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Sample records for measured fuel centerline

  1. Cracked pellet gap conductance model: comparison of FRAP-S calculations with measured fuel centerline temperatures

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Broughton, J.M.

    1975-03-01

    Fuel pellets crack extensively upon irradiation due both to thermal stresses induced by power changes and at high burnup, to accumulation of gaseous fission products at grain boundaries. Therefore, the distance between the fuel and cladding will be circumferentially nonuniform; varying between that calculated for intact operating fuel pellets and essentially zero (fuel segments in contact with the cladding wall). A model for calculation of temperatures in cracked pellets is proposed wherein the effective fuel to cladding gap conductance is calculated by taking a zero pressure contact conductance in series with an annular gap conductance. Comparisons of predicted and measured fuel centerline temperatures at beginning of life and at extended burnup are presented in support of the model. 13 references

  2. Fabrication and use of zircaloy/tantalum-sheathed cladding thermocouples and molybdenum/rhenium-sheathed fuel centerline thermocouples

    International Nuclear Information System (INIS)

    Wilkins, S.C.; Sepold, L.K.

    1985-01-01

    The thermocouples described in this report are zircaloy/tantalum-sheathed and molybdenum/rhenium alloy-sheathed instruments intended for fuel rod cladding and fuel centerline temperature measurements, respectively. Both types incorporate beryllium oxide insulation and tungsten/rhenium alloy thermoelements. These thermocouples, operated at temperatures of 2000 0 C and above, were developed for use in the internationally sponsored Severe Fuel Damage test series in the Power Burst Facility. The fabrication steps for both thermocouple types are described in detail. A laser-welding attachment technique for the cladding-type thermocouple is presented, and experience with alternate materials for cladding and fuel therocouples is discussed

  3. Effect of power variations across a fuel bundle and within a fuel element on fuel centerline temperature in PHWR bundles in uncrept and crept pressure tubes

    International Nuclear Information System (INIS)

    Onder, E.N.; Roubtsov, D.; Rao, Y.F.; Wilhelm, B.

    2017-01-01

    Highlights: • Pressure tube creep effect on fuel pin power and temperatures was investigated. • Noticeable effects were observed for 5.1% crept pressure tube. • Bundle eccentricity effect on power variations was insignificant for uncrept channels. • Difference of 112 °C was observed between top & bottom elements in 5.1% crept channel. • Not discernible fission gas release was expected with temperature difference of 112 °C. - Abstract: The neutron flux and fission power profiles through a fuel bundle and across a fuel element are important aspects of nuclear fuel analysis in multi-scale/multi-physics modelling of Pressurized Heavy Water Reactors (PHWRs) with advanced fuel bundles. Fuel channels in many existing PHWRs are horizontal. With ageing, pressure tubes creep and fuel bundles in these pressure tubes are eccentrically located, which results in an asymmetric coolant flow distribution between the top and bottom of the fuel bundles. The diametral change of the pressure tube due to creep is not constant along the fuel channel; it reaches a maximum in the vicinity of the maximum neutron flux location. The cross-sectional asymmetric positioning of fuel bundles in a crept pressure tube contributes to an asymmetric power distribution within a ring of fuel elements. Modern reactor physics lattice codes (such as WIMS-AECL) are capable of predicting the details of power distribution from basic principles. Thermalhydraulics subchannel codes (such as ASSERT-PV) use models to describe inhomogeneous power distribution within and across fuel elements (e.g., flux tilt model, different powers in different ring elements, or radial power profiles). In this work, physics and thermalhydraulics codes are applied to quantify the effect of eccentricity of a fuel bundle on power variations across it and within a fuel element, and ultimately on the fuel temperature distribution and fuel centerline temperature, which is one of the indicators of fuel performance under normal

  4. W-1 Sodium Loop Safety Facility experiment centerline fuel thermocouple performance

    International Nuclear Information System (INIS)

    Meyers, S.C.; Henderson, J.M.

    1980-05-01

    The W-1 Sodium Loop Safety Facility (SLSF) experiment is the fifth in a series of experiments sponsored by the Department of Energy (DOE) as part of the National Fast Breeder Reactor (FBR) Safety Assurance Program. The experiments are being conducted under the direction of Argonne National Laboratory (ANL) and Hanford Engineering Development Laboratory (HEDL). The irradiation phase of the W-1 SLSF experiment was conducted between May 27 and July 20, 1979, and terminated with incipient fuel pin cladding failure during the final boiling transient. Experimental hardware and facility performed as designed, allowing completion of all planned tests and test objectives. This paper focuses on high temperature in-fuel thermocouples and discusses their development, fabrication, and performance in the W-1 experiment

  5. Allegheny County Street Centerlines

    Data.gov (United States)

    Allegheny County / City of Pittsburgh / Western PA Regional Data Center — This dataset contains the locations of the street centerlines for vehicular and foot traffic in Allegheny County. Street Centerlines are classified as Primary Road,...

  6. VT Road Centerline

    Data.gov (United States)

    Vermont Center for Geographic Information — (Link to Metadata)(User Guide)(Symbology layer files: aotclass_only.lyr aotclass_surfacetyp.lyr)The Vermont Road Centerline data layer (TransRoad_RDS) contains all...

  7. Kansas Road Centerline Fle (KRCF)

    Data.gov (United States)

    Kansas Data Access and Support Center — This version of the Kansas Road Centerline File (0801) represents the first effort to create a statewide roads layer from best available data sources. KGS integrated...

  8. Optical fuel spray measurements

    Energy Technology Data Exchange (ETDEWEB)

    Hillamo, H.

    2011-07-01

    Diesel fuel sprays, including fuel/air mixing and the physics of two-phase jet formation, are discussed in the thesis. The fuel/air mixing strongly affects emissions formation in spray combustion processes where the local combustion conditions dictate the emission formation. This study comprises optical measurements both in pressurized spray test rigs and in a running engine.The studied fuel injection was arranged with a common rail injection system and the injectors were operated with a solenoid-based injection valve. Both marine and heavy-duty diesel engine injectors were used in the study. Optical fuel spray measurements were carried out with a laser-based double-framing camera system. This kind of equipments is usually used for flow field measurements with Particle Image Velocimetry technique (PIV) as well as for backlight imaging. Fundamental fuel spray properties and spray formation were studied in spray test rigs. These measurements involved studies of mixing, atomization, and the flow field. Test rig measurements were used to study the effect of individual injection parameters and component designs. Measurements of the fuel spray flow field, spray penetration, spray tip velocity, spray angle, spray structure, droplet accumulation, and droplet size estimates are shown. Measurement campaign in a running optically accessible large-bore medium-speed engine was also carried out. The results from engine tests were compared with equivalent test rig measurements, as well as computational results, to evaluate the level of understanding of sprays. It was shown that transient spray has an acceleration and a deceleration phase. Successive flow field measurements (PIV) in optically dense diesel spray resulted in local and average velocity data of diesel sprays. Processing fuel spray generates a flow field to surrounding gas and entrainment of surrounding gas into fuel jet was also seen at the sides of the spray. Laser sheet imaging revealed the inner structure of diesel

  9. Thermal Conductivity Measurement and Analysis of Fully Ceramic Microencapsulated fuel

    International Nuclear Information System (INIS)

    Lee, H. G.; Kim, D. J.; Park, J. Y.; Kim, W. J.; Lee, S. J.

    2015-01-01

    FCM nuclear fuel is composed of tristructural isotropic(TRISO) fuel particle and SiC ceramic matrix. SiC ceramic matrix play an essential part in protecting fission product. In the FCM fuel concept, fission product is doubly protected by TRISO coating layer and SiC ceramic matrix in comparison with the current commercial UO2 fuel system of LWR. In addition to a safety enhancement of FCM fuel, thermal conductivity of SiC ceramic matrix is better than that of UO2 fuel. Because the centerline temperature of FCM fuel is lower than that of the current UO2 fuel due to the difference of thermal conductivity of fuel, an operational release of fission products from the fuel can be reduced. SiC ceramic has attracted for nuclear fuel application due to its high thermal conductivity properties with good radiation tolerant properties, a low neutron absorption cross-section and a high corrosion resistance. Thermal conductivity of ceramic matrix composite depends on the thermal conductivity of each component and the morphology of reinforcement materials such as fibers and particles. There are many results about thermal conductivity of fiber-reinforced composite like as SiCf/SiC composite. Thermal conductivity of SiC ceramics and FCM pellets with the volume fraction of TRISO particles were measured and analyzed by analytical models. Polycrystalline SiC ceramics and FCM pellets with TRISO particles were fabricated by hot press sintering with sintering additives. Thermal conductivity of the FCM pellets with TRISO particles of 0 vol.%, 10 vol.%, 20 vol.%, 30 vol.% and 40 vol.% show 68.4, 52.3, 46.8, 43.0 and 34.5 W/mK, respectively. As the volume fraction of TRISO particles increased, the measured thermal conductivity values closely followed the prediction of Maxwell's equation

  10. Evaluation of the Centerline Temperature for the Irradiated DUPIC Pellet

    International Nuclear Information System (INIS)

    Park, Chang Je; Lee, Cheol Yong; Kang, Kweon Ho; Song, Kee Chan

    2007-01-01

    The DUPIC (Direct Use of spent PWR fuels In a CANDU reactor) fuel has a proliferation-resistant property and provides an efficient utilization of a spent fuel through a direct fabrication with the OREOX process in which most of the fission products remain and some volatile elements such as Xe, Kr, Cs, and I are reduced significantly. It is expected that the performance of the DUPIC fuel exhibits different behavior when compared with the fresh uranium oxide fuel. To evaluate the performance of the DUPIC fuel, total five irradiation tests have been performed in the HANARO reactor since May 2000. Recently, the fifth irradiation test of the DUPIC fuel was successfully completed for a total of three cycles from March 2006 to July 2006. The important characteristics of the first irradiation test are a high power test and a validation of a remote assembly of an irradiation rig. The second irradiation test was instrumented with a SPND (self-powered neutron detector) first for a typical CANDU burnup test. The third test was an extensive irradiation test of the second test and the total burnup was estimated as 6,700 MWd/tU. The forth test was a remote instrumented test of the pellet centerline temperature and the inlet and outlet coolant temperatures. The first remote instrumentation test was achieved with our own technology. The fifth test was a remote-instrumented test of the pellet centerline temperature by extending the technology of the forth irradiation test. In this paper, a DUPIC fuel performance code (KAOS, KAERI Advanced Oxide fuel performance code System) was used to compare the main simulation results of the irradiation tests in the HANARO

  11. Road and Street Centerlines - COUNTY_STREET_CENTERLINES_IDHS_IN: Street Centerlines Maintained by County Agencies in Indiana (Indiana Department of Homeland Security, Line Shapefile)

    Data.gov (United States)

    NSGIC State | GIS Inventory — COUNTY_STREET_CENTERLINES_IDHS_IN is a line feature class that contains street centerlines maintained by county agencies in Indiana, provided by personnel of Indiana...

  12. Fabrication of internally instrumented reactor fuel rods

    International Nuclear Information System (INIS)

    Schmutz, J.D.; Meservey, R.H.

    1975-01-01

    Procedures are outlined for fabricating internally instrumented reactor fuel rods while maintaining the original quality assurance level of the rods. Instrumented fuel rods described contain fuel centerline thermocouples, ultrasonic thermometers, and pressure tubes for internal rod gas pressure measurements. Descriptions of the thermocouples and ultrasonic thermometers are also contained

  13. Automatic Road Centerline Extraction from Imagery Using Road GPS Data

    Directory of Open Access Journals (Sweden)

    Chuqing Cao

    2014-09-01

    Full Text Available Road centerline extraction from imagery constitutes a key element in numerous geospatial applications, which has been addressed through a variety of approaches. However, most of the existing methods are not capable of dealing with challenges such as different road shapes, complex scenes, and variable resolutions. This paper presents a novel method for road centerline extraction from imagery in a fully automatic approach that addresses the aforementioned challenges by exploiting road GPS data. The proposed method combines road color feature with road GPS data to detect road centerline seed points. After global alignment of road GPS data, a novel road centerline extraction algorithm is developed to extract each individual road centerline in local regions. Through road connection, road centerline network is generated as the final output. Extensive experiments demonstrate that our proposed method can rapidly and accurately extract road centerline from remotely sensed imagery.

