the global damage is evaluated by finite element analysis. In many cases, those two damage modes are evaluated separately. In this research, a series of numerical simulations are performed using finite element analysis to evaluate the global damage of concrete overpack as well as its local damage under high speed missile impact. We consider two types of concrete overpack, one with steel in-cased concrete without reinforcement and the other with partially-confined reinforced concrete. The numerical simulation results are compared with test results and it is shown that appropriate modeling of material failure is crucial in this analysis and the results are highly dependent on the choice of failure parameters. (authors)« less

Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

DOE PAGES

Poulson, Daniel Cris; Durham, J. Matthew; Guardincerri, Elena; ...

2016-10-22

Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This article describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casksmore » is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ~18σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Finally, we discuss potential detector technologies and geometries.« less

Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poulson, Daniel Cris; Durham, J. Matthew; Guardincerri, Elena

Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This article describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casksmore » is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ~18σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Finally, we discuss potential detector technologies and geometries.« less

Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

NASA Astrophysics Data System (ADS)

Poulson, D.; Durham, J. M.; Guardincerri, E.; Morris, C. L.; Bacon, J. D.; Plaud-Ramos, K.; Morley, D.; Hecht, A. A.

2017-01-01

Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This paper describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casks is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ∼ 18 σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Potential detector technologies and geometries are discussed.

Hybrid Wing Body Configuration System Studies

NASA Technical Reports Server (NTRS)

Nickol, Craig L.; McCullers, Arnie

2009-01-01

The objective of this study was to develop a hybrid wing body (HWB) sizing and analysis capability, apply that capability to estimate the fuel burn potential for an HWB concept, and identify associated technology requirements. An advanced tube with wings concept was also developed for comparison purposes. NASA s Flight Optimization System (FLOPS) conceptual aircraft sizing and synthesis software was modified to enable the sizing and analysis of HWB concepts. The noncircular pressurized centerbody of the HWB concept was modeled, and several options were created for defining the outboard wing sections. Weight and drag estimation routines were modified to accommodate the unique aspects of an HWB configuration. The resulting capability was then utilized to model a proprietary Boeing blended wing body (BWB) concept for comparison purposes. FLOPS predicted approximately a 15 percent greater drag, mainly caused by differences in compressibility drag estimation, and approximately a 5 percent greater takeoff gross weight, mainly caused by the additional fuel required, as compared with the Boeing data. Next, a 777-like reference vehicle was modeled in FLOPS and calibrated to published Boeing performance data; the same mission definition was used to size an HWB in FLOPS. Advanced airframe and propulsion technology assumptions were applied to the HWB to develop an estimate for potential fuel burn savings from such a concept. The same technology assumptions, where applicable, were then applied to an advanced tube-with-wings concept. The HWB concept had a 39 percent lower block fuel burn than the reference vehicle and a 12 percent lower block fuel burn than the advanced tube-with-wings configuration. However, this fuel burn advantage is partially derived from assuming the high-risk technology of embedded engines with boundary-layer-ingesting inlets. The HWB concept does have the potential for significantly reduced noise as a result of the shielding advantages that are inherent with an over-body engine installation.

77 FR 16490 - Airworthiness Directives; Bombardier, Inc. Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-21

...-2C10 (Regional Jet Series 700, 701, & 702) airplanes, Model CL-600-2D15 (Regional Jet Series 705) airplanes, and Model CL-600-2D24 (Regional Jet Series 900) airplanes. This proposed AD was prompted by... and fuel tubes, and protective shields on the rudder quadrant support-beam in the aft equipment...

Structural Materials and Fuels for Space Power Plants

NASA Technical Reports Server (NTRS)

Bowman, Cheryl; Busby, Jeremy; Porter, Douglas

2008-01-01

A fission reactor combined with Stirling convertor power generation is one promising candidate in on-going Fission Surface Power (FSP) studies for future lunar and Martian bases. There are many challenges for designing and qualifying space-rated nuclear power plants. In order to have an affordable and sustainable program, NASA and DOE designers want to build upon the extensive foundation in nuclear fuels and structural materials. This talk will outline the current Fission Surface Power program and outline baseline design options for a lunar power plant with an emphasis on materials challenges. NASA first organized an Affordable Fission Surface Power System Study Team to establish a reference design that could be scrutinized for technical and fiscal feasibility. Previous papers and presentations have discussed this study process in detail. Considerations for the reference design included that no significant nuclear technology, fuels, or material development were required for near term use. The desire was to build upon terrestrial-derived reactor technology including conventional fuels and materials. Here we will present an overview of the reference design, Figure 1, and examine the materials choices. The system definition included analysis and recommendations for power level and life, plant configuration, shielding approach, reactor type, and power conversion type. It is important to note that this is just one concept undergoing refinement. The design team, however, understands that materials selection and improvement must be an integral part of the system development.

High efficiency Dual-Cycle Conversion System using Kr-85.

PubMed

Prelas, Mark A; Tchouaso, Modeste Tchakoua

2018-04-26

This paper discusses the use of one of the safest isotopes known isotopes, Kr-85, as a candidate fuel source for deep space missions. This isotope comes from 0.286% of fission events. There is a vast quantity of Kr-85 stored in spent fuel and it is continually being produced by nuclear reactors. In using Kr-85 with a novel Dual Cycle Conversion System (DCCS) it is feasible to boost the system efficiency from 26% to 45% over a single cycle device while only increasing the system mass by less than 1%. The Kr-85 isotope is the ideal fuel for a Photon Intermediate Direct Energy Conversion (PIDEC) system. PIDEC is an excellent choice for the top cycle in a DCCS. In the top cycle, ionization and excitation of the Kr-85:Cl gas mixture (99% Kr and 1% Cl) from beta particles creates KrCl* excimer photons which are efficiently absorbed by diamond photovoltaic cells on the walls of the pressure vessels. The benefit of using the DCCS is that Kr-85 is capable of operating at high temperatures in the primary cycle and the residual heat can then be converted into electrical power in the bottom cycle which uses a Stirling Engine. The design of the DCCS begins with a spherical pressure vessel of radius 13.7 cm with 3.7 cm thick walls and is filled with a Kr-85:Cl gas mixture. The inner wall has diamond photovoltaic cells attached to it and there is a sapphire window between the diamond photovoltaic cells and the Kr-85:Cl gas mixture which shields the photovoltaic cells from beta particles. The DCCS without a gamma ray shield has specific power of 6.49 W/kg. A removable 6 cm thick tungsten shield is used to safely limit the radiation exposure levels of personnel. A shadow shield remains in the payload to protect the radiation sensitive components in the flight package. The estimated specific power of the unoptimized system design in this paper is about 2.33 W/kg. The specific power of an optimized system should be higher. The Kr-85 isotope is relatively safe because it will disperse quickly in case of an accident and if it enters the lungs there is no significant biological half-life. Copyright © 2018 Elsevier Ltd. All rights reserved.

Analysis of Shield Construction in Spherical Weathered Granite Development Area

NASA Astrophysics Data System (ADS)

Cao, Quan; Li, Peigang; Gong, Shuhua

2018-01-01

The distribution of spherical weathered bodies (commonly known as "boulder") in the granite development area directly affects the shield construction of urban rail transit engineering. This paper is based on the case of shield construction of granite globular development area in Southern China area, the parameter control in shield machine selection and shield advancing during the shield tunneling in this special geological environment is analyzed. And it is suggested that shield machine should be selected for shield construction of granite spherical weathered zone. Driving speed, cutter torque, shield machine thrust, the amount of penetration and the speed of the cutter head of shield machine should be controlled when driving the boulder formation, in order to achieve smooth excavation and reduce the disturbance to the formation.

MTR MAIN FLOOR. NEUTRON TUNNEL (SPANNED BY STILELIKE STEPS) PROJECTS ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR MAIN FLOOR. NEUTRON TUNNEL (SPANNED BY STILE-LIKE STEPS) PROJECTS FROM THE SOUTHEAST CORNER OF THE MTR TOWARD SOUTHEAST CORNER OF BUILDING, WHERE SHIELDING BLOCKS BEGIN TO SURROUND THE TUNNEL AS IT NEARS DETECTING INSTRUMENTS NEAR THE BUILDING WALL. GEAR RELATED TO CRYSTAL NEUTRON SPECTROMETER IS IN FOREGROUND SURROUNDED BY SHIELDING. DATA CONSOLES ARE AT MID-LEVEL OF EAST FACE. OTHER WORK PROCEEDS ON TOP OF AND ELSEWHERE AROUND REACTOR. NOTE TOOLS HANGING AGAINST SOUTHEAST CORNER, USED TO CHANGE FUEL ELEMENTS AND OTHER REACTOR ITEMS DURING REFUELING CYCLES. INL NEGATIVE NO. 10439. Unknown Photographer, 4/20/1954 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Validation of Shielding Analysis Capability of SuperMC with SINBAD

NASA Astrophysics Data System (ADS)

Chen, Chaobin; Yang, Qi; Wu, Bin; Han, Yuncheng; Song, Jing

2017-09-01

Abstract: The shielding analysis capability of SuperMC was validated with the Shielding Integral Benchmark Archive Database (SINBAD). The SINBAD was compiled by RSICC and NEA, it includes numerous benchmark experiments performed with the D-T fusion neutron source facilities of OKTAVIAN, FNS, IPPE, etc. The results from SuperMC simulation were compared with experimental data and MCNP results. Very good agreement with deviation lower than 1% was achieved and it suggests that SuperMC is reliable in shielding calculation.

Basic and Applied Materials Science Research Efforts at MSFC Germane to NASA Goals

NASA Technical Reports Server (NTRS)

2003-01-01

Presently, a number of investigations are ongoing that blend basic research with engineering applications in support of NASA goals. These include (1) "Pore Formation and Mobility (PFMI) " An ISS Glovebox Investigation" NASA Selected Project - 400-34-3D; (2) "Interactions Between Rotating Bodies" Center Director's Discretionary Fund (CDDF) Project - 279-62-00-16; (3) "Molybdenum - Rhenium (Mo-Re) Alloys for Nuclear Fuel Containment" TD Collaboration - 800-11-02; (4) "Fabrication of Alumina - Metal Composites for Propulsion Components" ED Collaboration - 090-50-10; (5) "Radiation Shielding for Deep-Space Missions" SD Effort; (6) "Other Research". In brief, "Pore Formation and Mobility" is an experiment to be conducted in the ISS Microgravity Science Glovebox that will systematically investigate the development, movement, and interactions of bubbles (porosity) during the controlled directional solidification of a transparent material. In addition to promoting our general knowledge of porosity physics, this work will serve as a guide to future ISS experiments utilizing metal alloys. "Interactions Between Rotating Bodies" is a CDDF sponsored project that is critically examining, through theory and experiment, claims of "new" physics relating to gravity modification and electric field effects. "Molybdenum - Rhenium Alloys for Nuclear Fuel Containment" is a TD collaboration in support of nuclear propulsion. Mo-Re alloys are being evaluated and developed for nuclear fuel containment. "Fabrication of Alumina - Metal Composites for Propulsion Components" is an ED collaboration with the intent of increasing strength and decreasing weight of metal engine components through the incorporation of nanometer-sized alumina fibers. "Radiation Shielding for Deep-Space Missions" is an SD effort aimed at minimizing the health risk from radiation to human space voyagers; work to date has been primarily programmatic but experiments to develop hydrogen-rich materials for shielding are planned. "Other Research" includes: BUNDLE (Bridgman Unidirectional Dendrite in a Liquid Experiment) activities (primarily crucible development), vibrational float-zone processing (with Vanderbilt University), use of ultrasonics in materials processing (with UAH), rotational effects on microstructural development, and application of magnetic fields for mixing.

Securing the data economy: translating privacy and enacting security in the development of DataSHIELD.

PubMed

Murtagh, M J; Demir, I; Jenkings, K N; Wallace, S E; Murtagh, B; Boniol, M; Bota, M; Laflamme, P; Boffetta, P; Ferretti, V; Burton, P R

2012-01-01

Contemporary bioscience is seeing the emergence of a new data economy: with data as its fundamental unit of exchange. While sharing data within this new 'economy' provides many potential advantages, the sharing of individual data raises important social and ethical concerns. We examine ongoing development of one technology, DataSHIELD, which appears to elide privacy concerns about sharing data by enabling shared analysis while not actually sharing any individual-level data. We combine presentation of the development of DataSHIELD with presentation of an ethnographic study of a workshop to test the technology. DataSHIELD produced an application of the norm of privacy that was practical, flexible and operationalizable in researchers' everyday activities, and one which fulfilled the requirements of ethics committees. We demonstrated that an analysis run via DataSHIELD could precisely replicate results produced by a standard analysis where all data are physically pooled and analyzed together. In developing DataSHIELD, the ethical concept of privacy was transformed into an issue of security. Development of DataSHIELD was based on social practices as well as scientific and ethical motivations. Therefore, the 'success' of DataSHIELD would, likewise, be dependent on more than just the mathematics and the security of the technology. Copyright © 2012 S. Karger AG, Basel.

A Ballistic Limit Analysis Program for Shielding Against Micrometeoroids and Orbital Debris

NASA Technical Reports Server (NTRS)

Ryan, Shannon; Christiansen, Erie

2010-01-01

A software program has been developed that enables the user to quickly and simply perform ballistic limit calculations for common spacecraft structures that are subject to hypervelocity impact of micrometeoroid and orbital debris (MMOD) projectiles. This analysis program consists of two core modules: design, and; performance. The design module enables a user to calculate preliminary dimensions of a shield configuration (e.g., thicknesses/areal densities, spacing, etc.) for a ?design? particle (diameter, density, impact velocity, incidence). The performance module enables a more detailed shielding analysis, providing the performance of a user-defined shielding configuration over the range of relevant in-orbit impact conditions.

Modelling and Optimization of Four-Segment Shielding Coils of Current Transformers

PubMed Central

Gao, Yucheng; Zhao, Wei; Wang, Qing; Qu, Kaifeng; Li, He; Shao, Haiming; Huang, Songling

2017-01-01

Applying shielding coils is a practical way to protect current transformers (CTs) for large-capacity generators from the intensive magnetic interference produced by adjacent bus-bars. The aim of this study is to build a simple analytical model for the shielding coils, from which the optimization of the shielding coils can be calculated effectively. Based on an existing stray flux model, a new analytical model for the leakage flux of partial coils is presented, and finite element method-based simulations are carried out to develop empirical equations for the core-pickup factors of the models. Using the flux models, a model of the common four-segment shielding coils is derived. Furthermore, a theoretical analysis is carried out on the optimal performance of the four-segment shielding coils in a typical six-bus-bars scenario. It turns out that the “all parallel” shielding coils with a 45° starting position have the best shielding performance, whereas the “separated loop” shielding coils with a 0° starting position feature the lowest heating value. Physical experiments were performed, which verified all the models and the conclusions proposed in the paper. In addition, for shielding coils with other than the four-segment configuration, the analysis process will generally be the same. PMID:28587137

Modelling and Optimization of Four-Segment Shielding Coils of Current Transformers.

PubMed

Gao, Yucheng; Zhao, Wei; Wang, Qing; Qu, Kaifeng; Li, He; Shao, Haiming; Huang, Songling

2017-05-26

Applying shielding coils is a practical way to protect current transformers (CTs) for large-capacity generators from the intensive magnetic interference produced by adjacent bus-bars. The aim of this study is to build a simple analytical model for the shielding coils, from which the optimization of the shielding coils can be calculated effectively. Based on an existing stray flux model, a new analytical model for the leakage flux of partial coils is presented, and finite element method-based simulations are carried out to develop empirical equations for the core-pickup factors of the models. Using the flux models, a model of the common four-segment shielding coils is derived. Furthermore, a theoretical analysis is carried out on the optimal performance of the four-segment shielding coils in a typical six-bus-bars scenario. It turns out that the "all parallel" shielding coils with a 45° starting position have the best shielding performance, whereas the "separated loop" shielding coils with a 0° starting position feature the lowest heating value. Physical experiments were performed, which verified all the models and the conclusions proposed in the paper. In addition, for shielding coils with other than the four-segment configuration, the analysis process will generally be the same.

DETECTION OF COATING FAILURES IN A NEUTRONIC REACTOR

DOEpatents

Snell, A.H.; Allison, S.K.

1958-02-11

This patent relates to water-cooled reactor systems and discloses a means to detect leaks in the jackets of jacketed fuel elements comprising a neutron detector located in the cooling water discharge pipe,the pipe being provided with an enlarged portion for housing the detector so that the latter is completely surrounded by the water in its passage through the pipe, said enlarged portion and detector being shielded from the reactor for the purpose of detecting only those delayed neutrons emitted in the cooling water and due to the latter picking up fission fragments from the defective fuel elements.

Thermionic energy conversion technology - Present and future

NASA Technical Reports Server (NTRS)

Shimada, K.; Morris, J. F.

1977-01-01

Aerospace and terrestrial applications of thermionic direct energy conversion and advances in direct energy conversion (DEC) technology are surveyed. Electrode materials, the cesium plasma drop (the difference between the barrier index and the collector work function), DEC voltage/current characteristics, conversion efficiency, and operating temperatures are discussed. Attention is centered on nuclear reactor system thermionic DEC devices, for in-core or out-of-core operation. Thermionic fuel elements, the radiation shield, power conditions, and a waste heat rejection system are considered among the thermionic DEC system components. Terrestrial applications include topping power systems in fossil fuel and solar power generation.

Thermos reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Labrousse, M.; Lerouge, B.; Dupuy, G.

1978-04-01

THERMOS is a water reactor designed to provide hot water up to 120/sup 0/C for district heating or for desalination applications. It is a 100-MW reactor based on proven technology: oxide fuel plate elements, integrated primary circuit, and reactor vessel located in the bottom of a pool. As in swimming pool reactors, the pool is used for biological shielding, emergency core cooling, and fission product filtering (in case of an accident). Before economics, safety is the main characteristic of the concept: no fuel failure admitted, core under water in any accidental configuration, inspection of every ''nuclear'' component, and double-wall containment.

Neutron field characterization at the independent spent fuel storage installation of the Trillo nuclear power plant.

PubMed

Campo, Xandra; Méndez, Roberto; Embid, Miguel; Ortego, Alberto; Novo, Manuel; Sanz, Javier

2018-05-01

Neutron fields inside and outside the independent spent fuel storage installation of Trillo Nuclear Power Plant are characterized exhaustively in terms of neutron spectra and ambient dose equivalent, measured by Bonner sphere system and LB6411 monitor. Measurements are consistent with storage casks and building shield characteristics, and also with casks distribution inside the building. Outer values at least five times lower than dose limit for free access area are found. Measurements with LB6411 and spectrometer are consistent with each other. Copyright © 2018 Elsevier Ltd. All rights reserved.

77 FR 36129 - Airworthiness Directives; Bombardier, Inc. Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-18

... (AD) for certain Bombardier, Inc. Model CL-600-2C10 (Regional Jet Series 700, 701, & 702) airplanes, Model CL-600-2D15 (Regional Jet Series 705) airplanes, and Model CL-600-2D24 (Regional Jet Series 900... the wing box and fuel tubes, and protective shields on the rudder quadrant support-beam in the aft...

Shielding effectiveness of multiple-shield cables with arbitrary terminations via transmission line analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Campione, Salvatore; Basilio, Lorena I.; Warne, Larry Kevin

Our paper reports on a transmission-line model for calculating the shielding effectiveness of multiple-shield cables with arbitrary terminations. Since the shields are not perfect conductors and apertures in the shields permit external magnetic and electric fields to penetrate into the interior regions of the cable, we use this model to estimate the effects of the outer shield current and voltage (associated with the external excitation and boundary conditions associated with the external conductor) on the inner conductor current and voltage. It is commonly believed that increasing the number of shields of a cable will improve the shielding performance. But thismore » is not always the case, and a cable with multiple shields may perform similar to or worse than a cable with a single shield. Furthermore, we want to shed more light on these situations, which represent the main focus of this paper.« less

Shielding effectiveness of multiple-shield cables with arbitrary terminations via transmission line analysis

DOE PAGES

Campione, Salvatore; Basilio, Lorena I.; Warne, Larry Kevin; ...

2016-06-25

Our paper reports on a transmission-line model for calculating the shielding effectiveness of multiple-shield cables with arbitrary terminations. Since the shields are not perfect conductors and apertures in the shields permit external magnetic and electric fields to penetrate into the interior regions of the cable, we use this model to estimate the effects of the outer shield current and voltage (associated with the external excitation and boundary conditions associated with the external conductor) on the inner conductor current and voltage. It is commonly believed that increasing the number of shields of a cable will improve the shielding performance. But thismore » is not always the case, and a cable with multiple shields may perform similar to or worse than a cable with a single shield. Furthermore, we want to shed more light on these situations, which represent the main focus of this paper.« less

Structural Analysis of Thermal Shields During a Quench of a Torus Magnet for the 12 GeV Upgrade

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pastor, Orlando; Willard, Thomas; Ghoshal, Probir K.

A toroidal magnet system consisting of six superconducting coils is being built for the Jefferson Lab 12- GeV accelerator upgrade project. This paper details the analysis of eddy current effects during a quench event on the aluminum thermal shield. The shield has been analyzed for mechanical stresses induced as a result of a coil quench as well as a fast discharge of the complete magnet system. The shield has been designed to reduce the eddy current effects and result in stresses within allowable limits.

Secondary Impacts on Structures on the Lunar Surface

NASA Technical Reports Server (NTRS)

Christiansen, Eric; Walker, James D.; Grosch, Donald J.

2010-01-01

The Altair Lunar Lander is being designed for the planned return to the Moon by 2020. Since it is hoped that lander components will be re-used by later missions, studies are underway to examine the exposure threat to the lander sitting on the Lunar surface for extended periods. These threats involve both direct strikes of meteoroids on the vehicle as well as strikes from Lunar regolith and rock thrown by nearby meteorite strikes. Currently, the lander design is comprised of up to 10 different types of pressure vessels. These vessels included the manned habitation module, fuel, cryogenic fuel and gas storage containers, and instrument bays. These pressure vessels have various wall designs, including various aluminum alloys, honeycomb, and carbon-fiber composite materials. For some of the vessels, shielding is being considered. This program involved the test and analysis of six pressure vessel designs, one of which included a Whipple bumper shield. In addition to the pressure vessel walls, all the pressure vessels are wrapped in multi-layer insulation (MLI). Two variants were tested without the MLI to better understand the role of the MLI in the impact performance. The tests of performed were to examine the secondary impacts on these structures as they rested on the Lunar surface. If a hypervelocity meteor were to strike the surface nearby, it would throw regolith and rock debris into the structure at a much lower velocity. Also, when the manned module departs for the return to Earth, its rocket engines throw up debris that can impact the remaining lander components and cause damage. Glass spheres were used as a stimulant for the regolith material. Impact tests were performed with a gas gun to find the V50 of various sized spheres striking the pressure vessels. The impacts were then modeled and a fast-running approximate model for the V50 data was developed. This model was for performing risk analysis to assist in the vessel design and in the identification of ideal long-term mission sites. This paper reviews the impact tests and analysis and modeling examining the impact threat to various components in the lander design.

Remote Assessment of Lunar Resource Potential

NASA Technical Reports Server (NTRS)

Taylor, G. Jeffrey

1992-01-01

Assessing the resource potential of the lunar surface requires a well-planned program to determine the chemical and mineralogical composition of the Moon's surface at a range of scales. The exploration program must include remote sensing measurements (from both Earth's surface and lunar orbit), robotic in situ analysis of specific places, and eventually, human field work by trained geologists. Remote sensing data is discussed. Resource assessment requires some idea of what resources will be needed. Studies thus far have concentrated on oxygen and hydrogen production for propellant and life support, He-3 for export as fuel for nuclear fusion reactors, and use of bulk regolith for shielding and construction materials. The measurement requirements for assessing these resources are given and discussed briefly.

Assessment of the Performance Potential of Advanced Subsonic Transport Concepts for NASA's Environmentally Responsible Aviation Project

NASA Technical Reports Server (NTRS)

Nickol, Craig L.; Haller, William J.

2016-01-01

NASA's Environmentally Responsible Aviation (ERA) project has matured technologies to enable simultaneous reductions in fuel burn, noise, and nitrogen oxide (NOx) emissions for future subsonic commercial transport aircraft. The fuel burn reduction target was a 50% reduction in block fuel burn (relative to a 2005 best-in-class baseline aircraft), utilizing technologies with an estimated Technology Readiness Level (TRL) of 4-6 by 2020. Progress towards this fuel burn reduction target was measured through the conceptual design and analysis of advanced subsonic commercial transport concepts spanning vehicle size classes from regional jet (98 passengers) to very large twin aisle size (400 passengers). Both conventional tube-and-wing (T+W) concepts and unconventional (over-wing-nacelle (OWN), hybrid wing body (HWB), mid-fuselage nacelle (MFN)) concepts were developed. A set of propulsion and airframe technologies were defined and integrated onto these advanced concepts which were then sized to meet the baseline mission requirements. Block fuel burn performance was then estimated, resulting in reductions relative to the 2005 best-in-class baseline performance ranging from 39% to 49%. The advanced single-aisle and large twin aisle T+W concepts had reductions of 43% and 41%, respectively, relative to the 737-800 and 777-200LR aircraft. The single-aisle OWN concept and the large twin aisle class HWB concept had reductions of 45% and 47%, respectively. In addition to their estimated fuel burn reduction performance, these unconventional concepts have the potential to provide significant noise reductions due, in part, to engine shielding provided by the airframe. Finally, all of the advanced concepts also have the potential for significant NOx emissions reductions due to the use of advanced combustor technology. Noise and NOx emissions reduction estimates were also generated for these concepts as part of the ERA project.

Depleted uranium as a backfill for nuclear fuel waste package

DOEpatents

Forsberg, Charles W.

1998-01-01

A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

Depleted uranium as a backfill for nuclear fuel waste package

DOEpatents

Forsberg, C.W.

1998-11-03

A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

Modeling and Analysis of Geoelectric Fields: Extended Solar Shield

NASA Astrophysics Data System (ADS)

Ngwira, C. M.; Pulkkinen, A. A.

2016-12-01

In the NASA Applied Sciences Program Solar Shield project, an unprecedented first-principles-based system to forecast geomagnetically induced current (GIC) in high-voltage power transmission systems was developed. Rapid progress in the field of numerical physics-based space environment modeling has led to major developments over the past few years. In this study modeling and analysis of induced geoelectric fields is discussed. Specifically, we focus on the successful incorporation of 3-D EM transfer functions in the modeling of E-fields, and on the analysis of near real-time simulation outputs used in the Solar Shield forecast system. The extended Solar Shield is a collaborative project between DHS, NASA, NOAA, CUA and EPRI.

An Analysis of Radiation Penetration through the U-Shaped Cast Concrete Joints of Concrete Shielding in the Multipurpose Gamma Irradiator of BATAN

NASA Astrophysics Data System (ADS)

Ardiyati, Tanti; Rozali, Bang; Kasmudin

2018-02-01

An analysis of radiation penetration through the U-shaped joints of cast concrete shielding in BATAN’s multipurpose gamma irradiator has been carried out. The analysis has been performed by calculating the radiation penetration through the U-shaped joints of the concrete shielding using MCNP computer code. The U-shaped joints were a new design in massive concrete construction in Indonesia and, in its actual application, it is joined by a bonding agent. In the MCNP simulation model, eight detectors were located close to the observed irradiation room walls of the concrete shielding. The simulation results indicated that the radiation levels outside the concrete shielding was less than the permissible limit of 2.5 μSv/h so that the workers could safely access electrical room, control room, water treatment facility and outside irradiation room. The radiation penetration decreased as the density of material increased.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sarta, Jose A.; Castiblanco, Luis A

With cooperation of the International Atomic Energy Agency (IAEA) and the Department of Energy (DOE) of the United States, several calculations and tasks related to the waste disposal of spent MTR fuel enriched nominally to 93% were carried out for the conversion of the IAN-R1 Research Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to remove the spent MTR-HEU fuel of the core and store it safely a program was established at the Instituto de Ciencias Nucleares y Energias Alternativas (INEA). This program included training, acquisition of hardware and software, design and construction of a decay pool, transfer ofmore » the spent HEU fuel elements into the decay pool and his final transport to Savannah River in United States. In this paper are presented data of activities calculated for each relevant radionuclide present in spent MTR-HEU fuel elements of the IAN-R1 Research Reactor and the total activity. The total activity calculated takes in consideration contributions of fission, activation and actinides products. The data obtained were the base for shielding calculations for the decay pool concerning the storage of spent MTR-HEU fuel elements and the respective dosimetric evaluations in the transferring operations of fuel elements into the decay pool.« less

Shielding of substations against direct lightning strokes by shield wires

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chowdhuri, P.

1994-01-01

A new analysis for shielding outdoor substations against direct lightning strokes by shield wires is proposed. The basic assumption of this proposed method is that any lightning stroke which penetrates the shields will cause damage. The second assumption is that a certain level of risk of failure must be accepted, such as one or two failures per 100 years. The proposed method, using electrogeometric model, was applied to design shield wires for two outdoor substations: (1) 161-kV/69-kV station, and (2) 500-kV/161-kV station. The results of the proposed method were also compared with the shielding data of two other substations.

CAD-based stand-alone spacecraft radiation exposure analysis system: An application of the early man-tended Space Station

NASA Technical Reports Server (NTRS)

Appleby, M. H.; Golightly, M. J.; Hardy, A. C.

1993-01-01

Major improvements have been completed in the approach to analyses and simulation of spacecraft radiation shielding and exposure. A computer-aided design (CAD)-based system has been developed for determining the amount of shielding provided by a spacecraft and simulating transmission of an incident radiation environment to any point within or external to the vehicle. Shielding analysis is performed using a customized ray-tracing subroutine contained within a standard engineering modeling software package. This improved shielding analysis technique has been used in several vehicle design programs such as a Mars transfer habitat, pressurized lunar rover, and the redesigned international Space Station. Results of analysis performed for the Space Station astronaut exposure assessment are provided to demonastrate the applicability and versatility of the system.

A shielded measurement system for irradiated nuclear fuel measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mosby, W.R.; Aumeier, S.E.; Klann, R.T.

1999-07-01

The US Department of Energy (DOE) is driving a transition toward dry storage of irradiated nuclear fuel (INF), toward characterization of INF for final disposition, and toward resumption of measurement-based material control and accountability (MC and A) efforts for INF. For these reasons, the ability to efficiently acquire radiological measurements of INF in a dry environment is important. The DOE has recently developed a guidance document proposing MC and A requirements for INF. The intent of this document is to encourage the direct measurement of INF on inventory within DOE. The guidance document reinforces and clarifies existing material safeguards requirementsmore » as they pertain to INF. Validation of nuclear material contents of non-self-protecting INF must be accomplished by direct measurement, application of validated burnup codes using qualified initial fissile content, burnup data, and age or by other valid means. The fuel units must remain intact with readable identification numbers. INF may be subject to periodic inventories with visual item accountability checks. Quantitative measurements may provide greater assurance of the integrity of INF inventories at a lower cost and with less personnel exposure than visual item accountability checks. Currently, several different approaches are used to measure the radiological attributes of INF. Although these systems are useful for a wide variety of applications, there is currently no relatively inexpensive measurement system that is readily deployable for INF measurements for materials located in dry storage. The authors present the conceptual design of a shielded measurement system (SMS) that could be used for this purpose. The SMS consists of a shielded enclosure designed to house a collection of measurement systems to allow measurements on spent fuel outside of a hot cell. The phase 1 SMS will contain {sup 3}He detectors and ionization chambers to allow for gross neutron and gamma-ray measurements. The phase 2 SMS will be developed by adding additional measurement capabilities to the phase 1 SMS. Planned additions include medium-resolution gamma-ray detectors (CdZnTe or high-pressure Xe), additional {sup 3}He tubes to allow coincidence measurements, and a {sup 252}Cf neutron source and motion control system to allow active neutron interrogation measurements. The phase 2 SMS will be capable of performing more direct measurements of INF properties such as burnup, cooling time, spontaneous fission isotope contents, and total fissile contents.« less

Analysis of Broadband Metamaterial Shielding for Counter-Directed Energy Weapons

DTIC Science & Technology

2017-06-01

SHIELDING FOR COUNTER-DIRECTED ENERGY WEAPONS by Chester H. Hewitt III June 2017 Thesis Advisor: Dragoslav Grbovic Second Reader: James H...COVERED Master’s thesis 4. TITLE AND SUBTITLE ANALYSIS OF BROADBAND METAMATERIAL SHIELDING FOR COUNTER-DIRECTED ENERGY WEAPONS 5. FUNDING NUMBERS 6...high-power microwave (HPM) directed- energy weapons (DEWs), which can disrupt electronics remotely with great accuracy without the need to inflict

Designing dual-plate meteoroid shields: A new analysis

NASA Technical Reports Server (NTRS)

Swift, H. F.; Bamford, R.; Chen, R.

1982-01-01

Physics governing ultrahigh velocity impacts onto dual-plate meteor armor is discussed. Meteoroid shield design methodologies are considered: failure mechanisms, qualitative features of effective meteoroid shield designs, evaluating/processing meteoroid threat models, and quantitative techniques for optimizing effective meteoroid shield designs. Related investigations are included: use of Kevlar cloth/epoxy panels in meteoroid shields for the Halley's Comet intercept vehicle, mirror exposure dynamics, and evaluation of ion fields produced around the Halley Intercept Mission vehicle by meteoroid impacts.

Design Analysis of SNS Target StationBiological Shielding Monoligh with Proton Power Uprate

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bekar, Kursat B.; Ibrahim, Ahmad M.

2017-05-01

This report documents the analysis of the dose rate in the experiment area outside the Spallation Neutron Source (SNS) target station shielding monolith with proton beam energy of 1.3 GeV. The analysis implemented a coupled three dimensional (3D)/two dimensional (2D) approach that used both the Monte Carlo N-Particle Extended (MCNPX) 3D Monte Carlo code and the Discrete Ordinates Transport (DORT) two dimensional deterministic code. The analysis with proton beam energy of 1.3 GeV showed that the dose rate in continuously occupied areas on the lateral surface outside the SNS target station shielding monolith is less than 0.25 mrem/h, which compliesmore » with the SNS facility design objective. However, the methods and codes used in this analysis are out of date and unsupported, and the 2D approximation of the target shielding monolith does not accurately represent the geometry. We recommend that this analysis is updated with modern codes and libraries such as ADVANTG or SHIFT. These codes have demonstrated very high efficiency in performing full 3D radiation shielding analyses of similar and even more difficult problems.« less

SCALE: A modular code system for performing Standardized Computer Analyses for Licensing Evaluation. Volume 1, Part 2: Control modules S1--H1; Revision 5

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

SCALE--a modular code system for Standardized Computer Analyses Licensing Evaluation--has been developed by Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission. The SCALE system utilizes well-established computer codes and methods within standard analysis sequences that (1) allow an input format designed for the occasional user and/or novice, (2) automated the data processing and coupling between modules, and (3) provide accurate and reliable results. System development has been directed at problem-dependent cross-section processing and analysis of criticality safety, shielding, heat transfer, and depletion/decay problems. Since the initial release of SCALE in 1980, the code system hasmore » been heavily used for evaluation of nuclear fuel facility and package designs. This revision documents Version 4.3 of the system.« less

Nuclear Fuel Assay through analysis of Uranium L-shell by Hybrid L-edge/XRF Densitometer using a Surrogate Material

NASA Astrophysics Data System (ADS)

Park, Seunghoon; Joung, Sungyeop; Park, Jerry AB(; ), AC(; )

2018-01-01

Assay of L-series of nuclear material solution is useful for determination of amount of nuclear materials and ratio of minor actinide in the materials. The hybrid system of energy dispersive X-ray absorption edge spectrometry, i.e. L-edge densitometry, and X-ray fluorescence spectrometry is one of the analysis methods. The hybrid L-edge/XRF densitometer can be a promising candidate for a portable and compact equipment due to advantage of using low energy X-ray beams without heavy shielding systems and liquid nitrogen cooling compared to hybrid K-edge/XRF densitometer. A prototype of the equipment was evaluated for feasibility of the nuclear material assay using a surrogate material (lead) to avoid radiation effects from nuclear materials. The uncertainty of L-edge and XRF characteristics of the sample material and volume effects was discussed in the article.

Determination of the gamma-ray skyshine dose contribution in a Loss Of Shielding accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dennis, M.L.; Weiner, R.F.; Osborn, D.M.

2007-07-01

The goal of this research is to determine the gamma-ray dose contribution from skyshine. In a transportation accident involving the loss of lead gamma shielding, first responders to the accident will be exposed to both direct gamma radiation streaming from the exposed spent nuclear fuel and atmospherically reflected gamma radiation. The reflected radiation is referred to as skyshine and should contribute minimally to the overall dose; however, when there is minimal shielding above the exposed source, skyshine at large distances from the source must be considered. The program SKYDOSE developed by Shultis and Faw evaluates the gamma-ray skyshine dose frommore » a point, isotropic, polyenergetic, gamma-photon source. Assuming an infinite black wall shielding all direct radiation, the model assumes a first responder is located at varying distances from the wall. Skyshine doses are calculated both through SKYDOSE's integral line-beam method and an approximate approach prescribed by the National Council of Radiation Protection and Measurements. Initial results from SKYDOSE indicate nearly equivalent dose rates from either direct or skyshine radiation at nine meters from the wall, which seemed unusual and not readily explained. NCRP methodology, however, yields skyshine dose rates which are drastically smaller than direct dose rates at the same distance. Further investigation using the program MicroSkyshine{sup R}, which allows a variety of source configurations, suggests skyshine contributes minimally to dose in a loss-of-shielding accident. (authors)« less

Nuclear fuel element

DOEpatents

Armijo, Joseph S.; Coffin, Jr., Louis F.

1980-04-29

A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.

Summary of Prometheus Radiation Shielding Nuclear Design Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Stephens

2006-01-13

This report transmits a summary of radiation shielding nuclear design studies performed to support the Prometheus project. Together, the enclosures and references associated with this document describe NRPCT (KAPL & Bettis) shielding nuclear design analyses done for the project.

Occupational health and safety assessment of exposure to jet fuel combustion products in air medical transport.

PubMed

MacDonald, Russell D; Thomas, Laura; Rusk, Frederick C; Marques, Shauna D; McGuire, Dan

2010-01-01

Transport medicine personnel are potentially exposed to jet fuel combustion products. Setting-specific data are required to determine whether this poses a risk. This study assessed exposure to jet fuel combustion products, compared various engine ignition scenarios, and determined methods to minimize exposure. The Beechcraft King Air B200 turboprop aircraft equipped with twin turbine engines, using a kerosene-based jet fuel (Jet A-1), was used to measure products of combustion during boarding, engine startup, and flight in three separate engine start scenarios ("shielded": internal engine start, door closed; "exposed": ground power unit start, door open; and "minimized": ground power unit right engine start, door open). Real-time continuous monitoring equipment was used for oxygen, carbon dioxide, carbon monoxide, nitrogen dioxide, hydrogen sulfide, sulfur dioxide, volatile organic compounds, and particulate matter. Integrated methods were used for aldehydes, polycyclic aromatic hydrocarbons, volatile organic compounds, and aliphatic hydrocarbons. Samples were taken in the paramedic breathing zone for approximately 60 minutes, starting just before the paramedics boarded the aircraft. Data were compared against regulated time-weighted exposure thresholds to determine the presence of potentially harmful products of combustion. Polycyclic aromatic hydrocarbons, aldehydes, volatile organic compounds, and aliphatic hydrocarbons were found at very low concentrations or beneath the limits of detection. There were significant differences in exposures to particulates, carbon monoxide, and total volatile organic compound between the "exposed" and "minimized" scenarios. Elevated concentrations of carbon monoxide and total volatile organic compounds were present during the ground power unit-assisted dual-engine start. There were no appreciable exposures during the "minimized" or "shielded" scenarios. Air medical personnel exposures to jet fuel combustion products were generally low and did not exceed established U.S. or Canadian health and safety exposure limits. Avoidance of ground power unit-assisted dual-engine starts and closing the hangar door prior to start minimize or eliminate the occupational exposure.

Metal droplet erosion and shielding plasma layer under plasma flows typical of transient processes in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martynenko, Yu. V., E-mail: Martynenko-YV@nrcki.ru

It is shown that the shielding plasma layer and metal droplet erosion in tokamaks are closely interrelated, because shielding plasma forms from the evaporated metal droplets, while droplet erosion is caused by the shielding plasma flow over the melted metal surface. Analysis of experimental data and theoretical models of these processes is presented.

Performance of Fuels, Lubricants, and Associated Products Used during Operation Desert Shield/Storm

DTIC Science & Technology

1992-08-01

to have enabled the. units to perform their mission. However, most units reported taking enough hydi aulic fluid with them to Southwest Asia, while...Fluid Lubncant for Extreme Climates " Sand does not mick to at Did not work good on 30mm MG on Apache Did not work for "coax" guns on Bradley Crew

Albedo Neutron Dosimetry in a Deep Geological Disposal Repository for High-Level Nuclear Waste.

PubMed

Pang, Bo; Becker, Frank

2017-04-28

Albedo neutron dosemeter is the German official personal neutron dosemeter in mixed radiation fields where neutrons contribute to personal dose. In deep geological repositories for high-level nuclear waste, where neutrons can dominate the radiation field, it is of interest to investigate the performance of albedo neutron dosemeter in such facilities. In this study, the deep geological repository is represented by a shielding cask loaded with spent nuclear fuel placed inside a rock salt emplacement drift. Due to the backscattering of neutrons in the drift, issues concerning calibration of the dosemeter arise. Field-specific calibration of the albedo neutron dosemeter was hence performed with Monte Carlo simulations. In order to assess the applicability of the albedo neutron dosemeter in a deep geological repository over a long time scale, spent nuclear fuel with different ages of 50, 100 and 500 years were investigated. It was found out, that the neutron radiation field in a deep geological repository can be assigned to the application area 'N1' of the albedo neutron dosemeter, which is typical in reactors and accelerators with heavy shielding. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Numerical Simulation of Earth Pressure on Head Chamber of Shield Machine with FEM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li Shouju; Kang Chengang; Sun, Wei

2010-05-21

Model parameters of conditioned soils in head chamber of shield machine are determined based on tree-axial compression tests in laboratory. The loads acting on tunneling face are estimated according to static earth pressure principle. Based on Duncan-Chang nonlinear elastic constitutive model, the earth pressures on head chamber of shield machine are simulated in different aperture ratio cases for rotating cutterhead of shield machine. Relationship between pressure transportation factor and aperture ratio of shield machine is proposed by using aggression analysis.

Development and calibration of the shielded measurement system for fissile contents measurements on irradiated nuclear fuel in dry storage.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mosby, W. R.; Jensen, B. A.

2002-05-31

In recent years there has been a trend towards storage of Irradiated Nuclear Fuel (INF) in dry conditions rather than in underwater environments. At the same time, the Department of Energy (DOE) has begun encouraging custodians of INF to perform measurements on INF for which no recent fissile contents measurement data exists. INF, in the form of spent fuel from Experimental Breeder Reactor 2 (EBR-II), has been stored in close-fitting, dry underground storage locations at the Radioactive Scrap and Waste Facility (RSWF) at Argonne National Laboratory-West (ANL-W) for many years. In Fiscal Year 2000, funding was obtained from the DOEmore » Office of Safeguards and Security Technology Development Program to develop and prepare for deployment a Shielded Measurement System (SMS) to perform fissile content measurements on INF stored in the RSWF. The SMS is equipped to lift an INF item out of its storage location, perform scanning neutron coincidence and high-resolution gamma-ray measurements, and restore the item to its storage location. The neutron and gamma-ray measurement results are compared to predictions based on isotope depletion and Monte Carlo neutral-particle transport models to provide confirmation of the accuracy of the models and hence of the fissile material contents of the item as calculated by the same models. This paper describes the SMS and discusses the results of the first calibration and validation measurements performed with the SMS.« less

Hypervelocity impact testing above 10 km/s of advanced orbital debris shields

NASA Astrophysics Data System (ADS)

Christiansen, Eric L.; Crews, Jeanne Lee; Kerr, Justin H.; Chhabildas, Lalit C.

1996-05-01

NASA has developed enhanced performance shields to improve the protection of spacecraft from orbital debris and meteoroid impacts. One of these enhanced shields includes a blanket of Nextel™ ceramic fabric and Kevlar™ high strength fabric that is positioned midway between an aluminum bumper and the spacecraft pressure wall. As part of the evaluation of this new shielding technology, impact data above 10 km/sec has been obtained by NASA Johnson Space Center (JSC) from the Sandia National Laboratories HVL ("hypervelocity launcher") and the Southwest Research Institute inhibited shaped charge launcher (ISCL). The HVL launches flyer-plates in the velocity range of 10 to 15 km/s while the ISCL launches hollow cylinders at ˜11.5 km/s. The >10 km/s experiments are complemented by hydrocode analysis and light-gas gun testing at the JSC Hypervelocity Impact Test Facility (HIT-F) to assess the effects of projectile shape on shield performance. Results from the testing and analysis indicate that the Nextel™/Kevlar™ shield provides superior protection performance compared to an all-aluminum shield alternative.

Advanced Multifunctional MMOD Shield: Radiation Shielding Assessment

NASA Technical Reports Server (NTRS)

Rojdev, Kristina; Christiansen, Eric

2013-01-01

As NASA is looking to explore further into deep space, multifunctional materials are a necessity for decreasing complexity and mass. One area where multifunctional materials could be extremely beneficial is in the micrometeoroid orbital debris (MMOD) shield. A typical MMOD shield on the International Space Station (ISS) is a stuffed whipple shield consisting of multiple layers. One of those layers is the thermal blanket, or multi-layer insulation (MLI). Increasing the MMOD effectiveness of MLI blankets, while still preserving their thermal capabilities, could allow for a less massive MMOD shield. Thus, a study was conducted to evaluate a concept MLI blanket for an MMOD shield. In conjunction, this MLI blanket and the subsequent MMOD shield was also evaluated for its radiation shielding effectiveness towards protecting crew. The overall MMOD shielding system using the concept MLI blanket proved to only have a marginal increase in the radiation mitigating properties. Therefore, subsequent analysis was performed on various conceptual MMOD shields to determine the combination of materials that may prove superior for radiation mitigating purposes. The following paper outlines the evaluations performed and discusses the results and conclusions of this evaluation for radiation shielding effectiveness.