  14. Design and manufacturing of instrumented capsule(03F-05K) for nuclear fuel irradiation in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Goo; Sohn, J. M.; Shin, Y. T. [and others

    2004-06-01

    An instrumented capsule is being developed to be able to measure fuel characteristics, such as fuel temperature, internal pressure of fuel rod, fuel elongation, and neutron flux, etc., during the irradiation test of nuclear fuel in HANARO. The instrumented capsule(02F-11K) for measuring and monitoring fuel centerline temperature and neutron flux was designed and manufactured. The instrumented capsule includes three test fuel rods installed thermocouple to measure fuel centerline temperature and three SPNDs (self-powered neutron detector) to monitor the neutron flux. Its stability was verified by out-of-pile performance test, and its safety evaluation was also shown that the safety requirements were satisfied. And then, to verify the design of the instrumented capsule in the test hole, it was successfully irradiated in the test hole of HANARO from March 14, 2003 to June 1, 2003 (53.8 full power days at 24 MWth). During irradiation, the centerline temperature of PWR UO{sub 2} fuel pellets fabricated by KEPCO Nuclear Fuel Company and the neutron flux were continuously measured and monitored. In the year of 2004, 3 test fuel rods and the instrumented capsule(03F-05K) were designed and fabricated to measure fuel centerline temperature, internal pressure of fuel rod, and fuel axial deformation during irradiation test. This capsule is being irradiated in the test hole OR5 of HANARO reactor from April 26, 2004.

  15. Method of measuring distance between fuel element

    International Nuclear Information System (INIS)

    Urata, Megumu.

    1991-01-01

    The distance between fuel elements contained in a pool is measured in a contactless manner even for a narrow distance less than 1 mm. That is, the equipment for measuring the distance between spent fuel elements of a spent fuel assembly in a nuclear reactor comprises a optical fiber scope, a lens, an industrial TV camera and a monitor TV. The top end of the optical fiber scope is inserted between fuel elements to be measured. The state thereof is displayed on the TV screen to measure the distance between the fuel elements. The measured results are compared with a previously formed calibration curve to determine the value between the fuel elements. Then, the distance between the fuel elements can be determined in the pool of a power plant without dismantling the fuel assembly, to investigate the state of the bending and estimate the fuel working life. (I.S.)

  16. Design verification test of instrumented capsule (02F-11K) for nuclear fuel irradiation in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Goo; Sohn, J. M.; Oh, J. M. [and others

    2004-01-01

    An instrumented capsule is being developed to be able to measure fuel characteristics, such as fuel temperature, internal pressure of fuel rod, fuel elongation, and neutron flux, etc., during the irradiation test of nuclear fuel in HANARO. The instrumented capsule for measuring and monitoring fuel centerline temperature and neutron flux was designed and manufactured. The instrumented capsule includes three test fuel rods installed thermocouple to measure fuel centerline temperature and three SPNDs (Self-Powered Neutron Detector) to monitor the neutron flux. Its stability was verified by out-of-pile performance test, and its safety evaluation was also shown that the safety requirements were satisfied. And then, to verify the design of the instrumented capsule in the test hole, it was successfully irradiated in the test hole of HANARO from March 14, 2003 to June 1, 2003 (53.8 full power days at 24 MWth). During irradiation, the centerline temperature of PWR UO{sub 2} fuel pellets fabricated by KEPCO Nuclear Fuel Company and the neutron flux were continuously measured and monitored. The test fuel rods were irradiated at less than 350 W/cm to 5.13 GWD/MTU with fuel centerline peak temperature below 1,375 .deg. C. The structural stability of the capsule was satisfied by the naked eye in service pool of HANARO. The capsule and test fuel rods were dismantled and test fuel rods were examined at the hot cell of IMEF (Irradiated Material Examination Facility)

  17. Spent fuel critical masses and supportive measurements

    International Nuclear Information System (INIS)

    Toffer, H.; Wells, A.H.

    1987-01-01

    Critical masses for spent fuel are larger than for green fuel and therefore use of the increased masses could result in improved handling, storage, and transport of such materials. To apply spent fuel critical masses requires an assessment of fuel exposure and the corresponding isotopic compositions. The paper discusses several approaches at the Hanford N Reactor in establishing fuel exposure, including a direct measurement of spent to green fuel critical masses. The benefits derived from the use of spent fuel critical masses are illustrated for cask designs at the Nuclear Assurance Corporation. (author)

  18. Wide-field absolute transverse blood flow velocity mapping in vessel centerline

    Science.gov (United States)

    Wu, Nanshou; Wang, Lei; Zhu, Bifeng; Guan, Caizhong; Wang, Mingyi; Han, Dingan; Tan, Haishu; Zeng, Yaguang

    2018-02-01

    We propose a wide-field absolute transverse blood flow velocity measurement method in vessel centerline based on absorption intensity fluctuation modulation effect. The difference between the light absorption capacities of red blood cells and background tissue under low-coherence illumination is utilized to realize the instantaneous and average wide-field optical angiography images. The absolute fuzzy connection algorithm is used for vessel centerline extraction from the average wide-field optical angiography. The absolute transverse velocity in the vessel centerline is then measured by a cross-correlation analysis according to instantaneous modulation depth signal. The proposed method promises to contribute to the treatment of diseases, such as those related to anemia or thrombosis.

  19. LOFT fuel rod pressure measurement

    International Nuclear Information System (INIS)

    Billeter, T.R.

    1979-01-01

    Pressure sensors selected for measuring fuel rod pressure within the LOFT reactor exhibited stable, repeatable operating characteristics during calibrations at temperatures up to 800 0 F and pressures to 2500 psig. All sensors have a nominal sensitivity of .5 millivolts per psi, decreasing monotonically with temperature. Output signal increases linearly with increasing pressure up to 2000 psig. For imposed slow and rapid temperature variations and for pressure applied during these tests, the sensor indicates a pressure at variance with the actual value by up to 15% of reading. However, the imposed temperature rates of change often exceeded the value of -10 0 F/sec. specified for LOFT. The series of tests in an autoclave permit creation of an environment most closely resembling sensor operating conditions within LOFT. For multiple blowdowns and for longtime durations the sensor continued to provide pressure-related output signals. For temperature rates up to -87 0 F/sec, the indicated pressure measurement error remained less than 13% of reading. Adverse effects caused by heating the 1/16 inch O.D. signal cable to 800 0 F contributed only insignificantly to the noted pressure measurement error

  20. Measuring vibrations in fuel channels CNE

    International Nuclear Information System (INIS)

    Martín Ghiselli, A.; Fiori, J.; Sacchi, M.; Villabrille, G.

    2013-01-01

    This paper present a description of implementation and execution of vibration measurements made at the request of NUCLEOELECTRICA ARGENTINA S.A. on the ends of the reactor fuel channels of Embalse Nuclear Power Plant to explore possible differences between the dynamic behavior of empty fuel channel and with full charge of fuel elements inside. (author)

  1. A Robust Level-Set Algorithm for Centerline Extraction

    NARCIS (Netherlands)

    Telea, Alexandru; Vilanova, Anna

    2003-01-01

    We present a robust method for extracting 3D centerlines from volumetric datasets. We start from a 2D skeletonization method to locate voxels centered with respect to three orthogonal slicing directions. Next, we introduce a new detection criterion to extract the centerline voxels from the above

  2. Objective and subjective measures of fuel poverty

    International Nuclear Information System (INIS)

    Waddams Price, Catherine; Brazier, Karl; Wang, Wenjia

    2012-01-01

    As energy prices continue to rise to reflect the real cost of carbon, the numbers of households in fuel poverty is increasing rapidly. This paper uses a unique data set to explore one alternative measure of fuel poverty based on whether consumers feel able to afford their energy, similar to one introduced by the government a few years ago. We explore the links between an expenditure based and our subjective measure of fuel poverty and compare these differences with those between official measures. Amongst low income households, 28% spent more than a tenth of their income on energy in the home, and so would be likely to qualify as fuel poor under the most usual definition; but only 16% felt unable to afford sufficient energy to keep their homes warm. Amongst this group who felt they had difficulty in affording sufficient energy, less than half were ‘expenditure’ fuel poor. The paper argues that reintroduction of a self-reported measure by the government would be a valuable aid to policy development. - Highlights: ► Defines a subjective measure of fuel poverty (unable to afford heating). ► Fewer households feel fuel poor than meet the official definition of fuel poverty. ► Though they are positively related, different factors affect the two measures. ► Those on standard and prepayment metres more likely to feel fuel poor. ► Large families spend a higher proportion of income on energy but are no more likely to feel fuel poor.

  3. Inland Waters - Mississippi River Centerline - Headwaters to Gulf Coast

    Data.gov (United States)

    Army Corps of Engineers, Department of the Army, Department of Defense — Mississippi River centerline data derived from USACE navigation sailing line (recommended track) data and on-screen digitized in areas of no data. Data set extends...

  4. Pyrometric fuel particle measurements in pressurised reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hernberg, R.; Joutsenoja, T. [Tampere Univ. of Technology (Finland)

    1996-12-01

    A fiberoptic two-colour pyrometric technique for fuel particle temperature and size measurement is modified and applied to three pressurized reactors of different type in Finland, Germany and France. A modification of the pyrometric method for simultaneous in situ measurement of the temperature and size of individual pulverized coal particles at the pressurized entrained flow reactor in Jyvaeskylae was developed and several series of measurements were made. In Orleans a fiberoptic pyrometric device was installed to a pressurised thermogravimetric reactor and the two-colour temperatures of fuel samples were measured. Some results of these measurements are presented. The project belongs to EU`s Joule 2 extension research programme. (author)

  5. Pyrometric fuel particle measurements in pressurised reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hernberg, R; Joutsenoja, T [Tampere Univ. of Technology (Finland)

    1997-10-01

    A fibre-optic two-colour pyrometric technique for fuel particle temperature and size measurement is modified and applied to three pressurised reactors of different type in Finland, Germany and France. A modification of the pyrometric method for simultaneous in situ measurement of the temperature and size of individual pulverised coal particles at the pressurised entrained flow reactor of VTT Energy in Jyvaeskylae was developed and several series of measurements were made in order to study the effects of oxygen concentration (3-30 vol%) and pressure (0.2-1.0 MPa) on the particle temperature. The fuels used in the experiments were Westerholt, Polish and Goettelborn hvb coals, Gardanne lignite and Niederberg anthracite. The initial nominal fuel particle size varied in the experiments from 70 to 250 ,{mu}m and the gas temperature was typically 1173 K. For the anthracite also the effects of gas temperature (1073-1423K) and CO{sub 2} concentration (6-80 vol%) were studied. In Orleans a fibreoptic pyrometric device was installed to a pressurised thermogravimetric reactor of CNRS and the two-colour temperatures of fuel samples were measured. The fuel in the experiments was pulverised Goettelborn char. The reliability of optical temperature measurement in this particular application was analysed. In Essen a fibre-optic pyrometric technique that is capable to measure bed and fuel particle temperatures was applied to an atmospheric fluidised bed reactor of DMT. The effects of oxygen concentration (3-8 vol%) and bed temperature (1123-1193 K) on the fuel particle temperature were studied. The fuels in these were Westerholt coal and char and EBV-coal. Some results of these measurements are presented. The project belonged to EU`s Joule 2 extension research programme (contract JOU2-CT93-0331). (orig.)