Analytic Shielding Optimization to Reduce Crew Exposure to Ionizing Radiation Inside Space Vehicles

NASA Technical Reports Server (NTRS)

Gaza, Razvan; Cooper, Tim P.; Hanzo, Arthur; Hussein, Hesham; Jarvis, Kandy S.; Kimble, Ryan; Lee, Kerry T.; Patel, Chirag; Reddell, Brandon D.; Stoffle, Nicholas;

Analysis of Doppler Lidar Data Acquired During the Pentagon Shield Field Campaign

DTIC Science & Technology

2011-04-01

two coherent Doppler lidars deployed during the Pentagon Shield field campaign are analyzed in conjunction with other sensors to characterize the...Observations from two coherent Doppler lidars deployed during the Pentagon Shield field campaign are analyzed in conjunction with other sensors to... coherent Doppler lidars deployed during the Pentagon Shield field campaign are analyzed in conjunction with other sensors to characterize the overall

The ENABLER - Based on proven NERVA technology

NASA Astrophysics Data System (ADS)

Livingston, Julie M.; Pierce, Bill L.

The ENABLER reactor for use in a nuclear thermal propulsion engine uses the technology developed in the NERVA/Rover program, updated to incorporate advances in the technology. Using composite fuel, higher power densities per fuel element, improved radiation resistant control components and the advancements in use of carbon-carbon materials; the ENABLER can provide a specific impulse of 925 seconds, an engine thrust to weight (excluding reactor shield) approaching five, an improved initial mass in low Earth orbit and a consequent reduction in launch costs and logistics problems. This paper describes the 75,000 lbs thrust ENABLER design which is a low cost, low risk approach to meeting tommorrow's space propulsion needs.

Simulations of a PSD Plastic Neutron Collar for Assaying Fresh Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hausladen, Paul; Newby, Jason; McElroy, Robert Dennis

The potential performance of a notional active coincidence collar for assaying uranium fuel based on segmented detectors constructed from the new PSD plastic fast organic scintillator with pulse shape discrimination capability was investigated in simulation. Like the International Atomic Energy Agency's present Uranium Neutron Collar for LEU (UNCL), the PSD plastic collar would also function by stimulating fission in the 235U content of the fuel with a moderated 241Am/Li neutron source and detecting instances of induced fission via neutron coincidence counting. In contrast to the moderated detectors of the UNCL, the fast time scale of detection in the scintillator eliminatesmore » statistical errors due to accidental coincidences that limit the performance of the UNCL. However, the potential to detect a single neutron multiple times historically has been one of the properties of organic scintillator detectors that has prevented their adoption for international safeguards applications. Consequently, as part of the analysis of simulated data, a method was developed by which true neutron-neutron coincidences can be distinguished from inter-detector scatter that takes advantage of the position and timing resolution of segmented detectors. Then, the performance of the notional simulated coincidence collar was evaluated for assaying a variety of fresh fuels, including some containing burnable poisons and partial defects. In these simulations, particular attention was paid to the analysis of fast mode measurements. In fast mode, a Cd liner is placed inside the collar to shield the fuel from the interrogating source and detector moderators, thereby eliminating the thermal neutron flux that is most sensitive to the presence of burnable poisons that are ubiquitous in modern nuclear fuels. The simulations indicate that the predicted precision of fast mode measurements is similar to what can be achieved by the present UNCL in thermal mode. For example, the statistical accuracy of a ten-minute measurement of fission coincidences collected in fast mode will be approximately 1% for most fuels of interest, yielding a ~1.4% error after subtraction of a five minute measurement of the spontaneous fissions from 238U in the fuel, a ~2% error in analyzed linear density after accounting for the slope of the calibration curve, and a ~2.9% total error after addition of an assumed systematic error of 2%.« less

Radiation Analysis for the Human Lunar Return Mission

NASA Technical Reports Server (NTRS)

Wilson, J. W.; Simonsen, L. C.; Shinn, J. L.; Kim, M.; Dubey, R. R.; Jordan, W.

1997-01-01

An analysis of the radiation hazards that are anticipated on an early Human Lunar Return (HLR) mission in support of NASA deep space exploration activities is presented. The HLR mission study emphasized a low cost lunar return to expand human capabilities in exploration, to answer fundamental science questions, and to seek opportunities for commercial development. As such, the radiation issues are cost related because the parasitic shield mass is expensive due to high launch costs. The present analysis examines the shield requirements and their impact on shield design.

An analytical and experimental evaluation of shadow shields and their support members

NASA Technical Reports Server (NTRS)

Stochl, R. J.; Boyle, R. J.

1972-01-01

Experimental tests were performed on a model shadow shield thermal protection system to examine the effect of certain configuration variables. The experimental results were used to verify the ability of an analytical program to predict the shadow shield performance including the shield-support interaction. In general, the analysis (assuming diffuse surfaces) agreed well with the experimental support temperature profiles. The agreement for the shield profiles was not as good. The results demonstrated: (1) shadow shields can be effective in reducing the heat transfer into cryogenic propellant tanks, and (2) the conductive heat transfer through supports can be reduced by selective surface coatings.

Shield fields: Concentrations of small volcanic edifices on Venus

NASA Technical Reports Server (NTRS)

Aubele, J. C.; Crumpler, L. S.

1992-01-01

Pre-Magellan analysis of the Venera 15/16 data indicated the existence of abundant small volcanic edifices, each less than or equal to 20 km diameter, interpreted to be predominantly shield volcanoes and occurring throughout the plains terrain, most common in equidimensional clusters. With the analysis of Magellan data, these clusters of greater than average concentration of small volcanic edifices have been called 'shield fields'. Although individual small shields can and do occur almost everywhere on the plains terrain of Venus, they most commonly occur in fields that are well-defined, predominantly equant, clusters of edifices. Major questions include why the edifices are concentrated in this way, how they relate to the source of the eruptive material, and what the possible relationship of shield fields to plains terrain is. There are three possible models for the origin of fields and small shields: (1) a field represents an 'island' of higher topography subsequently surrounded by later plains material; and (2) a field represents the area of magma reservoir.

Nuclear reactor system study for NASA/JPL

NASA Technical Reports Server (NTRS)

Palmer, R. G.; Lundberg, L. B.; Keddy, E. S.; Koenig, D. R.

1982-01-01

Reactor shielding, safety studies, and heat pipe development work are described. Monte Carlo calculations of gamma and neutron shield configurations show that substantial weight penalties are incurred if exposure at 25 m to neutrons and gammas must be limited to 10 to the 12th power nvt and 10 to the 6th power rad, instead of the 10 to the 13th power nvt and 10 to the 7th power rad values used earlier. For a 1.6 MW sub t reactor, the required shield weight increases from 400 to 815 kg. Water immersion critically calculations were extended to study the effect of water in fuel void spaces as well as in the core heat pipes. These show that the insertion into the core of eight blades of B4C with a mass totaling 2.5 kg will guarantee subcriticality. The design, fabrication procedure, and testing of a 4m long molybdenum/lithium heat pipe are described. It appears that an excess of oxygen in the wick prevented the attainment of expected performance capability.

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

NASA Technical Reports Server (NTRS)

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

2016-01-01

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

CVTR PROJECT. CAROLINAS VIRGINIA NUCLEAR POWER ASSOCIATES, INC. MONTHLY PROGRESS REPORT, MAY 1961

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1961-10-31

The capsule A-2 was removed from the WTR reflector hole at the end of the WTR Cycle 13, and was stored in the WTR canal. The in-pile loop has operated for eight months and the test thimble was irradiated a total of 108 days. Tensile tests were completed on the extruded and annealed Zircaloy-4 Phase-II pressure tubes. The tensile properties varied with location in the pressure tube. The lowest values were obtained in the top flange where the material was fully annealed for ten hours at 800 deg C. Increased properties were achieved from working the material during extrusion operations.more » A shielding ring is provided to prevent streaming through a void generated by the rotating shield volley supports. It was determined that an additional thickness of iron or steel is required to compensate for the loss of shielding from the removal of one foot of concrete at the bottom of the trench. Various portions of the U-tube and fuel assemblies were homogenized in various axial regions for computer studies. The studies indicated a decrease of 500 hours in core life from non-uniform axial burnup. Pressure tube specimens are being tested under the impulsive test burst program. A test specimen experienced a 51% increase in O.D. under 20 impact blows before it failed. Observations of the tested specimens indicated ductilities far in excess of those predicted from the material's behavior in uniaxial tension. Teste on a Zircaloy-stainless steel joint were concluded after an extensive program of testing under various pressure, temperature and bending moment conditions. No sign of leakage was noted throughout the program. Subsequent inspection of the joint showed cracks in the sleeve portion of the joint. Analysis of the test water indicated a chloride content of approx 88 ppm. A test fuel assembly was dismantled and converted to a four baffle design. Modifications were made to the prototype control-rod-drive system. The alignment between ths vertical and horizontal miter gears was improved by charging the mounting of horizontal shaft and bearings. Scram tests were resumed; these tests indicated that the dashpot was acting too soon. The dashpot is being modified. (auth)« less

78 FR 63375 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-24

... include: adding a new transfer cask (TC), the OS197L, for use with the 32PT and 61BT dry shielded.... 1004. Specifically, Transnuclear, Inc. requested changes to: (1) add a new TC, the OS197L, for use with... with NUREG-1745 requirements. Deleting the TC dose rates for all currently licensed payloads (TSs 1.2...

RER SPECTRA OBTAINED WITH A MULTICRYSTAL SPECTROMETER

DOE Office of Scientific and Technical Information (OSTI.GOV)

Austin, W.E.; Champion, W.R.

1959-11-01

Relative gamma spectra were obtained twenty feet from the Hadiation Effects Reactor. The measurements were made using a multicry-stal spectrometer. This design incorporates pair and anticompton spectrometers in combination. Two reactor configurations were used; with shield tanks empty- and water filled. The spectra were obtained before the fuel elements were run at high power. Consequently very little of the fission product spectrum is tntermined. (J.R.D.)

TREE Simulation Facilities, Second Edition, Revision 2

DTIC Science & Technology

1979-01-01

included radiation effects on propellants , ordnance, electronics and chemicals, vehicle shielding, neutron radiography , dosimetry, and health physics...Special Capabilities 2.11.10.1 Radiography Facility 2.11.10.2 Flexo-Rabbit System Support Capabilities 2.11.11.1 Staff 2.11.11.2 Electronics...5,400-MW pulsing operation (experimental dosimetry values for a typical core loading of 94 fuel elements). 2-156 2-46 ACPR radiography facility

Shielding of the Hip Prosthesis During Radiation Therapy for Heterotopic Ossification is Associated with Increased Failure of Prophylaxis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Balboni, Tracy A.; Gaccione, Peter; Gobezie, Reuben

2007-04-01

Purpose: Radiation therapy (RT) is frequently administered to prevent heterotopic ossification (HO) after total hip arthroplasty (THA). The purpose of this study was to determine if there is an increased risk of HO after RT prophylaxis with shielding of the THA components. Methods and Materials: This is a retrospective analysis of THA patients undergoing RT prophylaxis of HO at Brigham and Women's Hospital between June 1994 and February 2004. Univariate and multivariate logistic regressions were used to assess the relationships of all variables to failure of RT prophylaxis. Results: A total of 137 patients were identified and 84 were eligiblemore » for analysis (61%). The median RT dose was 750 cGy in one fraction, and the median follow-up was 24 months. Eight of 40 unshielded patients (20%) developed any progression of HO compared with 21 of 44 shielded patients (48%) (p = 0.009). Brooker Grade III-IV HO developed in 5% of unshielded and 18% of shielded patients (p 0.08). Multivariate analysis revealed shielding (p = 0.02) and THA for prosthesis infection (p = 0.03) to be significant predictors of RT failure, with a trend toward an increasing risk of HO progression with age (p = 0.07). There was no significant difference in the prosthesis failure rates between shielded and unshielded patients. Conclusions: A significantly increased risk of failure of RT prophylaxis for HO was noted in those receiving shielding of the hip prosthesis. Shielding did not appear to reduce the risk of prosthesis failure.« less

Nuclear fuel element

DOEpatents

Armijo, Joseph S.; Coffin, Jr., Louis F.

1983-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a composite cladding having a substrate and a metal barrier metallurgically bonded on the inside surface of the substrate so that the metal barrier forms a shield between the substrate and the nuclear fuel material held within the cladding. The metal barrier forms about 1 to about 30 percent of the thickness of the cladding and is comprised of a low neutron absorption metal of substantially pure zirconium. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the substrate from contact and reaction with such impurities and fission products. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy. Methods of manufacturing the composite cladding are also disclosed.

Radiation Shielding for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Caffrey, Jarvis A.

2016-01-01

Design and analysis of radiation shielding for nuclear thermal propulsion has continued at Marshall Space Flight Center. A set of optimization tools are in development, and strategies for shielding optimization will be discussed. Considerations for the concurrent design of internal and external shielding are likely required for a mass optimal shield design. The task of reducing radiation dose to crew from a nuclear engine is considered to be less challenging than the task of thermal mitigation for cryogenic propellant, especially considering the likely implementation of additional crew shielding for protection from solar particles and cosmic rays. Further consideration is thus made for the thermal effects of radiation absorption in cryogenic propellant. Materials challenges and possible methods of manufacturing are also discussed.

A high-performance magnetic shield with large length-to-diameter ratio.

PubMed

Dickerson, Susannah; Hogan, Jason M; Johnson, David M S; Kovachy, Tim; Sugarbaker, Alex; Chiow, Sheng-wey; Kasevich, Mark A

2012-06-01

We have demonstrated a 100-fold improvement in the magnetic field uniformity on the axis of a large aspect ratio, cylindrical, mumetal magnetic shield by reducing discontinuities in the material of the shield through the welding and re-annealing of a segmented shield. The three-layer shield reduces Earth's magnetic field along an 8 m region to 420 μG (rms) in the axial direction, and 460 and 730 μG (rms) in the two transverse directions. Each cylindrical shield is a continuous welded tube which has been annealed after manufacture and degaussed in the apparatus. We present both experiments and finite element analysis that show the importance of uniform shield material for large aspect ratio shields, favoring a welded design over a segmented design. In addition, we present finite element results demonstrating the smoothing of spatial variations in the applied magnetic field by cylindrical magnetic shields. Such homogenization is a potentially useful feature for precision atom interferometric measurements.

Noise shielding by a hot subsonic jet

NASA Technical Reports Server (NTRS)

Vijayaraghavan, A.; Parthasarathy, S. P.

1981-01-01

An analysis is conducted of the shielding of the noise emitted by a high speed round jet by a hot, subsonic, semicircular jet. A plane wave front in the primary jet is resolved into elementary plane waves which undergo multiple reflections at the jet boundaries of the primary and the shielding jets. The jet boundaries are idealized to be vortex sheets. The far field sound is evaluated asymptotically by a superposition of the waves that penetrate the shielding jet. The angular directivities are plotted for several values of jet temperature and velocity to examine the effectiveness of shielding by the semicircular jet layer.

Effect of metal shielding on a wireless power transfer system

NASA Astrophysics Data System (ADS)

Li, Jiacheng; Huang, Xueliang; Chen, Chen; Tan, Linlin; Wang, Wei; Guo, Jinpeng

2017-05-01

In this paper, the effect of non-ferromagnetic metal shielding (NFMS) material on the resonator of wireless power transfer (WPT) is studied by modeling, simulation and experimental analysis. And, the effect of NFMS material on the power transfer efficiency (PTE) of WPT systems is investigated by circuit model. Meanwhile, the effect of ferromagnetic metal shielding material on the PTE of WPT systems is analyzed through simulation. A double layer metal shield structure is designed. Experimental results demonstrate that by applying the novel double layer metal shielding method, the system PTE increases significantly while the electromagnetic field of WPT systems declines dramatically.

Evaluation of Shielding Performance for Newly Developed Composite Materials

NASA Astrophysics Data System (ADS)

Evans, Beren Richard

This work details an investigation into the contributing factors behind the success of newly developed composite neutron shield materials. Monte Carlo simulation methods were utilized to assess the neutron shielding capabilities and secondary radiation production characteristics of aluminum boron carbide, tungsten boron carbide, bismuth borosilicate glass, and Metathene within various neutron energy spectra. Shielding performance and secondary radiation data suggested that tungsten boron carbide was the most effective composite material. An analysis of the macroscopic cross-section contributions from constituent materials and interaction mechanisms was then performed in an attempt to determine the reasons for tungsten boron carbide's success over the other investigated materials. This analysis determined that there was a positive correlation between a non-elastic interaction contribution towards a material's total cross-section and shielding performance within the thermal and epi-thermal energy regimes. This finding was assumed to be a result of the boron-10 absorption reaction. The analysis also determined that within the faster energy regions, materials featuring higher non-elastic interaction contributions were comparable to those exhibiting primarily elastic scattering via low Z elements. This allowed for the conclusion that composite shield success within higher energy neutron spectra does not necessitate the use elastic scattering via low Z elements. These findings suggest that the inclusion of materials featuring high thermal absorption properties is more critical to composite neutron shield performance than the presence of constituent materials more inclined to maximize elastic scattering energy loss.

Double-layer rotor magnetic shield performance analysis in high temperature superconducting synchronous generators under short circuit fault conditions

NASA Astrophysics Data System (ADS)

Hekmati, Arsalan; Aliahmadi, Mehdi

2016-12-01

High temperature superconducting, HTS, synchronous machines benefit from a rotor magnetic shield in order to protect superconducting coils against asynchronous magnetic fields. This magnetic shield, however, suffers from exerted Lorentz forces generated in light of induced eddy currents during transient conditions, e.g. stator windings short-circuit fault. In addition, to the exerted electromagnetic forces, eddy current losses and the associated effects on the cryogenic system are the other consequences of shielding HTS coils. This study aims at investigating the Rotor Magnetic Shield, RMS, performance in HTS synchronous generators under stator winding short-circuit fault conditions. The induced eddy currents in different circumferential positions of the rotor magnetic shield along with associated Joule heating losses would be studied using 2-D time-stepping Finite Element Analysis, FEA. The investigation of Lorentz forces exerted on the magnetic shield during transient conditions has also been performed in this paper. The obtained results show that double line-to-ground fault is of the most importance among different types of short-circuit faults. It was revealed that when it comes to the design of the rotor magnetic shields, in addition to the eddy current distribution and the associated ohmic losses, two phase-to-ground fault should be taken into account since the produced electromagnetic forces in the time of fault conditions are more severe during double line-to-ground fault.

The ENABLER—based on proven NERVA technology

NASA Astrophysics Data System (ADS)

Livingston, Julie M.; Pierce, Bill L.

1991-01-01

The ENABLER reactor for use in a nuclear thermal propulsion engine uses the technology developed in the NERVA/Rover program, updated to incorporate advances in the technology. Using composite fuel, higher power densities per fuel element, improved radiation resistant control components and the advancements in use of carbon-carbon materials; the ENABLER can provide a specific impulse of 925 seconds, an engine thrust to weight (excluding reactor shield) approaching five, an improved initial Mass In Low Earth Orbit (IMLEO) and a consequent reduction in launch costs and logistics problems. This paper describes the 75,000 lbs thrust ENABLER design which is a low cost, low risk approach to meeting tomorrow's space propulsion needs.

AIR COOLED NEUTRONIC REACTOR

DOEpatents

Fermi, E.; Szilard, L.

1958-05-27

A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

TRIGA MARK-II source term

DOE Office of Scientific and Technical Information (OSTI.GOV)

Usang, M. D., E-mail: mark-dennis@nuclearmalaysia.gov.my; Hamzah, N. S., E-mail: mark-dennis@nuclearmalaysia.gov.my; Abi, M. J. B., E-mail: mark-dennis@nuclearmalaysia.gov.my

ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences ofmore » results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel.« less

ETR, TRA642. CONSOLE FLOOR. CAMERA IS ON WEST SIDE OF ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR, TRA-642. CONSOLE FLOOR. CAMERA IS ON WEST SIDE OF FLOOR AND FACES NORTH. OUTER WALL OF STORAGE CANAL IS AT RIGHT. SHIELDING IS THICKER AT LOWER LEVEL, WHERE SPENT FUEL ELEMENTS WILL COOL AFTER REMOVAL FROM REACTOR. INL NEGATIVE NO. 56-1401. Jack L. Anderson, Photographer, 5/1/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR CRITICAL FACILITY, TRA654. SCIENTISTS STAND AT EDGE OF TANK ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR CRITICAL FACILITY, TRA-654. SCIENTISTS STAND AT EDGE OF TANK AND LIFT REMOVABLE BRIDGE ABOVE THE REACTOR. CONTROL RODS AND FUEL RODS ARE BELOW ENOUGH WATER TO SHIELD WORKERS ABOVE. NOTE CRANE RAILS ALONG WALLS, PUMICE BLOCK WALLS. INL NEGATIVE NO. 57-3690. R.G. Larsen, Photographer, 7/29/1957 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

IET. Aerial view during construction, facing southwest. Control building (TAN620) ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

IET. Aerial view during construction, facing southwest. Control building (TAN-620) in center. Retaining wall in place on west side. Tank building (TAN-627) and fuel transfer pump building (TAN-625) north of control building. Shielded roadway not yet built. Foundation of stack at right edge of view. Date: November 24, 1954. INEEL negative no. 13198 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

Applications in the Nuclear Industry for Thermal Spray Amorphous Metal and Ceramic Coatings

NASA Astrophysics Data System (ADS)

Blink, J.; Farmer, J.; Choi, J.; Saw, C.

2009-06-01

Amorphous metal and ceramic thermal spray coatings have been developed with excellent corrosion resistance and neutron absorption. These coatings, with further development, could be cost-effective options to enhance the corrosion resistance of drip shields and waste packages, and limit nuclear criticality in canisters for the transportation, aging, and disposal of spent nuclear fuel. Iron-based amorphous metal formulations with chromium, molybdenum, and tungsten have shown the corrosion resistance believed to be necessary for such applications. Rare earth additions enable very low critical cooling rates to be achieved. The boron content of these materials and their stability at high neutron doses enable them to serve as high efficiency neutron absorbers for criticality control. Ceramic coatings may provide even greater corrosion resistance for waste package and drip shield applications, although the boron-containing amorphous metals are still favored for criticality control applications. These amorphous metal and ceramic materials have been produced as gas-atomized powders and applied as near full density, nonporous coatings with the high-velocity oxy-fuel process. This article summarizes the performance of these coatings as corrosion-resistant barriers and as neutron absorbers. This article also presents a simple cost model to quantify the economic benefits possible with these new materials.

Morphometry of terrestrial shield volcanoes

NASA Astrophysics Data System (ADS)

Grosse, Pablo; Kervyn, Matthieu

2018-03-01

Shield volcanoes are described as low-angle edifices built primarily by the accumulation of successive lava flows. This generic view of shield volcano morphology is based on a limited number of monogenetic shields from Iceland and Mexico, and a small set of large oceanic islands (Hawaii, Galápagos). Here, the morphometry of 158 monogenetic and polygenetic shield volcanoes is analyzed quantitatively from 90-meter resolution SRTM DEMs using the MORVOLC algorithm. An additional set of 24 lava-dominated 'shield-like' volcanoes, considered so far as stratovolcanoes, are documented for comparison. Results show that there is a large variation in shield size (volumes from 0.1 to > 1000 km3), profile shape (height/basal width (H/WB) ratios mostly from 0.01 to 0.1), flank slope gradients (average slopes mostly from 1° to 15°), elongation and summit truncation. Although there is no clear-cut morphometric difference between shield volcanoes and stratovolcanoes, an approximate threshold can be drawn at 12° average slope and 0.10 H/WB ratio. Principal component analysis of the obtained database enables to identify four key morphometric descriptors: size, steepness, plan shape and truncation. Hierarchical cluster analysis of these descriptors results in 12 end-member shield types, with intermediate cases defining a continuum of morphologies. The shield types can be linked in terms of growth stages and shape evolution, related to (1) magma composition and rheology, effusion rate and lava/pyroclast ratio, which will condition edifice steepness; (2) spatial distribution of vents, in turn related to the magmatic feeding system and the tectonic framework, which will control edifice plan shape; and (3) caldera formation, which will condition edifice truncation.

The research Of Multilayer Thermal Insulation With Mechanical Properties Based On Model Analysis Test

NASA Astrophysics Data System (ADS)

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.

Rapid Analysis of Mass Distribution of Radiation Shielding

NASA Technical Reports Server (NTRS)

Zapp, Edward

2007-01-01

Radiation Shielding Evaluation Toolset (RADSET) is a computer program that rapidly calculates the spatial distribution of mass of an arbitrary structure for use in ray-tracing analysis of the radiation-shielding properties of the structure. RADSET was written to be used in conjunction with unmodified commercial computer-aided design (CAD) software that provides access to data on the structure and generates selected three-dimensional-appearing views of the structure. RADSET obtains raw geometric, material, and mass data on the structure from the CAD software. From these data, RADSET calculates the distribution(s) of the masses of specific materials about any user-specified point(s). The results of these mass-distribution calculations are imported back into the CAD computing environment, wherein the radiation-shielding calculations are performed.

Radioactive waste material melter apparatus

DOEpatents

Newman, D.F.; Ross, W.A.

1990-04-24

An apparatus for preparing metallic radioactive waste material for storage is disclosed. The radioactive waste material is placed in a radiation shielded enclosure. The waste material is then melted with a plasma torch and cast into a plurality of successive horizontal layers in a mold to form a radioactive ingot in the shape of a spent nuclear fuel rod storage canister. The apparatus comprises a radiation shielded enclosure having an opening adapted for receiving a conventional transfer cask within which radioactive waste material is transferred to the apparatus. A plasma torch is mounted within the enclosure. A mold is also received within the enclosure for receiving the melted waste material and cooling it to form an ingot. The enclosure is preferably constructed in at least two parts to enable easy transport of the apparatus from one nuclear site to another. 8 figs.

Evaluation of a large capacity heat pump concept for active cooling of hypersonic aircraft structure

NASA Technical Reports Server (NTRS)

Pagel, L. L.; Herring, R. L.

1978-01-01

Results of engineering analyses assessing the conceptual feasibility of a large capacity heat pump for enhancing active cooling of hypersonic aircraft structure are presented. A unique heat pump arrangement which permits cooling the structure of a Mach 6 transport to aluminum temperatures without the aid of thermal shielding is described. The selected concept is compatible with the use of conventional refrigerants, with Freon R-11 selected as the preferred refrigerant. Condenser temperatures were limited to levels compatible with the use of conventional refrigerants by incorporating a unique multipass condenser design, which extracts mechanical energy from the hydrogen fuel, prior to each subsequent pass through the condenser. Results show that it is technically feasible to use a large capacity heat pump in lieu of external shielding. Additional analyses are required to optimally apply this concept.

POWER REACTOR

DOEpatents

Zinn, W.H.

1958-07-01

A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

Radioactive waste material melter apparatus

DOEpatents

Newman, Darrell F.; Ross, Wayne A.

1990-01-01

An apparatus for preparing metallic radioactive waste material for storage is disclosed. The radioactive waste material is placed in a radiation shielded enclosure. The waste material is then melted with a plasma torch and cast into a plurality of successive horizontal layers in a mold to form a radioactive ingot in the shape of a spent nuclear fuel rod storage canister. The apparatus comprises a radiation shielded enclosure having an opening adapted for receiving a conventional transfer cask within which radioactive waste material is transferred to the apparatus. A plasma torch is mounted within the enclosure. A mold is also received within the enclosure for receiving the melted waste material and cooling it to form an ingot. The enclosure is preferably constructed in at least two parts to enable easy transport of the apparatus from one nuclear site to another.

Active shielding of cylindrical saddle-shaped coils: application to wire-wound RF coils for very low field NMR and MRI.

PubMed

Bidinosti, C P; Kravchuk, I S; Hayden, M E

2005-11-01

We provide an exact expression for the magnetic field produced by cylindrical saddle-shaped coils and their ideal shield currents in the low-frequency limit. The stream function associated with the shield surface current is also determined. The results of the analysis are useful for the design of actively shielded radio-frequency (RF) coils. Examples pertinent to very low field nuclear magnetic resonance (NMR) and magnetic resonance imaging (MRI) are presented and discussed.

Development of Ultra-Fine Multigroup Cross Section Library of the AMPX/SCALE Code Packages

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jeon, Byoung Kyu; Sik Yang, Won; Kim, Kang Seog

The Consortium for Advanced Simulation of Light Water Reactors Virtual Environment for Reactor Applications (VERA) neutronic simulator MPACT is being developed by Oak Ridge National Laboratory and the University of Michigan for various reactor applications. The MPACT and simplified MPACT 51- and 252-group cross section libraries have been developed for the MPACT neutron transport calculations by using the AMPX and Standardized Computer Analyses for Licensing Evaluations (SCALE) code packages developed at Oak Ridge National Laboratory. It has been noted that the conventional AMPX/SCALE procedure has limited applications for fast-spectrum systems such as boiling water reactor (BWR) fuels with very highmore » void fractions and fast reactor fuels because of its poor accuracy in unresolved and fast energy regions. This lack of accuracy can introduce additional error sources to MPACT calculations, which is already limited by the Bondarenko approach for resolved resonance self-shielding calculation. To enhance the prediction accuracy of MPACT for fast-spectrum reactor analyses, the accuracy of the AMPX/SCALE code packages should be improved first. The purpose of this study is to identify the major problems of the AMPX/SCALE procedure in generating fast-spectrum cross sections and to devise ways to improve the accuracy. For this, various benchmark problems including a typical pressurized water reactor fuel, BWR fuels with various void fractions, and several fast reactor fuels were analyzed using the AMPX 252-group libraries. Isotopic reaction rates were determined by SCALE multigroup (MG) calculations and compared with continuous energy (CE) Monte Carlo calculation results. This reaction rate analysis revealed three main contributors to the observed differences in reactivity and reaction rates: (1) the limitation of the Bondarenko approach in coarse energy group structure, (2) the normalization issue of probability tables, and (3) neglect of the self-shielding effect of resonance-like cross sections at high energy range such as (n,p) cross section of Cl35. The first error source can be eliminated by an ultra-fine group (UFG) structure in which the broad scattering resonances of intermediate-weight nuclides can be represented accurately by a piecewise constant function. A UFG AMPX library was generated with modified probability tables and tested against various benchmark problems. The reactivity and reaction rates determined with the new UFG AMPX library agreed very well with respect to Monte Carlo Neutral Particle (MCNP) results. To enhance the lattice calculation accuracy without significantly increasing the computational time, performing the UFG lattice calculation in two steps was proposed. In the first step, a UFG slowing-down calculation is performed for the corresponding homogenized composition, and UFG cross sections are collapsed into an intermediate group structure. In the second step, the lattice calculation is performed for the intermediate group level using the condensed group cross sections. A preliminary test showed that the condensed library reproduces the results obtained with the UFG cross section library. This result suggests that the proposed two-step lattice calculation approach is a promising option to enhance the applicability of the AMPX/SCALE system to fast system analysis.« less

Open Rotor Noise Shielding by Blended-Wing-Body Aircraft

NASA Technical Reports Server (NTRS)

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.

Instrumentation concepts and requirements for a space vacuum research facility. [molecular shield for spaceborne experiments

NASA Technical Reports Server (NTRS)

Norton, H. N.

1979-01-01

An earth-orbiting molecular shield that offers a unique opportunity for conducting physics, chemistry, and material processing experiments under a combination of environmental conditions that are not available in terrestrial laboratories is equipped with apparatus for forming a molecular beam from the freestream. Experiments are carried out using a moderate energy, high flux density, high purity atomic oxygen beam in the very low density environment within the molecular shield. As a minimum, the following instruments are required for the molecular shield: (1) a mass spectrometer; (2) a multifunction material analysis instrumentation system; and (3) optical spectrometry equipment. The design is given of a furlable molecular shield that allows deployment and retrieval of the system (including instrumentation and experiments) to be performed without contamination. Interfaces between the molecular shield system and the associated spacecraft are given. An in-flight deployment sequence is discussed that minimizes the spacecraft-induced contamination in the vicinity of the shield. Design approaches toward a precursor molecular shield system are shown.

A qualitative analysis of power take-off driveline shields: barriers and motivators to shield use for New York State farmers.

PubMed

Weil, R; Mellors, P; Fiske, T; Sorensen, J A

2014-01-01

Machinery entanglements are one of the top three causes of death in farming. Education on the risks of unshielded power take-off (PTO) equipment does not appear to significantly alter farmers' willingness to replace missing or broken shielding. Different assessments conducted in various regions of the U.S. indicate that as many as one-third to one-half of PTOs are inadequately shielded. Qualitative research was conducted with New York farmers to identify the factors that influence the decision to replace damaged or missing PTO driveline shields. Interview topics included: knowledge of entanglement risks, decisions regarding safety in general, decisions relating to PTO driveline shielding specifically, and the barriers and motivators to replacing missing or broken PTO driveline shields. Interviews with 38 farmers revealed the following themes: (1) farmers are fully aware of PTO entanglement risk, (2) insufficient time and money are primary barriers to purchasing or replacing damaged or missing PTO driveline shields, (3) PTO driveline shield designs are problematic and have led to negative experiences with shielding, and (4) risk acceptance and alternate work strategies are preferred alternatives to replacing shields. Our findings indicate that more innovative approaches will be required to make PTO driveline shield use a viable and attractive choice for farmers. New shield designs that address the practical barriers farmers face, as well as the provision of logistical and financial assistance for shield replacement, may alter the decision environment sufficiently to make replacing PTO driveline shielding a more attractive option for farmers.

Magnetic shielding structure optimization design for wireless power transmission coil

NASA Astrophysics Data System (ADS)

Dai, Zhongyu; Wang, Junhua; Long, Mengjiao; Huang, Hong; Sun, Mingui

2017-09-01

In order to improve the performance of the wireless power transmission (WPT) system, a novel design scheme with magnetic shielding structure on the WPT coil is presented in this paper. This new type of shielding structure has great advantages on magnetic flux leakage reduction and magnetic field concentration. On the basis of theoretical calculation of coil magnetic flux linkage and characteristic analysis as well as practical application feasibility consideration, a complete magnetic shielding structure was designed and the whole design procedure was represented in detail. The simulation results show that the coil with the designed shielding structure has the maximum energy transmission efficiency. Compared with the traditional shielding structure, the weight of the new design is significantly decreased by about 41%. Finally, according to the designed shielding structure, the corresponding experiment platform is built to verify the correctness and superiority of the proposed scheme.

Parametric analysis: SOC meteoroid and debris protection

NASA Technical Reports Server (NTRS)

Kowalski, R.

1985-01-01

The meteoroid and man made space debris environments of an Earth orbital manned space operations center are discussed. Protective shielding thickness and design configurations for providing given levels of no penetration probability were also calculated. Meteoroid/debris protection consists of a radiator/shield thickness, which is actually an outer skin, separated from the pressure wall, thickness by a distance. An ideal shield thickness, will, upon impact with a particle, cause both the particle and shield to vaporize, allowing a minimum amount of debris to impact the pressure wall itself. A shield which is too thick will crater on the outside, and release small particles of shield from the inside causing damage to the pressure wall. Inversely, if the shield is too thin, it will afford no protection, and the backup must provide all necessary protection. It was concluded that a double wall concept is most effective.

Electromagnetic interference and shielding: An introduction (revised version of 1991-23)

NASA Astrophysics Data System (ADS)

Dehoop, A. T.; Quak, D.

The basic equations of the electromagnetic field are summarized as far as they are needed in the theory of electromagnetic interference and shielding. Through the analysis of the planar electric current emitter, the propagation coefficient, attenuation coefficient, phase coefficient, wave-speed, wavelength, wave impedance, wave admittance, and power flow density of a wave are introduced. Next, the shielding effectiveness of a shielding plate and the shielding effectiveness of a shielding parallel-plate box are determined. In the latter, particular attention is given to the occurrence of internal resonance effects, which may degrade the shielding effectiveness. Further, a survey of some fundamental properties of a system of low frequency, multiconductor transmission lines is given. For a three conductor system with a plane of symmetry, the decomposition into the common mode and the differential mode of operation is discussed. Finally, expressions for the voltages and electric currents induced by external sources along a single transmission line are derived.

NASA Tech Briefs, April 2009

NASA Technical Reports Server (NTRS)

2009-01-01

Topics covered include: Direct-Solve Image-Based Wavefront Sensing; Use of UV Sources for Detection and Identification of Explosives; Using Fluorescent Viruses for Detecting Bacteria in Water; Gradiometer Using Middle Loops as Sensing Elements in a Low-Field SQUID MRI System; Volcano Monitor: Autonomous Triggering of In-Situ Sensors; Wireless Fluid-Level Sensors for Harsh Environments; Interference-Detection Module in a Digital Radar Receiver; Modal Vibration Analysis of Large Castings; Structural/Radiation-Shielding Epoxies; Integrated Multilayer Insulation; Apparatus for Screening Multiple Oxygen-Reduction Catalysts; Determining Aliasing in Isolated Signal Conditioning Modules; Composite Bipolar Plate for Unitized Fuel Cell/Electrolyzer Systems; Spectrum Analyzers Incorporating Tunable WGM Resonators; Quantum-Well Thermophotovoltaic Cells; Bounded-Angle Iterative Decoding of LDPC Codes; Conversion from Tree to Graph Representation of Requirements; Parallel Hybrid Vehicle Optimal Storage System; and Anaerobic Digestion in a Flooded Densified Leachbed.

Remote fabrication and irradiation test of recycled nuclear fuel prepared by the oxidation and reduction of spent oxide fuel

NASA Astrophysics Data System (ADS)

Jin Ryu, Ho; Chan Song, Kee; Il Park, Geun; Won Lee, Jung; Seung Yang, Myung

2005-02-01

A direct dry recycling process was developed in order to reuse spent pressurized light water reactor (LWR) nuclear fuel in CANDU reactors without the separation of sensitive nuclear materials such as plutonium. The benefits of the dry recycling process are the saving of uranium resources and the reduction of spent fuel accumulation as well as a higher proliferation resistance. In the process of direct dry recycling, fuel pellets separated from spent LWR fuel rods are oxidized from UO2 to U3O8 at 500 °C in an air atmosphere and reduced into UO2 at 700 °C in a hydrogen atmosphere, which is called OREOX (oxidation and reduction of oxide fuel). The pellets are pulverized during the oxidation and reduction processes due to the phase transformation between cubic UO2 and orthorhombic U3O8. Using the oxide powder prepared from the OREOX process, the compaction and sintering processes are performed in a remote manner in a shielded hot cell due to the high radioactivity of the spent fuel. Most of the fission gas and volatile fission products are removed during the OREOX and sintering processes. The mini-elements fabricated by the direct dry recycling process are irradiated in the HANARO research reactor for the performance evaluation of the recycled fuel pellets. Post-irradiation examination of the irradiated fuel showed that microstructural evolution and fission gas release behavior of the dry-recycled fuel were similar to high burnup UO2 fuel.

On the morphometry of terrestrial shield volcanoes

NASA Astrophysics Data System (ADS)

Grosse, Pablo; Kervyn, Matthieu

2016-04-01

Shield volcanoes are described as low angle edifices that have convex up topographic profiles and are built primarily by the accumulation of lava flows. This generic view of shields' morphology is based on a limited number of monogenetic shields from Iceland and Mexico, and a small set of large oceanic islands (Hawaii, Galapagos). Here, the morphometry of over 150 monogenetic and polygenetic shield volcanoes, identified inthe Global Volcanism Network database, are analysed quantitatively from 90-meter resolution DEMs using the MORVOLC algorithm. An additional set of 20 volcanoes identified as stratovolcanoes but having low slopes and being dominantly built up by accumulation of lava flows are documented for comparison. Results show that there is a large variation in shield size (volumes range from 0.1 to >1000 km3), profile shape (height/basal width ratios range from 0.01 to 0.1), flank slope gradients, elongation and summit truncation. Correlation and principal component analysis of the obtained quantitative database enables to identify 4 key morphometric descriptors: size, steepness, plan shape and truncation. Using these descriptors through clustering analysis, a new classification scheme is proposed. It highlights the control of the magma feeding system - either central, along a linear structure, or spatially diffuse - on the resulting shield volcano morphology. Genetic relationships and evolutionary trends between contrasted morphological end-members can be highlighted within this new scheme. Additional findings are that the Galapagos-type morphology with a central deep caldera and steep upper flanks are characteristic of other shields. A series of large oceanic shields have slopes systematically much steeper than the low gradients (<4-8°) generally attributed to large Hawaiian-type shields. Finally, the continuum of morphologies from flat shields to steeper complex volcanic constructs considered as stratovolcanoes calls for a revision of this oversimplified distinction, taking into account the lava/pyroclasts ratio and the spatial distribution of eruptive vents.

Applicability of 100kWe-class of space reactor power systems to NASA manned space station missions

NASA Technical Reports Server (NTRS)

Silverman, S. W.; Willenberg, H. J.; Robertson, C.

1985-01-01

An assessment is made of a manned space station operating with sufficiently high power demands to require a multihundred kilowatt range electrical power system. The nuclear reactor is a competitor for supplying this power level. Load levels were selected at 150kWe and 300kWe. Interactions among the reactor electrical power system, the manned space station, the space transportation system, and the mission were evaluated. The reactor shield and the conversion equipment were assumed to be in different positions with respect to the station; on board, tethered, and on a free flyer platform. Mission analyses showed that the free flyer concept resulted in unacceptable costs and technical problems. The tethered reactor providing power to an electrolyzer for regenerative fuel cells on the space station, results in a minimum weight shield and can be designed to release the reactor power section so that it moves to a high altitude orbit where the decay period is at least 300 years. Placing the reactor on the station, on a structural boom is an attractive design, but heavier than the long tethered reactor design because of the shield weight for manned activity near the reactor.

Dynamic Open-Rotor Composite Shield Impact Test Report

NASA Technical Reports Server (NTRS)

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

2015-01-01

The Federal Aviation Administration (FAA) is working with the European Aviation Safety Agency to determine the certification base for proposed new engines that would not have a containment structure on large commercial aircraft. Equivalent safety to the current fleet is desired by the regulators, which means that loss of a single fan blade will not cause hazard to the aircraft. NASA Glenn and Naval Air Warfare Center (NAWC) China Lake collaborated with the FAA Aircraft Catastrophic Failure Prevention Program to design and test a shield that would protect the aircraft passengers and critical systems from a released blade that could impact the fuselage. This report documents the live-fire test from a full-scale rig at NAWC China Lake. NASA provided manpower and photogrammetry expertise to document the impact and damage to the shields. The test was successful: the blade was stopped from penetrating the shield, which validates the design analysis method and the parameters used in the analysis. Additional work is required to implement the shielding into the aircraft.

SIRTF thermal design modifications to increase lifetime

NASA Astrophysics Data System (ADS)

Petrick, S. W.

1993-01-01

An effort was made to increase the predicted lifetime of the SIRTF dewar by lowering the exterior shell temperature, increasing the radiated energy from the vapor cooled shields and reconfiguring the vapor cooled shields. The lifetime increases can be used to increase the scientific return from the mission and as a trade-off against mass and cost. This paper describes the configurations studied, the steady state thermal model used, the analytical methods and the results of the analysis. Much of the heat input to the outside dewar shell is radiative heat transfer from the solar panel. To lower the shell temperature, radiative cooled shields were placed between the solar panel and the dewar shell and between the bus and the dewar shell. Analysis showed that placing a radiator on the outer vapor cooled shield had a significant effect on lifetime. Lengthening the distance between the outer shell and the point where the vapor cooled shields are attached to the support straps also improved lifetime.

Radiation protection design considerations for man in geosynchronous orbits

NASA Technical Reports Server (NTRS)

Rossi, M. L.; Stauber, M. C.

1977-01-01

A description is presented of preliminary studies which have been carried out to identify design requirements and mission constraints imposed by the geosynchronous radiation environment. The radiation species of dominant impact are the trapped electrons and solar flare particles. The criterion used in the conducted shielding design analysis has been to limit the skin dose to 100 rems for 3 months. The analysis included the optimization of an electron/bremsstrahlung shield for residence within the vehicle, the minimization of the dose received in extravehicular activity, and the calculation of special shield requirements for solar flares. An investigation was conducted of the potential benefits accruing from a three-layered composite shield with part of the aluminum layer replaced with a lower atomic number material. The materials considered were polyethylene, carbon, beryllium, and lithium hydride.

Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks

DOE PAGES

Banerjee, Kaushik; Robb, Kevin R.; Radulescu, Georgeta; ...

2016-06-15

We completed a novel assessment to determine the unquantified and uncredited safety margins (i.e., the difference between the licensing basis and as-loaded calculations) available in as-loaded spent nuclear fuel (SNF) casks. This assessment was performed as part of a broader effort to assess issues and uncertainties related to the continued safety of casks during extended storage and transportability following extended storage periods. Detailed analyses crediting the actual as-loaded cask inventory were performed for each of the casks at three decommissioned pressurized water reactor (PWR) sites to determine their characteristics relative to regulatory safety criteria for criticality, thermal, and shielding performance.more » These detailed analyses were performed in an automated fashion by employing a comprehensive and integrated data and analysis tool—Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). Calculated uncredited criticality margins from 0.07 to almost 0.30 Δk eff were observed; calculated decay heat margins ranged from 4 to almost 22 kW (as of 2014); and significant uncredited transportation dose rate margins were also observed. The results demonstrate that, at least for the casks analyzed here, significant uncredited safety margins are available that could potentially be used to compensate for SNF assembly and canister structural performance related uncertainties associated with long-term storage and subsequent transportation. The results also suggest that these inherent margins associated with how casks are loaded could support future changes in cask licensing to directly or indirectly credit the margins. Work continues to quantify the uncredited safety margins in the SNF casks loaded at other nuclear reactor sites.« less

Resources of Near-Earth Space: Abstracts

NASA Technical Reports Server (NTRS)

1991-01-01

The objectives are by theory, experiment, and bench-level testing of small systems, to develop scientifically-sound engineering processes and facility specifications for producing propellants and fuels, construction and shielding materials, and life support substances from the lithospheres and atmospheres of lunar, planetary, and asteroidal bodies. Current emphasis is on the production of oxygen, other usefull gases, metallic, ceramic/composite, and related byproducts from lunar regolith, carbonaceous chrondritic asteroids, and the carbon dioxide rich Martian atmosphere.

Apparatus and method for controlling plating uniformity

DOEpatents

Hachman Jr., John T.; Kelly, James J.; West, Alan C.

2004-10-12

The use of an insulating shield for improving the current distribution in an electrochemical plating bath is disclosed. Numerical analysis is used to evaluate the influence of shield shape and position on plating uniformity. Simulation results are compared to experimental data for nickel deposition from a nickel--sulfamate bath. The shield is shown to improve the average current density at a plating surface.

Protecting personal data in epidemiological research: DataSHIELD and UK law.

PubMed

Wallace, Susan E; Gaye, Amadou; Shoush, Osama; Burton, Paul R

2014-01-01

Data from individual collections, such as biobanks and cohort studies, are now being shared in order to create combined datasets which can be queried to ask complex scientific questions. But this sharing must be done with due regard for data protection principles. DataSHIELD is a new technology that queries nonaggregated, individual-level data in situ but returns query data in an anonymous format. This raises questions of the ability of DataSHIELD to adequately protect participant confidentiality. An ethico-legal analysis was conducted that examined each step of the DataSHIELD process from the perspective of UK case law, regulations, and guidance. DataSHIELD reaches agreed UK standards of protection for the sharing of biomedical data. All direct processing of personal data is conducted within the protected environment of the contributing study; participating studies have scientific, ethics, and data access approvals in place prior to the analysis; studies are clear that their consents conform with this use of data, and participants are informed that anonymisation for further disclosure will take place. DataSHIELD can provide a flexible means of interrogating data while protecting the participants' confidentiality in accordance with applicable legislation and guidance. © 2014 S. Karger AG, Basel.

Conceptual Design and Structural Analysis of an Open Rotor Hybrid Wing Body Aircraft

NASA Technical Reports Server (NTRS)

Gern, Frank H.

2013-01-01

Through a recent NASA contract, Boeing Research and Technology in Huntington Beach, CA developed and optimized a conceptual design of an open rotor hybrid wing body aircraft (HWB). Open rotor engines offer a significant potential for fuel burn savings over turbofan engines, while the HWB configuration potentially allows to offset noise penalties through possible engine shielding. Researchers at NASA Langley converted the Boeing design to a FLOPS model which will be used to develop take-off and landing trajectories for community noise analyses. The FLOPS model was calibrated using Boeing data and shows good agreement with the original Boeing design. To complement Boeing s detailed aerodynamics and propulsion airframe integration work, a newly developed and validated conceptual structural analysis and optimization tool was used for a conceptual loads analysis and structural weights estimate. Structural optimization and weight calculation are based on a Nastran finite element model of the primary HWB structure, featuring centerbody, mid section, outboard wing, and aft body. Results for flight loads, deformations, wing weight, and centerbody weight are presented and compared to Boeing and FLOPS analyses.

Galileo Probe forebody thermal protection

NASA Technical Reports Server (NTRS)

Green, M. J.; Davy, W. C.

1981-01-01

Material response solutions for the forebody heat shield on the candidate 310-kg Galileo Probe are presented. A charring material ablation analysis predicts thermochemical surface recession, insulation thickness, and total required heat shield mass. Benchmark shock layer solutions provide the imposed entry heating environments on the ablating surface. Heat shield sizing results are given for a nominal entry into modeled nominal and cool-heavy Jovian atmospheres, and for two heat-shield property models. The nominally designed heat shield requires a mass of at least 126 kg and would require an additional 13 kg to survive entry into the less probable cool-heavy atmosphere. The material-property model with a 30% surface reflectance reduces these mass requirements by as much as 16%.

Environment Impact Analysis of Shield Passing Alongside Bridge Pile Platform Using Three Dimensional Numerical Simulation

NASA Astrophysics Data System (ADS)

Shang, Yanliang; Shi, Wenjun; Han, Tongyin; Qin, Zhichao; Du, Shouji

2017-10-01

The shield method has many advantages in the construction of urban subway, and has become the preferred method for the construction of urban subway tunnel. Taking Shijiazhuang metro line 3 (administrative center station - garden park station interval) Passing alongside bridge as the engineering background, double shield crossing the bridge pile foundation model was set up. The deformation and internal force of the pile foundation during the construction of the shield were analyzed. Pile stress caused by shield construction increases, but the maximum stress is less than the design strength; the maximum surface settlement caused by the construction of 10.2 mm, the results meet the requirements of construction.

Depletion optimization of lumped burnable poisons in pressurized water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kodah, Z.H.

1982-01-01

Techniques were developed to construct a set of basic poison depletion curves which deplete in a monotonical manner. These curves were combined to match a required optimized depletion profile by utilizing either linear or non-linear programming methods. Three computer codes, LEOPARD, XSDRN, and EXTERMINATOR-2 were used in the analyses. A depletion routine was developed and incorporated into the XSDRN code to allow the depletion of fuel, fission products, and burnable poisons. The Three Mile Island Unit-1 reactor core was used in this work as a typical PWR core. Two fundamental burnable poison rod designs were studied. They are a solidmore » cylindrical poison rod and an annular cylindrical poison rod with water filling the central region.These two designs have either a uniform mixture of burnable poisons or lumped spheroids of burnable poisons in the poison region. Boron and gadolinium are the two burnable poisons which were investigated in this project. Thermal self-shielding factor calculations for solid and annular poison rods were conducted. Also expressions for overall thermal self-shielding factors for one or more than one size group of poison spheroids inside solid and annular poison rods were derived and studied. Poison spheroids deplete at a slower rate than the poison mixture because each spheroid exhibits some self-shielding effects of its own. The larger the spheroid, the higher the self-shielding effects due to the increase in poison concentration.« less

Interior of the Plum Brook Reactor Facility

NASA Image and Video Library

1961-02-21

A view inside the 55-foot high containment vessel of the National Aeronautics and Space Administration (NASA) Plum Brook Reactor Facility in Sandusky, Ohio. The 60-megawatt test reactor went critical for the first time in 1961 and began its full-power research operations in 1963. From 1961 to 1973, this reactor performed some of the nation’s most advanced nuclear research. The reactor was designed to determine the behavior of metals and other materials after long durations of irradiation. The materials would be used to construct a nuclear-powered rocket. The reactor core, where the chain reaction occurred, sat at the bottom of the tubular pressure vessel, seen here at the center of the shielding pool. The core contained fuel rods with uranium isotopes. A cooling system was needed to reduce the heat levels during the reaction. A neutron-impervious reflector was also employed to send many of the neutrons back to the core. The Plum Brook Reactor Facility was constructed from high-density concrete and steel to prevent the excess neutrons from escaping the facility, but the water in the pool shielded most of the radiation. The water, found in three of the four quadrants served as a reflector, moderator, and coolant. In this photograph, the three 20-ton protective shrapnel shields and hatch have been removed from the top of the pressure tank revealing the reactor tank. An overhead crane could be manipulated to reach any section of this room. It was used to remove the shrapnel shields and transfer equipment.

Audio Script for Information Center Transportation Display

DOE Office of Scientific and Technical Information (OSTI.GOV)

NA

2003-05-26

Can waste be transported safely to Yucca Mountain? Both the Department of Energy and the Nuclear Regulatory Commission have found that spent nuclear fuel can be shipped safely and securely. In fact, over the last 30 years there have been more than 2,700 shipments of spent nuclear fuel traveling more than 1.7 million miles, and there has never been a release of radioactive material harmful to the public or the environment--not one. Spent nuclear fuel is a solid material--it cannot leak, burn, or explode. The shipping containers, called casks, are the most robust in the transportation industry and must bemore » certified by the Nuclear Regulatory Commission. They are designed to protect public health and safety under normal and severe accident conditions. Typically, every ton of shipped spent fuel is contained within approximately 4 tons of protective shielding and structural materials. How many shipments would be made to Yucca Mountain? DOE would use mainly trains and some legal-weight trucks to move spent nuclear fuel and high-level radioactive waste to Yucca Mountain. Once the repository opens, DOE estimates and average of 130 rail shipments and 45 truck shipments per year for 24 years.« less

Thulium-170 heat source

DOEpatents

Walter, Carl E.; Van Konynenburg, Richard; VanSant, James H.

1992-01-01

An isotopic heat source is formed using stacks of thin individual layers of a refractory isotopic fuel, preferably thulium oxide, alternating with layers of a low atomic weight diluent, preferably graphite. The graphite serves several functions: to act as a moderator during neutron irradiation, to minimize bremsstrahlung radiation, and to facilitate heat transfer. The fuel stacks are inserted into a heat block, which is encased in a sealed, insulated and shielded structural container. Heat pipes are inserted in the heat block and contain a working fluid. The heat pipe working fluid transfers heat from the heat block to a heat exchanger for power conversion. Single phase gas pressure controls the flow of the working fluid for maximum heat exchange and to provide passive cooling.

Galactic and Solar Cosmic Ray Shielding in Deep Space

NASA Technical Reports Server (NTRS)

Wilson, John W.; Cucinotta, Francis A.; Tai, H.; Simonsen, Lisa C.; Shinn, Judy L.; Thibeault, Shelia; Kim, M. Y.

1997-01-01

An analysis of the radiation hazards in support of NASA deep space exploration activities is presented. The emphasis is on materials required for radiation protection shielding. Aluminum has been found to be a poor shield material when dose equivalent is used with exposure limits for low Earth orbit (LEO) as a guide for shield requirements. Because the radiation issues are cost related-the parasitic shield mass has high launch costs, the use of aluminum as a basic construction material is clearly not cost-effective and alternate materials need to be developed. In this context, polyethylene is examined as a potentially useful material and demonstrates important advantages as an alternative to aluminum construction. Although polyethylene is useful as a shield material, it may not meet other design criteria (strength, stability, thermal); other polymer materials must be examined.

Parasitic heat loss reduction in AMTEC cells by heat shield optimization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Borkowski, C.A.; Svedberg, R.C.; Hendricks, T.J.

1997-12-31

Alkali metal thermal to electric conversion (AMTEC) cell performance can be increased by the proper design of thermal radiative shielding internal to the AMTEC cell. These heat shields essentially lower the radiative heat transfer between the heat input zone of the cell and the heat rejection zone of the cell. In addition to lowering the radiative heat transfer between the heat input and heat rejection surfaces of the cell, the shields raise the AMTEC cell performance by increasing the temperature of the beta alumina solid electrolyte (BASE). This increase in temperature of the BASE tube allows the evaporator temperature tomore » be increased without sodium condensing within the BASE tubes. Experimental testing and theoretical analysis have been performed to compare the relative merits of two candidate heat shield packages: (1) chevron, and (2) cylindrical heat shields. These two heat shield packages were compared to each other and a baseline cell which had no heat shields installed. For the two heat shield packages, the reduction in total heat transfer is between 17--27% for the heat input surface temperature varying from 700 C, 750 C, and 800 C with the heat rejection surface temperature kept at 300 C.« less

Analytical and experimental studies of flow-induced vibration of SSME components

NASA Technical Reports Server (NTRS)

Chen, S. S.; Jendrzejczyk, J. A.; Wambsganss, M. W.

1987-01-01

Components of the Space Shuttle Main Engines (SSMEs) are subjected to a severe environment that includes high-temperature, high-velocity flows. Such flows represent a source of energy that can induce and sustain large-amplitude vibratory stresses and/or result in fluidelastic instabilities. Three components are already known to have experienced failures in evaluation tests as a result of flow-induced structural motion. These components include the liquid-oxygen (LOX) posts, the fuel turbine bellows shield, and the internal inlet tee splitter vane. Researchers considered the dynamic behavior of each of these components with varying degrees of effort: (1) a theoretical and experimental study of LOX post vibration excited by a fluid flow; (2) an assessment of the internal inlet tee splitter vane vibration (referred to as the 4000-Hz vibration problem); and (3) a preliminary consideration of the bellows shield problem. Efforts to resolve flow-induced vibration problems associated with the SSMEs are summarized.

Optimization of custom cementless stem using finite element analysis and elastic modulus distribution for reducing stress-shielding effect.

PubMed

Saravana Kumar, Gurunathan; George, Subin Philip

2017-02-01

This work proposes a methodology involving stiffness optimization for subject-specific cementless hip implant design based on finite element analysis for reducing stress-shielding effect. To assess the change in the stress-strain state of the femur and the resulting stress-shielding effect due to insertion of the implant, a finite element analysis of the resected femur with implant assembly is carried out for a clinically relevant loading condition. Selecting the von Mises stress as the criterion for discriminating regions for elastic modulus difference, a stiffness minimization method was employed by varying the elastic modulus distribution in custom implant stem. The stiffness minimization problem is formulated as material distribution problem without explicitly penalizing partial volume elements. This formulation enables designs that could be fabricated using additive manufacturing to make porous implant with varying levels of porosity. Stress-shielding effect, measured as difference between the von Mises stress in the intact and implanted femur, decreased as the elastic modulus distribution is optimized.

Meta-Analysis of the Effect of Chest Shielding on Preventing Patent Ductus Arteriosus in Premature Infants.

PubMed

Mannan, Javed; Amin, Sanjiv B

2017-03-01

Objective  This study aims to perform a meta-analysis of randomized studies to evaluate if chest shielding during phototherapy is associated with decreased incidence of patent ductus arteriosus (PDA) in premature infants. Design/Methods  We used published guidelines for the meta-analysis of clinical trials. The search strategy included electronic searches of CINAHL, CENTRAL Cochrane Library, MEDLINE, PubMed, and abstracts presented at the Pediatric Academic Societies. Inclusion criteria were randomized controlled trials (RCTs), quasi-RCTs or cluster RCTs published in English and involving chest shielding during phototherapy in premature infants with PDA as an outcome. Exclusion criteria involved case reports, case series, and multiple publications from the same author. Heterogeneity testing using Q statistics was performed to evaluate the variance between studies. Results  Two RCTs met study criteria. There was heterogeneity (I 2 : 55.4%) between the two trials. Meta-analysis of RCTs using the random effect model demonstrated that chest shielding during phototherapy was associated with decreased incidence of PDA (odds ratio: 0.47, 95% confidence interval: 0.23-0.96). There was no publication bias on Eggers test. Heterogeneity was seen in gestational age, gender, prophylactic use of postnatal indomethacin, duration of phototherapy, and assessment of PDA. Conclusion  Chest shielding during phototherapy may be associated with decreased incidence of PDA among premature infants. Thieme Medical Publishers 333 Seventh Avenue, New York, NY 10001, USA.

Z-Pinch Magneto-Inertial Fusion Propulsion Engine Design Concept

NASA Technical Reports Server (NTRS)

Miernik, Janie H.; Statham, Geoffrey; Adams, Robert B.; Polsgrove, Tara; Fincher, Sharon; Fabisinski, Leo; Maples, C. Dauphne; Percy, Thomas K.; Cortez, Ross J.; Cassibry, Jason

2011-01-01

Fusion-based nuclear propulsion has the potential to enable fast interplanetary transportation. Due to the great distances between the planets of our solar system and the harmful radiation environment of interplanetary space, high specific impulse (Isp) propulsion in vehicles with high payload mass fractions must be developed to provide practical and safe vehicles for human spaceflight missions. Magneto-Inertial Fusion (MIF) is an approach which has been shown to potentially lead to a low cost, small fusion reactor/engine assembly (1). The Z-Pinch dense plasma focus method is an MIF concept in which a column of gas is compressed to thermonuclear conditions by an estimated axial current of approximately 100 MA. Recent advancements in experiments and the theoretical understanding of this concept suggest favorable scaling of fusion power output yield as I(sup 4) (2). The magnetic field resulting from the large current compresses the plasma to fusion conditions, and this is repeated over short timescales (10(exp -6) sec). This plasma formation is widely used in the field of Nuclear Weapons Effects (NWE) testing in the defense industry, as well as in fusion energy research. There is a wealth of literature characterizing Z-Pinch physics and existing models (3-5). In order to be useful in engineering analysis, a simplified Z-Pinch fusion thermodynamic model was developed to determine the quantity of plasma, plasma temperature, rate of expansion, energy production, etc. to calculate the parameters that characterize a propulsion system. The amount of nuclear fuel per pulse, mixture ratio of the D-T and nozzle liner propellant, and assumptions about the efficiency of the engine, enabled the sizing of the propulsion system and resulted in an estimate of the thrust and Isp of a Z-Pinch fusion propulsion system for the concept vehicle. MIF requires a magnetic nozzle to contain and direct the nuclear pulses, as well as a robust structure and radiation shielding. The structure, configuration, and materials of the nozzle must meet many severe requirements. The configuration would focus, in a conical manner, the Deuterium-Tritium (D-T) fuel and Lithium-6/7 liner fluid to meet at a specific point that acts as a cathode so the Li-6 can serve as a current return path to complete the circuit. In addition to serving as a current return path, the Li liner also serves as a radiation shield. The advantage to this configuration is the reaction between neutrons and Li-6 results in the production of additional Tritium, thus adding further fuel to the fusion reaction and boosting the energy output. To understand the applicability of Z-Pinch propulsion to interplanetary travel, it is necessary to design a concept vehicle that uses it. The propulsion system significantly impacts the design of the electrical, thermal control, avionics, radiation shielding, and structural subsystems of a vehicle. The design reference mission is the transport of crew and cargo to Mars and back, with the intention that the vehicle be reused for other missions. Several aspects of this vehicle are based on a previous crewed fusion vehicle study called Human Outer Planet Exploration (HOPE), which employed a Magnetized Target Fusion (MTF) propulsion concept. Analysis of this propulsion system concludes that a 40-fold increase of Isp over chemical propulsion is predicted. This along with a greater than 30% predicted payload mass fraction certainly warrants further development of enabling technologies. The vehicle is designed for multiple interplanetary missions and conceivably may be suited for an automated one-way interstellar voyage.

Parametric Thermal and Flow Analysis of ITER Diagnostic Shield Module

DOE Office of Scientific and Technical Information (OSTI.GOV)

Khodak, A.; Zhai, Y.; Wang, W.

As part of the diagnostic port plug assembly, the ITER Diagnostic Shield Module (DSM) is designed to provide mechanical support and the plasma shielding while allowing access to plasma diagnostics. Thermal and hydraulic analysis of the DSM was performed using a conjugate heat transfer approach, in which heat transfer was resolved in both solid and liquid parts, and simultaneously, fluid dynamics analysis was performed only in the liquid part. ITER Diagnostic First Wall (DFW) and cooling tubing were also included in the analysis. This allowed direct modeling of the interface between DSM and DFW, and also direct assessment of themore » coolant flow distribution between the parts of DSM and DFW to ensure DSM design meets the DFW cooling requirements. Design of the DSM included voids filled with Boron Carbide pellets, allowing weight reduction while keeping shielding capability of the DSM. These voids were modeled as a continuous solid with smeared material properties using analytical relation for thermal conductivity. Results of the analysis lead to design modifications improving heat transfer efficiency of the DSM. Furthermore, the effect of design modifications on thermal performance as well as effect of Boron Carbide will be presented.« less

Parametric Thermal and Flow Analysis of ITER Diagnostic Shield Module

DOE PAGES

Khodak, A.; Zhai, Y.; Wang, W.; ...

2017-06-19

As part of the diagnostic port plug assembly, the ITER Diagnostic Shield Module (DSM) is designed to provide mechanical support and the plasma shielding while allowing access to plasma diagnostics. Thermal and hydraulic analysis of the DSM was performed using a conjugate heat transfer approach, in which heat transfer was resolved in both solid and liquid parts, and simultaneously, fluid dynamics analysis was performed only in the liquid part. ITER Diagnostic First Wall (DFW) and cooling tubing were also included in the analysis. This allowed direct modeling of the interface between DSM and DFW, and also direct assessment of themore » coolant flow distribution between the parts of DSM and DFW to ensure DSM design meets the DFW cooling requirements. Design of the DSM included voids filled with Boron Carbide pellets, allowing weight reduction while keeping shielding capability of the DSM. These voids were modeled as a continuous solid with smeared material properties using analytical relation for thermal conductivity. Results of the analysis lead to design modifications improving heat transfer efficiency of the DSM. Furthermore, the effect of design modifications on thermal performance as well as effect of Boron Carbide will be presented.« less

Preliminary analyses of space radiation protection for lunar base surface systems

NASA Technical Reports Server (NTRS)

Nealy, John E.; Wilson, John W.; Townsend, Lawrence W.

1989-01-01

Radiation shielding analyses are performed for candidate lunar base habitation modules. The study primarily addresses potential hazards due to contributions from the galactic cosmic rays. The NASA Langley Research Center's high energy nucleon and heavy ion transport codes are used to compute propagation of radiation through conventional and regolith shield materials. Computed values of linear energy transfer are converted to biological dose-equivalent using quality factors established by the International Commision of Radiological Protection. Special fluxes of heavy charged particles and corresponding dosimetric quantities are computed for a series of thicknesses in various shield media and are used as an input data base for algorithms pertaining to specific shielded geometries. Dosimetric results are presented as isodose contour maps of shielded configuration interiors. The dose predictions indicate that shielding requirements are substantial, and an abbreviated uncertainty analysis shows that better definition of the space radiation environment as well as improvement in nuclear interaction cross-section data can greatly increase the accuracy of shield requirement predictions.

The Thermochemical Degradation of Hot Section Materials for Gas Turbine Engines in Alternative-Fuel Combustion Environments

NASA Astrophysics Data System (ADS)

Montalbano, Timothy

Gas turbine engines remain an integral part of providing the world's propulsion and power generation needs. The continued use of gas turbines requires increased temperature operation to reach higher efficiencies and the implementation of alternative fuels for a lower net-carbon footprint. This necessitates evaluation of the material coatings used to shield the hot section components of gas turbines in these new extreme environments in order to understand how material degradation mechanisms change. Recently, the US Navy has sought to reduce its use of fossil fuels by implementing a blended hydroprocessed renewable diesel (HRD) derived from algae in its fleet. To evaluate the material degradation in this alternative environment, metal alloys are exposed in a simulated combustion environment using this blended fuel or the traditional diesel-like fuel. Evaluation of the metal alloys showed the development of thick, porous scales with a large depletion of aluminum for the blend fuel test. A mechanism linking an increased solubility of the scale to the blend fuel test environment will be discussed. For power generation applications, Integrated Gasification Combined Cycle (IGCC) power plants can provide electricity with 45% efficiency and full carbon capture by using a synthetic gas (syngas) derived from coal, biomass, or another carbon feedstock. However, the combustion of syngas is known to cause high water vapor content levels in the exhaust stream with unknown material consequences. To evaluate the effect of increased humidity, air-plasma sprayed (APS), yttria-stabilized zirconia (YSZ) is thermally aged in an environment with and without humidity. An enhanced destabilization of the parent phase by humid aging is revealed by x-ray diffraction (XRD) and Raman spectroscopy. Microstructural analysis by transmission electron microscopy (TEM) and scanning-TEM (STEM) indicate an enhanced coarsening of the domain structure of the YSZ in the humid environment. The enhanced destabilization and coarsening in the humid aging environment is explained mechanistically by water-derived species being incorporated into the YSZ structure and altering the anion sublattice. The characterization of the metal alloy and ceramic coatings exposed in these alternative environments allows for a deeper understanding of the mechanisms behind the material evolution in these environments.

Wide Area Recovery and Resiliency Program (WARRP) Knowledge Enhancement Events: Waste Management Workshop After Action Report

DTIC Science & Technology

2012-03-01

approximately 2,300 curies of 137Cs (CsCl), and 1.5 tons of Ammonium nitrate / Fuel oil (ANFO). The explosive and the shielded CsCl sources are packaged into...previous findings. Experts also presented case studies on Hurricane Katrina, The British Petroleum (BP) Oil Spill, Fukushima Japan, Foot and Mouth...containers, conduct environmental monitoring. The waste streams were very organized into distinct categories. 1) oil , gasoline, pesticides, 2) batteries

Indirect drive targets for fusion power

DOEpatents

Amendt, Peter A.; Miles, Robin R.

2016-10-11

A hohlraum for an inertial confinement fusion power plant is disclosed. The hohlraum includes a generally cylindrical exterior surface, and an interior rugby ball-shaped surface. Windows over laser entrance holes at each end of the hohlraum enclose inert gas. Infrared reflectors on opposite sides of the central point reflect fusion chamber heat away from the capsule. P2 shields disposed on the infrared reflectors help assure an enhanced and more uniform x-ray bath for the fusion fuel capsule.

Environmental Compliance Assessment and Management Program (ECAMP)

DTIC Science & Technology

1994-06-01

square yard mg milligram yr year mi mile Chemicals CO carbon monoxide NO 2 nitrogen dioxide CO2 carbon dioxide NOx nitrogen oxides Hg mercury SO2 sulfur...installation intentionally shielded themselves from information which would have revealed a leak. (!)(3X)5)(7)(8) A.77. Facilities on Verify that facilities...released from the largest tank within the diked area, assuming a fuel tank. Verify that walls of diked areas are of earth , concrete, steel, or solid

Acoustic Prediction Methodology and Test Validation for an Efficient Low-Noise Hybrid Wing Body Subsonic Transport

NASA Technical Reports Server (NTRS)

Kawai, Ronald T. (Compiler)

2011-01-01

This investigation was conducted to: (1) Develop a hybrid wing body subsonic transport configuration with noise prediction methods to meet the circa 2007 NASA Subsonic Fixed Wing (SFW) N+2 noise goal of -52 dB cum relative to FAR 36 Stage 3 (-42 dB cum re: Stage 4) while achieving a -25% fuel burned compared to current transports (re :B737/B767); (2) Develop improved noise prediction methods for ANOPP2 for use in predicting FAR 36 noise; (3) Design and fabricate a wind tunnel model for testing in the LaRC 14 x 22 ft low speed wind tunnel to validate noise predictions and determine low speed aero characteristics for an efficient low noise Hybrid Wing Body configuration. A medium wide body cargo freighter was selected to represent a logical need for an initial operational capability in the 2020 time frame. The Efficient Low Noise Hybrid Wing Body (ELNHWB) configuration N2A-EXTE was evolved meeting the circa 2007 NRA N+2 fuel burn and noise goals. The noise estimates were made using improvements in jet noise shielding and noise shielding prediction methods developed by UC Irvine and MIT. From this the Quiet Ultra Integrated Efficient Test Research Aircraft #1 (QUIET-R1) 5.8% wind tunnel model was designed and fabricated.

New shielding material development for compact accelerator-driven neutron source

NASA Astrophysics Data System (ADS)

Hu, Guang; Hu, Huasi; Wang, Sheng; Han, Hetong; Otake, Y.; Pan, Ziheng; Taketani, A.; Ota, H.; Hashiguchi, Takao; Yan, Mingfei

2017-04-01

The Compact Accelerator-driven Neutron Source (CANS), especially the transportable neutron source is longing for high effectiveness shielding material. For this reason, new shielding material is researched in this investigation. The component of shielding material is designed and many samples are manufactured. Then the attenuation detection experiments were carried out. In the detections, the dead time of the detector appeases when the proton beam is too strong. To grasp the linear range and nonlinear range of the detector, two currents of proton are employed in Pb attenuation detections. The transmission ratio of new shielding material, polyethylene (PE), PE + Pb, BPE + Pb is detected under suitable current of proton. Since the results of experimental neutrons and γ-rays appear as together, the MCNP and PHITS simulations are applied to assisting the analysis. The new shielding material could reduce of the weight and volume compared with BPE + Pb and PE + Pb.

Research of glass fibre used in the electromagnetic wave shielding and absorption composite material

NASA Astrophysics Data System (ADS)

Xu, M.; Jia, F.; Bao, H. Q.; Cui, K.; Zhang, F.

2016-07-01

Electromagnetic shielding and absorption composite material plays an important role in the defence and economic field. Comparing with other filler, Glass fibre and its processed product—metal-coated glass fibre can greatly reduce the material's weight and costs, while it still remains the high strength and the electromagnetic shielding effectiveness. In this paper, the electromagnetic absorption mechanism and the reflection mechanism have been investigated as a whole, and the shielding effectiveness of the double-layer glass fibre composite material is mainly focused. The relationship between the shielding effectiveness and the filled glass fibre as well as its metal-coated product's parameters has also been studied. From the subsequent coaxial flange and anechoic chamber analysis, it can be confirmed that the peak electromagnetic shielding effectiveness of this double-layer material can reach -78dB while the bandwidth is from 2GHz to 18GHz.

Source Correlated Prompt Neutron Activation Analysis for Material Identification and Localization

NASA Astrophysics Data System (ADS)

Canion, Bonnie; McConchie, Seth; Landsberger, Sheldon

2017-07-01

This paper investigates the energy spectrum of photon signatures from an associated particle imaging deuterium tritium (API-DT) neutron generator interrogating shielded uranium. The goal is to investigate if signatures within the energy spectrum could be used to indirectly characterize shielded uranium when the neutron signature is attenuated. By utilizing the correlated neutron cone associated with each pixel of the API-DT neutron generator, certain materials can be identified and located via source correlated spectrometry of prompt neutron activation gamma rays. An investigation is done to determine if fission neutrons induce a significant enough signature within the prompt neutron-induced gamma-ray energy spectrum in shielding material to be useful for indirect nuclear material characterization. The signature deriving from the induced fission neutrons interacting with the shielding material was slightly elevated in polyethylene-shielding depleted uranium (DU), but was more evident in some characteristic peaks from the aluminum shielding surrounding DU.

An Analysis of Ablation-Shield Requirements for Manned Reentry Vehicles

NASA Technical Reports Server (NTRS)

Roberts, Leonard

1960-01-01

The problem of sublimation of material and accumulation of heat in an ablation shield is analyzed and the results are applied to the reentry of manned vehicles into the earth's atmosphere. The parameters which control the amount of sublimation and the temperature distribution within the ablation shield are determined and presented in a manner useful for engineering calculation. It is shown that the total mass loss from the shield during reentry and the insulation requirements may be given very simply in terms of the maximum deceleration of the vehicle or the total reentry time.

Gadolinium Oxide / Silicon Thin Film Heterojunction Solid-State Neutron Detector

DTIC Science & Technology

2010-03-01

PRODUCED AS A MEDICAL APPLICATOR SHOWN IN „A‟. THE SOURCE, PICTURED IN „B‟ HAS A PLASTIC SHIELD THAT SLIDES UP AND DOWN THE SHAFT WHICH IS DESIGNED TO...down the shaft which is designed to shield the operator from radiation. The source is sitting head-down and is covered by a thick aluminum shield for...EXPERIMENT, RESULTS, AND ANALYSIS ........................................................ 37 4.1 Experimental Design & Apparatus

Molecular orbital analysis of the inverse halogen dependence of nuclear magnetic shielding in LaX₃, X = F, Cl, Br, I.

PubMed

Moncho, Salvador; Autschbach, Jochen

2010-12-01

The NMR nuclear shielding tensors for the series LaX(3), with X = F, Cl, Br and I, have been computed using two-component relativistic density functional theory based on the zeroth-order regular approximation (ZORA). A detailed analysis of the inverse halogen dependence (IHD) of the La shielding was performed via decomposition of the shielding tensor elements into contributions from localized and delocalized molecular orbitals. Both spin-orbit and paramagnetic shielding terms are important, with the paramagnetic terms being dominant. Major contributions to the IHD can be attributed to the La-X bonding orbitals, as well as to trends associated with the La core and halogen lone pair orbitals, the latter being related to X-La π donation. An 'orbital rotation' model for the in-plane π acceptor f orbital of La helps to rationalize the significant magnitude of deshielding associated with the in-plane π donation. The IHD goes along with a large increase in the shielding tensor anisotropy as X becomes heavier, which can be associated with trends for the covalency of the La-X bonds, with a particularly effective transfer of spin-orbit coupling induced spin density from iodine to La in LaI(3). Copyright © 2010 John Wiley & Sons, Ltd.

Magnetic Shielding Design for Coupler of Wireless Electric Vehicle Charging Using Finite Element Analysis

NASA Astrophysics Data System (ADS)

Zhao, W. N.; Yang, X. J.; Yao, C.; Ma, D. G.; Tang, H. J.

2017-10-01

Inductive power transfer (IPT) is a practical and preferable method for wireless electric vehicle (EV) charging which proved to be safe, convenient and reliable. Due to the air gap between the magnetic coupler, the magnetic field coupling decreases and the magnetic leakage increases significantly compared to traditional transformer, and this may lead to the magnetic flux density around the coupler more than the safety limit for human. So magnetic shielding should be adding to the winding made from litz wire to enhance the magnetic field coupling effect in the working area and reduce magnetic field strength in non-working area. Magnetic shielding can be achieved by adding high-permeability material or high-conductivity material. For high-permeability material its magnetic reluctance is much lower than the surrounding air medium so most of the magnetic line goes through the high-permeability material rather than surrounding air. For high-conductivity material the eddy current in the material can produce reverse magnetic field to achieve magnetic shielding. This paper studies the effect of the two types of shielding material on coupler for wireless EV charging and designs combination shielding made from high-permeability material and high-conductivity material. The investigation of the paper is done with the help of finite element analysis.

Micrometeoroid and Orbital Debris (MMOD) Shield Ballistic Limit Analysis Program

NASA Technical Reports Server (NTRS)

Ryan, Shannon

2013-01-01

This software implements penetration limit equations for common micrometeoroid and orbital debris (MMOD) shield configurations, windows, and thermal protection systems. Allowable MMOD risk is formulated in terms of the probability of penetration (PNP) of the spacecraft pressure hull. For calculating the risk, spacecraft geometry models, mission profiles, debris environment models, and penetration limit equations for installed shielding configurations are required. Risk assessment software such as NASA's BUMPERII is used to calculate mission PNP; however, they are unsuitable for use in shield design and preliminary analysis studies. The software defines a single equation for the design and performance evaluation of common MMOD shielding configurations, windows, and thermal protection systems, along with a description of their validity range and guidelines for their application. Recommendations are based on preliminary reviews of fundamental assumptions, and accuracy in predicting experimental impact test results. The software is programmed in Visual Basic for Applications for installation as a simple add-in for Microsoft Excel. The user is directed to a graphical user interface (GUI) that requires user inputs and provides solutions directly in Microsoft Excel workbooks.

Thermal analyses of the IF-300 shipping cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Meier, J.K.

1978-07-01

In order to supply temperature data for structural testing and analysis of shipping casks, a series of thermal analyses using the TRUMP thermal analyzer program were performed on the GE IF-300 spent fuel shipping cask. Major conclusions of the analyses are: (1) Under normal cooling conditions and a cask heat load of 262,000 BTU/h, the seal area of the cask will be roughly 100/sup 0/C (180/sup 0/F) above the ambient surroundings. (2) Under these same conditions the uranium shield at the midpoint of the cask will be between 69/sup 0/C (125/sup 0/F) and 92/sup 0/C (166/sup 0/F) above the ambientmore » surroundings. (3) Significant thermal gradients are not likely to develop between the head studs and the surrounding metal. (4) A representative time constant for the cask as a whole is on the order of one day.« less

Structural Design and Thermal Analysis for Thermal Shields of the MICE Coupling Magnets

DOE Office of Scientific and Technical Information (OSTI.GOV)

Green, Michael A.; Pan, Heng; Liu, X. K.

2009-07-01

A superconducting coupling magnet made from copper matrix NbTi conductors operating at 4 K will be used in the Muon Ionization Cooling Experiment (MICE) to produce up to 2.6 T on the magnet centerline to keep the muon beam within the thin RF cavity indows. The coupling magnet is to be cooled by two cryocoolers with a total cooling capacity of 3 W at 4.2 K. In order to keep a certain operating temperature margin, the most important is to reduce the heat leakage imposed on cold surfaces of coil cold mass assembly. An ntermediate temperature shield system placed betweenmore » the coupling coil and warm vacuum chamber is adopted. The shield system consists of upper neck shield, main shields, flexible connections and eight supports, which is to be cooled by the first stage cold heads of two ryocoolers with cooling capacity of 55 W at 60 K each. The maximum temperature difference on the shields should be less than 20 K, so the thermal analyses for the shields with different thicknesses, materials, flexible connections for shields' cooling and structure design for heir supports were carried out. 1100 Al is finally adopted and the maximum temperature difference is around 15 K with 4 mm shield thickness. The paper is to present detailed analyses on the shield system design.« less

Gravity Scaling of a Power Reactor Water Shield

NASA Technical Reports Server (NTRS)

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.

Neutron dose rate analysis on HTGR-10 reactor using Monte Carlo code

NASA Astrophysics Data System (ADS)

Suwoto; Adrial, H.; Hamzah, A.; Zuhair; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 °C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman’s Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has been fulfilled. From the result analysis, it can be concluded that the model of calculation result of neutron dose rate for HTGR-10 core has met the required radiation safety standards.

Comparative thermal analysis of the Space Station Freedom photovoltaic deployable boom structure using TRASYS, NEVADA, and SINDA programs

NASA Technical Reports Server (NTRS)

Baumeister, Joseph F.; Beach, Duane E.; Armand, Sasan C.

1989-01-01

The proposed Space Station Photovoltaic Deployable Boom was analyzed for operating temperatures. The boom glass/epoxy structure design needs protective shielding from environmental degradation. The protective shielding optical properties (solar absorptivity and emissivity) dictate the operating temperatures of the boom components. The Space Station Boom protective shielding must also withstand the effects of the extendible/retractable coiling acting within the mast canister. A thermal analysis method was developed for the Space Station Deployable Boom to predict transient temperatures for a variety of surface properties. The modeling procedures used to evaluate temperatures within the boom structure incorporated the TRASYS, NEVADA, and SINDA thermal analysis programs. Use of these programs led to a comparison between TRASYS and NEVADA analysis methods. Comparing TRASYS and NEVADA results exposed differences in the environmental solar flux predictions.

Comparative thermal analysis of the space station Freedom photovoltaic deployable boom structure using TRASYS, NEVADA, and SINDA programs

NASA Technical Reports Server (NTRS)

Baumeister, Joseph F.; Beach, Duane E.; Armand, Sasan C.

1989-01-01

The proposed Space Station Photovoltaic Deployable Boom was analyzed for operating temperatures. The boom glass/epoxy structure design needs protective shielding from environmental degradation. The protective shielding optical properties (solar absorptivity and emissivity) dictate the operating temperatures of the boom components. The Space Station Boom protective shielding must also withstand the effects of the extendible/retractable coiling action within the mast canister. A thermal analysis method was developed for the Space Station Deployable Boom to predict transient temperatures for a variety of surface properties. The modeling procedures used to evaluate temperatures within the boom structure incorporated the TRASYS, NEVADA, and SINDA thermal analysis programs. Use of these programs led to a comparison between TRASYS and NEVADA analysis methods. Comparing TRASYS and NEVADA results exposed differences in the environmental solar flux predictions.

Analytical-HZETRN Model for Rapid Assessment of Active Magnetic Radiation Shielding

NASA Technical Reports Server (NTRS)

Washburn, S. A.; Blattnig, S. R.; Singleterry, R. C.; Westover, S. C.

2014-01-01

The use of active radiation shielding designs has the potential to reduce the radiation exposure received by astronauts on deep-space missions at a significantly lower mass penalty than designs utilizing only passive shielding. Unfortunately, the determination of the radiation exposure inside these shielded environments often involves lengthy and computationally intensive Monte Carlo analysis. In order to evaluate the large trade space of design parameters associated with a magnetic radiation shield design, an analytical model was developed for the determination of flux inside a solenoid magnetic field due to the Galactic Cosmic Radiation (GCR) radiation environment. This analytical model was then coupled with NASA's radiation transport code, HZETRN, to account for the effects of passive/structural shielding mass. The resulting model can rapidly obtain results for a given configuration and can therefore be used to analyze an entire trade space of potential variables in less time than is required for even a single Monte Carlo run. Analyzing this trade space for a solenoid magnetic shield design indicates that active shield bending powers greater than 15 Tm and passive/structural shielding thicknesses greater than 40 g/cm2 have a limited impact on reducing dose equivalent values. Also, it is shown that higher magnetic field strengths are more effective than thicker magnetic fields at reducing dose equivalent.

Analysis and measurements of low frequency lightning component penetration through aerospace vehicle metal and graphite skins

NASA Technical Reports Server (NTRS)

Robb, J. D.; Chen, T.

1980-01-01

An analysis of the shielding properties of mixed metal and graphite composite structures has illustrated some important aspects of electromagnetic field penetration into the interior. These include: (1) that graphite access doors on metallic structures will attenuate lightning magnetic fields very little; conversely, metal doors on a graphite structure will also attenuate fields from lightning strike currents very little, i.e., homogeneity of the shield is a critical factor in shielding and (2) that continuous conductors between two points inside a graphite skin such as an air data probe metallic tubing connection to an air data computer can allow large current penetrations into a vehicle interior. The true weight savings resulting from the use of composite materials can only be evaluated after the resulting electromagnetic problems such as current penetrations have been solved, and this generally requires weight addition in the form of cable shields, conductor bonding or external metallization.

Skeletal mechanism generation for surrogate fuels using directed relation graph with error propagation and sensitivity analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Niemeyer, Kyle E.; Sung, Chih-Jen; Raju, Mandhapati P.

2010-09-15

A novel implementation for the skeletal reduction of large detailed reaction mechanisms using the directed relation graph with error propagation and sensitivity analysis (DRGEPSA) is developed and presented with examples for three hydrocarbon components, n-heptane, iso-octane, and n-decane, relevant to surrogate fuel development. DRGEPSA integrates two previously developed methods, directed relation graph-aided sensitivity analysis (DRGASA) and directed relation graph with error propagation (DRGEP), by first applying DRGEP to efficiently remove many unimportant species prior to sensitivity analysis to further remove unimportant species, producing an optimally small skeletal mechanism for a given error limit. It is illustrated that the combination ofmore » the DRGEP and DRGASA methods allows the DRGEPSA approach to overcome the weaknesses of each, specifically that DRGEP cannot identify all unimportant species and that DRGASA shields unimportant species from removal. Skeletal mechanisms for n-heptane and iso-octane generated using the DRGEP, DRGASA, and DRGEPSA methods are presented and compared to illustrate the improvement of DRGEPSA. From a detailed reaction mechanism for n-alkanes covering n-octane to n-hexadecane with 2115 species and 8157 reactions, two skeletal mechanisms for n-decane generated using DRGEPSA, one covering a comprehensive range of temperature, pressure, and equivalence ratio conditions for autoignition and the other limited to high temperatures, are presented and validated. The comprehensive skeletal mechanism consists of 202 species and 846 reactions and the high-temperature skeletal mechanism consists of 51 species and 256 reactions. Both mechanisms are further demonstrated to well reproduce the results of the detailed mechanism in perfectly-stirred reactor and laminar flame simulations over a wide range of conditions. The comprehensive and high-temperature n-decane skeletal mechanisms are included as supplementary material with this article. (author)« less

Radiation Engineering Analysis of Shielding Materials to Assess Their Ability to Protect Astronauts in Deep Space From Energetic Particle Radiation

NASA Technical Reports Server (NTRS)

Singleterry, R. C.

2013-01-01

An analysis is performed on four typical materials (aluminum, liquid hydrogen, polyethylene, and water) to assess their impact on the length of time an astronaut can stay in deep space and not exceed a design basis radiation exposure of 150 mSv. A large number of heavy lift launches of pure shielding mass are needed to enable long duration, deep space missions to keep astronauts at or below the exposure value with shielding provided by the vehicle. Therefore, vehicle mass using the assumptions in the paper cannot be the sole shielding mechanism for long duration, deep space missions. As an example, to enable the Mars Design Reference Mission 5.0 with a 400 day transit to and from Mars, not including the 500 day stay on the surface, a minimum of 24 heavy lift launches of polyethylene at 89,375 lbm (40.54 tonnes) each are needed for the 1977 galactic cosmic ray environment. With the assumptions used in this paper, a single heavy lift launch of water or polyethylene can protect astronauts for a 130 day mission before exceeding the exposure value. Liquid hydrogen can only protect the astronauts for 160 days. Even a single launch of pure shielding material cannot protect an astronaut in deep space for more than 180 days using the assumptions adopted in the analysis. It is shown that liquid hydrogen is not the best shielding material for the same mass as polyethylene for missions that last longer than 225 days.

Closeup view of the bottom area of Space Shuttle Main ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

Close-up view of the bottom area of Space Shuttle Main Engine (SSME) 2052 engine assembly mounted in a SSME Engine Handler in the Horizontal Processing area of the SSME Processing Facility at Kennedy Space Center. The most prominent features in this view are the Low-Pressure Oxidizer Discharge Duct toward the bottom of the assembly, the SSME Engine Controller and the Main Fuel Valve Hydraulic Actuator are in the approximate center of the assembly in this view, the Low-Pressure Fuel Turbopump (LPFTP), the LPFTP Discharge Duct are to the left on the assembly in this view and the High-Pressure Fuel Turbopump is located toward the top of the engine assembly in this view. The ring of tabs in the right side of the image, at the approximate location of the Nozzle and the Coolant Outlet Manifold interface is the Heat Shield Support Ring. - Space Transportation System, Space Shuttle Main Engine, Lyndon B. Johnson Space Center, 2101 NASA Parkway, Houston, Harris County, TX

Behavior analysis in consumer affairs: encouraging dental professionals to provide consumers with shielding from unnecessary X-ray exposure.

PubMed

Greene, B F; Neistat, M D

1983-01-01

An unobtrusive observation system was developed to determine the extent to which dental professionals in two communities provided lead shielding to patients during X-ray exams. A lengthy baseline revealed low and irregular provision of shielding among half of these professionals. Subsequently, a program was undertaken by a consumer's group in which these professionals were requested to provide shielding and were given confidential feedback regarding its use during the baseline period. The provision of shielding dramatically increased at all offices and was maintained throughout a follow-up period extending to more than 9 months after the program's implementation. Little or no generalized effect was observed in the occurrence of three collateral behaviors that were also assessed throughout the study.

Behavior analysis in consumer affairs: encouraging dental professionals to provide consumers with shielding from unnecessary X-ray exposure.

PubMed Central

Greene, B F; Neistat, M D

1983-01-01

An unobtrusive observation system was developed to determine the extent to which dental professionals in two communities provided lead shielding to patients during X-ray exams. A lengthy baseline revealed low and irregular provision of shielding among half of these professionals. Subsequently, a program was undertaken by a consumer's group in which these professionals were requested to provide shielding and were given confidential feedback regarding its use during the baseline period. The provision of shielding dramatically increased at all offices and was maintained throughout a follow-up period extending to more than 9 months after the program's implementation. Little or no generalized effect was observed in the occurrence of three collateral behaviors that were also assessed throughout the study. PMID:6833165

Descent Stage of Mars Science Laboratory During Assembly

NASA Technical Reports Server (NTRS)

2008-01-01

This image from early October 2008 shows personnel working on the descent stage of NASA's Mars Science Laboratory inside the Spacecraft Assembly Facility at NASA's Jet Propulsion Laboratory, Pasadena, Calif.

The descent stage will provide rocket-powered deceleration for a phase of the arrival at Mars after the phases using the heat shield and parachute. When it nears the surface, the descent stage will lower the rover on a bridle the rest of the way to the ground. The larger three of the orange spheres in the descent stage are fuel tanks. The smaller two are tanks for pressurant gas used for pushing the fuel to the rocket engines.

JPL, a division of the California Institute of Technology, manages the Mars Science Laboratory Project for the NASA Science Mission Directorate, Washington.

Testimony of Fred R. Mynatt before the Energy Research and Development Subcommittee of the Committee on Science, Space, and Technology, US House of Representatives. [Advanced fuel technology, gas-cooled reactor technology, and liquid metal-cooled reactor technology programs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mynatt, F.R.