  6. Calibration of spent fuel measurement assembly

    International Nuclear Information System (INIS)

    Koleska, Michal; Viererbl, Ladislav; Marek, Milan

    2014-01-01

    The LVR-15 research reactor (Czech Republic) had been converted from the highly enriched IRT-2M to the low enriched IRT-4M fuel. For the possibility of the independent pre-transport evaluation of IRT-2M burnup, a spectrometric system was developed. This spectrometric system consists of the fuel holder, the collimator and the portable Canberra Big MAC HPGe (High Purity Germanium) detector. In order to have well reproducible and reliable experimental data for modeling of the measurement system, calibration with the 110m Ag isotope with known activity was performed. This isotope was chosen for having energies similar to isotopes measured in fuel assemblies. The 110m Ag isotope was prepared by irradiating of the silver foil in LVR-15 research reactor; its activity was evaluated in the LVR-15's spectrometric laboratory. From the measured data, an efficiency curve of the spectrometric system has been determined. The experimental data were compared to the calculation results with the MCNPX model of the spectrometric system. - Highlights: • Calibration of research reactor spent fuel measurement assembly. • On-site prepared 110m Ag isotope used for the measurement. • Calculated self-shielding factor for the IRT-2M fuel. • Applicable to other research reactor fuel geometries

  7. Enrichment measurement in TRIGA type fuels

    International Nuclear Information System (INIS)

    Aguilar H, F.; Mazon R, R.

    2001-05-01

    The Department of Energy of the United States of North America, through the program 'Idaho Operations Nuclear Spent Fuel Program' of the Idaho National Engineering and Environmental Laboratory (INEEL), in Idaho Falls; Idaho USA, hires to Global Technologies Inc. (GTI) to develop a prototype device of detection enrichment uranium (DEU Detection of Enrichment of Uranium) to determine quantitatively the enrichment in remainder U-235 in a TRIGA fuel element at the end of it useful life. The characteristics of the prototype developed by GTI are the following ones: It allows to carry out no-destructive measurements of TRIGA type fuel. Easily transportable due to that reduced of it size. The determination of the enrichment (in grams of U-235) it is obtained with a precision of 5%. The National Institute of Nuclear Research (ININ), in its facilities of the Nuclear Center of Mexico, it has TRIGA type fuel of high and low enrichment (standard and FLIP) fresh and with burnt, it also has the infrastructure (hot cells, armor-plating of transport, etc) and qualified personnel to carry out the necessary maneuvers to prove the operation of the DEU prototype. For this its would be used standard type fuel elements and FLIP, so much fresh as with certain burnt one. In the case of the fresh fuels the measurement doesn't represent any risk, the fuels before and after the measurement its don't contain a quantity of fission products that its represent a radiological risk in its manipulation; but in the case of the fuels with burnt the handling of the same ones represents an important radiological risk reason why for its manipulation it was used the transport armor-plating and the hot cells. (Author)

  8. Recalculation of measured fuel nuclide concentrations

    International Nuclear Information System (INIS)

    Moeller, W.

    1984-01-01

    The concentrations and concentration ratios of heavy fuel nuclides determined in the Central Institute for Nuclear Research Rossendorf on the basis of destructive burnup measurements are compared with the results of microburnup calculations. The possibility is discussed to improve the results by taking into account the spectral characteristics at the positions of the measuring samples. (author)

  9. Synergistic Smart Fuel For Microstructure Mediated Measurements

    Energy Technology Data Exchange (ETDEWEB)

    James A. Smith; Dale K. Kotter; Steven L. Garrett; Randall A. Ali

    2013-07-01

    Advancing the Nuclear Fuel Cycle and Next Generation Nuclear Power Plants requires enhancing our basic understanding of fuel and materials behavior under irradiation. The two most significant issues limiting the effectiveness and lifespan of the fuel are the loss of thermal conductivity of the fuel and the mechanical strength of both fuel and cladding. The core of a nuclear reactor presents an extremely harsh and challenging environment for both sensors and telemetry due to elevated temperatures and large fluxes of energetic and ionizing particles from radioactive decay processes. The majority of measurements are made in reactors using “radiation hardened” sensors and materials. A different approach has been pursued in this research that exploits high temperatures and materials that are robust with respect to ionizing radiation. This synergistically designed thermoacoustic sensor will be self-powered, wireless, and provide telemetry. The novel sensor will be able to provide reactor process information even if external electrical power and communication are unavailable. In addition, the form-factor for the sensor is identical to the existing fuel rods within reactors and contains no moving parts. Results from initial proof of concept experiments designed to characterize porosity, surface properties and monitor gas composition will be discussed.

  10. Coincidence measurements of FFTF breeder fuel subassemblies

    International Nuclear Information System (INIS)

    Eccleston, G.W.; Foley, J.E.; Krick, M.; Menlove, H.O.; Goris, P.; Ramalho, A.

    1984-04-01

    A prototype coincidence counter developed to assay fast breeder reactor fuel was used to measure four fast-flux test facility subassemblies at the Hanford Engineering Development Laboratory in Richland, Washington. Plutonium contents in the four subassemblies ranged between 7.4 and 9.7 kg with corresponding 240 Pu-effective contents between 0.9 and 1.2 kg. Large count rates were observed from the measurements, and plots of the data showed significant multiplication in the fuel. The measured data were corrected for deadtime and multiplication effects using established formulas. These corrections require accurate knowledge of the plutonium isotopics and 241 Am content in the fuel. Multiplication-corrected coincidence count rates agreed with the expected count rates based on spontaneous fission-neutron emission rates. These measurements indicate that breeder fuel subassemblies with 240 Pu-effective contents up to 1.2 kg can be nondestructively assayed using the shift-register electronics with the prototype counters. Measurements using the standard Los Alamos National Laboratory shift-register coincidence electronics unit can produce an assay value accurate to +-1% in 1000 s. The uncertainty results from counting statistics and deadtime-correction errors. 3 references, 8 figures, 8 tables

  11. Burnup measurements of leader fuel elements

    International Nuclear Information System (INIS)

    Henriquez, C; Navarro, G; Pereda, C

    2000-01-01

    Some time ago the CCHEN authorities decided to produce a set of 50 low enrichment fuel elements. These elements were produced in the PEC (Fuel Elements Plant), located at CCHEN offices in Lo Aguirre. These new fuel elements have basically the same geometrical characteristics of previous ones, which were British and made with raw material from the U.S. The principal differences between our fuel elements and the British ones is the density of fissile material, U-235, which was increased to compensate the reduction in enrichment. Last year, the Fuel Elements Plant (PEC) delivered the shipment's first four (4) fuel elements, called leaders, to the RECH1. A test element was delivered too, and the complete set was introduced into the reactor's nucleus, following the normal routine, but performing a special follow-up on their behavior inside the nucleus. This experimental element has only one outside fuel plate, and the remaining (15) structural plates are aluminum. In order to study the burnup, the test element was taken out of the nucleus, in mid- November 1999, and left to decay until June 2000, when it was moved to the laboratory (High Activity Cell), to start the burnup measurements, with a gamma spectroscopy system. This work aims to show the results of these measurements and in addition to meet the following objectives: (a) Visual test of the plate's general condition; (b) Sipping test of fission products; (c) Study of burn-up distribution in the plate; (d) Check and improve the calculus algorithm; (e) Comparison of the results obtained from the spectroscopy with the ones from neutron calculus

  12. Local Measurement of Fuel Energy Deposition and Heat Transfer Environment During Fuel Lifetime Using Controlled Calorimetry

    International Nuclear Information System (INIS)

    Don W. Miller; Andrew Kauffmann; Eric Kreidler; Dongxu Li; Hanying Liu; Daniel Mills; Thomas D. Radcliff; Joseph Talnagi

    2001-01-01

    A comprehensive description of the accomplishments of the DOE grant titled, ''Local Measurement of Fuel Energy Deposition and Heat Transfer Environment During Fuel Lifetime using Controlled Calorimetry''

  13. An Augmented Fast Marching Method for Computing Skeletons and Centerlines

    NARCIS (Netherlands)

    Telea, Alexandru; Wijk, Jarke J. van

    2002-01-01

    We present a simple and robust method for computing skeletons for arbitrary planar objects and centerlines for 3D objects. We augment the Fast Marching Method (FMM) widely used in level set applications by computing the paramterized boundary location every pixel came from during the boundary

  14. CENTERLINE RUMBLE STRIPS SAFETY IMPACT EVALUATION—PHASE 2

    Science.gov (United States)

    2018-02-02

    Centerline rumble strips are used by various states as a low-cost countermeasure for mitigating cross-over crashes on two-way highways. This study performs a safety impact evaluation using an empirical Bayesian analysis. The researchers obtained a cr...

  15. Factors Influencing Pitot Probe Centerline Displacement in a Turbulent Supersonic Boundary Layer

    Science.gov (United States)

    Grosser, Wendy I.

    1997-01-01

    When a total pressure probe is used for measuring flows with transverse total pressure gradients, a displacement of the effective center of the probe is observed (designated Delta). While this phenomenon is well documented in incompressible flow and supersonic laminar flow, there is insufficient information concerning supersonic turbulent flow. In this study, three NASA Lewis Research Center Supersonic Wind Tunnels (SWT's) were used to investigate pitot probe centerline displacement in supersonic turbulent boundary layers. The relationship between test conditions and pitot probe centerline displacement error was to be determined. For this investigation, ten circular probes with diameter-to-boundary layer ratios (D/delta) ranging from 0.015 to 0.256 were tested in the 10 ft x 10 ft SWT, the 15 cm x 15 cm SWT, and the 1 ft x 1 ft SWT. Reynolds numbers of 4.27 x 10(exp 6)/m, 6.00 x 10(exp 6)/in, 10.33 x 10(exp 6)/in, and 16.9 x 10(exp 6)/m were tested at nominal Mach numbers of 2.0 and 2.5. Boundary layer thicknesses for the three tunnels were approximately 200 mm, 13 mm, and 30 mm, respectively. Initial results indicate that boundary layer thickness, delta, and probe diameter, D/delta play a minimal role in pitot probe centerline offset error, Delta/D. It appears that the Mach gradient, dM/dy, is an important factor, though the exact relationship has not yet been determined. More data is needed to fill the map before a conclusion can be drawn with any certainty. This research provides valuable supersonic, turbulent boundary layer data from three supersonic wind tunnels with three very different boundary layers. It will prove a valuable stepping stone for future research into the factors influencing pitot probe centerline offset error.

  16. Burnup degree measuring device for spent fuel

    International Nuclear Information System (INIS)

    Doi, Hideo; Imaizumi, Hideki; Endo, Yasumi; Itahara, Kuniyuki.

    1994-01-01

    The present invention provides a small-sized and convenient device for measuring a burnup degree of spent fuels, which can be installed without remodelling an existent fuel storage pool. Namely, a gamma-ray detecting portion incorporates a Cd-Te detector for measuring intensity ratio of gamma-rays. A neutron detecting portion incorporates a fission counter tube. The Cd-Te detector comprises a neutron shielding member for reducing radiation damages and a background controlling plate for reducing low energy gamma-rays entering from a collimator. Since the Cd-Td detector for use in a gamma-ray spectroscopy can be used at a normal temperature and can measure even a relatively strong radiation field, it can measure the intensity of gamma-rays from Cs-137 and Cs-134 in spent fuels accurately at a resolving power of less than 10 keV. Further, in a case where a cooling period is less than one year, gamma-rays from Rh-106 and Nb-95 can also be measured. (I.S.)

  17. Holdup measurement for nuclear fuel manufacturing plants

    International Nuclear Information System (INIS)

    Zucker, M.S.; Degen, M.; Cohen, I.; Gody, A.; Summers, R.; Bisset, P.; Shaub, E.; Holody, D.

    The assay of nuclear material holdup in fuel manufacturing plants is a laborious but often necessary part of completing the material balance. A range of instruments, standards, and a methodology for assaying holdup has been developed. The objectives of holdup measurement are ascertaining the amount, distribution, and how firmly fixed the SNM is. The purposes are reconciliation of material unbalance during or after a manufacturing campaign or plant decommissioning, to decide security requirements, or whether further recovery efforts are justified

  18. Dimensional measurement of fresh fuel bundle for CANDU reactor

    International Nuclear Information System (INIS)

    Jo, Chang Keun; Cho, Moon Sung; Suk, Ho Chun; Koo, Dae Seo; Jun, Ji Su; Jung, Jong Yeob

    2005-01-01

    This report describes the results of the dimensional measurement of fresh fuel bundles for the CANDU reactor in order to estimate the integrity of fuel bundle in two-phase flow in the CANDU-6 fuel channel. The dimensional measurements of fuel bundles are performed by using the 'CANDU Fuel In-Bay Inspection and Dimensional Measurement System', which was developed by this project. The dimensional measurements are done from February 2004 to March 2004 in the CANDU fuel storage of KNFC for the 36 fresh fuel bundles, which are produced by KNFC and are waiting for the delivery to the Wolsong-3 plant. The detail items of dimensional measurements are included fuel rod and bearing pad profiles of the outer ring in fuel bundle, diameter of fuel bundle, bowing of fuel bundle, fuel rod length, and surface profile of end plate profile. The measurement data will be compared with those of the post-irradiated bundles cooled in Wolsong-3 NPP spent fuel pool by using the same bundles and In-Bay Measurement System. So, this analysis of data will be applied for the evaluation of fuel bundle integrity in two-phase flow of the CANDU-6 fuel channel

  19. Shock Tube Measurements for Liquid Fuels Combustion

    National Research Council Canada - National Science Library

    Hanson, Ronald K

    2006-01-01

    ...) fundamental studies of fuel spray evaporation rates and ignition times of low-vapor pressure fuels such as JP-8, diesel fuel and normal alkane surrogates in a new aerosol shock tube using state...

  20. On the evaluation of a fuel assembly design by means of uncertainty and sensitivity measures

    International Nuclear Information System (INIS)

    Jaeger, Wadim; Sanchez Espinoza, Victor Hugo

    2012-01-01

    This paper will provide results of an uncertainty and sensitivity study in order to calculate parameters of safety related importance like the fuel centerline temperature, the cladding temperature and the fuel assembly pressure drop of a lead-alloy cooled fast system. Applying best practice guidelines, a list of uncertain parameters has been identified. The considered parameter variations are based on the experience gained during fabrication and operation of former and existing liquid metal cooled fast systems as well as on experimental results and on engineering judgment. (orig.)

  1. On the evaluation of a fuel assembly design by means of uncertainty and sensitivity measures

    Energy Technology Data Exchange (ETDEWEB)

    Jaeger, Wadim; Sanchez Espinoza, Victor Hugo [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Neutron Physics and Reactor Technology

    2012-11-15

    This paper will provide results of an uncertainty and sensitivity study in order to calculate parameters of safety related importance like the fuel centerline temperature, the cladding temperature and the fuel assembly pressure drop of a lead-alloy cooled fast system. Applying best practice guidelines, a list of uncertain parameters has been identified. The considered parameter variations are based on the experience gained during fabrication and operation of former and existing liquid metal cooled fast systems as well as on experimental results and on engineering judgment. (orig.)

  2. Instrument for measuring fuel cladding strain

    International Nuclear Information System (INIS)

    Billeter, T.R.

    1976-01-01

    Development work to provide instrumentation for the continuous measurement of strain of material specimens such as nuclear fuel cladding has shown that a microwave sensor and associated instrumentation hold promise. The cylindrical sensor body enclosing the specimen results in a coaxial resonator absorbing microwave energy at frequencies dependent upon the diameter of the specimen. Diametral changes of a microinch can be resolved with use of the instrumentation. Very reasonable values of elastic strain were measured at 75 0 F and 1000 0 F for an internally pressurized 20 percent C.W. 316 stainless steel specimen simulating nuclear fuel cladding. The instrument also indicated the creep strain of the same specimen pressurized at 6500 psi and at a temperature of 1000 0 F for a period of 700 hours. Although the indicated strain appears greater than actual, the sensor/specimen unit experienced considerable oxidation even though an inert gas purge persisted throughout the test duration. By monitoring at least two modes of resonance, the measured strain was shown to be nearly independent of sensor temperature. To prevent oxidation, a second test was performed in which the specimen/sensor units were contained in an evacuated enclosure. The strain of the two prepressurized specimens as indicated by the microwave instrumentation agreed very closely with pre- and post-test measurements obtained with use of a laser interferometer

  3. 40 CFR 610.42 - Fuel economy measurement.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Fuel economy measurement. 610.42... ECONOMY RETROFIT DEVICES Test Procedures and Evaluation Criteria General Vehicle Test Procedures § 610.42 Fuel economy measurement. (a) Fuel consumption will be measured by: (1) The carbon balance method, or...

  4. Design and manufacturing of 05F-01K instrumented capsule for nuclear fuel irradiation in Hanaro

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, J. M.; Shin, Y. T.; Park, S. J. (and others)

    2007-07-15

    An instrumented capsule was developed to be able to measure fuel characteristics, such as fuel temperature, internal pressure of fuel rod, fuel pellet elongation, and neutron flux, etc., during the irradiation test of nuclear fuel in Hanaro. The instrumented capsule(02F-11K) for measuring and monitoring and monitoring fuel centerline temperature and neutron flux was designed and manufactured. It was successfully irradiated in the test hole OR5 of Hanaro from March 14, 2003 to June 1, 2003 (53.84 full power days at 24 MW). In the year of 2004, 3 test fuel rods and the instrumented capsule(03F-05K) were designed and manufactured to measure fuel centerline temperature, internal pressure of fuel rod, and fuel axial deformation during irradiation test. This capsule was irradiated in the test hole OR5 of Hanaro reactor from April 26, 2004 to October 1, 2004 (59.5 EFPD at 24 {approx} 30 MW). The six typed dual instrumented fuel rods, which allow for two characteristics to be measured simultaneously in one fuel rod, have been designed and manufactured to enhance the efficiency of the irradiation test using the instrumented fuel capsule. The 05F-01K instrumented fuel capsule was designed and manufactured for a design verification test of the three dual instrumented fuel rods. The irradiation test of the 05F-01K instrumented fuel capsule will be carried out at the OR5 vertical experimental hole of Hanaro.

  5. Premiering SAFE for Safety Added Fuel Element - 15020

    International Nuclear Information System (INIS)

    Bhowmik, P.K.; Shamim, J.A.; Suh, K.Y.; Suh, K.S.

    2015-01-01

    The impact of the Fukushima accident has been the willingness to implement passive safety measures in reactor design and to simplify reactor design itself. Within this framework, a new fuel element, named SAFE (Safety Added Fuel Element) based on the concept of accident tolerant fuel, is presented. SAFE is a new type of fuel element cooled internally and externally by light water and with stainless steel as the cladding material. The removal of boron may trigger a series of changes which may simplify the system greatly. A simplified thermal analysis of SAFE shows that the fuel centerline temperature is well below the maximal limit during the normal operation of the plant

  6. A shielded measurement system for irradiated nuclear fuel measurements

    International Nuclear Information System (INIS)

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

    1999-01-01

    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 requirements 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 3 He detectors and ionization chambers to allow for gross neutron and gamma-ray measurements. The phase 2 SMS

  7. LOFT fuel rod surface temperature measurement testing

    International Nuclear Information System (INIS)

    Eaton, A.M.; Tolman, E.L.; Solbrig, C.W.

    1978-01-01

    Testing of the LOFT fuel rod cladding surface thermocouples has been performed to evaluate how accurately the LOFT thermocouples measure the cladding surface temperature during a loss-of-coolant accident (LOCA) sequence and what effect, if any, the thermocouple would have on core performance. Extensive testing has been done to characterize the thermocouple design. Thermal cycling and corrosion testing of the thermocouple weld design have provided an expected lifetime of 6000 hours when exposed to reactor coolant conditions of 620 K and 15.9 MPa and to sixteen thermal cycles with an initial temperature of 480 K and peak temperatures ranging from 870 to 1200K. Departure from nucleate boiling (DNB) tests have indicated a DNB penalty (5 to 28% lower) during steady state operation and negligible effects during LOCA blowdown caused by the LOFT fuel rod surface thermocouple arrangement. Experience with the thermocouple design in Power Burst Facility (PBF) and LOFT nonnuclear blowdown testing has been quite satisfactory. Tests discussed here were conducted using both stainless steel and zircaloy-clad electrically heated rod in the LOFT Test Support Facility (LTSF) blowdown simulation loop

  8. Automatic three-dimensional rib centerline extraction from CT scans for enhanced visualization and anatomical context

    Science.gov (United States)

    Ramakrishnan, Sowmya; Alvino, Christopher; Grady, Leo; Kiraly, Atilla

    2011-03-01

    We present a complete automatic system to extract 3D centerlines of ribs from thoracic CT scans. Our rib centerline system determines the positional information for the rib cage consisting of extracted rib centerlines, spinal canal centerline, pairing and labeling of ribs. We show an application of this output to produce an enhanced visualization of the rib cage by the method of Kiraly et al., in which the ribs are digitally unfolded along their centerlines. The centerline extraction consists of three stages: (a) pre-trace processing for rib localization, (b) rib centerline tracing, and (c) post-trace processing to merge the rib traces. Then we classify ribs from non-ribs and determine anatomical rib labeling. Our novel centerline tracing technique uses the Random Walker algorithm to segment the structural boundary of the rib in successive 2D cross sections orthogonal to the longitudinal direction of the ribs. Then the rib centerline is progressively traced along the rib using a 3D Kalman filter. The rib centerline extraction framework was evaluated on 149 CT datasets with varying slice spacing, dose, and under a variety of reconstruction kernels. The results of the evaluation are presented. The extraction takes approximately 20 seconds on a modern radiology workstation and performs robustly even in the presence of partial volume effects or rib pathologies such as bone metastases or fractures, making the system suitable for assisting clinicians in expediting routine rib reading for oncology and trauma applications.

  9. Fabrication of the instrumented fuel rods for the 3-Pin Fuel Test Loop at HANARO

    International Nuclear Information System (INIS)

    Sohn, Jae Min; Park, Sung Jae; Shin, Yoon Tag; Lee, Jong Min; Ahn, Sung Ho; Kim, Soo Sung; Kim, Bong Goo; Kim, Young Ki; Lee, Ki Hong; Kim, Kwan Hyun

    2008-09-01

    The 3-Pin Fuel Test Loop(hereinafter referred to as the '3-Pin FTL') facility has been installed at HANARO(High-flux Advanced Neutron Application Reactor) and the 3-Pin FTL is under a test operation. The purpose of this report is to fabricate the instrumented fuel rods for the 3-Pin FTL. The fabrication of these fuel rods was based on experiences and technologies of the instrumented fuel rods for an irradiation fuel capsule. The three instrumented fuel rods of the 3-Pin FTL have been designed. The one fuel rod(180 .deg. ) was designed to measure the centerline temperature of the nuclear fuels and the internal pressure of the fuel rod, and others(60 .deg. and 300 .deg. ) were designed to measure the centerline temperature of the fuel pellets. The claddings were made of the reference material 1 and 2 and new material 1 and 2. And nuclear fuel was used UO 2 (2.0w/o) pellet type with large grain and standard grain. The major procedures of fabrication are followings: (1) the assembling and weld of fuel rods with the pellet mockups and the sensor mockups for the qualification tests, (2) the qualification tests(dimension measurements, tensile tests, metallography examinations and helium leak tests) of weld, (3) the assembling and weld of instrumented fuel rods with the nuclear pellets and the sensors for the irradiation test, and (4) the qualification tests(the helium leak test, the dimensional measurement, electric resistance measurements of sensors) of test fuel rods. Satisfactory results were obtained for all the qualification tests of the instrumented fuel rods for the 3-Pin FTL. Therefore the three instrumented fuel rods for the 3-Pin FTL have been fabricated successfully. These will be installed in the In-Pile Section of 3-Pin FTL. And the irradiation test of these fuel rods is planned from the early next year for about 3 years at HANARO

  10. Risk assessment of sleeping disorder breathing based on upper airway centerline evaluation

    Science.gov (United States)

    Alsufyani, Noura; Shen, Rui; Cheng, Irene; Major, Paul

    2013-02-01

    One of the most important breathing disorders in childhood is obstructive sleep apnea syndrome which affects 2-3% of children, and the reported failure rate of surgical treatment was as high as 54%. A possible reason in respiratory complications is having reduced dimensions of the upper airway which are further compressed when muscle tone is decreased during sleep. In this study, we use Cone-beam computed tomography (CBCT) to assess the location or cause of the airway obstruction. To date, all studies analyzing the upper airway in subjects with Sleeping Disorder Breathing were based on linear, area, or volumetric measurements, which are global computations and can easily ignore local significance. Skeletonization was initially introduced as a 3D modeling technique by which representative medial points of a model are extracted to generate centerlines for evaluations. Although centerlines have been commonly used in guiding surgical procedures, our novelty lies in comparing its geometric properties before and after surgeries. We apply 3D data refinement, registration and projection steps to quantify and localize the geometric deviation in target airway regions. Through cross validation with corresponding subjects' therapy data, we expect to quantify the tolerance threshold beyond which reduced dimensions of the upper airway are not clinically significant. The ultimate goal is to utilize this threshold to identify patients at risk of complications. Outcome from this research will also help establish a predictive model for training and to estimate treatment success based on airway measurements prior to intervention. Preliminary results demonstrate the feasibility of our approach.