1987-03-18

This report provides a description of the statements submitted for the record to the committee on Science, Space, and Technology of the United States House of Representatives. These statements describe three principal areas of activity of the Advanced Reactor Technology Program of the Department of Energy (DOE). These areas are advanced fuel cycle technology, modular high-temperature gas-cooled reactor technology, and liquid metal-cooled reactor. The areas of automated reactor control systems, robotics, materials and structural design shielding and international cooperation were included in these statements describing the Oak Ridge National Laboratory's efforts in these areas. (FI)

Wide Area Recovery and Resiliency Program (WARRP) Knowledge Enhancement Events: Damage Assessment Building Abandonment After Action Report

DTIC Science & Technology

2012-04-17

nitrate / Fuel oil (ANFO). The explosive and the shielded CsCl sources are packaged into bombs and loaded onto a truck. The total explosive yield in...Transportation Eric Jacobs State of Colorado Laura Johnston Dewberry Carl Miller Colorado Springs OEM John Miller American Red Cross Matthew...the health effects of the contamination will be. Scenario Terrorist obtain approximately 2,300 curies of 137 Cs (CsCl), and 1.5 tons of Ammonium

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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.

Strike Rescue - Are We on the Right Path?

DTIC Science & Technology

1990-11-16

Vietnam, a pilot had more to worry about than just enemy SAM’s and flak. More than one POW was bagged by his own weapon system. CDR Jeremiah A. Denton ...Squadron (SOS) based at Eglin AFB in Florida. The MH-60G features an air-to-air refueling probe and auxiliary internal fuel tanks. This allows for 4.5...missions in a contaminated NBC environment. Specifically, let’s consider SC𔄁 missions to rescue "human shields" that Hussein has inserted to protect

Inhibited Shaped Charge Launcher Testing of Spacecraft Shield Designs

NASA Technical Reports Server (NTRS)

Grosch, Donald J.

1996-01-01

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

Radiation resistant concrete for applications in nuclear power and radioactive waste industries

NASA Astrophysics Data System (ADS)

Burnham, Steven Robert

Elemental components of ordinary concrete contain a variety of metals and rare earth elements that are susceptible to neutron activation. This activation occurs by means of radiative capture, a neutron interaction that results in formation of radioisotopes such as Co-60, Eu-152, and Eu-154. Studies have shown that these three radioisotopes are responsible for the residual radioactivity found in nuclear power plant concrete reactor dome and shielding walls. Such concrete is classified as Low Level Radioactive Waste (LLRW) and Very Low Level Waste (VLLW) by International Atomic Energy Agency (IAEA) standards and requires disposal at appropriate disposal sites. There are only three such sites in the USA, and every nuclear power plant will produce at the time of decommissioning approximately 1,500 tonnes of activated concrete classified as LLRW and VLLW. NAVA ALIGA (ancient word for a new stone) is a new concrete mixture developed mainly by research as presented in this thesis. The purpose of NAVA ALIGA is to satisfy IAEA clearance levels if used as a material for reactor dome, spent fuel pool, or radioactive waste canisters. NAVA ALIGA will never be activated above the IAEA clearance level after long-term exposure to neutron radiation when used as a material for reactor dome, spent fuel pool, and radioactive waste canisters. Components of NAVA ALIGA were identified using Instrumental Neutron Activation Analysis (INAA) and Inductively Coupled Plasma Mass Spectrometry (ISP-MS) to determine trace element composition. In addition, it was tested for compressive strength and permeability, important for nuclear infrastructure. The studied mixture had a high water to cement ratio of 0.56, which likely resulted in the high measured permeability, yet the mixture also showed a compressive strength greater than 6 000 psi after 28 days. In addition to this experimental analysis, which goal was to develop a standard approach to define the concrete mixtures in satisfying the IAEA radiation clearance levels, the NAVA ALIGA concrete was analyzed as to potentially be used together with depleted uranium. This study was purely computational (based on MCNP6 models) and was twofold: to find if this new concrete mix would enhance the radiation shielding properties when combined with depleted uranium and to find if this will be an effective and useful way of using the existing large quantities of disposed depleted uranium.

Experimental and Analytical Studies of Shielding Concepts for Point Sources and Jet Noises.

NASA Astrophysics Data System (ADS)

Wong, Raymond Lee Man

This analytical and experimental study explores concepts for jet noise shielding. Model experiments centre on solid planar shields, simulating engine-over-wing installations, and 'sugar scoop' shields. Tradeoff on effective shielding length is set by interference 'edge noise' as the shield trailing edge approaches the spreading jet. Edge noise is minimized by (i) hyperbolic cutouts which trim off the portions of most intense interference between the jet flow and the barrier and (ii) hybrid shields--a thermal refractive extension (a flame); for (ii) the tradeoff is combustion noise. In general, shielding attenuation increases steadily with frequency, following low frequency enhancement by edge noise. Although broadband attenuation is typically only several dB, the reduction of the subjectively weighted perceived noise levels is higher. In addition, calculated ground contours of peak PN dB show a substantial contraction due to shielding: this reaches 66% for one of the 'sugar scoop' shields for the 90 PN dB contour. The experiments are complemented by analytical predictions. They are divided into an engineering scheme for jet noise shielding and more rigorous analysis for point source shielding. The former approach combines point source shielding with a suitable jet source distribution. The results are synthesized into a predictive algorithm for jet noise shielding: the jet is modelled as a line distribution of incoherent sources with narrow band frequency (TURN)(axial distance)('-1). The predictive version agrees well with experiment (1 to 1.5 dB) up to moderate frequencies. The insertion loss deduced from the point source measurements for semi-infinite as well as finite rectangular shields agrees rather well with theoretical calculation based on the exact half plane solution and the superposition of asymptotic closed-form solutions. An approximate theory, the Maggi-Rubinowicz line integral, is found to yield reasonable predictions for thin barriers including cutouts if a certain correction is applied. The more exact integral equation approach (solved numerically) is applied to a more demanding geometry: a half round sugar scoop shield. It is found that the solutions of integral equation derived from Helmholtz formula in normal derivative form show satisfactory agreement with measurements.

HVI Ballistic Performance Characterization of Non-Parallel Walls

NASA Technical Reports Server (NTRS)

Bohl, William; Miller, Joshua; Christiansen, Eric

2012-01-01

The Double-Wall, "Whipple" Shield [1] has been the subject of many hypervelocity impact studies and has proven to be an effective shield system for Micro-Meteoroid and Orbital Debris (MMOD) impacts for spacecraft. The US modules of the International Space Station (ISS), with their "bumper shields" offset from their pressure holding rear walls provide good examples of effective on-orbit use of the double wall shield. The concentric cylinder shield configuration with its large radius of curvature relative to separation distance is easily and effectively represented for testing and analysis as a system of two parallel plates. The parallel plate double wall configuration has been heavily tested and characterized for shield performance for normal and oblique impacts for the ISS and other programs. The double wall shield and principally similar Stuffed Whipple Shield are very common shield types for MMOD protection. However, in some locations with many spacecraft designs, the rear wall cannot be modeled as being parallel or concentric with the outer bumper wall. As represented in Figure 1, there is an included angle between the two walls. And, with a cylindrical outer wall, the effective included angle constantly changes. This complicates assessment of critical spacecraft components located within outer spacecraft walls when using software tools such as NASA's BumperII. In addition, the validity of the risk assessment comes into question when using the standard double wall shield equations, especially since verification testing of every set of double wall included angles is impossible.

Thermal Examination of an Orbiting Cryogenic Fuel Depot

NASA Technical Reports Server (NTRS)

Hull, Patrick V.; Canfield, Steven L.; Carrington, Connie; Fikes, John

2002-01-01

For many years NASA has been interested in the storage and transfer of cryogenic fuels in space. Lunar, L2 and other chemical propulsive space vehicle missions now have staged refueling needs that a fuel depot would satisfy. The depot considered is located in lower earth orbit. Many considerations must go into designing and building such a station. Multi-layer insulation systems, thermal shielding and low conductive structural supports are the principal means of protecting the system from excessive heat loss due to boiloff. This study focuses on the thermal losses associated with storing LH2 in a passively cooled fuel depot in a lower earth equatorial orbit. The corresponding examination looks at several configurations of the fuel depot. An analytical model has been developed to determine the thermal advantages and disadvantages of three different fuel depot configurations. Each of the systems consists of three Boeing rocket bodies arranged in various configurations. The first two configurations are gravity gradient stabilized while the third one is a spin-stabilized concept. Each concept was chosen for self-righting capabilities as well as the fuel settling capabilities, however the purpose of this paper is to prove which of the three concepts is the most efficient passively cooled system. The specific areas to be discussed are the heating time from the fusion temperature to the vaporization temperature and the amount of boiloff for a specific number of orbits. Each of the previous points is compared using various sun exposed surface areas of the tanks.

Gravity Scaling of a Power Reactor Water Shield

NASA Technical Reports Server (NTRS)

Reid, Robert S.; Pearson, J. Boise

2007-01-01

A similarity analysis on a water-based reactor shield examined the effect of gravity on free convection between a reactor shield inner and outer vessel boundaries. Two approaches established similarity between operation on the Earth and the Moon: 1) direct scaling of Rayleigh number equating gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant. Nusselt number for natural convection (laminar and turbulent) is assumed of form Nu = CRa(sup n).

Two-dimensional over-all neutronics analysis of the ITER device

NASA Astrophysics Data System (ADS)

Zimin, S.; Takatsu, Hideyuki; Mori, Seiji; Seki, Yasushi; Satoh, Satoshi; Tada, Eisuke; Maki, Koichi

1993-07-01

The present work attempts to carry out a comprehensive neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) developed during the Conceptual Design Activities (CDA). The two-dimensional cylindrical over-all calculational models of ITER CDA device including the first wall, blanket, shield, vacuum vessel, magnets, cryostat and support structures were developed for this purpose with a help of the DOGII code. Two dimensional DOT 3.5 code with the FUSION-40 nuclear data library was employed for transport calculations of neutron and gamma ray fluxes, tritium breeding ratio (TBR), and nuclear heating in reactor components. The induced activity calculational code CINAC was employed for the calculations of exposure dose rate after reactor shutdown around the ITER CDA device. The two-dimensional over-all calculational model includes the design specifics such as the pebble bed Li2O/Be layered blanket, the thin double wall vacuum vessel, the concrete cryostat integrated with the over-all ITER design, the top maintenance shield plug, the additional ring biological shield placed under the top cryostat lid around the above-mentioned top maintenance shield plug etc. All the above-mentioned design specifics were included in the employed calculational models. Some alternative design options, such as the water-rich shielding blanket instead of lithium-bearing one, the additional biological shield plug at the top zone between the poloidal field (PF) coil No. 5, and the maintenance shield plug, were calculated as well. Much efforts have been focused on analyses of obtained results. These analyses aimed to obtain necessary recommendations on improving the ITER CDA design.

Micrometeoroid and Orbital Debris Threat Assessment: Mars Sample Return Earth Entry Vehicle

NASA Technical Reports Server (NTRS)

Christiansen, Eric L.; Hyde, James L.; Bjorkman, Michael D.; Hoffman, Kevin D.; Lear, Dana M.; Prior, Thomas G.

2011-01-01

This report provides results of a Micrometeoroid and Orbital Debris (MMOD) risk assessment of the Mars Sample Return Earth Entry Vehicle (MSR EEV). The assessment was performed using standard risk assessment methodology illustrated in Figure 1-1. Central to the process is the Bumper risk assessment code (Figure 1-2), which calculates the critical penetration risk based on geometry, shielding configurations and flight parameters. The assessment process begins by building a finite element model (FEM) of the spacecraft, which defines the size and shape of the spacecraft as well as the locations of the various shielding configurations. This model is built using the NX I-deas software package from Siemens PLM Software. The FEM is constructed using triangular and quadrilateral elements that define the outer shell of the spacecraft. Bumper-II uses the model file to determine the geometry of the spacecraft for the analysis. The next step of the process is to identify the ballistic limit characteristics for the various shield types. These ballistic limits define the critical size particle that will penetrate a shield at a given impact angle and impact velocity. When the finite element model is built, each individual element is assigned a property identifier (PID) to act as an index for its shielding properties. Using the ballistic limit equations (BLEs) built into the Bumper-II code, the shield characteristics are defined for each and every PID in the model. The final stage of the analysis is to determine the probability of no penetration (PNP) on the spacecraft. This is done using the micrometeoroid and orbital debris environment definitions that are built into the Bumper-II code. These engineering models take into account orbit inclination, altitude, attitude and analysis date in order to predict an impacting particle flux on the spacecraft. Using the geometry and shielding characteristics previously defined for the spacecraft and combining that information with the environment model calculations, the Bumper-II code calculates a probability of no penetration for the spacecraft.

[The economics of preventing psycho-social risks].

PubMed

Golzio, Luigi

2014-01-01

The aim of the essay is to show the SHIELD methodology for helping the firm management to improve the risks prevention policy. It has been tested in the field with positive results. SHIELD is a cost-benefit analysis application to compare prevention and non-prevention costs, which arise from non-market risks. In the economic perspective safety risks (which include psycho-social risks) are non-market ones as they cause injures to workers during the job. SHIELD (Social Health Indicators for Economic Labour Decisions), is the original method proposed by the author. It is a cost benefits analysis application, which compares safety prevention and non-prevention costs. The comparison allow stop management to evaluate the efficiency of the current safety prevention policy as it helps top management to answer to the policy question: how much to invest in prevention costs? The costs comparison is obtained through the reclassification of safety costs between prevention and non-prevention costs (which are composed by claim damages and penalty sanction costs). SHIELD has been tested empirically in four companies operating in the agribusiness sector during a research financed by the Assessorato all'Agricoltura and INAI Regionale of Emilia Romagna Region. Results are postive: it has been found that the increase of prevention costs causes the cut of non-prevention costs in all companies looked into, as assumed by the high reliability organization theory. SHIELD can be applied to all companies which must have an accounting system by law, no matter of the industry they act. Its application has limited costs as SHIELD doesn't need changes in the accounting system. Safety costs sustained by the company are simply reclassified in prevention and non-prevention costs. The comparison of these two costs categories has been appreciated by top management of companies investigated as a useful support to decide the risks prevention policy for the company. The SHIELD original feature compared with others cost benefit analysis application is to compute registered costs in the company accounting system.

Shielding Development for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Caffrey, Jarvis A.; Gomez, Carlos F.; Scharber, Luke L.

2015-01-01

Radiation shielding analysis and development for the Nuclear Cryogenic Propulsion Stage (NCPS) effort is currently in progress and preliminary results have enabled consideration for critical interfaces in the reactor and propulsion stage systems. Early analyses have highlighted a number of engineering constraints, challenges, and possible mitigating solutions. Performance constraints include permissible crew dose rates (shared with expected cosmic ray dose), radiation heating flux into cryogenic propellant, and material radiation damage in critical components. Design strategies in staging can serve to reduce radiation scatter and enhance the effectiveness of inherent shielding within the spacecraft while minimizing the required mass of shielding in the reactor system. Within the reactor system, shield design is further constrained by the need for active cooling with minimal radiation streaming through flow channels. Material selection and thermal design must maximize the reliability of the shield to survive the extreme environment through a long duration mission with multiple engine restarts. A discussion of these challenges and relevant design strategies are provided for the mitigation of radiation in nuclear thermal propulsion.

Quantum diffraction and shielding effects on the low-energy electron-ion bremsstrahlung in two-component semiclassical plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Myoung-Jae; Jung, Young-Dae, E-mail: ydjung@hanyang.ac.kr; Department of Physics, Applied Physics, and Astronomy, Rensselaer Polytechnic Institute, 110 8th Street, Troy, New York 12180-3590

2015-10-15

The quantum diffraction and shielding effects on the low-energy bremsstrahlung process are investigated in two-component semiclassical plasmas. The impact-parameter analysis with the micropotential taking into account the quantum diffraction and shielding effects is employed to obtain the electron-ion bremsstrahlung radiation cross section as a function of the de Broglie wavelength, density parameter, impact parameter, photon energy, and projectile energy. The result shows that the influence of quantum diffraction and shielding strongly suppresses the bremsstrahlung radiation spectrum in semiclassical plasmas. It is found that the quantum diffraction and shielding effects have broaden the photon emission domain. It is also found thatmore » the photon emission domain is almost independent of the radiation photon energy. In addition, it is found that the influence of quantum diffraction and shielding on the bremsstrahlung spectrum decreases with an increase of the projectile energy. The density effect on the electron-ion bremsstrahlung cross section is also discussed.« less

Analysis of the bond-valence method for calculating (29) Si and (31) P magnetic shielding in covalent network solids.

PubMed

Holmes, Sean T; Alkan, Fahri; Iuliucci, Robbie J; Mueller, Karl T; Dybowski, Cecil

2016-07-05

(29) Si and (31) P magnetic-shielding tensors in covalent network solids have been evaluated using periodic and cluster-based calculations. The cluster-based computational methodology employs pseudoatoms to reduce the net charge (resulting from missing co-ordination on the terminal atoms) through valence modification of terminal atoms using bond-valence theory (VMTA/BV). The magnetic-shielding tensors computed with the VMTA/BV method are compared to magnetic-shielding tensors determined with the periodic GIPAW approach. The cluster-based all-electron calculations agree with experiment better than the GIPAW calculations, particularly for predicting absolute magnetic shielding and for predicting chemical shifts. The performance of the DFT functionals CA-PZ, PW91, PBE, rPBE, PBEsol, WC, and PBE0 are assessed for the prediction of (29) Si and (31) P magnetic-shielding constants. Calculations using the hybrid functional PBE0, in combination with the VMTA/BV approach, result in excellent agreement with experiment. © 2016 Wiley Periodicals, Inc. © 2016 Wiley Periodicals, Inc.

SOLID PHASE MICROEXTRACTION SAMPLING OF FIRE DEBRIS RESIDUES IN THE PRESENCE OF RADIONUCLIDE SURROGATE METALS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Duff, M; Keisha Martin, K; S Crump, S

2007-03-23

The Federal Bureau of Investigation (FBI) Laboratory currently does not have on site facilities for handling radioactive evidentiary materials and there are no established FBI methods or procedures for decontaminating highly radioactive fire debris (FD) evidence while maintaining evidentiary value. One experimental method for the isolation of FD residue from radionuclide metals involves using solid phase microextraction (SPME) fibers to remove the residues of interest. Due to their high affinity for organics, SPME fibers should have little affinity for most (radioactive) metals. The focus of this research was to develop an examination protocol that was applicable to safe work inmore » facilities where high radiation doses are shielded from the workers (as in radioactive shielded cells or ''hot cells''). We also examined the affinity of stable radionuclide surrogate metals (Co, Ir, Re, Ni, Ba, Cs, Nb, Zr and Nd) for sorption by the SPME fibers. This was done under exposure conditions that favor the uptake of FD residues under conditions that will provide little contact between the SPME and the FD material (such as charred carpet or wood that contains commonly-used accelerants). Our results from mass spectrometric analyses indicate that SPME fibers show promise for use in the room temperature head space uptake of organic FD residue (namely, diesel fuel oil, kerosene, gasoline and paint thinner) with subsequent analysis by gas chromatography (GC) with mass spectrometric (MS) detection. No inorganic forms of ignitable fluids were included in this study.« less

The 3D Radiation Dose Analysis For Satellite

NASA Astrophysics Data System (ADS)

Cai, Zhenbo; Lin, Guocheng; Chen, Guozhen; Liu, Xia

2002-01-01

the earth. These particles come from the Van Allen Belt, Solar Cosmic Ray and Galaxy Cosmic Ray. They have different energy and flux, varying with time and space, and correlating with solar activity tightly. These particles interact with electrical components and materials used on satellites, producing various space radiation effects, which will damage satellite to some extent, or even affect its safety. orbit. Space energy particles inject into components and materials used on satellites, and generate radiation dose by depositing partial or entire energy in them through ionization, which causes their characteristic degradation or even failure. As a consequence, the analysis and protection for radiation dose has been paid more attention during satellite design and manufacture. Designers of satellites need to analyze accurately the space radiation dose while satellites are on orbit, and use the results as the basis for radiation protection designs and ground experiments for satellites. can be calculated, using the model of the trapped proton and the trapped electron in the Van Allen Belt (AE8 and AP8). This is the 1D radiation dose analysis for satellites. Obviously, the mass shielding from the outside space to the computed point in all directions is regarded as a simple sphere shell. The actual structure of satellites, however, is very complex. When energy particles are injecting into a given equipment inside satellite from outside space, they will travel across satellite structure, other equipment, the shell of the given equipment, and so on, which depends greatly on actual layout of satellite. This complex radiation shielding has two characteristics. One is that the shielding masses for the computed point are different in different injecting directions. The other is that for different computed points, the shielding conditions vary in all space directions. Therefore, it is very difficult to tell the differences described above using the 1D radiation analysis, and hence, it is too simple to guide satellite radiation protection and ground experiments only based on the 1D radiation analysis results. To comprehend the radiation dose status of satellite adequately, it's essential to perform 3D radiation analysis for satellites. using computer software. From this 3D layout, the satellite model can be simplified appropriately. First select the point to be analyzed in the simplified satellite model, and extend many lines to the outside space, which divides the 4 space into many corresponding small areas with a certain solid angle. Then the shielding masses through the satellite equipment and structures along each direction are calculated, resulting in the shielding mass distribution in all space directions based on the satellite layout. Finally, using the relationship between radiation dose and shielding thickness from the 1D analysis, calculate the radiation dose in each area represented by each line. After we obtain the radiation dose and its space distribution for the point of interest, the 3D satellite radiation analysis is completed. radiation analysis based on satellite 3D CAD layout has larger benefit for engineering applications than the 1D analysis based on the solid sphere shielding model. With the 3D model, the analysis of space environment and its effect is combined closely with actual satellite engineering. The 3D radiation analysis not only provides valuable engineering data for satellite radiation design and protection, but also provides possibility to apply new radiation protection approaches, which expands technology horizon and broadens ways for technology development.

PRESSURIZED WATER REACTOR PROGRAM TECHNICAL PROGRESS REPORT FOR THE PERIOD MAY 5, 1955 TO JUNE 16, 1955

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

The current PWR plant and core parameters are listed. Resign requirements are briefly summarized for a radiation monitoring system, a fuel handling water system, a coolant purification system, an electrical power distribution system, and component shielding. Results of studies on thermal bowing and stressing of UO/sub 2/ are reported. A graph is presented of reactor power vs. reactor flow for various hot channel conditions. Development of U-- Mo and U-Nb alloys has been stopped because of the recent selection of UO/sub 2/ fuel material for the PWR core and blanket. The fabrication characteristics of UO/sub 2/ powders are being studied.more » Seamless Zircaloy-2 tubing has been tested to determine elastic limits, bursting pressures, and corrosion resistance. Fabrication techniques and tests for corrosion and defects in Zircaloy-clad U-Mo and UO/sub 2/ fuel rods are described. The preparation of UO/sub 2/ by various methods is being studied to determine which method produces a material most suitable for PWR fuel elements. The stability of UO/sub 2/ compacts in high temperature water and steam is being determined. Surface area and density measurements have been performed on samples of UO/sub 2/ powder prepared by various methods. Revelopment work on U-- Mo and U--Nb alloys has included studies of the effect on corrosion behavior of additions to the test water, additions to the alloys, homogenization of the alloys, annealing times, cladding, and fabrication techniques. Data are presented on relaxation in spring materials after exposure to a corrosive environment. Results are reported from loop and autoclave tests on fission product and crud deposition. Results of irradiation and corrosion testing of clad and unclad U--Mo and U-Nh alloys are described. The UO/sub 2/ irradiation program has included studies of dimensional changes, release of fission gases, and activity in the water surrounding the samples. A review of the methods of calculating reactor physics parameters has been completed, and the established procedures have been applied to determination of PWR reference design parameters. Critical experiments and primary loop shielding analyses are described. (D.E.B.)« less

Comparison of fresh fuel experimental measurements to MCNPX calculations using self-interrogation neutron resonance densitometry

NASA Astrophysics Data System (ADS)

LaFleur, Adrienne M.; Charlton, William S.; Menlove, Howard O.; Swinhoe, Martyn T.

2012-07-01

A new non-destructive assay technique called Self-Interrogation Neutron Resonance Densitometry (SINRD) is currently being developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for Light Water Reactor (LWR) fuel assemblies. SINRD consists of four 235U fission chambers (FCs): bare FC, boron carbide shielded FC, Gd covered FC, and Cd covered FC. Ratios of different FCs are used to determine the amount of resonance absorption from 235U in the fuel assembly. The sensitivity of this technique is based on using the same fissile materials in the FCs as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. In this work, experimental measurements were performed in air with SINRD using a reference Pressurized Water Reactor (PWR) 15×15 low enriched uranium (LEU) fresh fuel assembly at LANL. The purpose of this experiment was to assess the following capabilities of SINRD: (1) ability to measure the effective 235U enrichment of the PWR fresh LEU fuel assembly and (2) sensitivity and penetrability to the removal of fuel pins from an assembly. These measurements were compared to Monte Carlo N-Particle eXtended transport code (MCNPX) simulations to verify the accuracy of the MCNPX model of SINRD. The reproducibility of experimental measurements via MCNPX simulations is essential to validating the results and conclusions obtained from the simulations of SINRD for LWR spent fuel assemblies.

Assessing the shielding of engine noise by the wings for current aircraft using model predictions and measurements.

PubMed

Vieira, Ana; Snellen, Mirjam; Simons, Dick G

2018-01-01

Reducing aircraft noise is a major issue to be dealt with by the aerospace industry. In addition to lowering noise emissions from the engine and airframe, also the shielding of engine noise by the aircraft is considered as a promising means for reducing the perceived noise on the ground. In literature, noise shielding predictions indicate significant reductions in received noise levels for blended wing body configurations, but also for conventional aircraft with the engines placed above the wings. Little work has been done in assessing these potential shielding effects for full aircraft under real operational conditions. Therefore, in this work, noise shielding for current aircraft is investigated using both measurements and model predictions. The predictions are based on the Kirchhoff integral theory and the Modified Theory of Physical Optics. For the comparison between the predictions and measurements, Twenty Fokker 70 flyovers are considered. The data analysis approach for the extraction of shielding levels for aircraft under these operational conditions is presented. Directly under the flight path, the simulations predict an engine noise shielding of 6 dB overall sound pressure level. This is confirmed by some of the flyover data. On average, the measurements show somewhat lower shielding levels.

Self-Shielding Analysis of the Zap-X System

PubMed Central

Schneider, M. Bret; Adler, John R.

2017-01-01

The Zap-X is a self-contained and first-of-its-kind self-shielded therapeutic radiation device dedicated to brain as well as head and neck stereotactic radiosurgery (SRS). By utilizing an S-band linear accelerator (linac) with a 2.7 megavolt (MV) accelerating potential and incorporating radiation-shielded mechanical structures, the Zap-X does not typically require a radiation bunker, thereby saving SRS facilities considerable cost. At the same time, the self-shielded features of the Zap-X are designed for more consistency of radiation protection, reducing the risk to radiation workers and others potentially exposed from a poorly designed or constructed radiotherapy vault. The hypothesis of the present study is that a radiosurgical system can be self-shielded such that it produces radiation exposure levels deemed safe to the public while operating under a full clinical workload. This study summarizes the Zap-X system shielding and found that the overall system radiation leakage values are reduced by a factor of 50 compared to the occupational radiation limit stipulated by the Nuclear Regulatory Commission (NRC) or agreement states. The goal of self-shielding is achieved under all but the most exceptional conditions for which additional room shielding or a larger restricted area in the vicinity of the Zap-X system would be required. PMID:29441251

SCALE: A modular code system for performing standardized computer analyses for licensing evaluation. Functional modules F1--F8 -- Volume 2, Part 1, Revision 4

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greene, N.M.; Petrie, L.M.; Westfall, R.M.

SCALE--a modular code system for Standardized Computer Analyses Licensing Evaluation--has been developed by Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission. The SCALE system utilizes well-established computer codes and methods within standard analysis sequences that (1) allow an input format designed for the occasional user and/or novice, (2) automate the data processing and coupling between modules, and (3) provide accurate and reliable results. System development has been directed at problem-dependent cross-section processing and analysis of criticality safety, shielding, heat transfer, and depletion/decay problems. Since the initial release of SCALE in 1980, the code system hasmore » been heavily used for evaluation of nuclear fuel facility and package designs. This revision documents Version 4.2 of the system. The manual is divided into three volumes: Volume 1--for the control module documentation; Volume 2--for functional module documentation; and Volume 3--for documentation of the data libraries and subroutine libraries.« less

In-Plane Shielding for CT: Effect of Off-Centering, Automatic Exposure Control and Shield-to-Surface Distance

PubMed Central

Dang, Pragya; Singh, Sarabjeet; Saini, Sanjay; Shepard, Jo-Anne O.

2009-01-01

Objective To assess effects of off-centering, automatic exposure control, and padding on attenuation values, noise, and radiation dose when using in-plane bismuth-based shields for CT scanning. Materials and Methods A 30 cm anthropomorphic chest phantom was scanned on a 64-multidetector CT, with the center of the phantom aligned to the gantry isocenter. Scanning was repeated after placing a bismuth breast shield on the anterior surface with no gap and with 1, 2, and 6 cm of padding between the shield and the phantom surface. The "shielded" phantom was also scanned with combined modulation and off-centering of the phantom at 2 cm, 4 cm and 6 cm below the gantry isocenter. CT numbers, noise, and surface radiation dose were measured. The data were analyzed using an analysis of variance. Results The in-plane shield was not associated with any significant increment for the surface dose or CT dose index volume, which was achieved by comparing the radiation dose measured by combined modulation technique to the fixed mAs (p > 0.05). Irrespective of the gap or the surface CT numbers, surface noise increased to a larger extent compared to Hounsfield unit (HU) (0-6 cm, 26-55%) and noise (0-6 cm, 30-40%) in the center. With off-centering, in-plane shielding devices are associated with less dose savings, although dose reduction was still higher than in the absence of shielding (0 cm off-center, 90% dose reduction; 2 cm, 61%) (p < 0.0001). Streak artifacts were noted at 0 cm and 1 cm gaps but not at 2 cm and 6 cm gaps of shielding to the surface distances. Conclusion In-plane shields are associated with greater image noise, artifactually increased attenuation values, and streak artifacts. However, shields reduce radiation dose regardless of the extent of off-centering. Automatic exposure control did not increase radiation dose when using a shield. PMID:19270862

77 FR 67678 - Content Specifications and Shielding Evaluations for Type B Transportation Packages

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-13

...The U.S. Nuclear Regulatory Commission (NRC or the Commission) is issuing for public comment Draft Regulatory Issue Summary (RIS) 2012-XX, ``Content Specifications and Shielding Evaluations for Type B Transportation Packages.'' This RIS clarifies the NRC's use of staff guidance in NUREG-1609, ``Standard Review Plan for Transport Packages for Radioactive Material,'' for the review of content specifications and shielding evaluations included in the Certificates of Compliance (CoC) and safety analysis reports (SARs) for Type B transportation packages.

Army Pacific Pathways: Comprehensive Assessment and Planning Needed to Capture Benefits Relative to Costs and Enhance Value for Participating Units

DTIC Science & Technology

2016-11-01

Strike ( Malaysia ) Garuda Shield (Indonesia) Orient Shield (Japan) 15-1 Jan. – May 2015 2-25 Stryker BCT ~880 personnel Cobra Gold (Thailand) Foal...Keris Strike ( Malaysia ) 15-3 June-July, Aug. – Nov. 2015 1-25 Stryker BCT ~420 personnel Khan Quest (Mongolia) Orient Shield (Japan) Hoguk...Strike ( Malaysia ) Source: GAO analysis of U.S. Army Pacific data. | GAO-17-126 aIn addition to the brigades identified here, other units and personnel

Analysis methods for Kevlar shield response to rotor fragments

NASA Technical Reports Server (NTRS)

Gerstle, J. H.

1977-01-01

Several empirical and analytical approaches to rotor burst shield sizing are compared and principal differences in metal and fabric dynamic behavior are discussed. The application of transient structural response computer programs to predict Kevlar containment limits is described. For preliminary shield sizing, present analytical methods are useful if insufficient test data for empirical modeling are available. To provide other information useful for engineering design, analytical methods require further developments in material characterization, failure criteria, loads definition, and post-impact fragment trajectory prediction.

Multi-physics design and analyses of long life reactors for lunar outposts

NASA Astrophysics Data System (ADS)

Schriener, Timothy M.

Future human exploration of the solar system is likely to include establishing permanent outposts on the surface of the Moon. These outposts will require reliable sources of electrical power in the range of 10's to 100's of kWe to support exploration and resource utilization activities. This need is best met using nuclear reactor power systems which can operate steadily throughout the long ˜27.3 day lunar rotational period, irrespective of location. Nuclear power systems can potentially open up the entire lunar surface for future exploration and development. Desirable features of nuclear power systems for the lunar surface include passive operation, the avoidance of single point failures in reactor cooling and the integrated power system, moderate operating temperatures to enable the use of conventional materials with proven irradiation experience, utilization of the lunar regolith for radiation shielding and as a supplemental neutron reflector, and safe post-operation decay heat removal and storage for potential retrieval. In addition, it is desirable for the reactor to have a long operational life. Only a limited number of space nuclear reactor concepts have previously been developed for the lunar environment, and these designs possess only a few of these desirable design and operation features. The objective of this research is therefore to perform design and analyses of long operational life lunar reactors and power systems which incorporate the desirable features listed above. A long reactor operational life could be achieved either by increasing the amount of highly enriched uranium (HEU) fuel in the core or by improving the neutron economy in the reactor through reducing neutron leakage and parasitic absorption. The amount of fuel in surface power reactors is constrained by the launch safety requirements. These include ensuring that the bare reactor core remains safely subcritical when submerged in water or wet sand and flooded with seawater in the unlikely event of a launch abort accident. Increasing the amount of fuel in the reactor core, and hence its operational life, would be possible by launching the reactor unfueled and fueling it on the Moon. Such a reactor would, thus, not be subject to launch criticality safety requirements. However, loading the reactor with fuel on the Moon presents a challenge, requiring special designs of the core and the fuel elements, which lend themselves to fueling on the lunar surface. This research investigates examples of both a solid core reactor that would be fueled at launch as well as an advanced concept which could be fueled on the Moon. Increasing the operational life of a reactor fueled at launch is exercised for the NaK-78 cooled Sectored Compact Reactor (SCoRe). A multi-physics design and analyses methodology is developed which iteratively couples together detailed Monte Carlo neutronics simulations with 3-D Computational Fluid Dynamics (CFD) and thermal-hydraulics analyses. Using this methodology the operational life of this compact, fast spectrum reactor is increased by reconfiguring the core geometry to reduce neutron leakage and parasitic absorption, for the same amount of HEU in the core, and meeting launch safety requirements. The multi-physics analyses determine the impacts of the various design changes on the reactor's neutronics and thermal-hydraulics performance. The option of increasing the operational life of a reactor by loading it on the Moon is exercised for the Pellet Bed Reactor (PeBR). The PeBR uses spherical fuel pellets and is cooled by He-Xe gas, allowing the reactor core to be loaded with fuel pellets and charged with working fluid on the lunar surface. The performed neutronics analyses ensure the PeBR design achieves a long operational life, and develops safe launch canister designs to transport the spherical fuel pellets to the lunar surface. The research also investigates loading the PeBR core with fuel pellets on the Moon using a transient Discrete Element Method (DEM) analysis in lunar gravity. In addition, this research addresses the post-operation storage of the SCoRe and PeBR concepts, below the lunar surface, to determine the time required for the radioactivity in the used fuel to decrease to a low level to allow for its safe recovery. The SCoRe and PeBR concepts are designed to operate at coolant temperatures ≤ 900 K and use conventional stainless steels and superalloys for the structure in the reactor core and power system. They are emplaced below grade on the Moon to take advantage of the regolith as a supplemental neutron reflector and as shielding of the lunar outpost from the reactors' neutron and gamma radiation.

Thermal investigation of nuclear waste disposal in space

NASA Technical Reports Server (NTRS)

Wilkinson, C. L.

1981-01-01

A thermal analysis has been conducted to determine the allowable size and response of bare and shielded nuclear waste forms in both low earth orbit and at 0.85 astronomical units. Contingency conditions of re-entry with a 45 deg and 60 deg aeroshell are examined as well as re-entry of a spherical shielded waste form. A variety of shielded schemes were examined and the waste form thermal response for each determined. Two optimum configurations were selected. The thermal response of these two shielded waste configurations to indefinite exposure to ground conditions following controlled and uncontrolled re-entry is determined. In all cases the prime criterion is that waste containment must be maintained.

An investigation of MAGSAT and complementary data emphasizing precambrian shields and adjacent areas of West Africa and South America

NASA Technical Reports Server (NTRS)

Hastings, D. A. (Principal Investigator)

1981-01-01

Accomplishments with regard to the mapping and analysis of MAGSAT data for the investigation of correlations between the magnetic field characteristics of South American and African shields are reported. Significant results in the interpretation of the global total-field anomalies and the anomaly patterns of Africa and South America are discussed. The central position of the Brazilian shield tends to form a negative total-field anomaly, consistent with findings for shields in equatorial Africa. Sedimentary sequences in the Amazon basin and in the Rio de Janeiro-Sao Paolo areas exhibit positive anomalies, also consistent with equatorial Africa. Results for the Caribbean Sea and Guyana regions are also described.

Study on textile comfort properties of polypropylene blended stainless steel woven fabric for the application of electromagnetic shielding effectiveness

NASA Astrophysics Data System (ADS)

Palanisamy, S.; Tunakova, V.; Karthik, D.; Ali, A.; Militky, J.

2017-10-01

In this study, the different proportion of conductive component blended with polypropylene yarn were taken for making conductive textile samples for analysis of electromagnetic shielding effectiveness, fabric bending moment and air permeability. The ASTM D4935 coaxial transmission line method was used to study the electromagnetic shielding. Electromagnetic shielding effectiveness of textile structures containing different percentage of metal content ranges from 1 to 50 dB at high frequency range. Breathability of structures, more precisely air permeability was considered as one of important parameters for designing of electromagnetic radiation protective fabrics for certain applications. The bending moment of samples is decreases with increasing metal component percent.

Impact of Spacecraft Shielding on Direct Ionization Soft Error Rates for sub-130 nm Technologies

NASA Technical Reports Server (NTRS)

Pellish, Jonathan A.; Xapsos, Michael A.; Stauffer, Craig A.; Jordan, Michael M.; Sanders, Anthony B.; Ladbury, Raymond L.; Oldham, Timothy R.; Marshall, Paul W.; Heidel, David F.; Rodbell, Kenneth P.

2010-01-01

We use ray tracing software to model various levels of spacecraft shielding complexity and energy deposition pulse height analysis to study how it affects the direct ionization soft error rate of microelectronic components in space. The analysis incorporates the galactic cosmic ray background, trapped proton, and solar heavy ion environments as well as the October 1989 and July 2000 solar particle events.

Impact of Spacecraft Shielding on Direct Ionization Soft Error Rates for Sub-130 nm Technologies

NASA Technical Reports Server (NTRS)

Pellish, Jonathan A.; Xapsos, Michael A.; Stauffer, Craig A.; Jordan, Thomas M.; Sanders, Anthony B.; Ladbury, Raymond L.; Oldham, Timothy R.; Marshall, Paul W.; Heidel, David F.; Rodbell, Kenneth P.

2010-01-01

We use ray tracing software to model various levels of spacecraft shielding complexity and energy deposition pulse height analysis to study how it affects the direct ionization soft error rate of microelectronic components in space. The analysis incorporates the galactic cosmic ray background, trapped proton, and solar heavy ion environments as well as the October 1989 and July 2000 solar particle events.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ketusky, E.

The U.S. Department of Energy’s (DOE’s) Used Fuel Disposition Campaign (UFDC) Program has transported high-burnup nuclear sister fuel rods from a commercial nuclear power plant for purposes of evaluation and testing. The evaluation and testing of high-burnup used nuclear fuel is integral to DOE initiatives to collect information useful in determining the integrity of fuel cladding for future safe transportation of the fuel, and for determining the effects of aging, on the integrity of UNF subjected to extended storage and subsequent transportation. The UFDC Program, in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry, has obtainedmore » individual used nuclear fuel rods for testing. The rods have been received at Oak Ridge National Laboratory (ORNL) for both separate effects testing (SET) and small-scale testing (SST). To meet the research objectives, testing on multiple 6 inch fuel rod pins cut from the rods at ORNL will be performed at Pacific Northwest National Laboratory (PNNL). Up to 10 rod equivalents will be shipped. Options were evaluated for multiple shipments using the 10-160B (based on 4.5 rod equivalents) and a single shipment using the NAC-LWT. Based on the original INL/Virginia Power transfer agreement, the rods are assumed to 152 inches in length with a 0.374-inch diameter. This report provides a preliminary content evaluation for use of the 10-160B and NAC-LWT for transporting those fuel rod pins from ORNL to PNNL. This report documents the acceptability of using these packagings to transport the fuel segments from ORNL to PNNL based on the following evaluations: enrichment, A2 evaluation, Pu-239 FGE evaluation, heat load, shielding (both gamma and neutron), and content weight/structural evaluation.« less

Morphometric comparison of Icelandic lava shield volcanoes versus selected Venusian edifices

NASA Technical Reports Server (NTRS)

Garvin, James B.; Williams, Richard S., Jr.

1993-01-01

Shield volcanoes are common landforms on the silicate planets of the inner Solar System, and a wide variety have recently been documented on Venus by means of Magellan observations. In this report, we emphasize our recently completed morphometric analysis of three representative Icelandic lava shields: the classic Skjaldbreidur edifice, the low-reflief Lambahraun feature, and the monogenetic Sandfellshaed shield, as the basis for comparison with representative venusian edifices (greater than 60 km in diameter). Our detailed morphometric measurements of a representative and well-studied set of Icelandic volcanoes permits us to make comparisons with our measurements of a reasonable subset of shield-like edifices on Venus on the basis of Magellan global radar altimetry. Our study has been restricted to venusian features larger than approximately 60 km in basal diameter, on the basis of the minimum intrinsic spatial resolution (8 km) of the Magellan radar altimetry data. Finally, in order to examine the implications of landform scaling from terrestrial simple and composite shields to larger venusian varieties, we have considered the morphometry of the subaerial component of Mauna Loa, a type-locality for a composite shield edifice on Earth.

Radiation Exposure Analyses Supporting the Development of Solar Particle Event Shielding Technologies

NASA Technical Reports Server (NTRS)

Walker, Steven A.; Clowdsley, Martha S.; Abston, H. Lee; Simon, Hatthew A.; Gallegos, Adam M.

2013-01-01

NASA has plans for long duration missions beyond low Earth orbit (LEO). Outside of LEO, large solar particle events (SPEs), which occur sporadically, can deliver a very large dose in a short amount of time. The relatively low proton energies make SPE shielding practical, and the possibility of the occurrence of a large event drives the need for SPE shielding for all deep space missions. The Advanced Exploration Systems (AES) RadWorks Storm Shelter Team was charged with developing minimal mass SPE storm shelter concepts for missions beyond LEO. The concepts developed included "wearable" shields, shelters that could be deployed at the onset of an event, and augmentations to the crew quarters. The radiation transport codes, human body models, and vehicle geometry tools contained in the On-Line Tool for the Assessment of Radiation In Space (OLTARIS) were used to evaluate the protection provided by each concept within a realistic space habitat and provide the concept designers with shield thickness requirements. Several different SPE models were utilized to examine the dependence of the shield requirements on the event spectrum. This paper describes the radiation analysis methods and the results of these analyses for several of the shielding concepts.

3D Space Radiation Transport in a Shielded ICRU Tissue Sphere

NASA Technical Reports Server (NTRS)

Wilson, John W.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.

2014-01-01

A computationally efficient 3DHZETRN code capable of simulating High Charge (Z) and Energy (HZE) and light ions (including neutrons) under space-like boundary conditions with enhanced neutron and light ion propagation was recently developed for a simple homogeneous shield object. Monte Carlo benchmarks were used to verify the methodology in slab and spherical geometry, and the 3D corrections were shown to provide significant improvement over the straight-ahead approximation in some cases. In the present report, the new algorithms with well-defined convergence criteria are extended to inhomogeneous media within a shielded tissue slab and a shielded tissue sphere and tested against Monte Carlo simulation to verify the solution methods. The 3D corrections are again found to more accurately describe the neutron and light ion fluence spectra as compared to the straight-ahead approximation. These computationally efficient methods provide a basis for software capable of space shield analysis and optimization.

Tunable Protein Stabilization In Vivo Mediated by Shield-1 in Transgenic Medaka

PubMed Central

Froschauer, Alexander; Kube, Lisa; Kegler, Alexandra; Rieger, Christiane; Gutzeit, Herwig O.

2015-01-01

Techniques for conditional gene or protein expression are important tools in developmental biology and in the analysis of physiology and disease. On the protein level, the tunable and reversible expression of proteins can be achieved by the fusion of the protein of interest to a destabilizing domain (DD). In the absence of its specific ligand (Shield-1), the protein is degraded by the proteasome. The DD-Shield system has proven to be an excellent tool to regulate the expression of proteins of interests in mammalian systems but has not been applied in teleosts like the medaka. We present the application of the DD-Shield technique in transgenic medaka and show the ubiquitous conditional expression throughout life. Shield-1 administration to the water leads to concentration-dependent induction of a YFP reporter gene in various organs and in spermatogonia at the cellular level. PMID:26148066

Evaluation of ilmenite serpentine concrete and ordinary concrete as nuclear reactor shielding

NASA Astrophysics Data System (ADS)

Abulfaraj, Waleed H.; Kamal, Salah M.

1994-07-01

The present study involves adapting a formal decision methodology to the selection of alternative nuclear reactor concretes shielding. Multiattribute utility theory is selected to accommodate decision makers' preferences. Multiattribute utility theory (MAU) is here employed to evaluate two appropriate nuclear reactor shielding concretes in terms of effectiveness to determine the optimal choice in order to meet the radiation protection regulations. These concretes are Ordinary concrete (O.C.) and Ilmenite Serpentile concrete (I.S.C.). These are normal weight concrete and heavy heat resistive concrete, respectively. The effectiveness objective of the nuclear reactor shielding is defined and structured into definite attributes and subattributes to evaluate the best alternative. Factors affecting the decision are dose received by reactor's workers, the material properties as well as cost of concrete shield. A computer program is employed to assist in performing utility analysis. Based upon data, the result shows the superiority of Ordinary concrete over Ilmenite Serpentine concrete.