  11. Use of GMM and SCMS for Accurate Road Centerline Extraction from the Classified Image

    Directory of Open Access Journals (Sweden)

    Zelang Miao

    2015-01-01

    Full Text Available The extraction of road centerline from the classified image is a fundamental image analysis technology. Common problems encountered in road centerline extraction include low ability for coping with the general case, production of undesired objects, and inefficiency. To tackle these limitations, this paper presents a novel accurate centerline extraction method using Gaussian mixture model (GMM and subspace constraint mean shift (SCMS. The proposed method consists of three main steps. GMM is first used to partition the classified image into several clusters. The major axis of the ellipsoid of each cluster is extracted and deemed to be taken as the initial centerline. Finally, the initial result is adjusted using SCMS to produce precise road centerline. Both simulated and real datasets are used to validate the proposed method. Preliminary results demonstrate that the proposed method provides a comparatively robust solution for accurate centerline extraction from a classified image.

  12. 1st Fire Behavior and Fuels Conference: Fuels Management-How to Measure Success

    Science.gov (United States)

    Patricia L. Andrews

    2006-01-01

    The 1st Fire Behavior and Fuels Conference: Fuels Management -- How to Measure Success was held in Portland, Oregon, March 28-30, 2006. The International Association of Wildland Fire (IAWF) initiated a conference on this timely topic primarily in response to the needs of the U.S. National Interagency Fuels Coordinating Group (http://www.nifc.gov/).

  13. Irradiation test plan of instrumented capsule(05F-01K) for nuclear fuel irradiation in Hanaro (Revision 1)

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Jae Min; Kim, B. G.; Choi, M. H. (and others)

    2006-09-15

    An instrumented capsule was developed to be able to measure fuel characteristics, such as fuel temperature, internal pressure of fuel rod, fuel pellet elongation, and neutron flux, etc., during the irradiation test of nuclear fuel in HANARO. The instrumented capsule for measuring and monitoring fuel centerline temperature and neutron flux was designed and manufactured. And then, to verify the design of the instrumented capsule in the test hole, it was successfully irradiated in the test hole of HANARO from March 14, 2003 to June 1, 2003 (53.84 full power days at 24 MW). In the year of 2004, 3 test fuel rods and the 03F-05K instrumented fuel capsule were designed and fabricated to measure fuel centerline temperature, internal pressure of fuel rod, and fuel axial deformation during irradiation test. Now, this capsule was successfully irradiated in the test hole OR5 of HANARO reactor from April 27, 2004 to October 1, 2004 (59.5 full power days at 24-30 MW). The capsule and fuel rods have been be dismantled and fuel rods have been examined at the hot cell of IMEF. The instrumented fuel capsule (05F-01K) was designed and manufactured for a design verification test of the dual instrumented fuel rods. The irradiation test of the 05F-01K instrumented fuel capsule will be carried out at the OR5 vertical experimental hole of HANARO.

  14. Measurement of the fuel temperature and the fuel-to-coolant heat transfer coefficient of Super Phenix 1 fuel elements

    International Nuclear Information System (INIS)

    Edelmann, M.

    1995-12-01

    A new measurement method for measuring the mean fuel temperature as well as the fuel-to-coolant heat transfer coefficient of fast breeder reactor subassemblies (SA) is reported. The method is based on the individual heat balance of fuel SA's after fast reactor shut-downs and uses only the plants normal SA outlet temperature and neutron power signals. The method was used successfully at the french breeder prototype Super Phenix 1. The mean SA fuel temperature as well as the heat transfer coefficient of all SPX SA's have been determined at power levels between 15 and 90% of nominal power and increasing fuel burn-up from 3 to 83 EFPD (Equivalent of Full Power-Days). The measurements also provided fuel and whole SA time constants. The estimated accuracy of measured fuel parameters is in the order of 10%. Fuel temperatures and SA outlet temperature transients were also calculated with the SPX1 systems code DYN2 for exactly the same fuel and reactor operating parameters as in the experiments. Measured fuel temperatures were higher than calculated ones in all cases. The difference between measured and calculated core mean values increases from 50 K at low power to 180 K at 90% n.p. This is about the double of the experimental error margins. Measured SA heat transfer coefficients are by nearly 20% lower than corresponding heat transfer parameters used in the calculations. Discrepancies found between measured and calculated results also indicate that either the transient heat transfer in the gap between fuel and cladding (gap conductance) might not be exactly reproduced in the computer code or that the gap in the fresh fuel was larger than assumed in the calculations. (orig.) [de

  15. Non-rigid point set registration of curves: registration of the superficial vessel centerlines of the brain

    Science.gov (United States)

    Marreiros, Filipe M. M.; Wang, Chunliang; Rossitti, Sandro; Smedby, Örjan

    2016-03-01

    In this study we present a non-rigid point set registration for 3D curves (composed by 3D set of points). The method was evaluated in the task of registration of 3D superficial vessels of the brain where it was used to match vessel centerline points. It consists of a combination of the Coherent Point Drift (CPD) and the Thin-Plate Spline (TPS) semilandmarks. The CPD is used to perform the initial matching of centerline 3D points, while the semilandmark method iteratively relaxes/slides the points. For the evaluation, a Magnetic Resonance Angiography (MRA) dataset was used. Deformations were applied to the extracted vessels centerlines to simulate brain bulging and sinking, using a TPS deformation where a few control points were manipulated to obtain the desired transformation (T1). Once the correspondences are known, the corresponding points are used to define a new TPS deformation(T2). The errors are measured in the deformed space, by transforming the original points using T1 and T2 and measuring the distance between them. To simulate cases where the deformed vessel data is incomplete, parts of the reference vessels were cut and then deformed. Furthermore, anisotropic normally distributed noise was added. The results show that the error estimates (root mean square error and mean error) are below 1 mm, even in the presence of noise and incomplete data.

  16. Computerized x-ray radiographic system for fuel pellet measurements

    International Nuclear Information System (INIS)

    Green, D.R.; Karnesky, R.A.; Bromley, C.

    1977-01-01

    The development and operation of a computerized system for determination of fuel pellet diameters from x-ray radiography is described. Actual fuel pellet diameter measurements made with the system are compared to micrometer measurements on the same pellets, and statistically evaluated. The advantages and limitations of the system are discussed, and recommendations are made for further development

  17. Measurement of fuel importance distribution in non-uniformly distributed fuel systems

    International Nuclear Information System (INIS)

    Yamane, Yoshihiro; Hirano, Yasushi; Yasui, Hazime; Izima, Kazunori; Shiroya, Seiji; Kobayashi, Keiji.

    1995-01-01

    A reactivity effect due to a spatial variation of nuclear fuel concentration is an important problem for nuclear criticality safety in a reprocessing plant. As a theory estimating this reactivity effect, the Goertzel and fuel importance theories are well known. It has been shown that the Goertzel's theory is valid in the range of our experiments based on measurements of reactivity effect and thermal neutron flux in non-uniformly distributed fuel systems. On the other hand, there have been no reports concerning systematic experimental studies on the flatness of fuel importance which is a more general index than the Goertzel's theory. It is derived from the perturbation theory that the fuel importance is proportional to the reactivity change resulting from a change of small amount of fuel mass. Using a uniform and three kinds of nonuniform fuel systems consisting of 93.2% enriched uranium plates and polyethylene plates, the fuel importance distributions were measured. As a result, it was found experimentally that the fuel importance distribution became flat, as its reactivity effect became large. Therefore it was concluded that the flatness of fuel importance distribution is the useful index for estimating reactivity effect of non-uniformly distributed fuel system. (author)

  18. Measurement of reactivity effect caused by nonuniform fuel distribution

    International Nuclear Information System (INIS)

    Yamane, Yoshihiro; Hirano, Yasushi; Yasui, Hazime; Nishina, Kojiro; Shiroya, Seiji

    1991-01-01

    A reactivity effect due to a spatial variation of nuclear fuel concentration is an important problem in a reprocessing plant. To estimate this reactivity effect theoretically, the ''Goertzel's necessary condition, and th Fuel Importance'' theory have been proposed. In order to verify these theories, we have performed systematic measurements of reactivity effect due to the nonuniformity in the fuel distribution within the Kyoto University Critical Assembly. Neutron flux distribution and Fuel Importance distribution were also determined. A nonuniform assembly whose fuel concentration in the center region was 40% higher than the uniform one was found to have an excess reactivity of 0.3%Δk/k, with the same total uranium mass for which the uniform assembly was just critical. Moreover, its spatial distribution of thermal neutron flux and of Fuel Importance were more flat than those of the uniform assembly, as expected by the Goertzel's condition and the Fuel Importance theory. (Author)

  19. Methods for measuring of fuel can deformation under radiation conditions

    International Nuclear Information System (INIS)

    Zelenchuk, A.V.; Fetisov, B.V.; Lakin, Yu.G.; Tonkov, V.Yu.

    1978-01-01

    The possibility for measuring fuel can deformation under radiation conditions by means of the acoustic method and tensoresistors is considered. The construction and operation of the in-pile facility for measuring creep of the fuel can specimen loaded by the internal pressure is described. The data on neutron radiation effect on changes in creep rate for zirconium fuel can are presented. The results obtained with tensoresistors are in a good agreement with those obtained by the acoustic method, which enables to recommend the use of both methods for the irradiation creep investigation of the fuel element cans

  20. Oxide thickness measurement for monitoring fuel performance at high burnup

    International Nuclear Information System (INIS)

    Jaeger, M.A.; Van Swam, L.F.P.; Brueck-Neufeld, K.

    1991-01-01

    For on-site monitoring of the fuel performance at high burnup, Advanced Nuclear Fuels uses the linear scan eddy current method to determine the oxide thickness of irradiated Zircaloy fuel cans. Direct digital data acquisition methods are employed to collect the data on magnetic storage media. This field-proven methodology allows oxide thickness measurements and rapid interpretation of the data during the reactor outages and makes it possible to immediately reinsert the assemblies for the next operating cycle. The accuracy of the poolside measurements and data acquisition/interpretation techniques have been verified through hot cell metallographic measurements of rods previously measured in the fuel pool. The accumulated data provide a valuable database against which oxide growth models have been benchmarked and allow for effective monitoring of fuel performance. (orig.) [de

  1. Accuracy of fuel motion measurements using in-core detectors

    International Nuclear Information System (INIS)

    Dupree, S.A.

    1975-01-01

    An initial assessment has been made as to how accurately fuel motion can be measured with in-core detectors. A portion of this assessment has involved the calculation of the response of various detectors to fuel motion and the development of a formalism for correlating uncertainties in a neutron flux measurement to uncertainties in the fuel motion. Initially, four idealized configurations were studied in one dimension. These configurations consisted of (1) a single fuel-pin test using ACPR, (2) a seven fuel-pin test using ACPR, (3) a full subassembly (271 pin) test using a Class I ANL-type SAREF, and (4) a full subassembly plus six partial subassemblies (approximately 1000 pin) test using a Class III GE-type SAREF. It was assumed that melt would occur symmetrically at the center of the test fuel and that fuel would therefore disappear from the center of the geometry. For each case of series of calculations was performed in which detector responses were determined at several radial locations for the unperturbed core and for the core with various fractions of the fuel replaced with Na. This fuel loss was assumed to occur essentially instantaneously such that the power level in the remaining portion of the test fuel remained unchanged from that of the initial unperturbed condition

  2. Irradiation effects on thermal properties of LWR hydride fuel

    Energy Technology Data Exchange (ETDEWEB)

    Terrani, Kurt, E-mail: terrani@berkeley.edu [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Balooch, Mehdi [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Carpenter, David; Kohse, Gordon [Massachusetts Institute of Technology, 138 Albany St., Cambridge, MA 02139 (United States); Keiser, Dennis; Meyer, Mitchell [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Olander, Donald [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States)

    2017-04-01

    Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH{sub 1.6}) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

  3. Precise measurement of fuel content of irradiated and nonirradiated materials

    International Nuclear Information System (INIS)

    Harker, Y.D.; Napper, P.R.; Proctor, A.E.