On thermal stress failure of the SNAP-19A RTG heat shield

NASA Technical Reports Server (NTRS)

Pitts, W. C.; Anderson, L. A.

1974-01-01

Results of a study on thermal stress problems in an amorphous graphite heat shield that is part of the launch-abort protect system for the SNAP-19A radio-isotope thermoelectric generators (RTG) that will be used on the Viking Mars Lander are presended. The first result is from a thermal stress analysis of a full-scale RTG heat source that failed to survive a suborbital entry flight test, possibly due to thermal stress failure. It was calculated that the maximum stress in the heat shield was only 50 percent of the ultimate strength of the material. To provide information on the stress failure criterion used for this calculation, some heat shield specimens were fractured under abort entry conditions in a plasma arc facility. It was found that in regions free of stress concentrations the POCO graphite heat shield material did fracture when the local stress reached the ultimate uniaxial stress of the material.

Heat pipe nuclear reactor for space power

NASA Technical Reports Server (NTRS)

Koening, D. R.

1976-01-01

A heat-pipe-cooled nuclear reactor has been designed to provide 3.2 MWth to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat-pipe temperature of 1675 K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum/lithium-vapor heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO2 pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber, and a BeO reflector containing boron-loaded control drums.

Reducing absorbed dose to eye lenses in head CT examinations: the effect of bismuth shielding.

PubMed

Ciarmatori, Alberto; Nocetti, L; Mistretta, G; Zambelli, G; Costi, T

2016-06-01

The eye lens is considered to be among the most radiosensitive human tissues. Brain CT scans may unnecessarily expose it to radiation even if the area of clinical interest is far from the eyes. The aim of this study is to implement a bismuth eye lens shielding system for Head-CT acquisitions in these cases. The study is focused on the assessment of the dosimetric characteristics of the shielding system as well as on its effect on image quality. The shielding system was tested in two set-ups which differ for distance ("contact" and "4 cm" Set up respectively). Scans were performed on a CTDI phantom and an anthropomorphic phantom. A reference set up without shielding system was acquired to establish a baseline. Image quality was assessed by signal (not HU converted), noise and contrast-to-noise ratio (CNR) evaluation. The overall dose reduction was evaluated by measuring the CTDIvol while the eye lens dose reduction was assessed by placing thermoluminescent dosimeters (TLDs) on an anthropomorphic phantom. The image quality analysis exhibits the presence of an artefact that mildly increases the CT number up to 3 cm below the shielding system. Below the artefact, the difference of the Signal and the CNR are negligible between the three different set-ups. Regarding the CTDI, the analysis demonstrates a decrease by almost 12 % (in the "contact" set-up) and 9 % (in the "4 cm" set-up). TLD measurements exhibit an eye lens dose reduction by 28.5 ± 5 and 21.1 ± 5 % respectively at the "contact" and the "4 cm" distance. No relevant artefact was found and image quality was not affected by the shielding system. Significant dose reductions were measured. These features make the shielding set-up useful for clinical implementation in both studied positions.

Lymphatic invasion and the Shields index in predicting melanoma metastases.

PubMed

Špirić, Zorica; Erić, Mirela; Eri, Živka

2017-11-01

Findings of the prognostic significance of lymphatic invasion are contradictory. To determine an as efficient cutaneous melanoma metastasis predictor as possible, Shields et al. created a new prognostic index. This study aimed to examine whether the lymphatic invasion analysis and the Shields index calculation can be used in predicting lymph node status in patients with cutaneous melanoma. Lymphatic invasion of 100 melanoma specimens was detected by dual immunohistochemistry staining for the lymphatic endothelial marker D2-40 and melanoma cell S-100 protein. The Shields index was calculated as a logarithm by multiplying the melanoma thickness, square of peritumoural lymphatic vessel density and the number "2" for the present lymphatic invasion. No statistically significant difference was observed between lymph node metastatic and nonmetastatic melanomas regarding the lymphatic invasion. Metastatic melanomas showed a significantly higher Shields index value than nonmetastatic melanomas (p = 0.00). Area under the receiver operator characteristic (ROC) curve (AUC) proved that the Shields index (AUC = 0.86, 95% confidence interval (CI) 0.79-0.93, p = 0.00) was the most accurate predictor of lymph node status, followed by the melanoma thickness (AUC = 0.76, 95% CI 0.67-0.86, p = 0.00) and American Joint Committee on Cancer (AJCC) staging (AUC = 0.75, 95% CI 0.66-0.85, p = 0.00), while lymphatic invasion was not successful in predicting (AUC = 0.56, 95% CI 0.45-0.67, p = 0.31). The Shields index achieved 81.3% sensitivity and 75% specificity (cut-off mean value). Our findings show that D2-40/S-100 immunohistochemical analysis of lymphatic invasion cannot be used for predicting the lymph node status, while the Shields index calculation predicts disease outcome more accurately than the melanoma thickness and AJCC staging. Copyright © 2017 British Association of Plastic, Reconstructive and Aesthetic Surgeons. Published by Elsevier Ltd. All rights reserved.

Neutron Radiation Shielding For The NIF Streaked X-Ray Detector (SXD) Diagnostic

DOE Office of Scientific and Technical Information (OSTI.GOV)

Song, P; Holder, J; Young, B

2006-11-02

The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory (LLNL) is preparing for the National Ignition Campaign (NIC) scheduled in 2010. The NIC is comprised of several ''tuning'' physics subcampaigns leading up to a demonstration of Inertial Confinement Fusion (ICF) ignition. In some of these experiments, time-resolved x-ray imaging of the imploding capsule may be required to measure capsule trajectory (shock timing) or x-ray ''bang-time''. A capsule fueled with pure tritium (T) instead of a deutriun-tritium (DT) mixture is thought to offer useful physics surrogacy, with reduced yields of up to 5e14 neutrons. These measurements will require the usemore » of the NIF streak x-ray detector (SXD). The resulting prompt neutron fluence at the planned SXD location ({approx}1.7 m from the target) would be {approx}1.4e9/cm{sup 2}. Previous measurements suggest the onset of significant background at a neutron fluence of {approx} 1e8/cm{sup 2}. The radiation damage and operational upsets which starts at {approx}1e8 rad-Si/sec must be factored into an integrated experimental campaign plan. Monte Carlo analyses were performed to predict the neutron and gamma/x-ray fluences and radiation doses for the proposed diagnostic configuration. A possible shielding configuration is proposed to mitigate radiation effects. The primary component of this shielding is an 80 cm thickness of Polyethylene (PE) between target chamber center (TCC) and the SXD diagnostic. Additionally, 6-8 cm of PE around the detector provide from the large number of neutrons that scatter off the inside of the target chamber. This proposed shielding configuration reduces the high-energy neutron fluence at the SXD by approximately a factor {approx}50.« less

Nuclear Rocket Facility Decommissioning Project: Controlled Explosive Demolition of Neutron-Activated Shield Wall

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael R. Kruzic

2008-06-01

Located in Area 25 of the Nevada Test Site (NTS), the Test Cell A (TCA) Facility (Figure 1) was used in the early to mid-1960s for testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program, to further space travel. Nuclear rocket testing resulted in the activation of materials around the reactors and the release of fission products and fuel particles. The TCA facility, known as Corrective Action Unit 115, was decontaminated and decommissioned (D&D) from December 2004 to July 2005 using the Streamlined Approach for Environmental Restoration (SAFER) process, under the Federal Facility Agreement and Consentmore » Order. The SAFER process allows environmental remediation and facility closure activities (i.e., decommissioning) to occur simultaneously, provided technical decisions are made by an experienced decision maker within the site conceptual site model. Facility closure involved a seven-step decommissioning strategy. First, preliminary investigation activities were performed, including review of process knowledge documentation, targeted facility radiological and hazardous material surveys, concrete core drilling and analysis, shield wall radiological characterization, and discrete sampling, which proved to be very useful and cost-effective in subsequent decommissioning planning and execution and worker safety. Second, site setup and mobilization of equipment and personnel were completed. Third, early removal of hazardous materials, including asbestos, lead, cadmium, and oil, was performed ensuring worker safety during more invasive demolition activities. Process piping was to be verified void of contents. Electrical systems were de-energized and other systems were rendered free of residual energy. Fourth, areas of high radiological contamination were decontaminated using multiple methods. Contamination levels varied across the facility. Fixed beta/gamma contamination levels ranged up to 2 million disintegrations per minute (dpm)/100 centimeters squared (cm2) beta/gamma. Removable beta/gamma contamination levels seldom exceeded 1,000 dpm/100 cm2, but, in railroad trenches on the reactor pad containing soil on the concrete pad in front of the shield wall, the beta dose rates ranged up to 120 milli-roentgens per hour from radioactivity entrained in the soil. General area dose rates were less than 100 micro-roentgens per hour. Prior to demolition of the reactor shield wall, removable and fixed contaminated surfaces were decontaminated to the best extent possible, using traditional decontamination methods. Fifth, large sections of the remaining structures were demolished by mechanical and open-air controlled explosive demolition (CED). Mechanical demolition methods included the use of conventional demolition equipment for removal of three main buildings, an exhaust stack, and a mobile shed. The 5-foot (ft), 5-inch (in.) thick, neutron-activated reinforced concrete shield was demolished by CED, which had never been performed at the NTS.« less

DataSHIELD: taking the analysis to the data, not the data to the analysis.

PubMed

Gaye, Amadou; Marcon, Yannick; Isaeva, Julia; LaFlamme, Philippe; Turner, Andrew; Jones, Elinor M; Minion, Joel; Boyd, Andrew W; Newby, Christopher J; Nuotio, Marja-Liisa; Wilson, Rebecca; Butters, Oliver; Murtagh, Barnaby; Demir, Ipek; Doiron, Dany; Giepmans, Lisette; Wallace, Susan E; Budin-Ljøsne, Isabelle; Oliver Schmidt, Carsten; Boffetta, Paolo; Boniol, Mathieu; Bota, Maria; Carter, Kim W; deKlerk, Nick; Dibben, Chris; Francis, Richard W; Hiekkalinna, Tero; Hveem, Kristian; Kvaløy, Kirsti; Millar, Sean; Perry, Ivan J; Peters, Annette; Phillips, Catherine M; Popham, Frank; Raab, Gillian; Reischl, Eva; Sheehan, Nuala; Waldenberger, Melanie; Perola, Markus; van den Heuvel, Edwin; Macleod, John; Knoppers, Bartha M; Stolk, Ronald P; Fortier, Isabel; Harris, Jennifer R; Woffenbuttel, Bruce H R; Murtagh, Madeleine J; Ferretti, Vincent; Burton, Paul R

2014-12-01

Research in modern biomedicine and social science requires sample sizes so large that they can often only be achieved through a pooled co-analysis of data from several studies. But the pooling of information from individuals in a central database that may be queried by researchers raises important ethico-legal questions and can be controversial. In the UK this has been highlighted by recent debate and controversy relating to the UK's proposed 'care.data' initiative, and these issues reflect important societal and professional concerns about privacy, confidentiality and intellectual property. DataSHIELD provides a novel technological solution that can circumvent some of the most basic challenges in facilitating the access of researchers and other healthcare professionals to individual-level data. Commands are sent from a central analysis computer (AC) to several data computers (DCs) storing the data to be co-analysed. The data sets are analysed simultaneously but in parallel. The separate parallelized analyses are linked by non-disclosive summary statistics and commands transmitted back and forth between the DCs and the AC. This paper describes the technical implementation of DataSHIELD using a modified R statistical environment linked to an Opal database deployed behind the computer firewall of each DC. Analysis is controlled through a standard R environment at the AC. Based on this Opal/R implementation, DataSHIELD is currently used by the Healthy Obese Project and the Environmental Core Project (BioSHaRE-EU) for the federated analysis of 10 data sets across eight European countries, and this illustrates the opportunities and challenges presented by the DataSHIELD approach. DataSHIELD facilitates important research in settings where: (i) a co-analysis of individual-level data from several studies is scientifically necessary but governance restrictions prohibit the release or sharing of some of the required data, and/or render data access unacceptably slow; (ii) a research group (e.g. in a developing nation) is particularly vulnerable to loss of intellectual property-the researchers want to fully share the information held in their data with national and international collaborators, but do not wish to hand over the physical data themselves; and (iii) a data set is to be included in an individual-level co-analysis but the physical size of the data precludes direct transfer to a new site for analysis. © The Author 2014; all rights reserved. Published by Oxford University Press on behalf of the International Epidemiological Association.

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

NASA Technical Reports Server (NTRS)

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

2015-01-01

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

Effect of Alternate Supply of Shielding Gases of Tungsten Inert Gas Welding on Mechanical Properties of Austenitic Stainless Steel

NASA Astrophysics Data System (ADS)

Shinde, Neelam Vilas; Telsang, Martand Tamanacharya

2016-07-01

In the present study, an attempt is made to study the effect of alternate supply of the shielding gas in comparison with the conventional method of TIG welding with pure argon gas. The two sets of combination are used as 10-10 and 40-20 s for alternate supply of the Argon and Helium shielding gas respectively. The effect of alternate supply of shielding gas is studied on the mechanical properties like bend test, tensile test and impact test. The full factorial experimental design is applied for three set of combinations. The ANOVA is used to find significant parameters for the process and regression analysis used to develop the mathematical model. The result shows that the alternate supply of the shielding gas for 10-10 s provides better result for the bend, tensile and impact test as compared with the conventional argon gas and the alternate supply of 40-20 s argon and helium gas respectively. Welding speed can be increased for alternate supply of the shielding gas that can reduce the total welding cost.

Synthesis of calculational methods for design and analysis of radiation shields for nuclear rocket systems

NASA Technical Reports Server (NTRS)

Capo, M. A.; Disney, R. K.; Jordan, T. A.; Soltesz, R. G.; Woodsum, H. C.

1969-01-01

Eight computer programs make up a nine volume synthesis containing two design methods for nuclear rocket radiation shields. The first design method is appropriate for parametric and preliminary studies, while the second accomplishes the verification of a final nuclear rocket reactor design.

Stagnation-Point Shielding by Melting and Vaporization

NASA Technical Reports Server (NTRS)

Roberts, Leonard

1959-01-01

An approximate theoretical analysis was made of the shielding mechanism whereby the rate of heat transfer to the forward stagnation point of blunt bodies is reduced by melting and evaporation. General qualitative results are given and a numerical example, the melting and evaporation of ice, is presented and discussed in detail.

Theoretical analysis and experimental investigation on performance of the thermal shield of accelerator cryomodules by thermo-siphon cooling of liquid nitrogen

NASA Astrophysics Data System (ADS)

Datta, T. S.; Kar, S.; Kumar, M.; Choudhury, A.; Chacko, J.; Antony, J.; Babu, S.; Sahu, S. K.

2015-12-01

Five beam line cryomodules with total 27 superconducting Radio Frequency (RF) cavities are installed and commissioned at IUAC to enhance the energy of heavy ion from 15 UD Pelletron. To reduce the heat load at 4.2 K, liquid nitrogen (LN2) cooled intermediate thermal shield is used for all these cryomodules. For three linac cryomodules, concept of forced flow LN2 cooling is used and for superbuncher and rebuncher, thermo-siphon cooling is incorporated. It is noticed that the shield temperature of superbuncher varies from 90 K to 110 K with respect to liquid nitrogen level. The temperature difference can't be explained by using the basic concept of thermo-siphon with the heat load on up flow line. A simple thermo-siphon experimental set up is developed to simulate the thermal shield temperature profile. Mass flow rate of liquid nitrogen is measured with different heat load on up flow line for different liquid levels. It is noticed that small amount of heat load on down flow line have a significant effect on mass flow rate. The present paper will be investigating the data generated from the thermosiphon experimental set up and a theoretical analysis will be presented here to validate the measured temperature profile of the cryomodule shield.

Multiplate Radiation Shields: Investigating Radiational Heating Errors

NASA Astrophysics Data System (ADS)

Richardson, Scott James

1995-01-01

Multiplate radiation shield errors are examined using the following techniques: (1) analytic heat transfer analysis, (2) optical ray tracing, (3) numerical fluid flow modeling, (4) laboratory testing, (5) wind tunnel testing, and (6) field testing. Guidelines for reducing radiational heating errors are given that are based on knowledge of the temperature sensor to be used, with the shield being chosen to match the sensor design. Small, reflective sensors that are exposed directly to the air stream (not inside a filter as is the case for many temperature and relative humidity probes) should be housed in a shield that provides ample mechanical and rain protection while impeding the air flow as little as possible; protection from radiation sources is of secondary importance. If a sensor does not meet the above criteria (i.e., is large or absorbing), then a standard Gill shield performs reasonably well. A new class of shields, called part-time aspirated multiplate radiation shields, are introduced. This type of shield consists of a multiplate design usually operated in a passive manner but equipped with a fan-forced aspiration capability to be used when necessary (e.g., low wind speed). The fans used here are 12 V DC that can be operated with a small dedicated solar panel. This feature allows the fan to operate when global solar radiation is high, which is when the largest radiational heating errors usually occur. A prototype shield was constructed and field tested and an example is given in which radiational heating errors were reduced from 2 ^circC to 1.2 ^circC. The fan was run continuously to investigate night-time low wind speed errors and the prototype shield reduced errors from 1.6 ^ circC to 0.3 ^circC. Part-time aspirated shields are an inexpensive alternative to fully aspirated shields and represent a good compromise between cost, power consumption, reliability (because they should be no worse than a standard multiplate shield if the fan fails), and accuracy. In addition, it is possible to modify existing passive shields to incorporate part-time aspiration, thus making them even more cost-effective. Finally, a new shield is described that incorporates a large diameter top plate that is designed to shade the lower portion of the shield. This shield increases flow through it by 60%, compared to the Gill design and it is likely to reduce radiational heating errors, although it has not been tested.

Why do mothers use nipple shields and how does this influence duration of exclusive breastfeeding?

PubMed

Kronborg, Hanne; Foverskov, Else; Nilsson, Ingrid; Maastrup, Ragnhild

2017-01-01

The present study addressed the contentious discussions about the benefits and risks of nipple shield use. The objective was to explore self-reported reasons for using a nipple shield and examine associations pertaining to the mother, the infant and duration of breastfeeding. Data were collected from 4815 Danish mothers (68%) who filled out a self-administered questionnaire with open and closed question. Data were analyzed by content and statistical descriptive and multivariable analysis. Results showed that 22% of the mothers used nipple shields in the beginning and 7% used it the entire breastfeeding period. Primiparae used nipple shields more often than multiparae, and early breastfeeding problems as well as background factors like lower age, education and higher body mass index were associated with a higher likelihood of using nipple shields. Characteristics of infants associated with introducing nipple shields were lower- gestational age and birthweight. The use of nipple shields was furthermore found to be associated with a threefold increased risk of earlier cessation of exclusive breastfeeding: among primiparae odds ratio = 3.80 (confidence interval 2.61-5.53); among multiparae odds ratio = 3.33 (confidence interval 1.88-5.93). Mothers' own descriptions underlined how various early breastfeeding problems led to the use of nipple shields. Some mothers were helped through a difficult period; others described the use creating a kind of dependence. The results highlight how nipple shields may help breastfeeding mothers in the early period but is not necessarily a supportive solution to the inexperienced mother who needs extra support in the early process of learning to breastfeed. © 2016 John Wiley & Sons Ltd.

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

NASA Astrophysics Data System (ADS)

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.

Monte Carlo analysis of a time-dependent neutron and secondary gamma-ray integral experiment on a thick concrete and steel shield

NASA Astrophysics Data System (ADS)

Cramer, S. N.; Roussin, R. W.

1981-11-01

A Monte Carlo analysis of a time-dependent neutron and secondary gamma-ray integral experiment on a thick concrete and steel shield is presented. The energy range covered in the analysis is 15-2 MeV for neutron source energies. The multigroup MORSE code was used with the VITAMIN C 171-36 neutron-gamma-ray cross-section data set. Both neutron and gamma-ray count rates and unfolded energy spectra are presented and compared, with good general agreement, with experimental results.

Historical perspectives - The role of the NASA Lewis Research Center in the national space nuclear power programs

NASA Technical Reports Server (NTRS)

Bloomfield, H. S.; Sovie, R. J.

1991-01-01

The history of the NASA Lewis Research Center's role in space nuclear power programs is reviewed. Lewis has provided leadership in research, development, and the advancement of space power and propulsion systems. Lewis' pioneering efforts in nuclear reactor technology, shielding, high temperature materials, fluid dynamics, heat transfer, mechanical and direct energy conversion, high-energy propellants, electric propulsion and high performance rocket fuels and nozzles have led to significant technical and management roles in many natural space nuclear power and propulsion programs.

Thermoelectric generator having a resiliently mounted removable thermoelectric module

DOEpatents

Purdy, David L.; Shapiro, Zalman M.; Hursen, Thomas F.; Maurer, Gerould W.

1976-11-02

An electrical generator having an Isotopic Heat Capsule including radioactive fuel rod 21 as a primary heat source and Thermoelectric Modules 41 and 43 as converters. The Biological Shield for the Capsule is suspended from Spiders at each end each consisting of pretensioned rods 237 and 239 defining planes at right angles to each other. The Modules are mounted in cups 171 of transition members 173 of a heat rejection Fin Assembly whose fins 195 and 197 extend from both sides of the transition member 173 for effective cooling.

Historical perspectives: The role of the NASA Lewis Research Center in the national space nuclear power programs

NASA Technical Reports Server (NTRS)

Bloomfield, H. S.; Sovie, R. J.

1991-01-01

The history of the NASA Lewis Research Center's role in space nuclear power programs is reviewed. Lewis has provided leadership in research, development, and the advancement of space power and propulsion systems. Lewis' pioneering efforts in nuclear reactor technology, shielding, high temperature materials, fluid dynamics, heat transfer, mechanical and direct energy conversion, high-energy propellants, electric propulsion and high performance rocket fuels and nozzles have led to significant technical and management roles in many national space nuclear power and propulsion programs.

78 FR 26090 - Content Specifications and Shielding Evaluations for Type B Transportation Packages

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-03

...The U.S. Nuclear Regulatory Commission (NRC) is issuing Regulatory Issue Summary (RIS) 2013-04, ``Content Specifications and Shielding Evaluations for Type B Transportation Packages.'' This RIS clarifies the NRC's use of staff guidance in NUREG-1609, ``Standard Review Plan for Transport Packages for Radioactive Material,'' for the review of content specifications and shielding evaluations included in the Certificates of Compliance (CoC) and safety analysis reports (SARs) for Type B transportation packages. The RIS does not impose any additional regulatory requirements on NRC licensees.

Monte Carlo simulations of safeguards neutron counter for oxide reduction process feed material

NASA Astrophysics Data System (ADS)

Seo, Hee; Lee, Chaehun; Oh, Jong-Myeong; An, Su Jung; Ahn, Seong-Kyu; Park, Se-Hwan; Ku, Jeong-Hoe

2016-10-01

One of the options for spent-fuel management in Korea is pyroprocessing whose main process flow is the head-end process followed by oxide reduction, electrorefining, and electrowining. In the present study, a well-type passive neutron coincidence counter, namely, the ACP (Advanced spent fuel Conditioning Process) safeguards neutron counter (ASNC), was redesigned for safeguards of a hot-cell facility related to the oxide reduction process. To this end, first, the isotopic composition, gamma/neutron emission yield and energy spectrum of the feed material ( i.e., the UO2 porous pellet) were calculated using the OrigenARP code. Then, the proper thickness of the gammaray shield was determined, both by irradiation testing at a standard dosimetry laboratory and by MCNP6 simulations using the parameters obtained from the OrigenARP calculation. Finally, the neutron coincidence counter's calibration curve for 100- to 1000-g porous pellets, in consideration of the process batch size, was determined through simulations. Based on these simulation results, the neutron counter currently is under construction. In the near future, it will be installed in a hot cell and tested with spent fuel materials.

Recalculation with SEACAB of the activation by spent fuel neutrons and residual dose originated in the racks replaced at Cofrentes NPP

NASA Astrophysics Data System (ADS)

Ortego, Pedro; Rodriguez, Alain; Töre, Candan; Compadre, José Luis de Diego; Quesada, Baltasar Rodriguez; Moreno, Raul Orive

2017-09-01

In order to increase the storage capacity of the East Spent Fuel Pool at the Cofrentes NPP, located in Valencia province, Spain, the existing storage stainless steel racks were replaced by a new design of compact borated stainless steel racks allowing a 65% increase in fuel storing capacity. Calculation of the activation of the used racks was successfully performed with the use of MCNP4B code. Additionally the dose rate at contact with a row of racks in standing position and behind a wall of shielding material has been calculated using MCNP4B code as well. These results allowed a preliminary definition of the burnker required for the storage of racks. Recently the activity in the racks has been recalculated with SEACAB system which combines the mesh tally of MCNP codes with the activation code ACAB, applying the rigorous two-step method (R2S) developed at home, benchmarked with FNG irradiation experiments and usually applied in fusion calculations for ITER project.

Beta radiation shielding with lead and plastic: effect on bremsstrahlung radiation when switching the shielding order.

PubMed

Van Pelt, Wesley R; Drzyzga, Michael

2007-02-01

Lead and plastic are commonly used to shield beta radiation. Radiation protection literature is ubiquitous in advising the placement of plastic first to absorb all the beta particles before any lead shielding is used. This advice is based on the well established theory that radiative losses (bremsstrahlung production) are more prevalent in higher atomic number (Z) materials than in low Z materials. Using 32P beta radiation, we measured bremsstrahlung photons transmitted through lead and plastic (Lucite) shielding in different test configurations to determine the relative efficacy of lead alone, plastic alone, and the positional order of lead and plastic. With the source (32P) and detector held at a constant separation distance, we inserted lead and/or plastic absorbers and measured the reduction in bremsstrahlung radiation level measured by the detector. With these test conditions, analysis of measured bremsstrahlung radiation in various thicknesses and configurations of lead and plastic shielding shows the following: placing plastic first vs. lead first reduces the transmitted radiation level only marginally (10% to 40%); 2 mm of additional lead is sufficient to correct the "mistake" of placing the lead first; and for equal thicknesses or weights of lead and plastic, lead is a more efficient radiation shield than plastic.

Analysis of space radiation exposure levels at different shielding configurations by ray-tracing dose estimation method

NASA Astrophysics Data System (ADS)

Kartashov, Dmitry; Shurshakov, Vyacheslav

2018-03-01

A ray-tracing method to calculate radiation exposure levels of astronauts at different spacecraft shielding configurations has been developed. The method uses simplified shielding geometry models of the spacecraft compartments together with depth-dose curves. The depth-dose curves can be obtained with different space radiation environment models and radiation transport codes. The spacecraft shielding configurations are described by a set of geometry objects. To calculate the shielding probability functions for each object its surface is composed from a set of the disjoint adjacent triangles that fully cover the surface. Such description can be applied for any complex shape objects. The method is applied to the space experiment MATROSHKA-R modeling conditions. The experiment has been carried out onboard the ISS from 2004 to 2016. Dose measurements were realized in the ISS compartments with anthropomorphic and spherical phantoms, and the protective curtain facility that provides an additional shielding on the crew cabin wall. The space ionizing radiation dose distributions in tissue-equivalent spherical and anthropomorphic phantoms and for an additional shielding installed in the compartment are calculated. There is agreement within accuracy of about 15% between the data obtained in the experiment and calculated ones. Thus the calculation method used has been successfully verified with the MATROSHKA-R experiment data. The ray-tracing radiation dose calculation method can be recommended for estimation of dose distribution in astronaut body in different space station compartments and for estimation of the additional shielding efficiency, especially when exact compartment shielding geometry and the radiation environment for the planned mission are not known.

Orion Heat Shield Testing

NASA Image and Video Library

2015-05-26

THE ORION HEAT SHIELD, WHICH WAS AT NASA’S MARSHALL SPACE FLIGHT CENTER FROM MARCH-MAY 2015 FOR ENGINEERING AND ANALYSIS, IS READIED FOR DEPARTURE AT THE END OF ITS STAY. THE HEAT SHIELD’S ABLATED SURFACE MATERIAL WAS REMOVED AT MARSHALL FOR ANALYSIS, USING THE CENTER’S STATE-OF-THE-ART SEVEN-AXIS MILLING MACHINE. IT NEXT WILL GO TO NASA’S LANGLEY RESEARCH CENTER FOR WATER-IMPACT TESTING. NASA’S JOHNSON SPACE CENTER LEADS THE ORION PROGRAM FOR NASA.

Orion Heat Shield Testing

NASA Image and Video Library

2015-05-28

THE ORION HEAT SHIELD, WHICH WAS AT NASA’S MARSHALL SPACE FLIGHT CENTER FROM MARCH-MAY 2015 FOR ENGINEERING AND ANALYSIS, IS READIED FOR DEPARTURE AT THE END OF ITS STAY. THE HEAT SHIELD’S ABLATED SURFACE MATERIAL WAS REMOVED AT MARSHALL FOR ANALYSIS, USING THE CENTER’S STATE-OF-THE-ART SEVEN-AXIS MILLING MACHINE. IT NEXT WILL GO TO NASA’S LANGLEY RESEARCH CENTER FOR WATER-IMPACT TESTING. NASA’S JOHNSON SPACE CENTER LEADS THE ORION PROGRAM FOR NASA.

Radiation Exposure Effects and Shielding Analysis of Carbon Nanotube Materials

NASA Technical Reports Server (NTRS)

Wilkins, Richard; Armendariz, Lupita (Technical Monitor)

2002-01-01

Carbon nanotube materials promise to be the basis for a variety of emerging technologies with aerospace applications. Potential applications to human space flight include spacecraft shielding, hydrogen storage, structures and fixtures and nano-electronics. Appropriate risk analysis on the properties of nanotube materials is essential for future mission safety. Along with other environmental hazards, materials used in space flight encounter a hostile radiation environment for all mission profiles, from low earth orbit to interplanetary space.

Analysis and optimization of Love wave liquid sensors.

PubMed

Jakoby, B; Vellekoop, M J

1998-01-01

Love wave sensors are highly sensitive microacoustic devices, which are well suited for liquid sensing applications thanks to the shear polarization of the wave. The sensing mechanism thereby relies on the mechanical (or acoustic) interaction of the device with the liquid. The successful utilization of Love wave devices for this purpose requires proper shielding to avoid unwanted electric interaction of the liquid with the wave and the transducers. In this work we describe the effects of this electric interaction and the proper design of a shield to prevent it. We present analysis methods, which illustrate the impact of the interaction and which help to obtain an optimized design of the proposed shield. We also present experimental results for devices that have been fabricated according to these design rules.

Design and analysis of a personnel blast shield for different explosives applications

NASA Astrophysics Data System (ADS)

Lozano, Eduardo

The use of explosives brings countless benefits to our everyday lives in areas such as mining, oil and gas exploration, demolition, and avalanche control. However, because of the potential destructive power of explosives, strict safety procedures must be an integral part of any explosives operation. The goal of this work is to provide a solution to protect against the hazards that accompany the general use of explosives, specifically in avalanche control. For this reason, a blast shield was designed and tested to protect the Colorado Department of Transportation personnel against these unpredictable effects. This document will develop a complete analysis to answer the following questions: what are the potential hazards from the detonation of high explosives, what are their effects, and how can we protect ourselves against them. To answer these questions theoretical, analytical, and numerical calculations were performed. Finally, a full blast shield prototype was tested under different simulated operational environments proving its effectiveness as safety device. The Colorado Department of Transportation currently owns more than fifteen shields that are used during every operation involving explosive materials.

Design and Analysis of the Aperture Shield Assembly for a Space Solar Receiver

NASA Technical Reports Server (NTRS)

Strumpf, Hal J.; Trinh, Tuan; Westelaken, William; Krystkowiak, Christopher; Avanessian, Vahe; Kerslake, Thomas W.

1997-01-01

A joint U.S./Russia program has been conducted to design, develop, fabricate, launch, and operate the world's first space solar dynamic power system on the Russian Space Station Mir. The goal of the program was to demonstrate and confirm that solar dynamic power systems are viable for future space applications such as the International Space Station (ISS). The major components of the system include a solar receiver, a closed Brayton cycle power conversion unit, a power conditioning and control unit, a solar concentrator, a radiator, a thermal control system, and a Space Shuttle carrier. Unfortunately, the mission was demanifested from the ISS Phase 1 Space Shuttle Program in 1996. However, NASA Lewis is proposing to use the fabricated flight hardware as part of an all-American flight demonstration on the ISS in 2002. The present paper concerns the design and analysis of the solar receiver aperture shield assembly. The aperture shield assembly comprises the front face of the cylindrical receiver and is located at the focal plane of the solar concentrator. The aperture shield assembly is a critical component that protects the solar receiver structure from highly concentrated solar fluxes during concentrator off-pointing events. A full-size aperture shield assembly was fabricated. This unit was essentially identical to the flight configuration, with the exception of materials substitution. In addition, a thermal shock test aperture shield assembly was fabricated. This test article utilized the flight materials and was used for high-flux testing in the solar simulator test rig at NASA Lewis. This testing is described in a companion paper.

Radiation Protection for Lunar Mission Scenarios

NASA Technical Reports Server (NTRS)

Clowdsley, Martha S.; Nealy, John E.; Wilson, John W.; Anderson, Brooke M.; Anderson, Mark S.; Krizan, Shawn A.

2005-01-01

Preliminary analyses of shielding requirements to protect astronauts from the harmful effects of radiation on both short-term and long-term lunar missions have been performed. Shielding needs for both solar particle events (SPEs) and galactic cosmic ray (GCR) exposure are discussed for transit vehicles and surface habitats. This work was performed under the aegis of two NASA initiatives. The first study was an architecture trade study led by Langley Research Center (LaRC) in which a broad range of vehicle types and mission scenarios were compared. The radiation analysis for this study primarily focused on the additional shielding mass required to protect astronauts from the rare occurrence of a large SPE. The second study, led by Johnson Space Center (JSC), involved the design of lunar habitats. Researchers at LaRC were asked to evaluate the changes to mission architecture that would be needed if the surface stay were lengthened from a shorter mission duration of 30 to 90 days to a longer stay of 500 days. Here, the primary radiation concern was GCR exposure. The methods used for these studies as well as the resulting shielding recommendations are discussed. Recommendations are also made for more detailed analyses to minimize shielding mass, once preliminary vehicle and habitat designs have been completed. Here, methodologies are mapped out and available radiation analysis tools are described. Since, as yet, no dosimetric limits have been adopted for missions beyond low earth orbit (LEO), radiation exposures are compared to LEO limits. Uncertainties associated with the LEO career effective dose limits and the effects of lowering these limits on shielding mass are also discussed.

Physical analysis of the shielding capacity for a lightweight apron designed for shielding low intensity scattering X-rays

NASA Astrophysics Data System (ADS)

Kim, Seon Chil; Choi, Jeong Ryeol; Jeon, Byeong Kyou

2016-07-01

The purpose of this paper is to develop a lightweight apron that will be used for shielding low intensity radiation in medical imaging radiography room and to apply it to a custom-made effective shielding. The quality of existing aprons made for protecting our bodies from direct radiation are improved so that they are suitable for scattered X-rays. Textiles that prevent bodies from radiation are made by combining barium sulfate and liquid silicon. These materials have the function of shielding radiation in a manner like lead. Three kinds of textiles are produced. The thicknesses of each textile are 0.15 mm, 0.21 mm, and 0.29 mm and the corresponding lead equivalents are 0.039 mmPb, 0.095 mmPb, 0.22 mmPb for each. The rate of shielding space scattering rays are 80% from the distance of 0.5 m, 86% from 1.0 m, and 97% from 1.5 m. If we intend to approach with the purpose of shielding scattering X-rays and low intensity radiations, it is possible to reduce the weight of the apron to be 1/5 compared to that of the existing lead aprons whose weight is typically more than 4 kg. We confirm, therefore, that it is possible to produce lightweight aprons that are used for the purpose of shielding low dose radiations.

Fetal radiation monitoring and dose minimization during intensity modulated radiation therapy for glioblastoma in pregnancy.

PubMed

Horowitz, David P; Wang, Tony J C; Wuu, Cheng-Shie; Feng, Wenzheng; Drassinower, Daphnie; Lasala, Anita; Pieniazek, Radoslaw; Cheng, Simon; Connolly, Eileen P; Lassman, Andrew B

2014-11-01

We examined the fetal dose from irradiation of glioblastoma during pregnancy using intensity modulated radiation therapy (IMRT), and describe fetal dose minimization using mobile shielding devices. A case report is described of a pregnant woman with glioblastoma who was treated during the third trimester of gestation with 60 Gy of radiation delivered via a 6 MV photon IMRT plan. Fetal dose without shielding was estimated using an anthropomorphic phantom with ion chamber and diode measurements. Clinical fetal dose with shielding was determined with optically stimulated luminescent dosimeters and ion chamber. Clinical target volume (CTV) and planning target volume (PTV) coverage was 100 and 98 % receiving 95 % of the prescription dose, respectively. Normal tissue tolerances were kept below quantitative analysis of normal tissue effects in the clinic (QUANTEC) recommendations. Without shielding, anthropomorphic phantom measurements showed a cumulative fetal dose of 0.024 Gy. In vivo measurements with shielding in place demonstrated a cumulative fetal dose of 0.016 Gy. The fetal dose estimated without shielding was 0.04 % and with shielding was 0.026 % of the target dose. In vivo estimation of dose equivalent received by the fetus was 24.21 mSv. Using modern techniques, brain irradiation can be delivered to pregnant patients in the third trimester with very low measured doses to the fetus, without compromising target coverage or normal tissue dose constraints. Fetal dose can further be reduced with the use of shielding devices, in keeping with the principle of as low as reasonably achievable.

Physical analysis of the shielding capacity for a lightweight apron designed for shielding low intensity scattering X-rays

PubMed Central

Kim, Seon Chil; Choi, Jeong Ryeol; Jeon, Byeong Kyou

2016-01-01

The purpose of this paper is to develop a lightweight apron that will be used for shielding low intensity radiation in medical imaging radiography room and to apply it to a custom-made effective shielding. The quality of existing aprons made for protecting our bodies from direct radiation are improved so that they are suitable for scattered X-rays. Textiles that prevent bodies from radiation are made by combining barium sulfate and liquid silicon. These materials have the function of shielding radiation in a manner like lead. Three kinds of textiles are produced. The thicknesses of each textile are 0.15 mm, 0.21 mm, and 0.29 mm and the corresponding lead equivalents are 0.039 mmPb, 0.095 mmPb, 0.22 mmPb for each. The rate of shielding space scattering rays are 80% from the distance of 0.5 m, 86% from 1.0 m, and 97% from 1.5 m. If we intend to approach with the purpose of shielding scattering X-rays and low intensity radiations, it is possible to reduce the weight of the apron to be 1/5 compared to that of the existing lead aprons whose weight is typically more than 4 kg. We confirm, therefore, that it is possible to produce lightweight aprons that are used for the purpose of shielding low dose radiations. PMID:27461510

Physical analysis of the shielding capacity for a lightweight apron designed for shielding low intensity scattering X-rays.

PubMed

Kim, Seon Chil; Choi, Jeong Ryeol; Jeon, Byeong Kyou

2016-07-27

The purpose of this paper is to develop a lightweight apron that will be used for shielding low intensity radiation in medical imaging radiography room and to apply it to a custom-made effective shielding. The quality of existing aprons made for protecting our bodies from direct radiation are improved so that they are suitable for scattered X-rays. Textiles that prevent bodies from radiation are made by combining barium sulfate and liquid silicon. These materials have the function of shielding radiation in a manner like lead. Three kinds of textiles are produced. The thicknesses of each textile are 0.15 mm, 0.21 mm, and 0.29 mm and the corresponding lead equivalents are 0.039 mmPb, 0.095 mmPb, 0.22 mmPb for each. The rate of shielding space scattering rays are 80% from the distance of 0.5 m, 86% from 1.0 m, and 97% from 1.5 m. If we intend to approach with the purpose of shielding scattering X-rays and low intensity radiations, it is possible to reduce the weight of the apron to be 1/5 compared to that of the existing lead aprons whose weight is typically more than 4 kg. We confirm, therefore, that it is possible to produce lightweight aprons that are used for the purpose of shielding low dose radiations.

Pelvic infection: a comparison of the Dalkon shield and three other intrauterine devices.

PubMed Central

Snowden, R; Pearson, B

1984-01-01

A detailed analysis was undertaken of reports of possible pelvic infection in relation to the use of four commonly fitted intrauterine contraceptive devices during 1971 to 1978 in the United Kingdom. The four devices were the Dalkon shield, Lippes loops 3C and 2D, and the Gravigard (copper 7), and data used were those collected systematically through the UK intrauterine device research network. Prospective reports that the Dalkon shield was uniquely related to high levels of infection when compared with other intrauterine devices were not substantiated in this prospective study among 13 349 users. Though some factors such as social class and previous experience of abortion appeared to influence the rate of infection, the type of intrauterine device being worn did not appear to be a significant factor. Various methods of analysis were used including life table, regression, and discriminant analysis, using information relating to the type of intrauterine device worn, the characteristics of the user, the fitting centre, and the pattern of diagnosis and treatment of reported or suspected pelvic infection. The results of this study suggest that fears that the Dalkon shield may be associated with a higher incidence of pelvic infection than other intrauterine devices may have been unjustified. PMID:6426647

Protective Effectiveness of Porous Shields Under the Influence of High-Speed Impact Loading

NASA Astrophysics Data System (ADS)

Kramshonkov, E. N.; Krainov, A. V.; Shorohov, P. V.

2016-02-01

The results of numerical simulations of a compact steel impactor with the aluminum porous shields under high-speed shock loading are presented. The porosity of barrier varies in wide range provided that its mass stays the same, but the impactor has always equal (identical) mass. Here presented the final assessment of the barrier perforation speed depending on its porosity and initial shock speed. The range of initial impact speed varies from 1 to 10 km/s. Physical phenomena such as: destruction, melting, vaporization of a interacting objects are taken into account. The analysis of a shield porosity estimation disclosed that the protection effectiveness of porous shield reveals at the initial impact speed grater then 1.5 km/s, and it increases when initial impact speed growth.

L-Area STS MTR/NRU/NRX Grapple Assembly Closure Mechanics Review

DOE Office of Scientific and Technical Information (OSTI.GOV)

Huizenga, D. J.

2016-06-08

A review of the closure mechanics associated with the Shielded Transfer System (STS) MTR/NRU/NRX grapple assembly utilized at the Savannah River Site (SRS) was performed. This review was prompted by an operational event which occurred at the Canadian Nuclear Laboratories (CNL) utilizing a DTS-XL grapple assembly which is essentially identical to the STS MTR/NRU/NRX grapple assembly used at the SRS. The CNL operational event occurred when a NRU/NRX fuel basket containing spent nuclear fuel assemblies was inadvertently released by the DTS-XL grapple assembly during a transfer. The SM review of the STS MTR/NRU/NRX grapple assembly will examine the operational aspectsmore » of the STS and the engineered features of the STS which prevent such an event at the SRS. The design requirements for the STS NRU/NRX modifications and the overall layout of the STS are provided in other documents.« less

Engine System Model Development for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Nelson, Karl W.; Simpson, Steven P.

2006-01-01

In order to design, analyze, and evaluate conceptual Nuclear Thermal Propulsion (NTP) engine systems, an improved NTP design and analysis tool has been developed. The NTP tool utilizes the Rocket Engine Transient Simulation (ROCETS) system tool and many of the routines from the Enabler reactor model found in Nuclear Engine System Simulation (NESS). Improved non-nuclear component models and an external shield model were added to the tool. With the addition of a nearly complete system reliability model, the tool will provide performance, sizing, and reliability data for NERVA-Derived NTP engine systems. A new detailed reactor model is also being developed and will replace Enabler. The new model will allow more flexibility in reactor geometry and include detailed thermal hydraulics and neutronics models. A description of the reactor, component, and reliability models is provided. Another key feature of the modeling process is the use of comprehensive spreadsheets for each engine case. The spreadsheets include individual worksheets for each subsystem with data, plots, and scaled figures, making the output very useful to each engineering discipline. Sample performance and sizing results with the Enabler reactor model are provided including sensitivities. Before selecting an engine design, all figures of merit must be considered including the overall impacts on the vehicle and mission. Evaluations based on key figures of merit of these results and results with the new reactor model will be performed. The impacts of clustering and external shielding will also be addressed. Over time, the reactor model will be upgraded to design and analyze other NTP concepts with CERMET and carbide fuel cores.

Geological analysis of parts of the southern Arabian Shield based on Landsat imagery

NASA Astrophysics Data System (ADS)

Qari, Mohammed Yousef Hedaytullah T.

This thesis examines the capability and applicability of Landsat multispectral remote sensing data for geological analysis in the arid southern Arabian Shield, which is the eastern segment of the Nubian-Arabian Shield surrounding the Red Sea. The major lithologies in the study area are Proterozoic metavolcanics, metasediments, gneisses and granites. Three test-sites within the study area, located within two tectonic assemblages, the Asir Terrane and the Nabitah Mobile Belt, were selected for detailed comparison of remote sensing methods and ground geological studies. Selected digital image processing techniques were applied to full-resolution Landsat TM imagery and the results are interpreted and discussed. Methods included: image contrast improvement, edge enhancement for detecting lineaments and spectral enhancement for geological mapping. The last method was based on two principles, statistical analysis of the data and the use of arithmetical operators. New and detailed lithological and structural maps were constructed and compared with previous maps of these sites. Examples of geological relations identified using TM imagery include: recognition and mapping of migmatites for the first time in the Arabian Shield; location of the contact between the Asir Terrane and the Nabitah Mobile Belt; and mapping of lithologies, some of which were not identified on previous geological maps. These and other geological features were confirmed by field checking. Methods of lineament enhancement implemented in this study revealed structural lineaments, mostly mapped for the first time, which can be related to regional tectonics. Structural analysis showed that the southern Arabian Shield has been affected by at least three successive phases of deformation. The third phase is the most dominant and widespread. A crustal evolutionary model in the vicinity of the study area is presented showing four stages, these are: arc stage, accretion stage, collision stage and post-collision stage. The results of this study demonstrate that Landsat TM data can be used reliably for geological investigations in the Arabian Shield and comparable areas, particularly to generate detailed geological maps over large areas by using quantitative remote sensing methods, providing there is prior knowledge of part of the area.

Hybrid Wing Body Configuration Scaling Study

NASA Technical Reports Server (NTRS)

Nickol, Craig L.