    1984-01-01

    This paper discusses the application of precise reactivity measurements in the Advanced Reactivity Measurement Facility at Idaho National Engineering Laboratory (INEL) to determine th fuel content in irradiated and nonirradiated materials. Different methods of reactivity measurements and examples of how they have been are presented, which provides an insight in capabilities available to analyze samples with different geometrical sizes from small volumes approx. 100 cc to 12 ft long fuel pins and also samples with different fuel content ranges from approx. 2 mg to approx. 600 g. The overall accuracy of these measurements is approx. 0.5% (1sigma)

  4. Software to improve spent fuel measurements using the FDET

    International Nuclear Information System (INIS)

    Staples, P.; Beddingfield, D.; Lestone, J.; Pelowitz, D.; Sprinkle, J.; Bytchkov, V.; Starovich, Z.; Harizanov, I.; Vellejo-Luna, J.; Lavender, C.

    2001-01-01

    Full text: Vast quantities of spent fuel are available for safeguard measurements, primarily in Commonwealth of Independent States (CIS) of the former Soviet Union. This spent fuel, much of which consists of long cooling time material, is going to become less unique in the world safeguards arena as reprocessing projects or permanent repositories continue to be delayed or postponed. The long cooling time of many of the spent fuel assemblies in the CIS countries being prepared for intermediate term storage promotes the possibility of increased accuracy in spent fuel assays. An important point to consider for the future that could advance safeguards measurements for re-verification and inspection measurements would be to determine what safeguards requirements should be imposed upon this 'new' class of spent fuel. Improvements in measurement capability will obviously affect the safeguards requirements. What most significantly enables this progress in spent fuel measurements is the improvement in computer processing power and software enhancements leading to user-friendly Graphical User Interfaces (GUI's). The software used for these projects significantly reduces the IAEA inspector's time both learning and operating computer and data acquisition systems. While at the same time by standardizing the spent fuel measurements it is possible to increase reproducibility and reliability of the measurement data. The inclusion of various analysis algorithms into the operating environment, which can be performed in real time upon the measurement data, can also lead to increases in safeguard reliability and improvements in efficiency to plant operations. (author)

  5. Road and Street Centerlines, Buffalo County, WI road centerline file conflated with DOT WISLR attributes and updated continually based on survey plats, Published in 2002, 1:600 (1in=50ft) scale, Buffalo County Government.

    Data.gov (United States)

    NSGIC Local Govt | GIS Inventory — Road and Street Centerlines dataset current as of 2002. Buffalo County, WI road centerline file conflated with DOT WISLR attributes and updated continually based on...

  6. Fuels Management-How to Measure Success: Conference Proceedings

    Science.gov (United States)

    Patricia L. Andrews; Bret W. Butler

    2006-01-01

    Fuels management programs are designed to reduce risks to communities and to improve and maintain ecosystem health. The International Association of Wildland Fire initiated the 1st Fire Behavior and Fuels Conference to address development, implementation, and evaluation of these programs. The focus was on how to measure success. Over 500 participants from several...

  7. Measuring deformation of Fuel pin in a Nuclear Fuel Test Rig

    Energy Technology Data Exchange (ETDEWEB)

    Heo, S. H.; Yang, T. H.; Hong, J. T.; Joung, C. Y.; Ahn, S. H.; Jang, S. Y.; Kim, J. H. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, an LVDT core for measuring the longitudinal displacement of fuel pellets and clad was designed and produced. A signal processing method for the prepared core was investigated. The Nuclear Fuel Test Rig is used to observe changes in the characteristics of the fuel according to the neutron irradiation at HANARO (High-flux Advanced Neutron Application Reactor), which is a research reactor. Which are the strain and internal temperature of the irradiated nuclear fuel and the internal pressure of fuel due to fission gas, the characteristics of the fuel are measured using various sensors such as a thermocouple, SPND and LVDT. In this study, two shaped LVDT (Linear Variable Differential Transformer) cores for displacement measurements were designed and manufactured in order to measure the displacement of a fuel pellet and cladding tube using LVDT sensors for measuring electrical signals by converting the physical variation such as the force and displacement into a linear motion. In addition, signals from the manufactured LVDT sensor were collected and calibrated. Moreover, a method for obtaining the displacement in the core according to the sensing signal was planned. A derived equation can used to predict the change in the position of core. A following study should be conducted to test the output signal and real variation of out-pile system. For further work, a performance verification is required for an in-pile irradiation test.

  8. Nuclear reactor, fuel assembly and neutron measuring system

    International Nuclear Information System (INIS)

    Chaki, Masao; Murase, Michio; Zukeran, Atsushi; Moriya, Kimiaki

    1998-01-01

    The present invention provides a BWR type reactor improved with the efficiency of used fuels and fuel economy by increasing a rated power and reducing exchange fuels. Namely, in a BWR type reactor at present, a thermal limit value is determined by conducting nuclear calculation of the reactor core based on data of reactor flow rate measurement and data of neutron flux measurement. However, since the neutron calculation of the reactor core is based on fuel assemblies while the points for the neutron measurement are present at the outside of the fuel assemblies, errors are caused. A margin including the errors has been used as a thermal limit value during operation. In the present invention, neutron fluxes in the fuel assembly as a base of the nuclear calculation can be measured by the same number of neutron detector tubes, but the number of the measuring points is increased to four times. With such procedures, errors caused by the difference of the neutron calculation and values at neutron measuring points can be reduced. As a result, a margin of the thermal limit value is reduced to increase the degree of freedom of reactor operation. Then, the economical property of the reactor operation can be improved. (N.H.)

  9. Nondestructive nuclear measurement in the fuel cycle. Part 1

    International Nuclear Information System (INIS)

    Lyoussi, A.

    2005-01-01

    Nondestructive measurement techniques are today widely used in practically all steps of the fuel cycle. This article is devoted to the presentation of the control and characterization needs and to the main passive nondestructive nuclear methods used: 1 - nondestructive nuclear measurement, needs and motivation: nuclear fuel cycle, nondestructive nuclear measurements (passive and active methods), comments; 2 - main passive nondestructive nuclear measurement methods: gamma spectroscopy (principle, detectors, electronic systems, data acquisition and signal processing, domains of application, main limitations), passive neutronic measurements (needs and motivations, neutron detectors, total neutronic counting, neutronic coincidences counting, neutronic multiplicities counting, comments). (J.S.)

  10. Alternative Measuring Approaches in Gamma Scanning on Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sihm Kvenangen, Karen

    2007-06-15

    In the future, the demand for energy is predicted to grow and more countries plan to utilize nuclear energy as their source of electric energy. This gives rise to many important issues connected to nuclear energy, such as finding methods that can verify that the spent nuclear fuel has been handled safely and used in ordinary power producing cycles as stated by the operators. Gamma ray spectroscopy is one method used for identification and verification of spent nuclear fuel. In the specific gamma ray spectroscopy method called gamma scanning the gamma radiation from the fission products Cs-137, Cs-134 and Eu-154 are measured in a spent fuel assembly. From the results, conclusions can be drawn about the fuels characteristics. This degree project examines the possibilities of using alternative measuring approaches when using the gamma scanning method. The focus is on examining how to increase the quality of the measured data. How to decrease the measuring time as compared with the present measuring strategy, has also been investigated. The main part of the study comprises computer simulations of gamma scanning measurements. The simulations have been validated with actual measurements on spent nuclear fuel at the central interim storage, Clab. The results show that concerning the quality of the measuring data the conventional strategy is preferable, but with other starting positions and with a more optimized equipment. When focusing on the time aspect, the helical measuring strategy can be an option, but this needs further investigation.

  11. Alternative Measuring Approaches in Gamma Scanning on Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Sihm Kvenangen, Karen

    2007-06-01

    In the future, the demand for energy is predicted to grow and more countries plan to utilize nuclear energy as their source of electric energy. This gives rise to many important issues connected to nuclear energy, such as finding methods that can verify that the spent nuclear fuel has been handled safely and used in ordinary power producing cycles as stated by the operators. Gamma ray spectroscopy is one method used for identification and verification of spent nuclear fuel. In the specific gamma ray spectroscopy method called gamma scanning the gamma radiation from the fission products Cs-137, Cs-134 and Eu-154 are measured in a spent fuel assembly. From the results, conclusions can be drawn about the fuels characteristics. This degree project examines the possibilities of using alternative measuring approaches when using the gamma scanning method. The focus is on examining how to increase the quality of the measured data. How to decrease the measuring time as compared with the present measuring strategy, has also been investigated. The main part of the study comprises computer simulations of gamma scanning measurements. The simulations have been validated with actual measurements on spent nuclear fuel at the central interim storage, Clab. The results show that concerning the quality of the measuring data the conventional strategy is preferable, but with other starting positions and with a more optimized equipment. When focusing on the time aspect, the helical measuring strategy can be an option, but this needs further investigation

  12. Fission gas release in LWR fuel measured during nuclear operation

    International Nuclear Information System (INIS)

    Appelhans, A.D.; Skattum, E.; Osetek, D.J.

    1980-01-01

    A series of fuel behavior experiments are being conducted in the Heavy Boiling Water Reactor in Halden, Norway, to measure the release of Xe, Kr, and I fission products from typical light water reactor design fuel pellets. Helium gas is used to sweep the Xe and Kr fission gases out of two of the Instrumented Fuel Assembly 430 fuel rods and to a gamma spectrometer. The measurements of Xe and Kr are made during nuclear operation at steady state power, and for 135 I following reactor scram. The first experiments were conducted at a burnup of 3000 MWd/t UO 2 , at bulk average fuel temperatures of approx. 850 K and approx. 23 kW/m rod power. The measured release-to-birth ratios (R/B) of Xe and Kr are of the same magnitude as those observed in small UO 2 specimen experiments, when normalized to the estimated fuel surface-to-volume ratio. Preliminary analysis indicates that the release-to-birth ratios can be calculated, using diffusion coefficients determined from small specimen data, to within a factor of approx. 2 for the IFA-430 fuel. The release rate of 135 I is shown to be approximately equal to that of 135 Xe

  13. Measurement of soluble nuclide dissolution rates from spent fuel

    International Nuclear Information System (INIS)

    Wilson, C.N.; Gray, W.J.

    1990-01-01

    Gaining a better understanding of the potential release behavior of water-soluble radionuclides is the focus of new laboratory spent fuel dissolution studies being planned in support of the Yucca Mountain Project. Previous studies have suggested that maximum release rates for actinide nuclides, which account for most of the long-term radioactivity in spent fuel, should be solubility-limited and should not depend on the characteristics or durability of the spent fuel waste form. Maximum actinide concentrations should be sufficiently low to meet the NRC (Nuclear Regulatory Commission) annual release limits. Potential release rates for soluble nuclides such as 99 Tc, 135 Cs, 14 C and 129 I, which account for about 1-2% of the activity in spent fuel at 1,000 years, are less certain and may depend on processes such as oxidation of the fuel in the repository air environment. Dissolution rates for several soluble nuclides have been measured from spent fuel specimens using static and semi-static methods. However, such tests do not provide a direct measurement of fuel matrix dissolution rates that may ultimately control soluble-nuclide release rates. Flow-through tests are being developed as a potential supplemental method for determining the matrix component of soluble-nuclide dissolution. Advantages and disadvantages of both semi-static and flow-through methods are discussed. Tests with fuel specimens representing a range of potential fuel states that may occur in the repository, including oxidized fuel, are proposed. Preliminary results from flow-through tests with unirradiated UO 2 suggesting that matrix dissolution rates are very sensitive to water composition are also presented

  14. Ground measurements of fuel and fuel consumption from experimental and operational prescribed fires at Eglin Air Force Base, Florida

    Science.gov (United States)

    Roger D. Ottmar; Robert E. Vihnanek; Clinton S. Wright; Andrew T. Hudak

    2014-01-01

    Ground-level measurements of fuel loading, fuel consumption, and fuel moisture content were collected on nine research burns conducted at Eglin Air Force Base, Florida in November, 2012. A grass or grass-shrub fuelbed dominated eight of the research blocks; the ninth was a managed longleaf pine (Pinus palustrus) forest. Fuel loading ranged from 1.7 Mg ha-1 on a...