2012-01-01

The Hybrid Wing Body (HWB) configuration is a subsonic transport aircraft concept with the potential to simultaneously reduce fuel burn, noise and emissions compared to conventional concepts. Initial studies focused on very large applications with capacities for up to 800 passengers. More recent studies have focused on the large, twin-aisle class with passenger capacities in the 300-450 range. Efficiently scaling this concept down to the single aisle or smaller size is challenging due to geometric constraints, potentially reducing the desirability of this concept for applications in the 100-200 passenger capacity range or less. In order to quantify this scaling challenge, five advanced conventional (tube-and-wing layout) concepts were developed, along with equivalent (payload/range/technology) HWB concepts, and their fuel burn performance compared. The comparison showed that the HWB concepts have fuel burn advantages over advanced tube-and-wing concepts in the larger payload/range classes (roughly 767-sized and larger). Although noise performance was not quantified in this study, the HWB concept has distinct noise advantages over the conventional tube-and-wing configuration due to the inherent noise shielding features of the HWB. NASA s Environmentally Responsible Aviation (ERA) project will continue to investigate advanced configurations, such as the HWB, due to their potential to simultaneously reduce fuel burn, noise and emissions.

Progress in Open Rotor Research: A U.S. Perspective

NASA Technical Reports Server (NTRS)

Van Zante, Dale E.

2015-01-01

In response to the 1970s oil crisis, NASA created the Advanced Turboprop Project (ATP) to mature technologies for high-speed propellers to enable large reductions in fuel burn relative to turbofan engines of that era. Both single rotation and contra-rotation concepts were designed and tested in ground based facilities as well as flight. Some novel concepts configurations that were not well publicized at the time, were proposed as part of the effort. The high-speed propeller concepts did provide fuel burn savings, albeit with some acoustics and structural challenges to overcome. When fuel prices fell, the business case for radical new engine configurations collapsed and the research emphasis returned to high bypass ducted configurations. With rising oil prices and increased environmental concerns there is renewed interest in high-speed propeller based engine architectures. Contemporary analysis tools for aerodynamics and aeroacoustics have enabled a new era of blade designs that have both high efficiency and acceptable noise characteristics. A recent series of tests in the U.S. have characterized the aerodynamic performance and noise from these modern contra-rotating propeller designs. Additionally the installation and noise shielding aspects for conventional airframes and blended wing bodies have been studied. Historical estimates of propfan performance have relied on legacy propeller performance and acoustics data. Current system studies make use of the modern propeller data with higher fidelity installation effects data to estimate the performance of a contemporary aircraft system with favorable results. This paper presents the current state of high-speed propeller open rotor research within the U.S. from an overall viewpoint of the various efforts ongoing. The current projections for the technology are presented.

Gas-cooled reactor programs. High-temperature gas-cooled reactor technology development program. Annual progress report, December 31, 1983

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1984-06-01

ORNL continues to make significant contributions to the national program. In the HTR fuels area, we are providing detailed statistical information on the fission product retention performance of irradiated fuel. Our studies are also providing basic data on the mechanical, physical, and chemical behavior of HTR materials, including metals, ceramics, graphite, and concrete. The ORNL has an important role in the development of improved HTR graphites and in the specification of criteria that need to be met by commercial products. We are also developing improved reactor physics design methods. Our work in component development and testing centers in the Componentmore » Flow Test Loop (CFTL), which is being used to evaluate the performance of the HTR core support structure. Other work includes experimental evaluation of the shielding effectiveness of the lower portions of an HTR core. This evaluation is being performed at the ORNL Tower Shielding Facility. Researchers at ORNL are developing welding techniques for attaching steam generator tubing to the tubesheets and are testing ceramic pads on which the core posts rest. They are also performing extensive testing of aggregate materials obtained from potential HTR site areas for possible use in prestressed concrete reactor vessels. During the past year we continued to serve as a peer reviewer of small modular reactor designs being developed by GA and GE with balance-of-plant layouts being developed by Bechtel Group, Inc. We have also evaluated the national need for developing HTRs with emphasis on the longer term applications of the HTRs to fossil conversion processes.« less

A magnetic fluid seal for rotary blood pumps: Behaviors of magnetic fluids in a magnetic fluid seal.

PubMed

Mitamura, Yoshinori; Yano, Tetsuya; Nakamura, Wataru; Okamoto, Eiji

2013-01-01

A magnetic fluid (MF) seal has excellent durability. The performance of an MF seal, however, has been reported to decrease in liquids (several days). We have developed an MF seal that has a shield mechanism. The seal was perfect for 275 days in water. To investigate the effect of a shield, behaviors of MFs in a seal in water were studied both experimentally and computationally. (a) Two kinds of MF seals, one with a shield and one without a shield, were installed in a centrifugal pump. Behaviors of MFs in the seals in water were observed with a video camera and high-speed microscope. In the seal without a shield, the surface of the water in the seal waved and the turbulent flow affected behaviors of the MFs. In contrast, MFs rotated stably in the seal with a shield in water even at high rotational speeds. (b) Computational fluid dynamics analysis revealed that a stationary secondary flow pattern in the seal and small velocity difference between magnetic fluid and water at the interface. These MF behaviors prolonged the life of an MF seal in water.

A magnetic fluid seal for rotary blood pumps: image and computational analyses of behaviors of magnetic fluids.

PubMed

Mitamura, Yoshinori; Yano, Tetsuya; Okamoto, Eiji

2013-01-01

A magnetic fluid (MF) seal has excellent durability. The performance of an MF seal, however, has been reported to decrease in liquids (several days). We have developed an MF seal that has a shield mechanism. The seal was perfect for 275 days in water. To investigate the effect of a shield, behaviors of MFs in a seal in water were studied both experimentally and computationally. (a) Two kinds of MF seals, one with a shield and one without a shield, were installed in a centrifugal pump. Behaviors of MFs in the seals in water were observed with a video camera and high-speed microscope. In the seal without a shield, the surface of the water in the seal waved and the turbulent flow affected behaviors of the MFs. In contrast, MFs rotated stably in the seal with a shield in water even at high rotational speeds. (b) Computational fluid dynamics analysis revealed that a stationary secondary flow pattern in the seal and small velocity difference between magnetic fluid and water at the interface. These MF behaviors prolonged the life of an MF seal in water.

Cancer mortality in residents of the terrain-shielded area exposed to fallout from the Nagasaki atomic bombing

PubMed Central

Mine, Mariko; Kondo, Hisayoshi; Matsuda, Naoki; Shibata, Yoshisada; Takamura, Noboru

2018-01-01

Abstract The health effects of radiation exposure from the atomic bomb fallout remain unclear. The objective of the present study is to elucidate the association between low-dose radiation exposure from the atomic bomb fallout and cancer mortality among Nagasaki atomic bomb survivors. Of 77 884 members in the Nagasaki University Atomic Bomb Survivors Cohort, 610 residents in the terrain-shielded area with fallout were selected for this analysis; 1443 residents in the terrain-shielded area without fallout were selected as a control group; and 3194 residents in the direct exposure area were also selected for study. Fifty-two deaths due to cancer in the terrain-shielded fallout area were observed during the follow-up period from 1 January 1970 to 31 December 2012. The hazard ratio for cancer mortality in the terrain-shielded fallout area was 0.90 (95% confidence interval: 0.65–1.24). No increase in the risk of cancer mortality was observed, probably because the dose of the radiation exposure was low for residents in the terrain-shielded fallout areas of the Nagasaki atomic bomb, and also because the number of study subjects was small. PMID:29036510

Synthesis and electromagnetic interference shielding of cellulose-derived carbon aerogels functionalized with α-Fe2O3 and polypyrrole.

PubMed

Wan, Caichao; Li, Jian

2017-04-01

Eco-friendly cellulose-derived carbon aerogels (CDCA) were employed as porous substrate to integrate with α-Fe 2 O 3 and polypyrrole (PPy) via pyrolysis and vapor-phase polymerization. The SEM and TEM observations present that the wrinkled PPy sheets and the α-Fe 2 O 3 nanoparticles were well dispersed in CDCA. The strong interactions (such as hydrogen bonding) between the substrate and the nanomaterials were demonstrated by the FTIR and XPS analysis. When utilized as electromagnetic interference (EMI) shielding materials, the α-Fe 2 O 3 /PPy/CDCA (FPCA) composite has the highest total shielding effectiveness (SE total ) of 39.4dB, about 2.0, 2.9, and 1.3 times that of the acid-treated CDCA (19.3dB), PPy (13.6dB), and α-Fe 2 O 3 /CDCA (29.3dB), respectively. Moreover, the shielding effectiveness due to absorption accounts for 78.2%-84.2% of SE total for FPCA, indicative of the absorption-dominant shielding mechanism contributing to alleviating secondary radiation. These features make the composite a useful alternative candidate for EMI shielding. Copyright © 2017 Elsevier Ltd. All rights reserved.

A Fast Code for Jupiter Atmospheric Entry Analysis

NASA Technical Reports Server (NTRS)

Yauber, Michael E.; Wercinski, Paul; Yang, Lily; Chen, Yih-Kanq

1999-01-01

A fast code was developed to calculate the forebody heating environment and heat shielding that is required for Jupiter atmospheric entry probes. A carbon phenolic heat shield material was assumed and, since computational efficiency was a major goal, analytic expressions were used, primarily, to calculate the heating, ablation and the required insulation. The code was verified by comparison with flight measurements from the Galileo probe's entry. The calculation required 3.5 sec of CPU time on a work station, or three to four orders of magnitude less than for previous Jovian entry heat shields. The computed surface recessions from ablation were compared with the flight values at six body stations. The average, absolute, predicted difference in the recession was 13.7% too high. The forebody's mass loss was overpredicted by 5.3% and the heat shield mass was calculated to be 15% less than the probe's actual heat shield. However, the calculated heat shield mass did not include contingencies for the various uncertainties that must be considered in the design of probes. Therefore, the agreement with the Galileo probe's values was satisfactory in view of the code's fast running time and the methods' approximations.

Design Performance of Front Steering-Type Electron Cyclotron Launcher for ITER

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takahashi, K.; Imai, T.; Kobayashi, N.

2005-01-15

The performance of a front steering (FS)-type electron cyclotron launcher designed for the International Thermonuclear Experimental Reactor (ITER) is evaluated with a thermal, electromagnetic, and nuclear analysis of the components; a mechanical test of a spiral tube for the steering mirror; and a rotational test of bearings. The launcher consists of a front shield and a launcher plug where three movable optic mirrors to steer incident multimegawatt radio-frequency beam power, waveguide components, nuclear shields, and vacuum windows are installed. The windows are located behind a closure plate to isolate the transmission lines from the radioactivated circumstance (vacuum vessel). The waveguidemore » lines of the launcher are doglegged to reduce the direct neutron streaming toward the vacuum windows and other components. The maximum stresses on the critical components such as the steering mirror, its cooling tube, and the front shield are less than their allowable stresses. It was also identified that the stress on the launcher, which yielded from electromagnetic force caused by plasma disruption, was a little larger than the criteria, and a modification of the launcher plug structure was necessary. The nuclear analysis result shows that the neutron shield capability of the launcher satisfies the shield criteria of the ITER. It concludes that the design of the FS launcher is generally suitable for application to the ITER.« less

Analysis of low-dose radiation shield effectiveness of multi-gate polymeric sheets

NASA Astrophysics Data System (ADS)

Kim, S. C.; Lee, H. K.; Cho, J. H.

2014-07-01

Computed tomography (CT) uses a high dose of radiation to create images of the body. As patients are exposed to radiation during a CT scan, the use of shielding materials becomes essential in CT scanning. This study was focused on the radiation shielding materials used for patients during a CT scan. In this study, sheets were manufactured to shield the eyes and the thyroid, the most sensitive parts of the body, against radiation exposure during a CT scan. These sheets are manufactured using silicone polymers, barium sulfate (BaSO4) and tungsten, with the aim of making these sheets equally or more effective in radiation shielding and more cost-effective than lead sheets. The use of barium sulfate drew more attention than tungsten due to its higher cost-effectiveness. The barium sulfate sheets were coated to form a multigate structure by applying the maximum charge rate during the agitator and subsequent mixing processes and creating multilayered structures on the surface. To measure radiation shielding effectiveness, the radiation dose was measured around both eyes and the thyroid gland using sheets in three different thicknesses (1, 2 and 3 mm). Among the 1 and 2 mm sheets, the Pb sheets exhibited greater effectiveness in radiation shielding around both eyes, but the W sheets were more effective in radiation shielding around the thyroid gland. In the 3 mm sheets, the Pb sheet also attenuated a higher amount of radiation around both eyes while the W sheet was more effective around the thyroid gland. In conclusion, the sheets made from barium sulfate and tungsten proved highly effective in shielding against low-dose radiation in CT scans without causing ill-health effects, unlike lead.

Final Report for the “WSU Neutron Capture Therapy Facility Support”

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerald E. Tripard; Keith G. Fox

2006-08-24

The objective for the cooperative research program for which this report has been written was to provide separate NCT facility user support for the students, faculty and scientists who would be doing the U.S. Department of Energy Office (DOE) of Science supported advanced radiotargeted research at the WSU 1 megawatt TRIGA reactor. The participants were the Idaho National laboratory (INL, P.I., Dave Nigg), the Veterinary Medical Research Center of Washington State University (WSU, Janean Fidel and Patrick Gavin), and the Washington State University Nuclear Radiation Center (WSU, P.I., Gerald Tripard). A significant number of DOE supported modifications were made tomore » the WSU reactor in order to create an epithermal neutron beam while at the same time maintaining the other activities of the 1 MW reactor. These modifications were: (1) Removal of the old thermal column. (2) Construction and insertion of a new epithermal filter, collimator and shield. (3) Construction of a shielded room that could accommodate the very high radiation field created by an intense neutron beam. (4) Removal of the previous reactor core fuel cluster arrangement. (5) Design and loading of the new reactor core fuel cluster arrangement in order to optimize the neutron flux entering the epithermal neutron filter. (6) The integration of the shielded rooms interlocks and radiological controls into the SCRAM chain and operating electronics of the reactor. (7) Construction of a motorized mechanism for moving and remotely controlling the position of the entire reactor bridge. (8) The integration of the reactor bridge control electronics into the SCRAM chain and operating electronics of the reactor. (9) The design, construction and attachment to the support structure of the reactor of an irradiation box that could be inserted into position next to the face of the reactor. (Necessitated by the previously mentioned core rearrangement). All of the above modifications were successfully completed and tested. The resulting epithermal beam of 1 x 10{sup 9} n/sec-cm{sup 2} was measured by Idaho National Laboratory with assistance from WSU's Neutron Activation Analysis Group. The beam is as good as our initial proposals for the project had predicted. In addition to all of the design, construction and insertion of the hardware, shielding, electronics and radiation monitoring systems there was considerable manpower and effort put into changes in the Technical Specifications of the reactor and implementing procedures for use of the new facility. This staff involvement is one of the reasons we requested special facility support from the DOE. Once the facility was competed and all of the recalibrations and measurements made to characterize the differences between this reactor core and the previous core we began to assist INL in making their beam measurements with foils and phantoms. Although we proposed support for only one additional staff position to support this new NCT facility the staff support provided by the WSU Nuclear Radiation Center was greater than had been anticipated by our initial proposal. INL was also assisted in the testing of a heavy water (deuterated water) bladder that can be inserted into the collimator in order to produce an intense, external thermal neutron beam. The external epithermal and/or thermal neutron beam capability remains available for use, if funding becomes available for future research projects.« less

Cross Section Sensitivity and Propagated Errors in HZE Exposures

NASA Technical Reports Server (NTRS)

Heinbockel, John H.; Wilson, John W.; Blatnig, Steve R.; Qualls, Garry D.; Badavi, Francis F.; Cucinotta, Francis A.

2005-01-01

It has long been recognized that galactic cosmic rays are of such high energy that they tend to pass through available shielding materials resulting in exposure of astronauts and equipment within space vehicles and habitats. Any protection provided by shielding materials result not so much from stopping such particles but by changing their physical character in interaction with shielding material nuclei forming, hopefully, less dangerous species. Clearly, the fidelity of the nuclear cross-sections is essential to correct specification of shield design and sensitivity to cross-section error is important in guiding experimental validation of cross-section models and database. We examine the Boltzmann transport equation which is used to calculate dose equivalent during solar minimum, with units (cSv/yr), associated with various depths of shielding materials. The dose equivalent is a weighted sum of contributions from neutrons, protons, light ions, medium ions and heavy ions. We investigate the sensitivity of dose equivalent calculations due to errors in nuclear fragmentation cross-sections. We do this error analysis for all possible projectile-fragment combinations (14,365 such combinations) to estimate the sensitivity of the shielding calculations to errors in the nuclear fragmentation cross-sections. Numerical differentiation with respect to the cross-sections will be evaluated in a broad class of materials including polyethylene, aluminum and copper. We will identify the most important cross-sections for further experimental study and evaluate their impact on propagated errors in shielding estimates.

Ground penetrating radar imaging and time-domain modelling of the infiltration of diesel fuel in a sandbox experiment

NASA Astrophysics Data System (ADS)

Bano, Maksim; Loeffler, Olivier; Girard, Jean-François

2009-10-01

Ground penetrating radar (GPR) is a non-destructive method which, over the past 10 years, has been successfully used not only to estimate the water content of soil, but also to detect and monitor the infiltration of pollutants on sites contaminated by light non-aqueous phase liquids (LNAPL). We represented a model water table aquifer (72 cm depth) by injecting water into a sandbox that also contains several buried objects. The GPR measurements were carried out with shielded antennae of 900 and 1200 MHz, respectively, for common mid point (CMP) and constant offset (CO) profiles. We extended the work reported by Loeffler and Bano by injecting 100 L of diesel fuel (LNAPL) from the top of the sandbox. We used the same acquisition procedure and the same profile configuration as before fuel injection. The GPR data acquired on the polluted sand did not show any clear reflections from the plume pollution; nevertheless, travel times are very strongly affected by the presence of the fuel and the main changes are on the velocity anomalies. We can notice that the reflection from the bottom of the sandbox, which is recorded at a constant time when no fuel is present, is deformed by the pollution. The area close to the fuel injection point is characterized by a higher velocity than the area situated further away. The area farther away from the injection point shows a low velocity anomaly which indicates an increase in travel time. It seems that pore water has been replaced by fuel as a result of a lateral flow. We also use finite-difference time-domain (FDTD) numerical GPR modelling in combination with dielectric property mixing models to estimate the volume and the physical characteristics of the contaminated sand.

Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

NASA Astrophysics Data System (ADS)

Cisneros, Anselmo Tomas, Jr.

The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP and PEBBED for a high temperature gas cooled pebble bed reactor. Three parametric studies were performed for exploring the design space of the PB-FHR---to select a fuel design for the PB-FHR] to select a core configuration; and to optimize the PB-FHR design. These parametric studies investigated trends in the dependence of important reactor performance parameters such as burnup, temperature reactivity feedback, radiation damage, etc on the reactor design variables and attempted to understand the underlying reactor physics responsible for these trends. A pebble fuel parametric study determined that pebble fuel should be designed with a carbon to heavy metal ratio (C/HM) less than 400 to maintain negative coolant temperature reactivity coefficients. Seed and thorium blanket-, seed and inert pebble reflector- and seed only core configurations were investigated for annular FHR PBRs---the C/HM of the blanket pebbles and discharge burnup of the thorium blanket pebbles were additional design variable for core configurations with thorium blankets. Either a thorium blanket or graphite pebble reflector is required to shield the outer graphite reflector enough to extend its service lifetime to 60 EFPY. The fuel fabrication costs and long cycle lengths of the thorium blanket fuel limit the potential economic advantages of using a thorium blanket. Therefore, the seed and pebble reflector core configuration was adopted as the baseline core configuration. Multi-objective optimization with respect to economics was performed for the PB-FHR accounting for safety and other physical design constraints derived from the high-level safety regulatory criteria. These physical constraints were applied along in a design tool, Nuclear Application Value Estimator, that evaluated a simplified cash flow economics model based on estimates of reactor performance parameters calculated using correlations based on the results of parametric design studies for a specific PB-FHR design and a set of economic assumptions about the electricity market to evaluate the economic implications of design decisions. The optimal PB-FHR design---Mark 1 PB-FHR---is described along with a detailed summary of its performance characteristics including: the burnup, the burnup evolution, temperature reactivity coefficients, the power distribution, radiation damage distributions, control element worths, decay heat curves and tritium production rates. The Mk1 PB-FHR satisfies the PB-FHR safety criteria. The fuel, moderator (pebble core, pebble shell, graphite matrix, TRISO layers) and coolant have global negative temperature reactivity coefficients and the fuel temperatures are well within their limits.

Gamma-ray mirror technology for NDA of spent fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Descalle, M. A.; Ruz-Armendariz, J.; Decker, T.

Direct measurements of gamma rays emitted by fissile material have been proposed as an alternative to measurements of the gamma rays from fission products. From a safeguards applications perspective, direct detection of uranium (U) and plutonium (Pu) K-shell fluorescence emission lines and specific lines from some of their isotopes could lead to improved shipper-receiver difference or input accountability at the start of Pu reprocessing. However, these measurements are difficult to implement when the spent fuel is in the line-of-sight of the detector, as the detector is exposed to high rates dominated by fission product emissions. To overcome the combination ofmore » high rates and high background, grazing incidence multilayer mirrors have been proposed as a solution to selectively reflect U and Pu hard X-ray and soft gamma rays in the 90 to 420 keV energy into a high-purity germanium (HPGe) detector shielded from the direct line-of-sight of spent fuel. Several groups demonstrated that K-shell fluorescence lines of U and Pu in spent fuel could be detected with Ge detectors. In the field of hard X-ray optics the performance of reflective multilayer coated reflective optics was demonstrated up to 645 keV at the European Synchrotron Radiation Facility. Initial measurements conducted at Oak Ridge National Laboratory with sealed sources and scoping experiments conducted at the ORNL Irradiated Fuels Examination Laboratory (IFEL) with spent nuclear fuel further demonstrated the pass-band properties of multilayer mirrors for reflecting specific emission lines into 1D and 2D HPGe detectors, respectively.« less

NMR Spectra of Oriented Samples of Intercalated Fluorographite and 19F Chemical Shielding Anisotropy of the CIF 3 Molecule

NASA Astrophysics Data System (ADS)

Panich, A. M.

The analysis of 19F NMR spectra of polycrystalline and partially oriented samples of fluorinated graphite (C 2F) n intercalated with chlorine trifluoride has been carried out. Molecular mobility results in almost complete averaging of the dipole-dipole interactions of nuclei, while the essential chemical shielding anisotropy (CSA) is manifested. There is suggested molecular rotation about its C2 axes, which in turn rotates about the normal to the graphite plane. The CSA (σ || - σ ⊥) is determined to be 510 and -640 ppm, respectively, for the two inequivalent fluorine atoms of the molecule. The effect of the "antiparamagnetic" shielding leading to inversion of the chemical shielding tenser [(σ || - σ ⊥) < 0] for the equatorial F atom and anomalous line disposition in the NMR spectrum is discussed.

NASA-Lewis experiences with multigroup cross sections and shielding calculations

NASA Technical Reports Server (NTRS)

Lahti, G. P.

1972-01-01

The nuclear reactor shield analysis procedures employed at NASA-Lewis are described. Emphasis is placed on the generation, use, and testing of multigroup cross section data. Although coupled neutron and gamma ray cross section sets are useful in two dimensional Sn transport calculations, much insight has been gained from examination of uncoupled calculations. These have led to experimental and analytic studies of areas deemed to be of first order importance to reactor shield calculations. A discussion is given of problems encountered in using multigroup cross sections in the resolved resonance energy range. The addition to ENDF files of calculated and/or measured neutron-energy-dependent capture gamma ray spectra for shielding calculations is questioned for the resonance region. Anomalies inherent in two dimensional Sn transport calculations which may overwhelm any cross section discrepancies are illustrated.

Magnet Architectures and Active Radiation Shielding Study - SR2S Workshop

NASA Technical Reports Server (NTRS)

Westover, Shane; Meinke, Rainer; Burger, William; Ilin, Andrew; Nerolich, Shaun; Washburn, Scott

2014-01-01

Analyze new coil configurations with maturing superconductor technology -Develop vehicle-level concept solutions and identify engineering challenges and risks -Shielding performance analysis Recent advances in superconducting magnet technology and manufacturing have opened the door for re-evaluating active shielding solutions as an alternative to mass prohibitive passive shielding.Publications on static magnetic field environments and its bio-effects were reviewed. Short-term exposure information is available suggesting long term exposure may be okay. Further research likely needed. center dotMagnetic field safety requirements exist for controlled work environments. The following effects have been noted with little noted adverse effects -Magnetohydrodynamic (MHD) effects on ionized fluids (e.g. blood) creating an aortic voltage change -MHD interaction elevates blood pressure (BP) center dot5 Tesla equates to 5% BP elevation -Prosthetic devises and pacemakers are an issue (access limit of 5 gauss).

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Gyeong Won; Shim, Jaewon; Jung, Young-Dae, E-mail: ydjung@hanyang.ac.kr

The influence of renormalization plasma screening on the entanglement fidelity for the elastic electron-atom scattering is investigated in partially ionized dense hydrogen plasmas. The partial wave analysis and effective interaction potential are employed to obtain the scattering entanglement fidelity in dense hydrogen plasmas as functions of the collision energy, the Debye length, and the renormalization parameter. It is found that the renormalization plasma shielding enhances the scattering entanglement fidelity. Hence, we show that the transmission of the quantum information can be increased about 10% due to the renormalization shielding effect in dense hydrogen plasmas. It is also found that themore » renormalization shielding effect on the entanglement fidelity for the electron-atom collision increases with an increase of the collision energy. In addition, the renormalization shielding function increases with increasing collision energy and saturates to the unity with an increase of the Debye length.« less

Meeting Radiation Protection Requirements and Reducing Spacecraft Mass - A Multifunctional Materials Approach

NASA Technical Reports Server (NTRS)

Atwell, William; Koontz, Steve; Reddell, Brandon; Rojdev, Kristina; Franklin, Jennifer

2010-01-01

Both crew and radio-sensitive systems, especially electronics must be protected from the effects of the space radiation environment. One method of mitigating this radiation exposure is to use passive-shielding materials. In previous vehicle designs such as the International Space Station (ISS), materials such as aluminum and polyethylene have been used as parasitic shielding to protect crew and electronics from exposure, but these designs add mass and decrease the amount of usable volume inside the vehicle. Thus, it is of interest to understand whether structural materials can also be designed to provide the radiation shielding capability needed for crew and electronics, while still providing weight savings and increased useable volume when compared against previous vehicle shielding designs. In this paper, we present calculations and analysis using the HZETRN (deterministic) and FLUKA (Monte Carlo) codes to investigate the radiation mitigation properties of these structural shielding materials, which includes graded-Z and composite materials. This work is also a follow-on to an earlier paper, that compared computational results for three radiation transport codes, HZETRN, HETC, and FLUKA, using the Feb. 1956 solar particle event (SPE) spectrum. In the following analysis, we consider the October 1989 Ground Level Enhanced (GLE) SPE as the input source term based on the Band function fitting method. Using HZETRN and FLUKA, parametric absorbed doses at the center of a hemispherical structure on the lunar surface are calculated for various thicknesses of graded-Z layups and an all-aluminum structure. HZETRN and FLUKA calculations are compared and are in reasonable (18% to 27%) agreement. Both codes are in agreement with respect to the predicted shielding material performance trends. The results from both HZETRN and FLUKA are analyzed and the radiation protection properties and potential weight savings of various materials and materials lay-ups are compared.

Optimization of the EMI shielding effectiveness of fine and ultrafine POFA powder mix with OPC powder using Flower Pollination Algorithm

NASA Astrophysics Data System (ADS)

Narong, L. C.; Sia, C. K.; Yee, S. K.; Ong, P.; Zainudin, A.; Nor, N. H. M.; Kasim, N. A.

2017-01-01

In order to solve the electromagnetic interference (EMI) issue and provide a new application for palm oil fuel ash (POFA), POFA was used as the cement filler for enhancing the EMI absorption of cement-based composites. POFA was refined by using water precipitation for 24 hours to remove the filthiness and distinguish the layer 1 (floated) and layer 2 (sink) of POFA. Both layers POFA were dried for 24 hours at 100 ± 5 °C and grind separately for sieve at 140 μm (Fine) and 45 цш sizes (Ultrafine). The micro structure and element content of the both layers POFA were characterized by scanning electron microscope (SEM) and energy-dispersive X-ray spectroscopy (EDS) respectively. The results showed layer 1 POFA has potentialities for EMI shielding effectiveness (SE) due to its higher carbon content and porous structure. The study reveals that EMI SE also influenced by the particle size of POFA, where smaller particle size can increase 5 % to 13 % of EMI SE. When the specimen consists of 50% POFA with passing through 45 μm sieve, the EMI was shield -13.08 dB in between 50 MHz to 2 GHz range. Flower Pollination Algorithm (FPA) proves that POFA passing 45 μm sieve with 50% mixed to OPC is optimal parameter. The error between experimental and FPA simulation data is below 1.2 for both layers POFA.

Investigation of Lithium Metal Hydride Materials for Mitigation of Deep Space Radiation

NASA Technical Reports Server (NTRS)

Rojdev, Kristina; Atwell, William

2016-01-01

Radiation exposure to crew, electronics, and non-metallic materials is one of many concerns with long-term, deep space travel. Mitigating this exposure is approached via a multi-faceted methodology focusing on multi-functional materials, vehicle configuration, and operational or mission constraints. In this set of research, we are focusing on new multi-functional materials that may have advantages over traditional shielding materials, such as polyethylene. Metal hydride materials are of particular interest for deep space radiation shielding due to their ability to store hydrogen, a low-Z material known to be an excellent radiation mitigator and a potential fuel source. We have previously investigated 41 different metal hydrides for their radiation mitigation potential. Of these metal hydrides, we found a set of lithium hydrides to be of particular interest due to their excellent shielding of galactic cosmic radiation. Given these results, we will continue our investigation of lithium hydrides by expanding our data set to include dose equivalent and to further understand why these materials outperformed polyethylene in a heavy ion environment. For this study, we used HZETRN 2010, a one-dimensional transport code developed by NASA Langley Research Center, to simulate radiation transport through the lithium hydrides. We focused on the 1977 solar minimum Galactic Cosmic Radiation environment and thicknesses of 1, 5, 10, 20, 30, 50, and 100 g/cm2 to stay consistent with our previous studies. The details of this work and the subsequent results will be discussed in this paper.

Evaluation of laminar flow control systems for subsonic commercial transport aircraft: Executive summary

NASA Technical Reports Server (NTRS)

Pearce, W. E.

1982-01-01

An evaluation was made of laminar flow control (LFC) system concepts for subsonic commercial transport aircraft. Configuration design studies, performance analyses, fabrication development, structural testing, wind tunnel testing, and contamination-avoidance techniques were included. As a result of trade studies, a configuration with LFC on the upper wing surface only, utilizing an electron beam-perforated suction surface, and employing a retractable high-lift shield for contamination avoidance, was selected as the most practical LFC system. The LFC aircraft was then compared with an advanced turbulent aircraft designed for the same mission. This comparison indicated significant fuel savings.

Electrolytic cell with reference electrode

DOEpatents

Kessie, Robert W.

1989-01-01

A reference electrode device is provided for a high temperature electrolytic cell used to electrolytically recover uranium from spent reactor fuel dissolved in an anode pool, the device having a glass tube to enclose the electrode and electrolyte and serve as a conductive membrane with the cell electrolyte, and an outer metal tube about the glass tube to serve as a shield and basket for any glass sections broken by handling of the tube to prevent their contact with the anode pool, the metal tube having perforations to provide access between the bulk of the cell electrolyte and glass membrane.

Reference electrode for electrolytic cell

DOEpatents

Kessie, R.W.

1988-07-28

A reference electrode device is provided for a high temperature electrolytic cell used to electrolytically recover uranium from spent reactor fuel dissolved in an anode pool, the device having a glass tube to enclose the electrode and electrolyte and serve as a conductive membrane with the cell electrolyte, and an outer metal tube about the glass tube to serve as a shield and basket for any glass sections broken by handling of the tube to prevent their contact with the anode pool, the metal tube having perforations to provide access between the bulk of the cell electrolyte and glass membrane. 4 figs.

The resources of Mars for human settlement

NASA Astrophysics Data System (ADS)

Meyer, T. R.; McKay, C. P.

1989-04-01

Spacecraft exploration of Marshas shown that the essential resources necessary for life support are present on the Martian surface. The key life-support compounds O2, N2, and H2O are available on Mars. The soil could be used as radiation shielding and could provide many useful industrial and construction materials. Compounds with high chemical energy, such as rocket fuels, can be manufactured in-situ on Mars. Solar power, and possibly wind power, are available and practical on Mars. Preliminary engineering studies indicate that fairly autonomous processes can be designed to extract and stockpile Martian consumables.

IET. Aerial view of project, 95 percent complete. Camera facing ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

IET. Aerial view of project, 95 percent complete. Camera facing east. Left to right: stack, duct, mobile test cell building (TAN-624), four-rail track, dolly. Retaining wall between mobile test building and shielded control building (TAN-620) just beyond. North of control building are tank building (TAN-627) and fuel-transfer pump building (TAN-625). Guard house at upper right along exclusion fence. Construction vehicles and temporary warehouse in view near guard house. Date: June 6, 1955. INEEL negative no. 55-1462 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

NEUTRONIC REACTOR

DOEpatents

Wigner, E.P.; Weinberg, A.W.; Young, G.J.

1958-04-15

A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

System for handling and storing radioactive waste

DOEpatents

Anderson, J.K.; Lindemann, P.E.

1982-07-19

A system and method are claimed for handling and storing spent reactor fuel and other solid radioactive waste, including canisters to contain the elements of solid waste, storage racks to hold a plurality of such canisters, storage bays to store these racks in isolation by means of shielded doors in the bays. This system also includes means for remotely positioning the racks in the bays and an access tunnel within which the remotely operated means is located to position a rack in a selected bay. The modular type of these bays will facilitate the construction of additional bays and access tunnel extension.

System for handling and storing radioactive waste

DOEpatents

Anderson, John K.; Lindemann, Paul E.

1984-01-01

A system and method for handling and storing spent reactor fuel and other solid radioactive waste, including canisters to contain the elements of solid waste, storage racks to hold a plurality of such canisters, storage bays to store these racks in isolation by means of shielded doors in the bays. This system also includes means for remotely positioning the racks in the bays and an access tunnel within which the remotely operated means is located to position a rack in a selected bay. The modular type of these bays will facilitate the construction of additional bays and access tunnel extension.

The resources of Mars for human settlement

NASA Technical Reports Server (NTRS)

Meyer, Thomas R.; Mckay, Christopher P.

1989-01-01

Spacecraft exploration of Marshas shown that the essential resources necessary for life support are present on the Martian surface. The key life-support compounds O2, N2, and H2O are available on Mars. The soil could be used as radiation shielding and could provide many useful industrial and construction materials. Compounds with high chemical energy, such as rocket fuels, can be manufactured in-situ on Mars. Solar power, and possibly wind power, are available and practical on Mars. Preliminary engineering studies indicate that fairly autonomous processes can be designed to extract and stockpile Martian consumables.

Hybrid Skyshine Calculations for Complex Neutron and Gamma-Ray Sources

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shultis, J. Kenneth

2000-10-15

A two-step hybrid method is described for computationally efficient estimation of neutron and gamma-ray skyshine doses far from a shielded source. First, the energy and angular dependence of radiation escaping into the atmosphere from a source containment is determined by a detailed transport model such as MCNP. Then, an effective point source with this energy and angular dependence is used in the integral line-beam method to transport the radiation through the atmosphere up to 2500 m from the source. An example spent-fuel storage cask is analyzed with this hybrid method and compared to detailed MCNP skyshine calculations.

High purity silica reflective heat shield development

NASA Technical Reports Server (NTRS)

Blome, J. C.; Drennan, D. N.; Schmitt, R. J.

1974-01-01

Measurements were made of reflectance in the vacuum ultraviolet down to 0.15 micron. Scattering coefficients (S) and absorption coefficients (K) were also measured. These coefficients express the optical properties and are used directly in a thermodynamic analysis for sizing a heat shield. The effect of the thin silica melt layer formed during entry was also studied from the standpoint of trapped radiant energy.

CAD-Based Shielding Analysis for ITER Port Diagnostics

NASA Astrophysics Data System (ADS)

Serikov, Arkady; Fischer, Ulrich; Anthoine, David; Bertalot, Luciano; De Bock, Maartin; O'Connor, Richard; Juarez, Rafael; Krasilnikov, Vitaly

2017-09-01

Radiation shielding analysis conducted in support of design development of the contemporary diagnostic systems integrated inside the ITER ports is relied on the use of CAD models. This paper presents the CAD-based MCNP Monte Carlo radiation transport and activation analyses for the Diagnostic Upper and Equatorial Port Plugs (UPP #3 and EPP #8, #17). The creation process of the complicated 3D MCNP models of the diagnostics systems was substantially accelerated by application of the CAD-to-MCNP converter programs MCAM and McCad. High performance computing resources of the Helios supercomputer allowed to speed-up the MCNP parallel transport calculations with the MPI/OpenMP interface. The found shielding solutions could be universal, reducing ports R&D costs. The shield block behind the Tritium and Deposit Monitor (TDM) optical box was added to study its influence on Shut-Down Dose Rate (SDDR) in Port Interspace (PI) of EPP#17. Influence of neutron streaming along the Lost Alpha Monitor (LAM) on the neutron energy spectra calculated in the Tangential Neutron Spectrometer (TNS) of EPP#8. For the UPP#3 with Charge eXchange Recombination Spectroscopy (CXRS-core), an excessive neutron streaming along the CXRS shutter, which should be prevented in further design iteration.

A cargo inspection system based on pulsed fast neutron analysis (PFNA).

PubMed

Ipe, N E; Olsher, R; Ryge, P; Mrozack, J; Thieu, J

2005-01-01

A cargo inspection system based on pulsed fast neutron analysis (PFNA) is to be used at a border crossing to detect explosives and contraband hidden in trucks and cargo containers. Neutrons are produced by the interaction of deuterons in a deuterium target mounted on a moveable scan arm. The collimated pulsed fast neutron beam is used to determine the location and composition of objects in a cargo container. The neutrons produce secondary gamma rays that are characteristic of the object's elemental composition. The cargo inspection system building consists of an accelerator room and an inspection tunnel. The accelerator room is shielded and houses the injector, accelerator and the neutron production gas target. The inspection tunnel is partially shielded. The truck or container to be inspected will be moved through the inspection tunnel by a conveyor system. The facility and radiation source terms considered in the shielding design are described.

Finite Element Analysis of Absorbable Sheath to Prevent Stress Shielding of Tibial Interlocking Intramedullary Nail

NASA Astrophysics Data System (ADS)

Dong, Yansheng; Wang, Yongqing; Dong, Limin; Jia, Peng; Lu, Fengcheng

2017-07-01

The nail with absorbable sheath (AS nail) is designed to reduce the stress shielding effect of internal fixation with interlocking intramedullary nail. In order to verify its feasibility, two types of the finite element models of internal fixation of tibia with the AS nail and the common metal nail (CM nail) are established using the Softwares of Mimics, Geomagic, SolidWorks and ANSYS according to the CT scanning data of tibia. The result of the finite element analysis shows that the AS nail has great advantages compared with the CM nail in reducing the stress shielding effect in different periods of fracture healing. The conclusion is that the AS nail can realize the static fixation to the dynamic fixation from the early to the later automatically to shorten the time of fracture healing, which also provides a new technique to the interlocking intramedullary nail.

Use of satellite pictures for determining major shield fractures relevant for ore prospecting, northern Finland

NASA Technical Reports Server (NTRS)

Tuominen, H. V.; Aarnisalo, J. (Principal Investigator)

1976-01-01

The author has identified the following significant results. A combined analysis of LANDSAT 1 imagery, aeromagnetic and other maps, and aerial photos has revealed a dense network of bedrock fractures in northern Finland. They form several fracturing zones, which obviously represent surficial manifestations of major fractures. The fractures follow, in general, the eight main trends of crustal shear characteristics of the Baltic Shield, but show distinct deviations from them in detail. The major fracture zones divide the bedrock into a mosaic of polygonal blocks, which in many cases coincide with the main rock units of the area and are characterized by different patterns of internal fracturing. Known mineralizations show a tendency to concentrate along the fracture zones. Optical filtering of original LANDSAT images might provide a rapid tool for the analysis of major structural trends in extensive areas such as shields or entire continents.

Radiation shielding of the Fermilab 16 GeV proton driver

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nikolai V. Mokhov, Alexander I. Drozhdin and Oleg E. Krivosheev

2001-07-12

The radiation transport analysis in the proposed Fermi-lab 1.2 MWProton Driver (PD) [1] is fundamentally important because of the impact on machine performance, conventional facility design, maintenance operations, and related costs. The strategy adopted in the PD design is that the beam losses in the machine are localized and controlled as much as possible via the dedicated beam collimation system, with a high loss rate localized in that section and drastically lower uncontrolled beam loss rate in the rest of the lattice. Results of thorough Monte Carlo calculations of prompt and residual radiation in and around the PD components aremore » presented for realistic assumptions and geometry under normal operation and accidental conditions. This allowed one to conduct shielding design and analysis to meet regulatory requirements [2] for external shielding, hands-on maintenance and ground-water activation.« less

Space Debris Surfaces - Probability of no penetration versus impact velocity and obliquity

NASA Technical Reports Server (NTRS)

Elfer, N.; Meibaum, R.; Olsen, G.

1992-01-01

A collection of computer codes called Space Debris Surfaces (SD-SURF), have been developed to assist in the design and analysis of space debris protection systems. An SD-SURF analysis will show which obliquities and velocities are most likely to cause a penetration to help the analyst select a shield design best suited to the predominant penetration mechanism. Examples of the interaction between space vehicle geometry, the space debris environment, and the penetration and critical damage ballistic limit surfaces of the shield under consideration are presented.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brunett, A. J.; Fei, T.; Strons, P. S.

The Transient Reactor Test Facility (TREAT), located at Idaho National Laboratory (INL), is a test facility designed to evaluate the performance of reactor fuels and materials under transient accident conditions. The facility, an air-cooled, graphite-moderated reactor designed to utilize fuel containing high-enriched uranium (HEU), has been in non-operational standby status since 1994. Currently, in support of the missions of the Department of Energy (DOE) National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program, a new core design is being developed for TREAT that will utilize low-enriched uranium (LEU). The primary objective of this conversion effort ismore » to design an LEU core that is capable of meeting the performance characteristics of the existing HEU core. Minimal, if any, changes are anticipated for the supporting systems (e.g. reactor trip system, filtration/cooling system, etc.); therefore, the LEU core must also be able to function with the existing supporting systems, and must also satisfy acceptable safety limits. In support of the LEU conversion effort, a range of ancillary safety analyses are required to evaluate the LEU core operation relative to that of the existing facility. These analyses cover neutronics, shielding, and thermal hydraulic topics that have been identified as having the potential to have reduced safety margins due to conversion to LEU fuel, or are required to support the required safety analyses documentation. The majority of these ancillary tasks have been identified in [1] and [2]. The purpose of this report is to document the ancillary safety analyses that have been performed at Argonne National Laboratory during the early stages of the LEU design effort, and to describe ongoing and anticipated analyses. For all analyses presented in this report, methodologies are utilized that are consistent with, or improved from, those used in analyses for the HEU Final Safety Analysis Report (FSAR) [3]. Depending on the availability of historical data derived from HEU TREAT operation, results calculated for the LEU core are compared to measurements obtained from HEU TREAT operation. While all analyses in this report are largely considered complete and have been reviewed for technical content, it is important to note that all topics will be revisited once the LEU design approaches its final stages of maturity. For most safety significant issues, it is expected that the analyses presented here will be bounding, but additional calculations will be performed as necessary to support safety analyses and safety documentation. It should also be noted that these analyses were completed as the LEU design evolved, and therefore utilized different LEU reference designs. Preliminary shielding, neutronic, and thermal hydraulic analyses have been completed and have generally demonstrated that the various LEU core designs will satisfy existing safety limits and standards also satisfied by the existing HEU core. These analyses include the assessment of the dose rate in the hodoscope room, near a loaded fuel transfer cask, above the fuel storage area, and near the HEPA filters. The potential change in the concentration of tramp uranium and change in neutron flux reaching instrumentation has also been assessed. Safety-significant thermal hydraulic items addressed in this report include thermally-induced mechanical distortion of the grid plate, and heating in the radial reflector.« less

Analytical theory of coherent synchrotron radiation wakefield of short bunches shielded by conducting parallel plates

NASA Astrophysics Data System (ADS)

Stupakov, Gennady; Zhou, Demin

2016-04-01

We develop a general model of coherent synchrotron radiation (CSR) impedance with shielding provided by two parallel conducting plates. This model allows us to easily reproduce all previously known analytical CSR wakes and to expand the analysis to situations not explored before. It reduces calculations of the impedance to taking integrals along the trajectory of the beam. New analytical results are derived for the radiation impedance with shielding for the following orbits: a kink, a bending magnet, a wiggler of finite length, and an infinitely long wiggler. All our formulas are benchmarked against numerical simulations with the CSRZ computer code.

Analytical theory of coherent synchrotron radiation wakefield of short bunches shielded by conducting parallel plates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stupakov, Gennady; Zhou, Demin

2016-04-21

We develop a general model of coherent synchrotron radiation (CSR) impedance with shielding provided by two parallel conducting plates. This model allows us to easily reproduce all previously known analytical CSR wakes and to expand the analysis to situations not explored before. It reduces calculations of the impedance to taking integrals along the trajectory of the beam. New analytical results are derived for the radiation impedance with shielding for the following orbits: a kink, a bending magnet, a wiggler of finite length, and an infinitely long wiggler. All our formulas are benchmarked against numerical simulations with the CSRZ computer code.

Verification of Small Hole Theory for Application to Wire Chaffing Resulting in Shield Faults

NASA Technical Reports Server (NTRS)

Schuet, Stefan R.; Timucin, Dogan A.; Wheeler, Kevin R.

2011-01-01

Our work is focused upon developing methods for wire chafe fault detection through the use of reflectometry to assess shield integrity. When shielded electrical aircraft wiring first begins to chafe typically the resulting evidence is small hole(s) in the shielding. We are focused upon developing algorithms and the signal processing necessary to first detect these small holes prior to incurring damage to the inner conductors. Our approach has been to develop a first principles physics model combined with probabilistic inference, and to verify this model with laboratory experiments as well as through simulation. Previously we have presented the electromagnetic small-hole theory and how it might be applied to coaxial cable. In this presentation, we present our efforts to verify this theoretical approach with high-fidelity electromagnetic simulations (COMSOL). Laboratory observations are used to parameterize the computationally efficient theoretical model with probabilistic inference resulting in quantification of hole size and location. Our efforts in characterizing faults in coaxial cable are subsequently leading to fault detection in shielded twisted pair as well as analysis of intermittent faulty connectors using similar techniques.