  15. Shielding Performance Measurements of Spent Fuel Transportation Container

    Directory of Open Access Journals (Sweden)

    SUN Hong-chao

    2015-11-01

    Full Text Available The safety supervision of radioactive material transportation package has been further stressed and implemented. The shielding performance measurements of spent fuel transport container is the important content of supervision. However, some of the problems and difficulties reflected in practice need to be solved, such as the neutron dose rate on the surface of package is too difficult to measure exactly, the monitoring results are not always reliable, etc. The monitoring results using different spectrometers were compared and the simulation results of MCNP runs were considered. An improvement was provided to the shielding performance measurements technique and management of spent fuel transport.

  16. Measuring element for determining the internal pressure in fuel rods

    International Nuclear Information System (INIS)

    Deckers, H.; Drexler, H.; Reiser, H.

    1983-01-01

    A pressure cell is situated inside the fuel rod, which contains a magnetic core or a core influenced by magnetism, whose position relative to an outer front surface of an end stopper of the fuel rod can vary. The fuel rod contains a pressure cell directly above the lower end stopper or connected to it. This can consist of closed bellows, where if the internal pressure in the fuel rod rises, a ferrite core moves axially. When the pressure drops, this returns to the initial position, which is precisely defined by a stop. To detect a rod defect, the position of the soft iron core relative to the lower edge of the end stopper is scanned by a special measuring device. (orig./HP) [de

  17. Gamma densitometer for measuring Pu density in fuel tubes

    International Nuclear Information System (INIS)

    Winn, W.G.

    1982-01-01

    A fuel-gamma-densitometer (FGD) has been developed to examine nondestructively the uniformity of plutonium in aluminum-clad fuel tubes at the Savannah River Plant (SRP). The monitoring technique is γ-ray spectroscopy with a lead-collimated Ge(Li) detector. Plutonium density is correlated with the measured intensity of the 208 keV γ-ray from 237 U (7d) of the 241 Pu (15y) decay chain. The FGD measures the plutonium density within 0.125- or 0.25-inch-diameter areas of the 0.133- to 0.183-inch-thick tube walls. Each measurement yields a density ratio that relates the plutonium density of the measured area to the plutonium density in normal regions of the tube. The technique was used to appraise a series of fuel tubes to be irradated in an SRP reactor. High-density plutonium areas were initially identified by x-ray methods and then examined quantitatively with the FGD. The FGD reliably tested fuel tubes and yielded density ratios over a range of 0.0 to 2.5. FGD measurements examined (1) nonuniform plutonium densities or hot spots, (2) uniform high-density patches, and (3) plutonium density distribution in thin cladding regions. Measurements for tubes with known plutonium density agreed with predictions to within 2%. Attenuation measurements of the 208-keV γ-ray passage through the tube walls agreed to within 2 to 3% of calculated predictions. Collimator leakage measurements agreed with model calculations that predicted less than a 1.5% effect on plutonium density ratios. Finally, FGD measurements correlated well with x-ray transmission and fluoroscopic measurements. The data analysis for density ratios involved a small correction of about 10% for γ-shielding within the fuel tube. For hot spot examinations, limited information for this correction dictated a density ratio uncertainty of 3 to 5%

  18. Pressure drop ana velocity measurements in KMRR fuel rod bundles

    International Nuclear Information System (INIS)

    Yagn, Sun Kyu; Chung, Heung June; Chung, Chang Whan; Chun, Se Young; Song, Chul Wha; Won, Soon Yeun; Chung, Moon Ki

    1990-01-01

    The detailed hydraulic characteristic measurements in subchannels of longitudinally finned rod bundles using one-component LDV(Laser Doppler Velocimeter) were performed. Time mean axial velocity, turbulent intensity, and turbulent micro scales, such as time auto-correlation, Eulerian integral and micro scale, Kolmogorov length and time scale, and Taylor micro length scale were measured. The signals from LDV are inherently more or less discontinuous. The spectra of signals having such intermittent defects can be obtained by the fast Fourier transformation (FFT) of the auto-correlation function. The turbulent crossflow mixing rate between neighboring subchannels and dominant frequencies were evaluated from the measured data. Pressure drop data were obtained for the typical 36-element and 18-element fuel rod bundles fabricated by the design requirement of KMRR fuel and for other type of fuels assembled with 6-fin rods to investigate the fin effects on the pressure drop characteristics

  19. Nondestructive nuclear measurement in the fuel cycle. Part 2

    International Nuclear Information System (INIS)

    Lyoussi, A.

    2005-01-01

    Nondestructive measurement techniques are today widely used in practically all steps of the fuel cycle. This article is devoted to the presentation of the control and characterization needs and to the main active nondestructive nuclear methods used: 1 - main active nondestructive nuclear measurement methods: active neutronic measurement (needs and motivations, physical principle, measurement of delayed neutrons following a continuous irradiation, measurement of prompt neutrons (differential die-away technique - DDT), measurement of prompt and delayed neutrons (Sphincs method), neutronic method coupled to gamma spectroscopy), measurement by induced photo-fissions (needs and motivations, physical principle); 2 - conclusion. (J.S.)

  20. PWR and BWR spent fuel assembly gamma spectra measurements

    Energy Technology Data Exchange (ETDEWEB)

    Vaccaro, S. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Tobin, S.J.; Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hu, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Schwalbach, P. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden); Trellue, H.; Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2016-10-11

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of {sup 137}Cs, {sup 154}Eu, and {sup 134}Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  1. The interpretation of fuel centre temperature measurements on a suspected leaking fuel pin

    International Nuclear Information System (INIS)

    Ainscough, J.B.; Lang, C.; Clough, D.J.

    1983-01-01

    In order to study fuel densification a series of single instrumented pin irradiations has been carried out in the High Pressure Water Loop of DIDO at Harwell. The behaviour of two of these pins was different from that expected. In the fifth test, where the fuel was 95% dense pellet UO 2 and expected to densify readily in-reactor, the fuel centre temperature increased from its starting value of approx. 1300 deg. C at a rate somewhat higher than expected on the basis of predicted densification rates. After about six days, the temperature increased rapidly and unexpectedly to 2100-2200 deg. C and remained steady at this level for a further eight days until a reactor trip occurred and the pin was unloaded. Predictions made using the HOTROD code imply a maximum fuel temperature of less than 1500 deg. C after densification. Post-irradiation examination confirmed that fission gas release had occurred, that the measured temperatures were consistent with the fuel microstructure and that the pin had a high internal gas pressure. The fourth pin in the series contained 97% dense UO 2 which was also expected to be dimensionally unstable. Qualitatively its behaviour was similar to that of the fifth pin though the temperatures throughout were lower. This pin experienced a number of major power cycles and failed after about 30 days in-reactor. It is probable that coolant ingress occurred in both pins via the thermocouple Hoke seal, degrading the filling gas conductivity and allowing the fuel to densify rapidly with consequent increase in the fuel/clad gap and hence in fuel temperature. These irradiations show that, for a short time at least, an apparently unfailed pin could operate undetected with temperatures significantly higher than those predicted for normal operation. (author)

  2. Applications of a transportable spent-fuel measurement system

    International Nuclear Information System (INIS)

    Halbig, J.K.; Bosler, G.E.; Klosterbuer, S.F.; Rinard, P.M.

    1985-01-01

    A portable tool for making measurements on irradiated fuel has been developed to where in-plant installations having a 1 to 2% measurement uncertainty of relative exposure are feasible. The measurement uses a passive gross neutron signal and data from a gross gamma measurement as a consistency check of the neutron result and the operators declaration of cooling time. The uncertainties are about the same as those obtained using high-resolution gamma-ray techniques without the instrumentation being as obtrusive. The battery-operated microprocessor-based electronics package used with the irradiated fuel measurement system can also be used with single channel pulse counting detectors for other applications. This feature together with the large dynamic range of its current-mode ion chamber channel makes ION-I a good building block to be used in emergencies with an arsenal of detectors at a variety of nuclear plants. 8 figs., 3 tabs

  3. Passive measurements of mixed-oxide fuel for nuclear nonproliferation

    International Nuclear Information System (INIS)

    Dolan, Jennifer L.; Flaska, Marek; Pozzi, Sara A.; Chichester, David L.

    2013-01-01

    We present new results on passive measurements and simulations of mixed-oxide fuel-pin assemblies. Potential tools for mixed-oxide fuel pin characterization are discussed for future nuclear-nonproliferation applications. Four EJ-309 liquid scintillation detectors coupled with an accurate pulse timing and digital, offline and optimized pulse-shape discrimination method were used. Measurement analysis included pulse-height distributions to distinguish between purely fission neutron sources and alpha-n plus fission neutrons sources. Time-dependent cross-correlation functions were analyzed to measure the fission neutron contribution to the measured sample's neutron source. The use of Monte Carlo particle transport code MCNPX-PoliMi is discussed in conjunction with the measurements

  4. Fuel rod-to-support contact pressure and stress measurement for CHASNUPP-1(PWR) fuel

    International Nuclear Information System (INIS)

    Waseem; Elahi, N.; Siddiqui, A.; Murtaza, G.

    2011-01-01

    Research highlights: → A detailed finite element model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. → The spring hold-down force is calculated using the contact pressure obtained from the FE model. → Experiment has also been conducted in the same environment for the measurement of this force. → The spring hold-down force values obtained from both studies confirm the validation of this analysis. → The stress obtained through this analysis is less than the yield strength of spacer grid material, thus fulfils the structural integrity criteria of grid. - Abstract: This analysis has been made in an attempt to measure the contact pressure of the PWR fuel assembly spacer grid spring and to verify its structural integrity at room temperature in air. A detailed finite element (FE) model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. The FE model of a fuel rod-to-support system is produced with shell and contact elements. The spring hold-down force is calculated using the contact pressure obtained from the FE model. Experiment has also been conducted in the same environment for the measurement of this force. The spring hold-down force values obtained from both studies are compared, which show good agreement, and in turn confirm the validation of this analysis. The Stress obtained through this analysis is less than the yield strength of spacer grid material (Inconel-718), thus fulfils the structural integrity criteria of grid.

  5. Fuel rod-to-support contact pressure and stress measurement for CHASNUPP-1(PWR) fuel

    Energy Technology Data Exchange (ETDEWEB)

    Waseem, E-mail: wazim_me@hotmail.co [Directorate General Nuclear Power Fuel, Pakistan Atomic Energy Commission, P.O. Box No. 1847, Islamabad 44000 (Pakistan); Elahi, N.; Siddiqui, A.; Murtaza, G. [Directorate General Nuclear Power Fuel, Pakistan Atomic Energy Commission, P.O. Box No. 1847, Islamabad 44000 (Pakistan)

    2011-01-15

    Research highlights: A detailed finite element model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. The spring hold-down force is calculated using the contact pressure obtained from the FE model. Experiment has also been conducted in the same environment for the measurement of this force. The spring hold-down force values obtained from both studies confirm the validation of this analysis. The stress obtained through this analysis is less than the yield strength of spacer grid material, thus fulfils the structural integrity criteria of grid. - Abstract: This analysis has been made in an attempt to measure the contact pressure of the PWR fuel assembly spacer grid spring and to verify its structural integrity at room temperature in air. A detailed finite element (FE) model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. The FE model of a fuel rod-to-support system is produced with shell and contact elements. The spring hold-down force is calculated using the contact pressure obtained from the FE model. Experiment has also been conducted in the same environment for the measurement of this force. The spring hold-down force values obtained from both studies are compared, which show good agreement, and in turn confirm the validation of this analysis. The Stress obtained through this analysis is less than the yield strength of spacer grid material (Inconel-718), thus fulfils the structural integrity criteria of grid.