Influence of shielding gas on the mechanical and metallurgical properties of DP-GMA-welded 5083-H321 aluminum alloy

NASA Astrophysics Data System (ADS)

Koushki, Amin Reza; Goodarzi, Massoud; Paidar, Moslem

2016-12-01

In the present research, 6-mm-thick 5083-H321 aluminum alloy was joined by the double-pulsed gas metal arc welding (DP-GMAW) process. The objective was to investigate the influence of the shielding gas composition on the microstructure and properties of GMA welds. A macrostructural study indicated that the addition of nitrogen and oxygen to the argon shielding gas resulted in better weld penetration. Furthermore, the tensile strength and bending strength of the welds were improved when oxygen and nitrogen (at concentrations as high as approximately 0.1vol%) were added to the shielding gas; however, these properties were adversely affected when the oxygen and nitrogen contents were increased further. This behavior was attributed to the formation of excessive brown and black oxide films on the bead surface, the formation of intermetallic compounds in the weld metal, and the formation of thicker oxide layers on the bead surface with increasing nitrogen and oxygen contents in the argon-based shielding gas. Analysis by energy-dispersive X-ray spectroscopy revealed that most of these compounds are nitrides or oxides.

Cancer mortality in residents of the terrain-shielded area exposed to fallout from the Nagasaki atomic bombing.

PubMed

Yokota, Kenichi; Mine, Mariko; Kondo, Hisayoshi; Matsuda, Naoki; Shibata, Yoshisada; Takamura, Noboru

2018-01-01

The health effects of radiation exposure from the atomic bomb fallout remain unclear. The objective of the present study is to elucidate the association between low-dose radiation exposure from the atomic bomb fallout and cancer mortality among Nagasaki atomic bomb survivors. Of 77 884 members in the Nagasaki University Atomic Bomb Survivors Cohort, 610 residents in the terrain-shielded area with fallout were selected for this analysis; 1443 residents in the terrain-shielded area without fallout were selected as a control group; and 3194 residents in the direct exposure area were also selected for study. Fifty-two deaths due to cancer in the terrain-shielded fallout area were observed during the follow-up period from 1 January 1970 to 31 December 2012. The hazard ratio for cancer mortality in the terrain-shielded fallout area was 0.90 (95% confidence interval: 0.65-1.24). No increase in the risk of cancer mortality was observed, probably because the dose of the radiation exposure was low for residents in the terrain-shielded fallout areas of the Nagasaki atomic bomb, and also because the number of study subjects was small. © The Author 2017. Published by Oxford University Press on behalf of The Japan Radiation Research Society and Japanese Society for Radiation Oncology.

Simulation Study to Improve Focalization of a Figure Eight Coil by Using a Conductive Shield Plate and a Ferromagnetic Block.

PubMed

Zhao, Chen; Zhang, Shunqi; Liu, Zhipeng; Yin, Tao

2015-07-01

A new method to improve the focalization and efficiency of the Figure of Eight (FOE) coil in rTMS is discussed in this paper. In order to decrease the half width of the distribution curve (HWDC), as well to increase the ratio of positive peak value to negative peak value (RPN) of the induced electric field, a shield plate with a window and a ferromagnetic block are assumed to enhance the positive peak value of the induced electrical field. The shield is made of highly conductive copper, and the block is made of highly permeable soft magnetic ferrite. A computer simulation is conducted on ANSYS® software to conduct the finite element analysis (FEA). Two comparing coefficients were set up to optimize the sizes of the shield window and the block. Simulation results show that a shield with a 60 mm × 30 mm sized window, together with a block 40 mm thick, can decrease the focal area of a FOE coil by 46.7%, while increasing the RPN by 135.9%. The block enhances the peak value of the electrical field induced by a shield-FOE by 8.4%. A real human head model was occupied in this paper to further verify our method.

Spacecraft Solar Particle Event (SPE) Shielding: Shielding Effectiveness as a Function of SPE model as Determined with the FLUKA Radiation Transport Code

NASA Technical Reports Server (NTRS)

Koontz, Steve; Atwell, William; Reddell, Brandon; Rojdev, Kristina

2010-01-01

Analysis of both satellite and surface neutron monitor data demonstrate that the widely utilized Exponential model of solar particle event (SPE) proton kinetic energy spectra can seriously underestimate SPE proton flux, especially at the highest kinetic energies. The more recently developed Band model produces better agreement with neutron monitor data ground level events (GLEs) and is believed to be considerably more accurate at high kinetic energies. Here, we report the results of modeling and simulation studies in which the radiation transport code FLUKA (FLUktuierende KAskade) is used to determine the changes in total ionizing dose (TID) and single-event environments (SEE) behind aluminum, polyethylene, carbon, and titanium shielding masses when the assumed form (i. e., Band or Exponential) of the solar particle event (SPE) kinetic energy spectra is changed. FLUKA simulations have fully three dimensions with an isotropic particle flux incident on a concentric spherical shell shielding mass and detector structure. The effects are reported for both energetic primary protons penetrating the shield mass and secondary particle showers caused by energetic primary protons colliding with shielding mass nuclei. Our results, in agreement with previous studies, show that use of the Exponential form of the event

NMR absolute shielding scale and nuclear magnetic dipole moment of (207)Pb.

PubMed

Adrjan, Bożena; Makulski, Włodzimierz; Jackowski, Karol; Demissie, Taye B; Ruud, Kenneth; Antušek, Andrej; Jaszuński, Michał

2016-06-28

An absolute shielding scale is proposed for (207)Pb nuclear magnetic resonance (NMR) spectroscopy. It is based on ab initio calculations performed on an isolated tetramethyllead Pb(CH3)4 molecule and the assignment of the experimental resonance frequency from the gas-phase NMR spectra of Pb(CH3)4, extrapolated to zero density of the buffer gas to obtain the result for an isolated molecule. The computed (207)Pb shielding constant is 10 790 ppm for the isolated molecule, leading to a shielding of 10799.7 ppm for liquid Pb(CH3)4 which is the accepted reference standard for (207)Pb NMR spectra. The new experimental and theoretical data are used to determine μ((207)Pb), the nuclear magnetic dipole moment of (207)Pb, by applying the standard relationship between NMR frequencies, shielding constants and nuclear moments of two nuclei in the same external magnetic field. Using the gas-phase (207)Pb and (reference) proton results and the theoretical value of the Pb shielding in Pb(CH3)4, we find μ((207)Pb) = 0.59064 μN. The analysis of new experimental and theoretical data obtained for the Pb(2+) ion in water solutions provides similar values of μ((207)Pb), in the range of 0.59000-0.59131 μN.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wood, Elizabeth Sooby; Parker, Stephen Scott; Nelson, Andrew Thomas

The Fuel Cycle Research and Development program’s Advanced Fuels Campaign is currently supporting a range of experimental efforts aimed at the development and qualification of ‘accident tolerant’ nuclear fuel forms. One route to enhance the accident tolerance of nuclear fuel is to replace the zirconium alloy cladding, which is prone to rapid oxidation in steam at elevated temperatures, with a more oxidation-resistant cladding. Several cladding replacement solutions have been envisaged. The cladding can be completely replaced with a more oxidation resistant alloy, a layered approach can be used to optimize the strength, creep resistance, and oxidation tolerance of various materials,more » or the existing zirconium alloy cladding can be coated with a more oxidation-resistant material. Molybdenum is one candidate cladding material favored due to its high temperature creep resistance. However, it performs poorly under autoclave testing and suffers degradation under high temperature steam oxidation exposure. Development of composite cladding architectures consisting of a molybdenum core shielded by a molybdenum disilicide (MoSi 2) coating is hypothesized to improve the performance of a Mo-based cladding system. MoSi 2 was identified based on its high temperature oxidation resistance in O 2 atmospheres (e.g. air and “wet air”). However, its behavior in H 2O is less known. This report presents thermogravimetric analysis (TGA), scanning electron microscopy (SEM), and x-ray diffraction (XRD) results for MoSi 2 exposed to 670-1498 K water vapor. Synthetic air (80-20%, Ar-O 2) exposures were also performed, and those results are presented here for a comparative analysis. It was determined that MoSi 2 displays drastically different oxidation behavior in water vapor than in dry air. In the 670-1498 K temperature range, four distinct behaviors are observed. Parabolic oxidation is exhibited in only 670-773 K water vapor, a temperature range in which the material pests in dry O 2 environments. From 877-1084 K in water vapor, MoSi 2 undergoes rapid mass gain resulting in oxidation throughout the bulk of the sample at 980 K and 1084 K. The resulting material displays swelling and warping after the 980-1084 K exposures. A pre-passivation heat treatment performed at 1395 K was found capable of producing a coarse SiO 2 layer that limited pesting at lower temperatures in water vapor over the time periods investigated.« less

Electromagnetic simulation of helicon plasma antennas for their electrostatic shield design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stratakos, Yorgos, E-mail: y.stratakos@gmail.com; Zeniou, Angelos, E-mail: a.zeniou@inn.demokritos.gr; Gogolides, Evangelos, E-mail: e.gogolides@inn.demokritos.gr

A detailed electromagnetic parametric analysis of the helicon antenna (half Nagoya type) is shown at 13.56 MHz using a CST Microwave Studio 2012. The antenna is used to excite plasma inside a dielectric cylinder similar to a commercial reactor. Instead of focusing on the plasma state, the authors focus on the penetration and the three dimensional distribution of electric fields through the dielectric wall. Our aim is to reduce capacitive coupling which produces unwanted longitudinal and radial electric fields. Comparison of the helicon antenna electromagnetic performance under diverse boundary conditions shows that one is allowed to use vacuum simulations without plasmamore » present in the cylinder, or approximate the plasma as a column of gyrotropic material with a tensor dielectric permittivity and with a sheath of a few millimeters in order to qualitatively predict the electric field distribution, thus avoiding a full plasma simulation. This way the analysis of the full problem is much faster and allows an optimal shield design. A detailed study of various shields shows that one can reduce the radial and axial fields by more than 1 order of magnitude compared to the unshielded antenna, while the azimuthal field is reduced only by a factor of 2. Optimal shield design in terms of pitch and spacing of openings is determined. Finally, an experimental proof of concept of the effect of shielding on reduced wall sputtering is provided, by monitoring the roughness created during oxygen plasma etching of an organic polymer.« less

Shielding and activation calculations around the reactor core for the MYRRHA ADS design

NASA Astrophysics Data System (ADS)

Ferrari, Anna; Mueller, Stefan; Konheiser, J.; Castelliti, D.; Sarotto, M.; Stankovskiy, A.

2017-09-01

In the frame of the FP7 European project MAXSIMA, an extensive simulation study has been done to assess the main shielding problems in view of the construction of the MYRRHA accelerator-driven system at SCK·CEN in Mol (Belgium). An innovative method based on the combined use of the two state-of-the-art Monte Carlo codes MCNPX and FLUKA has been used, with the goal to characterize complex, realistic neutron fields around the core barrel, to be used as source terms in detailed analyses of the radiation fields due to the system in operation, and of the coupled residual radiation. The main results of the shielding analysis are presented, as well as the construction of an activation database of all the key structural materials. The results evidenced a powerful way to analyse the shielding and activation problems, with direct and clear implications on the design solutions.

BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bevard, Bruce Balkcom; Mertyurek, Ugur; Belles, Randy

The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been storedmore » on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is utilized or referenced, justification has been provided as to why the data can be utilized for BWR fuel.« less

CANDU in-reactor quantitative visual-based inspection techniques

NASA Astrophysics Data System (ADS)

Rochefort, P. A.

2009-02-01

This paper describes two separate visual-based inspection procedures used at CANDU nuclear power generating stations. The techniques are quantitative in nature and are delivered and operated in highly radioactive environments with access that is restrictive, and in one case is submerged. Visual-based inspections at stations are typically qualitative in nature. For example a video system will be used to search for a missing component, inspect for a broken fixture, or locate areas of excessive corrosion in a pipe. In contrast, the methods described here are used to measure characteristic component dimensions that in one case ensure ongoing safe operation of the reactor and in the other support reactor refurbishment. CANDU reactors are Pressurized Heavy Water Reactors (PHWR). The reactor vessel is a horizontal cylindrical low-pressure calandria tank approximately 6 m in diameter and length, containing heavy water as a neutron moderator. Inside the calandria, 380 horizontal fuel channels (FC) are supported at each end by integral end-shields. Each FC holds 12 fuel bundles. The heavy water primary heat transport water flows through the FC pressure tube, removing the heat from the fuel bundles and delivering it to the steam generator. The general design of the reactor governs both the type of measurements that are required and the methods to perform the measurements. The first inspection procedure is a method to remotely measure the gap between FC and other in-core horizontal components. The technique involves delivering vertically a module with a high-radiation-resistant camera and lighting into the core of a shutdown but fuelled reactor. The measurement is done using a line-of-sight technique between the components. Compensation for image perspective and viewing elevation to the measurement is required. The second inspection procedure measures flaws within the reactor's end shield FC calandria tube rolled joint area. The FC calandria tube (the outer shell of the FC) is sealed by rolling its ends into the rolled joint area. During reactor refurbishment, the original FC calandria tubes are removed, potentially scratching the rolled joint area and, thereby, compromising the seal with the new FC calandria tube. The procedure involves delivering an inspection module having a radiation-resistant camera, standard lighting, and a structured lighting projector. The surface is inspected by rotating the module within the rolled joint area. If a flaw is detected, its depth and width are gauged from the profile variation of the structured lighting in a captured image. As well, the diameter profile of the area is measured from the analysis of a series of captured circumferential images of the structured lighting profiles on the surface.

Graphite/Ultra-High Modulus Polyethylene Hybrid Fiber Composites with Epoxy and Polyethylene Matrices for Cosmic Radiation Shielding

NASA Technical Reports Server (NTRS)

2003-01-01

One of the most significant technical challenges in long-duration space missions is that of protecting the crew from harmful radiation. Protection against such radiation on a manned Mars mission will be of vital importance both during transit and while on the surface of the planet. The development of multifunctional materials that serve as integral structural members of the space vehicle and provide the necessary radiation shielding for the crew would be both mission enabling and cost effective. Additionally, combining shielding and structure could reduce total vehicle mass. Hybrid laminated composite materials having both ultramodulus polyethylene (PE) and graphite fibers in epoxy and PE matrices could meet such mission requirements. PE fibers have excellent physical properties, including the highest specific strength of any known fiber. Moreover, the high hydrogen (H) content of polyethylene makes the material an excellent shielding material for cosmic radiation. When such materials are incorporated into an epoxy or PE matrix a very effective shielding material is expected. Boron (B) may be added to the matrix resin or used as a coating to further increase the shielding effectiveness due to B s ability to slow thermal neutrons. These materials may also serve as micrometeorites shields due to PE s high impact energy absorption properties. It should be noted that such materials can be fabricated by existing equipment and methods. It is the objective of this work therefore to: (a) perform preliminary analysis of the radiation transport within these materials; (b) fabricate panels for mechanical property testing before and after radiation exposure. Preliminary determination on the effectiveness of the combinations of material components on both shielding and structural efficiency will be made.

Shielding of Turbomachinery Broadband Noise from a Hybrid Wing Body Aircraft Configuration

NASA Technical Reports Server (NTRS)

Hutcheson, Florence V.; Brooks, Thomas F.; Burley, Casey L.; Bahr, Christopher J.; Stead, Daniel J.; Pope, D. Stuart

2014-01-01

The results of an experimental study on the effects of engine placement and vertical tail configuration on shielding of exhaust broadband noise radiation are presented. This study is part of the high fidelity aeroacoustic test of a 5.8% scale Hybrid Wing Body (HWB) aircraft configuration performed in the 14- by 22-Foot Subsonic Tunnel at NASA Langley Research Center. Broadband Engine Noise Simulators (BENS) were used to determine insertion loss due to shielding by the HWB airframe of the broadband component of turbomachinery noise for different airframe configurations and flight conditions. Acoustics data were obtained from flyover and sideline microphones traversed to predefined streamwise stations. Noise measurements performed for different engine locations clearly show the noise benefit associated with positioning the engine nacelles further upstream on the HWB centerbody. Positioning the engine exhaust 2.5 nozzle diameters upstream (compared to 0.5 nozzle diameters downstream) of the HWB trailing edge was found of particular benefit in this study. Analysis of the shielding performance obtained with and without tunnel flow show that the effectiveness of the fuselage shielding of the exhaust noise, although still significant, is greatly reduced by the presence of the free stream flow compared to static conditions. This loss of shielding is due to the turbulence in the model near-wake/boundary layer flow. A comparison of shielding obtained with alternate vertical tail configurations shows limited differences in level; nevertheless, overall trends regarding the effect of cant angle and vertical location are revealed. Finally, it is shown that the vertical tails provide a clear shielding benefit towards the sideline while causing a slight increase in noise below the aircraft.

Container Management During Desert Shield/Storm: An Analysis and Critique of Lessons Learned

DTIC Science & Technology

1993-04-15

across the distribution spectrum.14 These issues were grouped into five major categories: Containerization and Packaging, Distribution Management , Automation...of containers is needed, according to TDAP. Distribution - Management issues. The Desert Shield experience identified three general distribution ...recommended the formation of a 19 Theater Distribution Management Center from the assets of the Movement Control Agency (MCA) and Material Management

Designing Shelter in New Buildings. A Manual for Architects on the Preliminary Designing of Shielding from Fallout Gamma Radiation in Normally Functioning Spaces in New Buildings.

ERIC Educational Resources Information Center

Knott, Albert

Analysis of radiation fallout prevention factors in new construction is presented with emphasis on architectural shielding principles. Numerous diagrams and charts illustrate--(1) radiation and fallout properties, (2) building protection principles, (3) details and planning suggestions, and (4) tabular data interpretation. A series of charts is…

Development of a mercuric iodide solid state spectrometer for X-ray astronomy

NASA Technical Reports Server (NTRS)

Vallerga, J.

1983-01-01

Mercuric iodide detectors, experimental development for astronomical use, X ray observations of the 1980 Cygnus X-1 High State, astronomical had X ray detectors in current use, detector development, balloon flight of large area (1500 sq cm) Phoswich detectors, had X ray telescope design, shielded mercuric iodide background measurement, Monte Carlo analysis, measurements with a shielded mercuric iodide detector are discussed.

Measurement of neutron spectra in the experimental reactor LR-0

DOE Office of Scientific and Technical Information (OSTI.GOV)

Prenosil, Vaclav; Mravec, Filip; Veskrna, Martin

2015-07-01

The measurement of fast neutron fluxes is important in many areas of nuclear technology. It affects the stability of the reactor structural components, performance of fuel, and also the fuel manner. The experiments performed at the LR-0 reactor were in the past focused on the measurement of neutron field far from the core, in reactor pressure vessel simulator or in biological shielding simulator. In the present the measurement in closer regions to core became more important, especially measurements in structural components like reactor baffle. This importance increases with both reactor power increase and also long term operation. Other important taskmore » is an increasing need for the measurement close to the fuel. The spectra near the fuel are aimed due to the planned measurements with the FLIBE salt, in FHR / MSR research, where one of the task is the measurement of the neutron spectra in it. In both types of experiments there is strong demand for high working count rate. The high count rate is caused mainly by high gamma background and by high fluxes. The fluxes in core or in its vicinity are relatively high to ensure safe reactor operation. This request is met in the digital spectroscopic apparatus. All experiments were realized in the LR-0 reactor. It is an extremely flexible light water zero-power research reactor, operated by the Research Center Rez (Czech Republic). (authors)« less

X-ray photoelectron spectroscopy analysis for the chemical impact of solvent addition rate on electromagnetic shielding effectiveness of HCl-doped polyaniline nanopowders

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tantawy, Hesham Ramzy; Aston, D. Eric, E-mail: aston@uidaho.edu; Kengne, Blaise-Alexis F.

2015-11-07

An in-depth analysis of the chemical functionality in HCl-doped polyaniline (PANI) nanopowders is discussed through interpretations of x-ray photoelectron spectra. The distinctions between three PANI sample types, produced under varied synthesis conditions, are compared on the basis correlations between newly collected electron spectra for chemical analysis (or also x-ray photoelectron spectroscopy) and electromagnetic (EM) shielding effectiveness (SE) within two frequency bands (100–1500 MHz and ∼2–14 GHz). The findings are discussed with reference to previous data analysis of electrical conductivities and Raman and UV-vis spectra analyzed from replicates of the same PANI nanopowders, where only the 8–12 GHz range for SE was tested.more » They further corroborate previous results for limited-solvent conditions that enhance EM shielding. The three nanopowder types show distinctive differences in polaron, bipolaron, and polar lattice contributions. The collective findings describe the chemical connections between controlling and, most importantly, limiting the available solvent for polymerization with simultaneously doping and how it is that the newly developed solvent-limited approach for HCl-PANI nanopowders provides better shielding than traditionally solvent-rich methods by having more extended and perhaps even faster polaron delocalization than other PANI-based products. The maximum oxidation (50%) and doping (49%) levels obtained in the solvent-free nanopowders also produced the highest SE values of 37.3 ± 3.7 dB (MHz band) and 68.6 ± 4.6 dB (GHz band)« less

Heterologous expression of equine CYP3A94 and investigation of a tunable system to regulate co-expressed NADPH P450 oxidoreductase levels.

PubMed

Dettwiler, Ramona; Schmitz, Andrea L; Plattet, Philippe; Zielinski, Jana; Mevissen, Meike

2014-01-01

The activity of cytochrome P450 enzymes depends on the enzyme NADPH P450 oxidoreductase (POR). The aim of this study was to investigate the activity of the equine CYP3A94 using a system that allows to regulate the POR protein levels in mammalian cells. CYP3A94 and the equine POR were heterologously expressed in V79 cells. In the system used, the POR protein regulation is based on a destabilizing domain (DD) that transfers its instability to a fused protein. The resulting fusion protein is therefore degraded by the ubiquitin-proteasome system (UPS). Addition of "Shield-1" prevents the DD fusion protein from degradation. The change of POR levels at different Shield-1 concentrations was demonstrated by cytochrome c reduction, Western immunoblot analysis, and immunocytochemistry. The alteration of CYP3A94 activity was investigated using a substrate (BFC) known to detect CYP3A4 activity. Equine CYP3A94 was demonstrated to be metabolically active and its activity could be significantly elevated by co-expression of POR. Cytochrome c reduction was significantly increased in V79-CYP3A94/DD-POR cells compared to V79-CYP3A94 cells. Surprisingly, incubation with different Shield-1 concentrations resulted in a decrease in POR protein shown by Western immunoblot analysis. Cytochrome c reduction did not change significantly, but the CYP3A94 activity decreased more than 4-fold after incubation with 500 nM and 1 µM Shield-1 for 24 hours. No differences were obtained when V79-CYP3A94 POR cells with and without Shield-1 were compared. The basal activity levels of V79-CYP3A94/DD-POR cells were unexpectedly high, indicating that DD/POR is not degraded without Shield-1. Shield-1 decreased POR protein levels and CYP3A94 activity suggesting that Shield-1 might impair POR activity by an unknown mechanism. Although regulation of POR with the pPTuner system could not be obtained, the cell line V79-CYP3A94/DD-POR system can be used for further experiments to characterize the equine CYP3A94 since the CYP activity was significantly enhanced with co-expressed POR.

Heterologous Expression of Equine CYP3A94 and Investigation of a Tunable System to Regulate Co-Expressed NADPH P450 Oxidoreductase Levels

PubMed Central

Dettwiler, Ramona; Schmitz, Andrea L.; Plattet, Philippe; Zielinski, Jana; Mevissen, Meike

2014-01-01

The activity of cytochrome P450 enzymes depends on the enzyme NADPH P450 oxidoreductase (POR). The aim of this study was to investigate the activity of the equine CYP3A94 using a system that allows to regulate the POR protein levels in mammalian cells. CYP3A94 and the equine POR were heterologously expressed in V79 cells. In the system used, the POR protein regulation is based on a destabilizing domain (DD) that transfers its instability to a fused protein. The resulting fusion protein is therefore degraded by the ubiquitin-proteasome system (UPS). Addition of “Shield-1” prevents the DD fusion protein from degradation. The change of POR levels at different Shield-1 concentrations was demonstrated by cytochrome c reduction, Western immunoblot analysis, and immunocytochemistry. The alteration of CYP3A94 activity was investigated using a substrate (BFC) known to detect CYP3A4 activity. Equine CYP3A94 was demonstrated to be metabolically active and its activity could be significantly elevated by co-expression of POR. Cytochrome c reduction was significantly increased in V79-CYP3A94/DD-POR cells compared to V79-CYP3A94 cells. Surprisingly, incubation with different Shield-1 concentrations resulted in a decrease in POR protein shown by Western immunoblot analysis. Cytochrome c reduction did not change significantly, but the CYP3A94 activity decreased more than 4-fold after incubation with 500 nM and 1 µM Shield-1 for 24 hours. No differences were obtained when V79-CYP3A94 POR cells with and without Shield-1 were compared. The basal activity levels of V79-CYP3A94/DD-POR cells were unexpectedly high, indicating that DD/POR is not degraded without Shield-1. Shield-1 decreased POR protein levels and CYP3A94 activity suggesting that Shield-1 might impair POR activity by an unknown mechanism. Although regulation of POR with the pPTuner system could not be obtained, the cell line V79-CYP3A94/DD-POR system can be used for further experiments to characterize the equine CYP3A94 since the CYP activity was significantly enhanced with co-expressed POR. PMID:25415624

S/sub n/ analysis of the TRX metal lattices with ENDF/B version III data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wheeler, F.J.; Pearlstein, S.

1975-03-01

Two critical assemblies, designated as thermal-reactor benchmarks TRX-1 and TRX-2 for ENDF/B data testing, were analyzed using the one-dimensional S/sub n/-theory code SCAMP. The two assemblies were simple lattices of aluminum-clad, uranium-metal fuel rods in triangular arrays with D$sub 2$O as moderator and reflector. The fuel was low-enriched (1.3 percent $sup 235$U), 0.387-inch in diameter and had an active height of 48 inches. The volume ratio of water to uranium was 2.35 for the TRX-1 lattice and 4.02 for TRX-2. Full-core S/sub n/ calculations based on Version III data were performed for these assemblies and the results obtained were comparedmore » with the measured values of the multiplication factors, the ratio of epithermal-to-thermal neutron capture in $sup 238$U, the ratio of epithermal-to-thermal fission in $sup 235$U, the ratio of $sup 238$U fission to $sup 235$U fission, and the ratio of capture in $sup 238$U to fission in $sup 235$U. Reaction rates were obtained from a central region of the full- core problems. Multigroup cross sections for the reactor calculation were obtained from S/sub n/ cell calculations with resonance self-shielding calculated using the RABBLE treatment. The results of the analyses are generally consistent with results obtained by other investigators. (auth)« less

Noise Scaling and Community Noise Metrics for the Hybrid Wing Body Aircraft

NASA Technical Reports Server (NTRS)

Burley, Casey L.; Brooks, Thomas F.; Hutcheson, Florence V.; Doty, Michael J.; Lopes, Leonard V.; Nickol, Craig L.; Vicroy, Dan D.; Pope, D. Stuart

2014-01-01

An aircraft system noise assessment was performed for the hybrid wing body aircraft concept, known as the N2A-EXTE. This assessment is a result of an effort by NASA to explore a realistic HWB design that has the potential to substantially reduce noise and fuel burn. Under contract to NASA, Boeing designed the aircraft using practical aircraft design princip0les with incorporation of noise technologies projected to be available in the 2020 timeframe. NASA tested 5.8% scale-mode of the design in the NASA Langley 14- by 22-Foot Subsonic Tunnel to provide source noise directivity and installation effects for aircraft engine and airframe configurations. Analysis permitted direct scaling of the model-scale jet, airframe, and engine shielding effect measurements to full-scale. Use of these in combination with ANOPP predictions enabled computations of the cumulative (CUM) noise margins relative to FAA Stage 4 limits. The CUM margins were computed for a baseline N2A-EXTE configuration and for configurations with added noise reduction strategies. The strategies include reduced approach speed, over-the-rotor line and soft-vane fan technologies, vertical tail placement and orientation, and modified landing gear designs with fairings. Combining the inherent HWB engine shielding by the airframe with added noise technologies, the cumulative noise was assessed at 38.7 dB below FAA Stage 4 certification level, just 3.3 dB short of the NASA N+2 goal of 42 dB. This new result shows that the NASA N+2 goal is approachable and that significant reduction in overall aircraft noise is possible through configurations with noise reduction technologies and operational changes.

Ultra high molecular weight polyethylene (UHMWPE) fiber epoxy composite hybridized with Gadolinium and Boron nanoparticles for radiation shielding

NASA Astrophysics Data System (ADS)

Mani, Venkat; Prasad, Narasimha S.; Kelkar, Ajit

2016-09-01

Deep space radiations pose a major threat to the astronauts and their spacecraft during long duration space exploration missions. The two sources of radiation that are of concern are the galactic cosmic radiation (GCR) and the short lived secondary neutron radiations that are generated as a result of fragmentation that occurs when GCR strikes target nuclei in a spacecraft. Energy loss, during the interaction of GCR and the shielding material, increases with the charge to mass ratio of the shielding material. Hydrogen with no neutron in its nucleus has the highest charge to mass ratio and is the element which is the most effective shield against GCR. Some of the polymers because of their higher hydrogen content also serve as radiation shield materials. Ultra High Molecular Weight Polyethylene (UHMWPE) fibers, apart from possessing radiation shielding properties by the virtue of the high hydrogen content, are known for extraordinary properties. An effective radiation shielding material is the one that will offer protection from GCR and impede the secondary neutron radiations resulting from the fragmentation process. Neutrons, which result from fragmentation, do not respond to the Coulombic interaction that shield against GCR. To prevent the deleterious effects of secondary neutrons, targets such as Gadolinium are required. In this paper, the radiation shielding studies that were carried out on the fabricated sandwich panels by vacuum-assisted resin transfer molding (VARTM) process are presented. VARTM is a manufacturing process used for making large composite structures by infusing resin into base materials formed with woven fabric or fiber using vacuum pressure. Using the VARTM process, the hybridization of Epoxy/UHMWPE composites with Gadolinium nanoparticles, Boron, and Boron carbide nanoparticles in the form of sandwich panels were successfully carried out. The preliminary results from neutron radiation tests show that greater than 99% shielding performance was achieved with these sandwich panels. Moreover, the mechanical testing and thermo-physical analysis performed show that core materials can preserve their thermo-physical and mechanical integrity after radiation.

Evaluation of coated columbium alloy heat shields for space shuttle thermal protection system application. Volume 3, phase 3: Full size TPS evaluation

NASA Technical Reports Server (NTRS)

Baer, J. W.; Black, W. E.

1974-01-01

The thermal protection system (TPS), designed for incorporation with space shuttle orbiter systems, consists of one primary heat shield thermally and structurally isolated from the test fixture by eight peripheral guard panels, all encompassing an area of approximately 12 sq ft. TPS components include tee-stiffened Cb 752/R-512E heat shields, bi-metallic support posts, panel retainers, and high temperature insulation blankets. The vehicle primary structure was simulated by a titanium skin, frames, and stiffeners. Test procedures, manufacturing processes, and methods of analysis are fully documented. For Vol. 1, see N72-30948; for Vol. 2, see N74-15660.

Analytical theory of coherent synchrotron radiation wakefield of short bunches shielded by conducting parallel plates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stupakov, Gennady; Zhou, Demin

2016-04-21

We develop a general model of coherent synchrotron radiation (CSR) impedance with shielding provided by two parallel conducting plates. This model allows us to easily reproduce all previously known analytical CSR wakes and to expand the analysis to situations not explored before. It reduces calculations of the impedance to taking integrals along the trajectory of the beam. New analytical results are derived for the radiation impedance with shielding for the following orbits: a kink, a bending magnet, a wiggler of finite length, and an infinitely long wiggler. Furthermore, all our formulas are benchmarked against numerical simulations with the CSRZ computermore » code.« less

Experiences in utilization of research reactors in Yugoslavia

DOE Office of Scientific and Technical Information (OSTI.GOV)

Copic, M.; Gabrovsek, Z.; Pop-Jordanov, J.

1971-06-15

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied bymore » means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro envisage the role of research reactors in the promotion of nuclear power programs in relation to the size of the program, the competence of domestic industries and the degree of independence where fuel supply is concerned. (author)« less

Medial Tibial Stress Shielding: A Limitation of Cobalt Chromium Tibial Baseplates.

PubMed

Martin, J Ryan; Watts, Chad D; Levy, Daniel L; Kim, Raymond H

2017-02-01

Stress shielding is a well-recognized complication associated with total knee arthroplasty. However, this phenomenon has not been thoroughly described. Specifically, no study to our knowledge has evaluated the radiographic impact of utilizing various tibial component compositions on tibial stress shielding. We retrospectively reviewed 3 cohorts of 50 patients that had a preoperative varus deformity and were implanted with a titanium, cobalt chromium (CoCr), or an all polyethylene tibial implant. A radiographic comparative analysis was performed to evaluate the amount of medial tibial bone loss in each cohort. In addition, a clinical outcomes analysis was performed on the 3 cohorts. The CoCr was noted to have a statistically significant increase in medial tibial bone loss compared with the other 2 cohorts. The all polyethylene cohort had a statistically significantly higher final Knee Society Score and was associated with the least amount of stress shielding. The CoCr tray is the most rigid of 3 implants that were compared in this study. Interestingly, this cohort had the highest amount of medial tibial bone loss. In addition, 1 patient in the CoCr cohort had medial soft tissue irritation which was attributed to a prominent medial tibial tray which required revision surgery to mitigate the symptoms. Copyright © 2016 Elsevier Inc. All rights reserved.

Detection of explosives, shielded nuclear materials and other hazardous substances in cargo containers

NASA Astrophysics Data System (ADS)

Kuznetsov, Andrey; Evsenin, Alexey; Vakhtin, Dmitry; Gorshkov, Igor; Osetrov, Oleg; Kalinin, Valery

2006-05-01

Nanosecond Neutron Analysis / Associated Particles Technique (NNA/APT) has been used to create devices for detection of explosives, radioactive and heavily shielded nuclear materials in cargo containers. Explosives and other hazardous materials are detected by analyzing secondary high-energy gamma-rays form reactions of fast neutrons with the materials inside the container. Depending on the dimensions of the inspected containers, the detecting system consists of one or several detection modules, each of which contains a small neutron generator with built-in position sensitive detector of associated alpha-particles and several scintillator-based gamma-ray detectors. The same gamma-ray detectors are used to detect unshielded radioactive and nuclear materials. Array of several detectors of fast neutrons is used to detect neutrons from spontaneous and induced fission of nuclear materials. These neutrons can penetrate thick layers of lead shielding, which can be used to conceal gamma-radioactivity from nuclear materials. Coincidence and timing analysis allows one to discriminate between fission neutrons and scattered probing neutrons. Mathematical modeling by MCNP5 code was used to estimate the sensitivity of the device and its optimal configuration. Capability of the device to detect 1 kg of explosive imitator inside container filled with suitcases and other baggage items has been confirmed experimentally. First experiments with heavily shielded nuclear materials have been carried out.

AIR SHIPMENT OF SPENT NUCLEAR FUEL FROM THE BUDAPEST RESEARCH REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dewes, J.

2014-02-24

The shipment of spent nuclear fuel is usually done by a combination of rail, road or sea, as the high activity of the SNF needs heavy shielding. Air shipment has advantages, e.g. it is much faster than any other shipment and therefore minimizes the transit time as well as attention of the public. Up to now only very few and very special SNF shipments were done by air, as the available container (TUK6) had a very limited capacity. Recently Sosny developed a Type C overpack, the TUK-145/C, compliant with IAEA Standard TS-R-1 for the VPVR/M type Skoda container. The TUK-145/Cmore » was first used in Vietnam in July 2013 for a single cask. In October and November 2013 a total of six casks were successfully shipped from Hungary in three air shipments using the TUK-145/C. The present paper describes the details of these shipments and formulates the lessons learned.« less

Radiological characterization of the pressure vessel internals of the BNL High Flux Beam Reactor.

PubMed

Holden, Norman E; Reciniello, Richard N; Hu, Jih-Perng

2004-08-01

In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, measurements and calculations of the decay gamma-ray dose-rate were performed in the reactor pressure vessel and on vessel internal structures such as the upper and lower thermal shields, the Transition Plate, and the Control Rod blades. Measurements of gamma-ray dose rates were made using Red Perspex polymethyl methacrylate high-dose film, a Radcal "peanut" ion chamber, and Eberline's RO-7 high-range ion chamber. As a comparison, the Monte Carlo MCNP code and MicroShield code were used to model the gamma-ray transport and dose buildup. The gamma-ray dose rate at 8 cm above the center of the Transition Plate was measured to be 160 Gy h (using an RO-7) and 88 Gy h at 8 cm above and about 5 cm lateral to the Transition Plate (using Red Perspex film). This compares with a calculated dose rate of 172 Gy h using Micro-Shield. The gamma-ray dose rate was 16.2 Gy h measured at 76 cm from the reactor core (using the "peanut" ion chamber) and 16.3 Gy h at 87 cm from the core (using Red Perspex film). The similarity of dose rates measured with different instruments indicates that using different methods and instruments is acceptable if the measurement (and calculation) parameters are well defined. Different measurement techniques may be necessary due to constraints such as size restrictions.

SU-E-T-495: Neutron Induced Electronics Failure Rate Analysis for a Single Room Proton Accelerator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Knutson, N; DeWees, T; Klein, E

2014-06-01

Purpose: To determine the failure rate as a function of neutron dose of the range modulator's servo motor controller system (SMCS) while shielded with Borated Polyethylene (BPE) and unshielded in a single room proton accelerator. Methods: Two experimental setups were constructed using two servo motor controllers and two motors. Each SMCS was then placed 30 cm from the end of the plugged proton accelerator applicator. The motor was then turned on and observed from outside of the vault while being irradiated to known neutron doses determined from bubble detector measurements. Anytime the motor deviated from the programmed motion a failuremore » was recorded along with the delivered dose. The experiment was repeated using 9 cm of BPE shielding surrounding the SMCS. Results: Ten SMCS failures were recorded in each experiment. The dose per monitor unit for the unshielded SMCS was 0.0211 mSv/MU and 0.0144 mSv/MU for the shielded SMCS. The mean dose to produce a failure for the unshielded SMCS was 63.5 ± 58.3 mSv versus 17.0 ±12.2 mSv for the shielded. The mean number of MUs between failures were 2297 ± 1891 MU for the unshielded SMCS and 2122 ± 1523 MU for the shielded. A Wilcoxon Signed Ranked test showed the dose between failures were significantly different (P value = 0.044) while the number of MUs between failures were not (P value = 1.000). Statistical analysis determined a SMCS neutron dose of 5.3 mSv produces a 5% chance of failure. Depending on the workload and location of the SMCS, this failure rate could impede clinical workflow. Conclusion: BPE shielding was shown to not reduce the average failure of the SMCS and relocation of the system outside of the accelerator vault was required to lower the failure rate enough to avoid impeding clinical work flow.« less

Spectral analysis of shielded gamma ray sources using precalculated library data

NASA Astrophysics Data System (ADS)

Holmes, Thomas Wesley; Gardner, Robin P.

2015-11-01

In this work, an approach has been developed for determining the intensity of a shielded source by first determining the thicknesses of three different shielding materials from a passively collected gamma-ray spectrum by making comparisons with predetermined shielded spectra. These evaluations are dependent on the accuracy and validity of the predetermined library spectra which were created by changing the thicknesses of the three chosen materials lead, aluminum and wood that are used to simulate any actual shielding. Each of the spectra produced was generated using MCNP5 with a sufficiently large number of histories to ensure a low relative error at each channel. The materials were held in the same respective order from source to detector, where each material consisted of three individual thicknesses and a null condition. This then produced two separate data sets of 27 total shielding material situations and subsequent predetermined libraries that were created for each radionuclide source used. The technique used to calculate the thicknesses of the materials implements a Levenberg-Marquardt nonlinear search that employs a tri-linear interpolation with the respective predetermined libraries within each channel for the supplied input unknown spectrum. Given that the nonlinear parameters require an initial guess for the calculations, the approach demonstrates first that when the correct values are input, the correct thicknesses are found. It then demonstrates that when multiple trials of random values are input for each of the nonlinear parameters, the average of the calculated solutions that successfully converges also produced the correct thicknesses. Under situations with sufficient information known about the detection situation at hand, the method was shown to behave in a manner that produces reasonable results and can serve as a good preliminary solution. This technique has the capability to be used in a variety of full spectrum inverse analysis problems including homeland security issues.

KSC-2011-1977

NASA Image and Video Library

2011-03-01

CAPE CANAVERAL, Fla. -- In the Vehicle Assembly Building at NASA's Kennedy Space Center in Florida, shuttle Endeavour is being lowered into a high bay to be attached to its external fuel tank and solid rocket boosters, already positioned on the mobile launcher platform. Endeavour and its STS-134 crew will deliver the Express Logistics Carrier-3, Alpha Magnetic Spectrometer, a high-pressure gas tank, additional spare parts for Dextre and micrometeoroid debris shields to the International Space Station. Endeavour's final launch is targeted for April 19 at 7:48 p.m. EDT. For more information visit, http://www.nasa.gov/mission_pages/shuttle/shuttlemissions/sts134/index.html. Photo credit: NASA/Jim Grossmann

KSC-2011-1991

NASA Image and Video Library

2011-03-01

CAPE CANAVERAL, Fla. -- In the Vehicle Assembly Building at NASA's Kennedy Space Center in Florida, shuttle Endeavour is lowered into place where it is being attached to its external fuel tank and solid rocket boosters, already positioned on the mobile launcher platform. Endeavour and its STS-134 crew will deliver the Express Logistics Carrier-3, Alpha Magnetic Spectrometer, a high-pressure gas tank, additional spare parts for Dextre and micrometeoroid debris shields to the International Space Station. Endeavour's final launch is targeted for April 19 at 7:48 p.m. EDT. For more information visit, http://www.nasa.gov/mission_pages/shuttle/shuttlemissions/sts134/index.html. Photo credit: NASA/Kim Shiflett

KSC-2011-1990

NASA Image and Video Library

2011-03-01

CAPE CANAVERAL, Fla. -- In the Vehicle Assembly Building at NASA's Kennedy Space Center in Florida, shuttle Endeavour is lowered into place where it is being attached to its external fuel tank and solid rocket boosters, already positioned on the mobile launcher platform. Endeavour and its STS-134 crew will deliver the Express Logistics Carrier-3, Alpha Magnetic Spectrometer, a high-pressure gas tank, additional spare parts for Dextre and micrometeoroid debris shields to the International Space Station. Endeavour's final launch is targeted for April 19 at 7:48 p.m. EDT. For more information visit, http://www.nasa.gov/mission_pages/shuttle/shuttlemissions/sts134/index.html. Photo credit: NASA/Kim Shiflett

KSC-2011-1981

NASA Image and Video Library

2011-03-01

CAPE CANAVERAL, Fla. -- In the Vehicle Assembly Building at NASA's Kennedy Space Center in Florida, shuttle Endeavour is lowered into place where it will be attached to its external fuel tank and solid rocket boosters, already positioned on the mobile launcher platform. Endeavour and its STS-134 crew will deliver the Express Logistics Carrier-3, Alpha Magnetic Spectrometer, a high-pressure gas tank, additional spare parts for Dextre and micrometeoroid debris shields to the International Space Station. Endeavour's final launch is targeted for April 19 at 7:48 p.m. EDT. For more information visit, http://www.nasa.gov/mission_pages/shuttle/shuttlemissions/sts134/index.html. Photo credit: NASA/Jim Grossmann

KSC-2011-1978

NASA Image and Video Library

2011-03-01

CAPE CANAVERAL, Fla. -- In the Vehicle Assembly Building at NASA's Kennedy Space Center in Florida, shuttle Endeavour is being lowered into place where it will be attached to its external fuel tank and solid rocket boosters, already positioned on the mobile launcher platform. Endeavour and its STS-134 crew will deliver the Express Logistics Carrier-3, Alpha Magnetic Spectrometer, a high-pressure gas tank, additional spare parts for Dextre and micrometeoroid debris shields to the International Space Station. Endeavour's final launch is targeted for April 19 at 7:48 p.m. EDT. For more information visit, http://www.nasa.gov/mission_pages/shuttle/shuttlemissions/sts134/index.html. Photo credit: NASA/Jim Grossmann

KSC-2011-1976

NASA Image and Video Library

2011-03-01

CAPE CANAVERAL, Fla. -- In the Vehicle Assembly Building at NASA's Kennedy Space Center in Florida, shuttle Endeavour is poised above a high bay where it will be lowered and attached to its external fuel tank and solid rocket boosters, already positioned on the mobile launcher platform. Endeavour and its STS-134 crew will deliver the Express Logistics Carrier-3, Alpha Magnetic Spectrometer, a high-pressure gas tank, additional spare parts for Dextre and micrometeoroid debris shields to the International Space Station. Endeavour's final launch is targeted for April 19 at 7:48 p.m. EDT. For more information visit, http://www.nasa.gov/mission_pages/shuttle/shuttlemissions/sts134/index.html. Photo credit: NASA/Jim Grossmann

KSC-2011-1993

NASA Image and Video Library

2011-03-01

CAPE CANAVERAL, Fla. -- In the Vehicle Assembly Building at NASA's Kennedy Space Center in Florida, shuttle Endeavour is lowered into place where it is being attached to its external fuel tank and solid rocket boosters, already positioned on the mobile launcher platform. Endeavour and its STS-134 crew will deliver the Express Logistics Carrier-3, Alpha Magnetic Spectrometer, a high-pressure gas tank, additional spare parts for Dextre and micrometeoroid debris shields to the International Space Station. Endeavour's final launch is targeted for April 19 at 7:48 p.m. EDT. For more information visit, http://www.nasa.gov/mission_pages/shuttle/shuttlemissions/sts134/index.html. Photo credit: NASA/Kim Shiflett

KSC-2011-1992

NASA Image and Video Library

2011-03-01

CAPE CANAVERAL, Fla. -- In the Vehicle Assembly Building at NASA's Kennedy Space Center in Florida, shuttle Endeavour is lowered into place where it is being attached to its external fuel tank and solid rocket boosters, already positioned on the mobile launcher platform. Endeavour and its STS-134 crew will deliver the Express Logistics Carrier-3, Alpha Magnetic Spectrometer, a high-pressure gas tank, additional spare parts for Dextre and micrometeoroid debris shields to the International Space Station. Endeavour's final launch is targeted for April 19 at 7:48 p.m. EDT. For more information visit, http://www.nasa.gov/mission_pages/shuttle/shuttlemissions/sts134/index.html. Photo credit: NASA/Kim Shiflett

KSC-2011-1980

NASA Image and Video Library

2011-03-01

CAPE CANAVERAL, Fla. -- In the Vehicle Assembly Building at NASA's Kennedy Space Center in Florida, shuttle Endeavour is lowered into place where it will be attached to its external fuel tank and solid rocket boosters, already positioned on the mobile launcher platform. Endeavour and its STS-134 crew will deliver the Express Logistics Carrier-3, Alpha Magnetic Spectrometer, a high-pressure gas tank, additional spare parts for Dextre and micrometeoroid debris shields to the International Space Station. Endeavour's final launch is targeted for April 19 at 7:48 p.m. EDT. For more information visit, http://www.nasa.gov/mission_pages/shuttle/shuttlemissions/sts134/index.html. Photo credit: NASA/Jim Grossmann

KSC-2011-1979

NASA Image and Video Library

2011-03-01

CAPE CANAVERAL, Fla. -- In the Vehicle Assembly Building at NASA's Kennedy Space Center in Florida, shuttle Endeavour is lowered into place where it will be attached to its external fuel tank and solid rocket boosters, already positioned on the mobile launcher platform. Endeavour and its STS-134 crew will deliver the Express Logistics Carrier-3, Alpha Magnetic Spectrometer, a high-pressure gas tank, additional spare parts for Dextre and micrometeoroid debris shields to the International Space Station. Endeavour's final launch is targeted for April 19 at 7:48 p.m. EDT. For more information visit, http://www.nasa.gov/mission_pages/shuttle/shuttlemissions/sts134/index.html. Photo credit: NASA/Jim Grossmann

Extensive Radiation Shielding Analysis for Different Spacecraft Orbits

NASA Astrophysics Data System (ADS)

Çay, Yiǧit; Kaymaz, Zerefsan

2016-07-01

Radiation environment around Earth poses a great danger for spacecraft and causes immature de-orbiting or loss of the spacecraft in near Earth space environment. In this study, a student project has been designed to build a CubeSat, PolarBeeSail (PBS), with an orbit having inclination of 80°, 4 Re in perigee and 20 Re in apogee to study the polar magnetospheric environment. An extensive radiation dose analyses were carried out for PBS orbit, and integral and differential fluxes were calculated using SPENVIS tools. A shielding analysis was performed and an optimum Aluminum thickness, 3 mm, was obtained. These results for PBS were then compared for other orbits at different altitudes both for polar and equatorial orbits. For this purpose, orbital characteristics of POES-19 and GOES-15 were used. The resulting proton flux analyses, TID analyses, and further shielding studies were conducted; comparisons and recommendations were made for future design of spacecraft that will use these environments.

The Exploration Atmospheres Working Group's Report on Space Radiation Shielding Materials

NASA Technical Reports Server (NTRS)

Barghouty, A. F.; Thibeault, S. A.

2006-01-01

This part of Exploration Atmospheres Working Group analyses focuses on the potential use of nonmetallic composites as the interior walls and structural elements exposed to the atmosphere of the spacecraft or habitat. The primary drive to consider nonmetallic, polymer-based composites as an alternative to aluminum structure is due to their superior radiation shielding properties. But as is shown in this analysis, these composites can also be made to combine superior mechanical properties with superior shielding properties. In addition, these composites can be made safe; i.e., with regard to flammability and toxicity, as well as "smart"; i.e., embedded with sensors for the continuous monitoring of material health and conditions. The analysis main conclusions are that (1) smart polymer-based composites are an enabling technology for safe and reliable exploration missions, and (2) an adaptive, synergetic systems approach is required to meet the missions requirements from structure, properties, and processes to crew health and protection for exploration missions.

Radiation shielding quality assurance

NASA Astrophysics Data System (ADS)

Um, Dallsun

For the radiation shielding quality assurance, the validity and reliability of the neutron transport code MCNP, which is now one of the most widely used radiation shielding analysis codes, were checked with lot of benchmark experiments. And also as a practical example, follows were performed in this thesis. One integral neutron transport experiment to measure the effect of neutron streaming in iron and void was performed with Dog-Legged Void Assembly in Knolls Atomic Power Laboratory in 1991. Neutron flux was measured six different places with the methane detectors and a BF-3 detector. The main purpose of the measurements was to provide benchmark against which various neutron transport calculation tools could be compared. Those data were used in verification of Monte Carlo Neutron & Photon Transport Code, MCNP, with the modeling for that. Experimental results and calculation results were compared in both ways, as the total integrated value of neutron fluxes along neutron energy range from 10 KeV to 2 MeV and as the neutron spectrum along with neutron energy range. Both results are well matched with the statistical error +/-20%. MCNP results were also compared with those of TORT, a three dimensional discrete ordinates code which was developed by Oak Ridge National Laboratory. MCNP results are superior to the TORT results at all detector places except one. This means that MCNP is proved as a very powerful tool for the analysis of neutron transport through iron & air and further it could be used as a powerful tool for the radiation shielding analysis. For one application of the analysis of variance (ANOVA) to neutron and gamma transport problems, uncertainties for the calculated values of critical K were evaluated as in the ANOVA on statistical data.

BUMPERII - DESIGN ANALYSIS CODE FOR OPTIMIZING SPACECRAFT SHIELDING AND WALL CONFIGURATION FOR ORBITAL DEBRIS AND METEOROID IMPACTS

NASA Technical Reports Server (NTRS)

Hill, S. A.

1994-01-01

BUMPERII is a modular program package employing a numerical solution technique to calculate a spacecraft's probability of no penetration (PNP) from man-made orbital debris or meteoroid impacts. The solution equation used to calculate the PNP is based on the Poisson distribution model for similar analysis of smaller craft, but reflects the more rigorous mathematical modeling of spacecraft geometry, orientation, and impact characteristics necessary for treatment of larger structures such as space station components. The technique considers the spacecraft surface in terms of a series of flat plate elements. It divides the threat environment into a number of finite cases, then evaluates each element of each threat. The code allows for impact shielding (shadowing) of one element by another in various configurations over the spacecraft exterior, and also allows for the effects of changing spacecraft flight orientation and attitude. Four main modules comprise the overall BUMPERII package: GEOMETRY, RESPONSE, SHIELD, and CONTOUR. The GEOMETRY module accepts user-generated finite element model (FEM) representations of the spacecraft geometry and creates geometry databases for both meteoroid and debris analysis. The GEOMETRY module expects input to be in either SUPERTAB Universal File Format or PATRAN Neutral File Format. The RESPONSE module creates wall penetration response databases, one for meteoroid analysis and one for debris analysis, for up to 100 unique wall configurations. This module also creates a file containing critical diameter as a function of impact velocity and impact angle for each wall configuration. The SHIELD module calculates the PNP for the modeled structure given exposure time, operating altitude, element ID ranges, and the data from the RESPONSE and GEOMETRY databases. The results appear in a summary file. SHIELD will also determine the effective area of the components and the overall model, and it can produce a data file containing the probability of penetration values per surface area for each element in the model. The SHIELD module writes this data file in either SUPERTAB Universal File Format or PATRAN Neutral File Format so threat contour plots can be generated as a post-processing feature of the FEM programs SUPERTAB and PATRAN. The CONTOUR module combines the functions of the RESPONSE module and most of the SHIELD module functions allowing determination of ranges of PNP's by looping over ranges of shield and/or wall thicknesses. A data file containing the PNP's for the corresponding shield and vessel wall thickness is produced. Users may perform sensitivity studies of two kinds. The effects of simple variations in orbital time, surface area, and flux may be analyzed by making changes to the terms in the equation representing the average number of penetrating particles per unit time in the PNP solution equation. The package analyzes other changes, including model environment, surface area, and configuration, by re-running the solution sequence with new GEOMETRY and RESPONSE data. BUMPERII can be run with no interactive output to the screen during execution. This can be particularly useful during batch runs. BUMPERII is written in FORTRAN 77 for DEC VAX series computers running under VMS, and was written for use with the finite-element model code SUPERTAB or PATRAN as both a pre-processor and a post-processor. Use of an alternate FEM code will require either development of a translator to change data format or modification of the GEOMETRY subroutine in BUMPERII. This program is available in DEC VAX BACKUP format on a 9-track 1600 BPI magnetic tape (standard distribution media) or on TK50 tape cartridge. The original BUMPER code was developed in 1988 with the BUMPERII revisions following in 1991 and 1992. SUPERTAB is a former name for I-DEAS. I-DEAS Finite Element Modeling is a trademark of Structural Dynamics Research Corporation. DEC, VAX, VMS and TK50 are trademarks of Digital Equipment Corporation.

Fuel Cell Technology Status Analysis | Hydrogen and Fuel Cells | NREL

Science.gov Websites

Technology Status Analysis Fuel Cell Technology Status Analysis Get Involved Fuel cell developers interested in collaborating with NREL on fuel cell technology status analysis should send an email to NREL's Technology Validation Team at techval@nrel.gov. NREL's analysis of fuel cell technology provides objective

Flame stabilization and mixing characteristics in a Stagnation Point Reverse Flow combustor

NASA Astrophysics Data System (ADS)

Bobba, Mohan K.

A novel combustor design, referred to as the Stagnation Point Reverse-Flow (SPRF) combustor, was recently developed that is able to operate stably at very lean fuel-air mixtures and with low NOx emissions even when the fuel and air are not premixed before entering the combustor. The primary objective of this work is to elucidate the underlying physics behind the excellent stability and emissions performance of the SPRF combustor. The approach is to experimentally characterize velocities, species mixing, heat release and flame structure in an atmospheric pressure SPRF combustor with the help of various optical diagnostic techniques: OH PLIF, chemiluminescence imaging, PIV and Spontaneous Raman Scattering. Results indicate that the combustor is primarily stabilized in a region downstream of the injector that is characterized by low average velocities and high turbulence levels; this is also the region where most of the heat release occurs. High turbulence levels in the shear layer lead to increased product entrainment levels, elevating the reaction rates and thereby enhancing the combustor stability. The effect of product entrainment on chemical timescales and the flame structure is illustrated with simple reactor models. Although reactants are found to burn in a highly preheated (1300 K) and turbulent environment due to mixing with hot product gases, the residence times are sufficiently long compared to the ignition timescales such that the reactants do not autoignite. Turbulent flame structure analysis indicates that the flame is primarily in the thin reaction zones regime throughout the combustor, and it tends to become more flamelet like with increasing distance from the injector. Fuel-air mixing measurements in case of non-premixed operation indicate that the fuel is shielded from hot products until it is fully mixed with air, providing nearly premixed performance without the safety issues associated with premixing. The reduction in NOx emissions in the SPRF combustor are primarily due to its ability to stably operate under ultra lean (and nearly premixed) condition within the combustor. Further, to extend the usefulness of this combustor configuration to various applications, combustor geometry scaling rules were developed with the help of simplified coaxial and opposed jet models.

Influence of Short Distance Super-large Diameter Shield Tunneling on Existing Tunnels in Sea Areas

NASA Astrophysics Data System (ADS)

Li, Zhuolin; Liu, Dagang; Wang, Mingnian; Xiao, Shihui; Yuan, Jiawei

2018-03-01

In oder to find out the influence of large diameter shield tunneling on the existing tunnel under the condition of compound strata in the sea area, taking the Maliuzhou traffic tunnel as the research background, numerical simulation and field test were combined to get the regulation of the additional internal force and deformation of the existing tunnel caused by the shield tunneling. Analysis of the data showed that: the shield construction caused the secondary additional internal force; The moment of the vault was most affected by the tunnel excavation; The axial force of the arch bottom was most affected by the excavation of the tunnel. The deformation of arch waist near excavation tunnel was more affected by tunnel excavation than that of the other side. Combined with the construction experience, the influence of the tunnel close-distance construction on the existing tunnel was within the control range, which could ensure the normal construction.

Resent Status of ITER Equatorial Launcher Development

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takahashi, K.; Kajiwara, K.; Kasugai, A.

2009-11-26

The ITER equatorial launcher is divided into a front shield and a port plug. The front shield is composed of fourteen blanket shield modules so as to form three openings for the injection of mm-wave beams into plasma. Twenty-four waveguide transmission lines, internal shields, cooling pipes and so on are installed in the port plug. The transmission lines consist of the corrugated waveguides, miter bends and the free space propagation region utilizing two mirrors in front of the waveguide outlet. The analysis of mm-wave beam propagation in the region shows that the transmission efficiency more than 99.5% is attained. Themore » high power experiments of the launcher mock-up have been carried out and the measured field patterns at each mirror and the outlet of the launcher are agreed with the calculations. It is concluded that the transmission line components in the launcher mock-up are fabricated as designed and the present mm-wave design in the launcher is feasible.« less

Shielding Gas and Heat Input Effects on the Mechanical and Metallurgical Characterization of Gas Metal Arc Welding of Super Martensitic Stainless Steel (12Cr5Ni2Mo) Joints

NASA Astrophysics Data System (ADS)

Prabakaran, T.; Prabhakar, M.; Sathiya, P.

This paper deals with the effects of shielding gas mixtures (100% CO2, 100% Ar and 80 % Ar + 20% CO2) and heat input (3.00, 3.65 and 4.33kJ/mm) on the mechanical and metallurgical characteristics of AISI 410S (American Iron and Steel Institute) super martensitic stainless steel (SMSS) by gas metal arc welding (GMAW) process. AISI 410S SMSS with 1.2mm diameter of a 410 filler wire was used in this study. A detailed microstructural analysis of the weld region as well as the mechanical properties (impact, microhardness and tensile tests at room temperature and 800∘C) was carried out. The tensile and impact fracture surfaces were further analyzed through scanning electron microscope (SEM). 100% Ar shielded welds have a higher amount of δ ferrite content and due to this fact the tensile strength of the joints is superior to the other two shielded welds.

A Practical Approach to Starting Fission Surface Power Development

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2006-01-01

The Prometheus Power and Propulsion Program has been reformulated to address NASA needs relative to lunar and Mars exploration. Emphasis has switched from the Jupiter Icy Moons Orbiter (JIMO) flight system development to more generalized technology development addressing Fission Surface Power (FSP) and Nuclear Thermal Propulsion (NTP). Current NASA budget priorities and the deferred mission need date for nuclear systems prohibit a fully funded reactor Flight Development Program. However, a modestly funded Advanced Technology Program can and should be conducted to reduce the risk and cost of future flight systems. A potential roadmap for FSP technology development leading to possible flight applications could include three elements: 1) Conceptual Design Studies, 2) Advanced Component Technology, and 3) Non-Nuclear System Testing. The Conceptual Design Studies would expand on recent NASA and DOE analyses while increasing the depth of study in areas of greatest uncertainty such as reactor integration and human-rated shielding. The Advanced Component Technology element would address the major technology risks through development and testing of reactor fuels, structural materials, primary loop components, shielding, power conversion, heat rejection, and power management and distribution (PMAD). The Non-Nuclear System Testing would provide a modular, technology testbed to investigate and resolve system integration issues.

Estimation of coolant void reactivity for CANDU-NG lattice using DRAGON and validation using MCNP5 and TRIPOLI-4.3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Karthikeyan, R.; Tellier, R. L.; Hebert, A.

2006-07-01

The Coolant Void Reactivity (CVR) is an important safety parameter that needs to be estimated at the design stage of a nuclear reactor. It helps to have an a priori knowledge of the behavior of the system during a transient initiated by the loss of coolant. In the present paper, we have attempted to estimate the CVR for a CANDU New Generation (CANDU-NG) lattice, as proposed at an early stage of the Advanced CANDU Reactor (ACR) development. We have attempted to estimate the CVR with development version of the code DRAGON, using the method of characteristics. DRAGON has several advancedmore » self-shielding models incorporated in it, each of them compatible with the method of characteristics. This study will bring to focus the performance of these self-shielding models, especially when there is voiding of such a tight lattice. We have also performed assembly calculations in 2 x 2 pattern for the CANDU-NG fuel, with special emphasis on checkerboard voiding. The results obtained have been validated against Monte Carlo codes MCNP5 and TRIPOLI-4.3. (authors)« less

Modified Laser and Thermos cell calculations on microcomputers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shapiro, A.; Huria, H.C.

1987-01-01

In the course of designing and operating nuclear reactors, many fuel pin cell calculations are required to obtain homogenized cell cross sections as a function of burnup. In the interest of convenience and cost, it would be very desirable to be able to make such calculations on microcomputers. In addition, such a microcomputer code would be very helpful for educational course work in reactor computations. To establish the feasibility of making detailed cell calculations on a microcomputer, a mainframe cell code was compiled and run on a microcomputer. The computer code Laser, originally written in Fortran IV for the IBM-7090more » class of mainframe computers, is a cylindrical, one-dimensional, multigroup lattice cell program that includes burnup. It is based on the MUFT code for epithermal and fast group calculations, and Thermos for the thermal calculations. There are 50 fast and epithermal groups and 35 thermal groups. Resonances are calculated assuming a homogeneous system and then corrected for self-shielding, Dancoff, and Doppler by self-shielding factors. The Laser code was converted to run on a microcomputer. In addition, the Thermos portion of Laser was extracted and compiled separately to have available a stand alone thermal code.« less

Spectroscopic determinations of carbon fluxes, sources, and shielding in the DIII-D divertors

NASA Astrophysics Data System (ADS)

Isler, R. C.; Colchin, R. J.; Brooks, N. H.; Evans, T. E.; West, W. P.; Whyte, D. G.

2001-10-01

The most important mechanisms for eroding plasma-facing components (PFCs) and introducing carbon into tokamak divertors are believed to be physical sputtering, chemical sputtering, sublimation, and radiation enhance sublimation (RES). The relative importance of these processes has been investigated by analyzing the spectral emission rates and the effective temperatures of CI, CD, and C2 under several operating conditions in the DIII-D tokamak [Plasma Physics Controlled Nuclear Fusion Research, 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159; Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque (Institute of Electrical and Electronic Engineers, Piscataway, 1999), p. 515]. Discrimination of chemical sputtering from physical sputtering is accomplished by quantitatively relating the fraction of CI influxes expected from dissociation of hydrocarbons to the measured CD and C2 influxes. Characteristics of sublimation are studied from carbon test samples heated to surface temperatures exceeding 2000 K. The shielding efficiency of carbon produced at the divertor target is assessed from comparison of fluxes of neutral atoms and ions; approximately 95% of the primary influx appears to be redeposited before being transported far enough upstream to fuel the core plasma.

Spectroscopic determinations of carbon fluxes, sources, and shielding in the DIII-D divertors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Isler, R. C.; Colchin, R. J.; Brooks, N. H.

2001-10-01

The most important mechanisms for eroding plasma-facing components (PFCs) and introducing carbon into tokamak divertors are believed to be physical sputtering, chemical sputtering, sublimation, and radiation enhance sublimation (RES). The relative importance of these processes has been investigated by analyzing the spectral emission rates and the effective temperatures of CI, CD, and C{sub 2} under several operating conditions in the DIII-D tokamak [Plasma Physics Controlled Nuclear Fusion Research, 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159; Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque (Institute of Electrical and Electronic Engineers, Piscataway, 1999), p. 515]. Discriminationmore » of chemical sputtering from physical sputtering is accomplished by quantitatively relating the fraction of CI influxes expected from dissociation of hydrocarbons to the measured CD and C{sub 2} influxes. Characteristics of sublimation are studied from carbon test samples heated to surface temperatures exceeding 2000 K. The shielding efficiency of carbon produced at the divertor target is assessed from comparison of fluxes of neutral atoms and ions; approximately 95% of the primary influx appears to be redeposited before being transported far enough upstream to fuel the core plasma.« less

Decontamination and deactivation of the power burst facility at the Idaho National Laboratory.

PubMed

Greene, Christy Jo

2007-05-01

Successful decontamination and deactivation of the Power Burst Facility located at the Idaho National Laboratory was accomplished through the use of extensive planning, job sequencing, engineering controls, continuous radiological support, and the use of a dedicated group of experienced workers. Activities included the removal and disposal of irradiated fuel, miscellaneous reactor components and debris stored in the canal, removal and disposition of a 15.6 curie Pu:Be start-up source, removal of an irradiated in-pile tube, and the removal of approximately 220,000 pounds of lead that was used as shielding primarily in Cubicle 13. The canal and reactor vessel were drained and water was transferred to an evaporation tank adjacent to the facility. The canal was decontaminated using underwater divers, and epoxy was affixed to the interior surfaces of the canal to contain loose contamination. The support structures and concrete or steel frame walls that form the confinement were left in place. The reactor core was left in place and a carbon steel shielding plate was placed over the reactor core to reduce radiation levels. All low-level waste and mixed low level waste generated as a result of the work activities was characterized and disposed.

Simbol-X Background Minimization: Mirror Spacecraft Passive Shielding Trade-off Study

NASA Astrophysics Data System (ADS)

Fioretti, V.; Malaguti, G.; Bulgarelli, A.; Palumbo, G. G. C.; Ferri, A.; Attinà, P.

2009-05-01

The present work shows a quantitative trade-off analysis of the Simbol-X Mirror Spacecraft (MSC) passive shielding, in the phase space of the various parameters: mass budget, dimension, geometry and composition. A simplified physical (and geometrical) model of the sky screen, implemented by means of a GEANT4 simulation, has been developed to perform a performance-driven mass optimization and evaluate the residual background level on Simbol-X focal plane.

The integral line-beam method for gamma skyshine analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shultis, J.K.; Faw, R.E.; Bassett, M.S.

1991-03-01

This paper presents a refinement of a simplified method, based on line-beam response functions, for performing skyshine calculations for shielded and collimated gamma-ray sources. New coefficients for an empirical fit to the line-beam response function are provided and a prescription for making the response function continuous in energy and emission direction is introduced. For a shielded source, exponential attenuation and a buildup factor correction for scattered photons in the shield are used. Results of the new integral line-beam method of calculation are compared to a variety of benchmark experimental data and calculations and are found to give generally excellent agreementmore » at a small fraction of the computational expense required by other skyshine methods.« less

Upper mantle structure under western Saudi Arabia from Rayleigh wave tomography and the origin of Cenozoic uplift and volcanism on the Arabian Shield

DOE Office of Scientific and Technical Information (OSTI.GOV)

Park, Y; Nyblade, A; Rodgers, A

2007-11-09

The shear velocity structure of the shallow upper mantle beneath the Arabian Shield has been modeled by inverting new Rayleigh wave phase velocity measurements between 45 and 140 s together with previously published Rayleigh wave group velocity measurement between 10 and 45 s. For measuring phase velocities, we applied a modified array method that minimizes the distortion of raypaths by lateral heterogeneity. The new shear velocity model shows a broad low velocity region in the lithospheric mantle across the Shield and a low velocity region at depths {ge} 150 km localized along the Red Sea coast and Makkah-Madinah-Nafud (MMN) volcanicmore » line. The velocity reduction in the upper mantle corresponds to a temperature anomaly of {approx}250-330 K. These finding, in particular the region of continuous low velocities along the Red Sea and MMN volcanic line, do not support interpretations for the origin of the Cenozoic plateau uplift and volcanism on the Shield invoking two separate plumes. When combined with images of the 410 and 660 km discontinuities beneath the southern part of the Arabian Shield, body wave tomographic models, a S-wave polarization analysis, and SKS splitting results, our new model supports an interpretation invoking a thermal upwelling of warm mantle rock originating in the lower mantle under Africa that crosses through the transition zone beneath Ethiopia and moves to the north and northwest under the eastern margin of the Red Sea and the Arabian Shield. In this interpretation, the difference in mean elevation between the Platform and Shield can be attributed to isostatic uplift caused by heating of the lithospheric mantle under the Shield, with significantly higher region along the Red Sea possibly resulting from a combination of lithosphere thinning and dynamic uplift.« less

Phanerozoic burial, uplift and denudation of the Equatorial Atlantic margin of South America

NASA Astrophysics Data System (ADS)

Japsen, Peter; Bonow, Johan M.; Green, Paul F.; dall'Asta, Massimo; Roig, Jean-Yves; Theveniaut, Hervé

2017-04-01

We have initiated a study aimed at understanding the history of burial, uplift and denudation of the South American Equatorial Atlantic Margin (SAEAM Uplift) including the Guiana Shield to provide a framework for investigating the hydrocarbon prospectivity of the offshore region. We report first results including observations from fieldwork at the northern and southern flank of the Guiana Shield. The study combines apatite fission-track analysis (AFTA) and vitrinite reflectance data from samples of outcrops and drillcores, sonic velocity data from drill holes and stratigraphic landscape analysis (mapping of peneplains) - all constrained by geological evidence, following the methods of Green et al. (2013). The study will thus combine the thermal history from AFTA data with the denudation history from stratigraphic landscape analysis to provide magnitudes and timing of vertical movements (Japsen et al. 2012, 2016). Along the Atlantic margin of Suriname and French Guiana, tilted and truncated Lower Cretaceous strata rest on Precambrian basement (Sapin et al. 2016). Our AFTA data show that the basement underwent Mesozoic exhumation prior to deposition of the Lower Cretaceous cover. Sub-horizontal peneplains define the landscape of the Guiana Shield at elevations up to 500 m a.s.l. As these sub-horizontal peneplains truncate the tilted, sub-Cretaceous surface along the Atlantic margin, these peneplains were therefore formed and uplifted in post-Cretaceous time. This interpretation is in good agreement with our AFTA data that define Paleogene exhumation along the margin and with the results of Theveniaut and Freyssinet (2002) who used palaeomagnetic data to conclude that bauxitic surfaces across basement at up to 400 m a.s.l. on the Guiana Shield formed during the Palaeogene. Integration of the results from AFTA with stratigraphic landscape analysis (currently in progress) and geological evidence will provide a robust reconstruction of the tectonic development of the onshore margin. References Green, Lidmar-Bergström, Japsen, Bonow & Chalmers 2013: Stratigraphic landscape analysis, thermochronology and the episodic development of elevated passive continental margins. GEUS Bulletin. Japsen, Green, Bonow & Erlström 2016: Episodic burial and exhumation of the southern Baltic Shield: Epeirogenic uplifts during and after break-up of Pangea. Gondwana Research. Japsen, Bonow, Green, Cobbold, Chiossi et al. 2012: Episodic burial and exhumation history of NE Brazil after opening of the South Atlantic. GSA Bulletin. Sapin, Davaux, dall'Asta et al. 2016: Post-rift subsidence of the French Guiana hyper-oblique margin: from rift-inherited subsidence to Amazon deposition effect. Geol. Soc. Spec. Publ. Theveniaut & Freyssinet 2002: Timing of lateritization on the Guiana Shield: synthesis of paleomagnetic results from French Guiana and Suriname. 3 x Palaeo.

Shielding analysis of the Microtron MT-25 bunker using the MCNP-4C code and NCRP Report 51.

PubMed

Casanova, A O; López, N; Gelen, A; Guevara, M V Manso; Díaz, O; Cimino, L; D'Alessandro, K; Melo, J C

2004-01-01

A cyclic electron accelerator Microtron MT-25 will be installed in Havana, Cuba. Electrons, neutrons and gamma radiation up to 25 MeV can be produced in the MT-25. A detailed shielding analysis for the bunker is carried out using two ways: the NCRP-51 Report and the Monte Carlo Method (MCNP-4C Code). The walls and ceiling thicknesses are estimated with dose constraints of 0.5 and 20 mSv y(-1), respectively, and an area occupancy factor of 1/16. Both results are compared and a preliminary bunker design is shown. Copyright 2004 Oxford University Press

Unified aeroacoustics analysis for high speed turboprop aerodynamics and noise. Volume 2: Development of theory for wing shielding

NASA Technical Reports Server (NTRS)

Amiet, R. K.

1991-01-01

A unified theory for aerodynamics and noise of advanced turboprops is presented. The theory and a computer code developed for evaluation at the shielding benefits that might be expected by an aircraft wing in a wing-mounted propeller installation are presented. Several computed directivity patterns are presented to demonstrate the theory. Recently with the advent of the concept of using the wing of an aircraft for noise shielding, the case of diffraction by a surface in a flow has been given attention. The present analysis is based on the case of diffraction of no flow. By combining a Galilean and a Lorentz transform, the wave equation with a mean flow can be reduced to the ordinary equation. Allowance is also made in the analysis for the case of a swept wing. The same combination of Galilean and Lorentz transforms lead to a problem with no flow but a different sweep. The solution procedures for the cases of leading and trailing edges are basically the same. Two normalizations of the solution are given by the computer program. FORTRAN computer programs are presented with detailed documentation. The output from these programs compares favorably with the results of other investigators.

Hypervelocity Impact Initiation of Explosive Transfer Lines

NASA Technical Reports Server (NTRS)

Bjorkman, Michael D.; Christiansen, Eric L.

2012-01-01

The Gemini, Apollo and Space Shuttle spacecraft utilized explosive transfer lines (ETL) in a number of applications. In each case the ETL was located behind substantial structure and the risk of impact initiation by micrometeoroids and orbital debris was negligible. A current NASA program is considering an ETL to synchronize the actuation of pyrobolts to release 12 capture latches in a contingency. The space constraints require placing the ETL 50 mm below the 1 mm thick 2024-T72 Whipple shield. The proximity of the ETL to the thin shield prompted analysts at NASA to perform a scoping analysis with a finite-difference hydrocode to calculate impact parameters that would initiate the ETL. The results suggest testing is required and a 12 shot test program with surplused Shuttle ETL is scheduled for February 2012 at the NASA White Sands Test Facility. Explosive initiation models are essential to the analysis and one exists in the CTH library for HNS I, but not the HNS II used in the Shuttle 2.5 gr/ft rigid shielded mild detonating cord (SMDC). HNS II is less sensitive than HNS I so it is anticipated that these results using the HNS I model are conservative. Until the hypervelocity impact test results are available, the only check on the analysis was comparison with the Shuttle qualification test result that a 22 long bullet would not initiate the SMDC. This result was reproduced by the hydrocode simulation. Simulations of the direct impact of a 7 km/s aluminum ball, impacting at 0 degree angle of incidence, onto the SMDC resulted in a 1.5 mm diameter ball initiating the SMDC and 1.0 mm ball failing to initiate it. Where one 1.0 mm ball could not initiate the SMDC, a cluster of six 1.0 mm diameter aluminum balls striking simultaneously could. Thus the impact parameters that will result in initiating SMDC located behind a Whipple shield will depend on how well the shield fragments the projectile and spreads the fragments. An end-to-end simulation of the impact of an aluminum ball onto a Whipple shield covering SMDC is problematic due to the hydrocode fracture models. Regardless, two simulations were performed resulting in a 5 mm ball initiating the SMDC and a 4 mm ball failing to initiate the SMDC.

Bitumoids in the crystalline rocks of the Kola superdeep drillhole

DOE Office of Scientific and Technical Information (OSTI.GOV)

Belokon, V.G.

1987-04-01

The genetic regularities in the distribution of organic fuels of various elemental compositions and molecular structures and their relationship to the processes of formation of regional structures present some of the most pressing and complicated problems of modern fossil-fuel geology. Regardless of the difference in molecular structure of the final products of these reactions and the phase state in nature (gases, petroleums, bitumens, lignite, bituminous coal or anthracite), fossil fuels manifest the property of carbon and hydrogen to yield a vast number of compounds with different extents of ordering of the structure, from simple linear compounds (methane and its homologs)more » to cyclic compounds of the graphite series. Karavayev worked out a classification diagram for solid fuels, based on calculation of the variation in the elemental composition of the organic matter. The ratio of hydrogen to carbon atoms, as a reflection of the extent of aromatization of the structure, is taken as a classification criterion. In investigating the earth's crust in the Baltic shield, the Kola superdeep drillhole found organic matter in the form of bitumoids, in the extractable part of which a broad spectrum of compounds was identified. Bitumoids are similar to humites in molecular structure, only somewhat more ordered. This paper applies Karavayev's principle to this type of compound. It was found that the elemental compositions of the organic matter from basement depths down to 10 km show patterns analogous to those from sedimentary basins. 9 references.« less

Nondestructive Evaluation of the VSC-17 Cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sheryl Morton; Al Carlson; Cecilia Hoffman

2006-01-01

In 2003, representatives from the Central Research Institute of Electric Power Industry (CRIEPI) requested development of a project with the objective of determining the performance of a concrete spent nuclear fuel storage cask. Radiation and environmental effects may cause chemical alteration of the concrete that could result in excessive cracking, spalling, and loss of compressive strength. The Idaho National Laboratory (INL) project team and CRIEPI representatives identified the Ventilated Storage Cask (VSC 17) spent nuclear fuel storage cask, originally located at the INL Test Area North, as a candidate to study cask performance because it had been used to storemore » fuel as part of a dry cask storage demonstration project for over 15 years. The project involved investigating the properties of the concrete shield. INL performed a survey of the cask in the summers of 2003 and 2004. The INL team met with the CRIEPI representatives in December of 2004 to discuss the next steps. As a result of that meeting, CRIEPI requested that in the summer 2005 INL perform additional surveys on the VSC 17 cask with participation of CRIEPI scientists. This document summarizes the evaluation methods used on the VSC 17 to evaluate the cask for compressive strength, concrete cracking, concrete thickness, and temperature distribution.« less

In vivo dosimetry and shielding disk alignment verification by EBT3 GAFCHROMIC film in breast IOERT treatment.

PubMed

Severgnini, Mara; de Denaro, Mario; Bortul, Marina; Vidali, Cristiana; Beorchia, Aulo

2014-01-08

Intraoperative electron radiation therapy (IOERT) cannot usually benefit, as conventional external radiotherapy, from software systems of treatment planning based on computed tomography and from common dose verify procedures. For this reason, in vivo film dosimetry (IVFD) proves to be an effective methodology to evaluate the actual radiation dose delivered to the target. A practical method for IVFD during breast IOERT was carried out to improve information on the dose actually delivered to the tumor target and on the alignment of the shielding disk with respect to the electron beam. Two EBT3 GAFCHROMIC films have been positioned on the two sides of the shielding disk in order to obtain the dose maps at the target and beyond the disk. Moreover the postprocessing analysis of the dose distribution measured on the films provides a quantitative estimate of the misalignment between the collimator and the disk. EBT3 radiochromic films have been demonstrated to be suitable dosimeters for IVD due to their linear dose-optical density response in a narrow range around the prescribed dose, as well as their capability to be fixed to the shielding disk without giving any distortion in the dose distribution. Off-line analysis of the radiochromic film allowed absolute dose measurements and this is indeed a very important verification of the correct exposure to the target organ, as well as an estimate of the dose to the healthy tissue underlying the shielding. These dose maps allow surgeons and radiation oncologists to take advantage of qualitative and quantitative feedback for setting more accurate treatment strategies and further optimized procedures. The proper alignment using elastic bands has improved the absolute dose accuracy and the collimator disk alignment by more than 50%.

In vivo dosimetry and shielding disk alignment verification by EBT3 GAFCHROMIC film in breast IOERT treatment

PubMed Central

de Denaro, Mario; Bortul, Marina; Vidali, Cristiana; Beorchia, Aulo

2014-01-01

Intraoperative electron radiation therapy (IOERT) cannot usually benefit, as conventional external radiotherapy, from software systems of treatment planning based on computed tomography and from common dose verify procedures. For this reason, in vivo film dosimetry (IVFD) proves to be an effective methodology to evaluate the actual radiation dose delivered to the target. A practical method for IVFD during breast IOERT was carried out to improve information on the dose actually delivered to the tumor target and on the alignment of the shielding disk with respect to the electron beam. Two EBT3 GAFCHROMIC films have been positioned on the two sides of the shielding disk in order to obtain the dose maps at the target and beyond the disk. Moreover the postprocessing analysis of the dose distribution measured on the films provides a quantitative estimate of the misalignment between the collimator and the disk. EBT3 radiochromic films have been demonstrated to be suitable dosimeters for IVD due to their linear dose‐optical density response in a narrow range around the prescribed dose, as well as their capability to be fixed to the shielding disk without giving any distortion in the dose distribution. Off‐line analysis of the radiochromic film allowed absolute dose measurements and this is indeed a very important verification of the correct exposure to the target organ, as well as an estimate of the dose to the healthy tissue underlying the shielding. These dose maps allow surgeons and radiation oncologists to take advantage of qualitative and quantitative feedback for setting more accurate treatment strategies and further optimized procedures. The proper alignment using elastic bands has improved the absolute dose accuracy and the collimator disk alignment by more than 50%. PACS number: 87.55.kh

A national survey of HDR source knowledge among practicing radiation oncologists and residents: Establishing a willingness-to-pay threshold for cobalt-60 usage.

PubMed

Mailhot Vega, Raymond; Talcott, Wesley; Ishaq, Omar; Cohen, Patrice; Small, Christina J; Duckworth, Tamara; Sarria Bardales, Gustavo; Perez, Carmen A; Schiff, Peter B; Small, William; Harkenrider, Matthew M

Ir-192 is the predominant source for high-dose-rate (HDR) brachytherapy in United States markets. Co-60, with longer half-life and fewer source exchanges, has piloted abroad with comparable clinical dosimetry but increased shielding requirements. We sought to identify practitioner knowledge of Co-60 and establish acceptable willingness-to-pay (WTP) thresholds for additional shielding requirements for use in future cost-benefit analysis. A nationwide survey of U.S. radiation oncologists was conducted from June to July 2015, assessing knowledge of HDR sources, brachytherapy unit shielding, and factors that may influence source-selection decision-making. Self-identified decision makers in radiotherapy equipment purchase and acquisition were asked their WTP on shielding should a more cost-effective source become available. Four hundred forty surveys were completed and included. Forty-four percent were ABS members. Twenty percent of respondents identified Co-60 as an HDR source. Respondents who identified Co-60 were significantly more likely to be ABS members, have attended a national brachytherapy conference, and be involved in brachytherapy selection. Sixty-six percent of self-identified decision makers stated that their facility would switch to a more cost-effective source than Ir-192, if available. Cost and experience were the most common reasons provided for not switching. The most common WTP value selected by respondents was <$25,000. A majority of respondents were unaware of Co-60 as a commercially available HDR source. This investigation was novel in directly assessing decision makers to establish WTP for shielding costs that source change to Co-60 may require. These results will be used to establish WTP threshold for future cost-benefit analysis. Copyright © 2017 American Brachytherapy Society. Published by Elsevier Inc. All rights reserved.

Double-layer neutron shield design as neutron shielding application

NASA Astrophysics Data System (ADS)

Sariyer, Demet; Küçer, Rahmi

2018-02-01

The shield design in particle accelerators and other high energy facilities are mainly connected to the high-energy neutrons. The deep penetration of neutrons through massive shield has become a very serious problem. For shielding to be efficient, most of these neutrons should be confined to the shielding volume. If the interior space will become limited, the sufficient thickness of multilayer shield must be used. Concrete and iron are widely used as a multilayer shield material. Two layers shield material was selected to guarantee radiation safety outside of the shield against neutrons generated in the interaction of the different proton energies. One of them was one meter of concrete, the other was iron-contained material (FeB, Fe2B and stainless-steel) to be determined shield thicknesses. FLUKA Monte Carlo code was used for shield design geometry and required neutron dose distributions. The resulting two layered shields are shown better performance than single used concrete, thus the shield design could leave more space in the interior shielded areas.

40 CFR Table 5 to Subpart Jjjjjj... - Fuel Analysis Requirements

Code of Federal Regulations, 2012 CFR

2012-07-01

... 40 Protection of Environment 15 2012-07-01 2012-07-01 false Fuel Analysis Requirements 5 Table 5... Part 63—Fuel Analysis Requirements As stated in § 63.11213, you must comply with the following requirements for fuel analysis testing for affected sources: To conduct a fuel analysis for the following...

SIR-B analysis of the Precambrian shield of Sudan and Egypt: Penetration studies and subsurface mapping

NASA Technical Reports Server (NTRS)

Dixon, T. H.; Roth, L.; Stern, R. J.; Almond, D. C.; Kroner, A.; Elshazly, E. M.

1984-01-01

A shuttle imaging radar-B (SIR-B) study is proposed for the Precambrian shield in southeast Egypt and northeast Sudan in an area east of the Nile. The phenomenon of radar penetration of thin, dry eolian/alluvial cover is to be confirmed and quantified. The penetration phenomenon is to be used to map structural and lithologic features. Field work to be done in conjunction with image acquisition is discussed.

Progress in Open Rotor Research: A U.S. Perspective

NASA Technical Reports Server (NTRS)

Van Zante, Dale E.

2015-01-01

In response to the 1970s oil crisis, NASA created the Advanced Turboprop Project (ATP) to mature technologies for high-speed propellers to enable large reductions in fuel burn relative to turbofan engines of that era. Both single rotation and contra- rotation concepts were designed and tested in ground based facilities as well as flight. Some novel concepts/configurations were proposed as part of the effort. The high-speed propeller concepts did provide fuel burn savings, albeit with some acoustics and structural challenges to overcome. When fuel prices fell, the business case for radical new engine configurations collapsed and the research emphasis returned to high bypass ducted configurations. With rising oil prices and increased environmental concerns there is renewed interest in high-speed propeller based engine architectures. Contemporary analysis tools for aerodynamics and aeroacoustics have enabled a new era of blade designs that have both high efficiency and lower noise characteristics. A recent series of tests in the U.S. have characterized the aerodynamic performance and noise from these modern contra-rotating propeller designs. Additionally the installation and noise shielding aspects for conventional airframes and blended wing bodies have been studied. Historical estimates of 'propfan' performance have relied on legacy propeller performance and acoustics data. Current system studies make use of the modern propeller data and higher fidelity installation effects data to estimate the performance of a contemporary aircraft system. Contemporary designs have demonstrated high net efficiency, approximately 86%, at 0.78 Mach, and low noise, greater than 15 EPNdB cumulative margin to Chapter 4 when analyzed on a NASA derived aircraft/mission. This paper presents the current state of high-speed propeller/open rotor research within the U.S. from an overall viewpoint of the various efforts ongoing. The remaining technical challenges to a production engine include propulsion airframe integration, acoustic sensitivity to aircraft weight and certification issues.

Design Features of the Neutral Particle Diagnostic System for the ITER Tokamak

NASA Astrophysics Data System (ADS)

Petrov, S. Ya.; Afanasyev, V. I.; Melnik, A. D.; Mironov, M. I.; Navolotsky, A. S.; Nesenevich, V. G.; Petrov, M. P.; Chernyshev, F. V.; Kedrov, I. V.; Kuzmin, E. G.; Lyublin, B. V.; Kozlovski, S. S.; Mokeev, A. N.

2017-12-01

The control of the deuterium-tritium (DT) fuel isotopic ratio has to ensure the best performance of the ITER thermonuclear fusion reactor. The diagnostic system described in this paper allows the measurement of this ratio analyzing the hydrogen isotope fluxes (performing neutral particle analysis (NPA)). The development and supply of the NPA diagnostics for ITER was delegated to the Russian Federation. The diagnostics is being developed at the Ioffe Institute. The system consists of two analyzers, viz., LENPA (Low Energy Neutral Particle Analyzer) with 10-200 keV energy range and HENPA (High Energy Neutral Particle Analyzer) with 0.1-4.0MeV energy range. Simultaneous operation of both analyzers in different energy ranges enables researchers to measure the DT fuel ratio both in the central burning plasma (thermonuclear burn zone) and at the edge as well. When developing the diagnostic complex, it was necessary to account for the impact of several factors: high levels of neutron and gamma radiation, the direct vacuum connection to the ITER vessel, implying high tritium containment, strict requirements on reliability of all units and mechanisms, and the limited space available for accommodation of the diagnostic hardware at the ITER tokamak. The paper describes the design of the diagnostic complex and the engineering solutions that make it possible to conduct measurements under tokamak reactor conditions. The proposed engineering solutions provide a safe—with respect to thermal and mechanical loads—common vacuum channel for hydrogen isotope atoms to pass to the analyzers; ensure efficient shielding of the analyzers from the ITER stray magnetic field (up to 1 kG); provide the remote control of the NPA diagnostic complex, in particular, connection/disconnection of the NPA vacuum beamline from the ITER vessel; meet the ITER radiation safety requirements; and ensure measurements of the fuel isotopic ratio under high levels of neutron and gamma radiation.

Naturally induced secondary radiation in interplanetary space: Preliminary analyses for gamma radiation and radioisotope production from thermal neutron activation

NASA Technical Reports Server (NTRS)

Plaza-Rosado, Heriberto

1991-01-01

Thermal neutron activation analyses were carried out for various space systems components to determine gamma radiation dose rates and food radiation contamination levels. The space systems components selected were those for which previous radiation studies existed. These include manned space vehicle radiation shielding, liquid hydrogen propellant tanks for a Mars mission, and a food supply used as space vehicle radiation shielding. The computational method used is based on the fast neutron distribution generated by the BRYNTRN and HZETRN transport codes for Galactic Cosmic Rays (GCR) at solar minimum conditions and intense solar flares in space systems components. The gamma dose rates for soft tissue are calculated for water and aluminum space vehicle slab shields considering volumetric source self-attenuation and exponential buildup factors. In the case of the lunar habitat with regolith shielding, a completely exposed spherical habitat was assumed for mathematical convenience and conservative calculations. Activation analysis of the food supply used as radiation shielding is presented for four selected nutrients: potassium, calcium, sodium, and phosphorus. Radioactive isotopes that could represent a health hazard if ingested are identified and their concentrations are identified. For nutrients soluble in water, it was found that all induced radioactivity was below the accepted maximum permissible concentrations.

Calcium-doped ceria/titanate tabular functional nanocomposite by layer-by-layer coating method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, Xiang W., E-mail: lxwluck@gmail.co; Devaraju, M.K.; Yin, Shu

2010-07-15

Ca-doped ceria (CDC)/tabular titanate (K{sub 0.8}Li{sub 0.27}Ti{sub 1.73}O{sub 4}, TT) UV-shielding functional nanocomposite with fairly uniform CDC coating layers was prepared through a polyelectrolyte-associated layer-by-layer (LbL) coating method. TT with lepidocrocite-like layered structure was used as the substrate, poly (diallyldimethylammonium chloride) (PDDA) was used as a coupling agent, CDC nanoparticles were used as the main UV-shielding component. CDC/TT nanocomposites with various coating layers of CDC were obtained through a multistep coating process. The phases were studied by X-ray diffraction. The morphology and coating quality were studied by scanning electron microscopy and element mapping of energy dispersive X-ray analysis. The oxidationmore » catalytic activity, UV-shielding ability and using comfort were characterized by Rancimat test, UV-vis spectra and dynamic friction test, respectively. CDC/TT nanocomposites with low oxidation catalytic activity, high UV-shielding ability and good using comfort were finally obtained. - Graphical abstract: Through the control of surface charge of particles calcium-doped ceria/titanate composites with low oxidation catalytic activity, higher UV-shielding ability and excellent comfort was obtained by a facile layer-by-layer coating method.« less