  6. Cerebral blood flow with [15O]water PET studies using an image-derived input function and MR-defined carotid centerlines

    Science.gov (United States)

    Fung, Edward K.; Carson, Richard E.

    2013-03-01

    Full quantitative analysis of brain PET data requires knowledge of the arterial input function into the brain. Such data are normally acquired by arterial sampling with corrections for delay and dispersion to account for the distant sampling site. Several attempts have been made to extract an image-derived input function (IDIF) directly from the internal carotid arteries that supply the brain and are often visible in brain PET images. We have devised a method of delineating the internal carotids in co-registered magnetic resonance (MR) images using the level-set method and applying the segmentations to PET images using a novel centerline approach. Centerlines of the segmented carotids were modeled as cubic splines and re-registered in PET images summed over the early portion of the scan. Using information from the anatomical center of the vessel should minimize partial volume and spillover effects. Centerline time-activity curves were taken as the mean of the values for points along the centerline interpolated from neighboring voxels. A scale factor correction was derived from calculation of cerebral blood flow (CBF) using gold standard arterial blood measurements. We have applied the method to human subject data from multiple injections of [15O]water on the HRRT. The method was assessed by calculating the area under the curve (AUC) of the IDIF and the CBF, and comparing these to values computed using the gold standard arterial input curve. The average ratio of IDIF to arterial AUC (apparent recovery coefficient: aRC) across 9 subjects with multiple (n = 69) injections was 0.49 ± 0.09 at 0-30 s post tracer arrival, 0.45 ± 0.09 at 30-60 s, and 0.46 ± 0.09 at 60-90 s. Gray and white matter CBF values were 61.4 ± 11.0 and 15.6 ± 3.0 mL/min/100 g tissue using sampled blood data. Using IDIF centerlines scaled by the average aRC over each subjects’ injections, gray and white matter CBF values were 61.3 ± 13.5 and 15.5 ± 3.4 mL/min/100 g tissue. Using global

  7. Cooling flow measurement in fuel elements of the RA-6

    International Nuclear Information System (INIS)

    Brollo, F; Silin, N

    2009-01-01

    Under the UBERA6 project for the core change and power increase of the RA-6 reactor, the total coolant flow was increased to meet the requirements imposed by the new operating conditions. The flow through the fuel elements is an important parameter and is difficult to determine due to the geometric complexity of the core. To ensure safe operation of the reactor, adequate safety margins must be kept for all operating conditions. In the present work we performed the direct measurement of the cooling flow rate of a fuel in the reactor core, for which we used a turbine flowmeter built specifically for this use. This helped to confirm previous results obtained during the launch, made by an indirect method based on measuring the pressure difference of the core. The turbine flowmeter was chosen due to its robustness, ease of operation and low disturbance of the input stream to the fuel. We describe the calibration of this instrument and the results of flow measurements made on some of the RA6 reactor fuel elements under conditions of zero power. [es

  8. Zircaloy sheathed thermocouples for PWR fuel rod temperature measurements

    International Nuclear Information System (INIS)

    Anderson, J.V.; Wesley, R.D.; Wilkins, S.C.

    1979-01-01

    Small diameter zircaloy sheathed thermocouples have been developed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory. Surface mounted thermocouples were developed to measure the temperature of zircaloy clad fuel rods used in the Thermal Fuels Behavior Program (TFBP), and embedded thermocouples were developed for use by the Loss-of-Fluid Test (LOFT) Program for support tests using zircaloy clad electrically heated nuclear fuel rod simulators. The first objective of this developmental effort was to produce zircaloy sheathed thermocouples to replace titanium sheathed thermocouples and thereby eliminate the long-term corrosion of the titanium-to-zircaloy attachment weld. The second objective was to reduce the sheath diameter to obtain faster thermal response and minimize cladding temperature disturbance due to thermocouple attachment

  9. Fuel element burnup measurements for the equilibrium LEU silicide RSG GAS (MPR-30) core under a new fuel management strategy

    International Nuclear Information System (INIS)

    Pinem, Surian; Liem, Peng Hong; Sembiring, Tagor Malem; Surbakti, Tukiran

    2016-01-01

    Highlights: • Burnup measurement of fuel elements comprising the new equilibrium LEU silicide core of RSG GAS. • The burnup measurement method is based on a linear relationship between reactivity and burnup. • Burnup verification was conducted using an in-house, in-core fuel management code BATAN-FUEL. • A good agreement between the measured and calculated burnup was confirmed. • The new fuel management strategy was confirmed and validated. - Abstract: After the equilibrium LEU silicide core of RSG GAS was achieved, there was a strong need to validate the new fuel management strategy by measuring burnup of fuel elements comprising the core. Since the regulatory body had a great concern on the safety limit of the silicide fuel element burnup, amongst the 35 burnt fuel elements we selected 22 fuel elements with high burnup classes i.e. from 20 to 53% loss of U-235 (declared values) for the present measurements. The burnup measurement method was based on a linear relationship between reactivity and burnup where the measurements were conducted under subcritical conditions using two fission counters of the reactor startup channel. The measurement results were compared with the declared burnup evaluated by an in-house in-core fuel management code, BATAN-FUEL. A good agreement between the measured burnup values and the calculated ones was found within 8% uncertainties. Possible major sources of differences were identified, i.e. large statistical errors (i.e. low fission counters’ count rates), variation of initial U-235 loading per fuel element and accuracy of control rod indicators. The measured burnup of the 22 fuel elements provided the confirmation of the core burnup distribution planned for the equilibrium LEU silicide core under the new fuel management strategy.

  10. LOFT advanced fuel rod instrumentation development

    International Nuclear Information System (INIS)

    Billeter, T.R.; Brown, R.L.; Chan, A.I.Y.; Day, C.K.; Meyers, S.C.; Sheen, E.M.; Stringer, J.L.

    1978-01-01

    Advanced fuel rod instrumentation for the Loss of Fluid Test (LOFT) reactor is being developed by the Hanford Engineering Development Laboratory for the Nuclear Regulatory Commission. This effort calls for development of sensors to measure fuel rod axial motion, fuel centerline temperature (to 2200 0 C), fuel rod plenum gas pressure (to 2500 psig), and plenum gas temperature (to 1500 0 F). A parallel test and evaluation of several modified commercial sensors was undertaken and will result in commercial availability of the final qualified sensors. Necessary test facilities were prepared for the development and evaluation effort. Tests to date indicate a three coil Linear Variable Differential Transformer (LVDT), operated from temperature compensating signal source and processing electronics, will meet the desired requirements

  11. Development of experimental methods for measuring fuel elements burnup

    International Nuclear Information System (INIS)

    PEREDA, C; HENRIQUEZ, C; NAVARRO, G; TORRES, H; KLEIN, J; CALDERON, D; MEDEL, J; MUTIS, O; DAIE, J; ITURRIETA, L; LONCOMILLA, M; ZAMBRANO, J; KESTELMAN, A

    2003-01-01

    This paper is a summary of the work carried out during the last two years in fuel burning measurements at RECH-1 for different enrichments, cooling times and burning rates. The measurements were made in two gamma-spectrometric facilities, one is installed in a hot cell and the other inside of the secondary pool of the RECH-1, where the element is under 2 meters of water. The hot cell measurements need at least 100 cooling days because of the problems generated by the transport of highly active fuel elements from the Reactor to the cell. This was the main reason for using the in-pool facility because of its capability to measure the burning of fuel elements without having to wait so long, that is with only 5 cooling days. The accumulated experience in measurements achieved in both facilities and the encouraging results show that this measuring method is reliable. The results agreed well with those obtained using the reactor's physics codes, which was the way they were obtained previously (Cw)

  12. Fission rate measurements in fuel plate type assembly reactor cores

    International Nuclear Information System (INIS)

    Rogers, J.W.

    1988-01-01

    The methods, materials and equipment have been developed to allow extensive and precise measurement of fission rate distributions in water moderated, U-Al fuel plate assembly type reactor cores. Fission rate monitors are accurately positioned in the reactor core, the reactor is operated at a low power for a short time, the fission rate monitors are counted with detectors incorporating automated sample changers and the measurements are converted to fission rate distributions. These measured fission rate distributions have been successfully used as baseline information related to the operation of test and experimental reactors with respect to fission power and distribution, fuel loading and fission experiments for approximately twenty years at the Idaho National Engineering Laboratory (INEL). 7 refs., 8 figs

  13. Heavy truck modeling for fuel consumption. Simulations and measurements

    Energy Technology Data Exchange (ETDEWEB)

    Sandberg, T.

    2001-12-01

    Fuel consumption for heavy trucks depends on many factors like roads, weather, and driver behavior that are hard for a manufacturer to influence. However, one design possibility is the power train configuration. Here a new simulation program for heavy trucks is created to find the configuration of the power train that gives the lowest fuel consumption for each transport task. For efficient simulations the model uses production code for speed and gear control, and it uses exchangeable data sets to allow simulation of the whole production range of engine types, on recorded road profiles from all over the world. Combined with a graphical user interface this application is called STARS (Scania Truck And Road Simulation). The forces of rolling resistance and air resistance in the model are validated through an experiment where the propeller shaft torque of a heavy truck is measured. It is found that the coefficient of rolling resistance is strongly dependent on tire temperature, not only on vehicle speed as expected. This led to the development of a new model for rolling resistance. The model includes the dynamic behavior of the tires and relates rolling resistance to tire temperature and vehicle speed. In another experiment the fuel consumption of a test truck in highway driving is measured. The altitude of the road is recorded with a barometer and used in the corresponding simulations. Despite of the limited accuracy of this equipment the simulation program manage to predict a level of fuel consumption only 2% lower than the real measurements. It is concluded that STARS is a good tool for predicting fuel consumption for trucks in highway driving and for comparing different power train configurations.

  14. Self-protection of FLIP fuel: Experimental measurements

    International Nuclear Information System (INIS)

    Johnson, A.G.; Ringle, J.C.; Anderson, T.V.; Dodd, B.; Bennett, S.; Carpenter, W.T.

    1980-01-01

    During the last year and one-half non-power reactor licensees of the Nuclear Regulatory Commission (NRC) have been repeatedly alerted to the fact that the Commission planned to tighten physical security regulations for their class of reactors. Essentially all of the new NRC physical security requirements for these reactors have been based on the type, quantity and enrichment of the special nuclear materials (e.g. enriched uranium and plutonium) possessed by the licensees. While each licensee's security classification will be evaluated individually, it is now apparent that non-power reactor licensees using or storing formula quantities of special nuclear material not meeting the 100 rem per hour at three feet self-protection exemption (e.g. 5 kilograms or more of non-self-protected fuel enriched to 20% or more with U-235) will be required to meet extremely stringent NRC physical security requirements patterned after the new Physical Protection Upgrade Rule. Implementation of these much stricter security requirements poses many economic and operational difficulties for university research reactors, to the point where facility closure could be the only alternative. TRIGA reactors utilizing a full FLIP fueled core, such as the Oregon State University TRIGA (OSTR), qualify for the highest physical protection category unless fuel can be maintained at self-protecting radiation levels. In order to demonstrate that OSTR fuel could be consistently kept above the 100 rem per hour threshold, a computer program was written which predicts the gamma radiation levels from an irradiated FLIP fuel element at 1, 2 and 3 feet in air and water. Furthermore, in order to verify the accuracy of the computer program, actual measurements of irradiated fuel elements were made at 3 feet in air and at 1, 2 and 3 feet in water, and the results compared very favorably to the predictions. The results of specific measurements, the instrumentation used and its calibration, the personnel doses

  15. Experience with W3Re/W25Re thermocouples in fuel pins of NS Otto Hahn's two cores

    International Nuclear Information System (INIS)

    Kolb, M.

    1976-01-01

    The paper first deals with the installation of 18 and 9 high-temperature sheathed thermocouples in fuel rods of the cores FDR-1 and FDR-2, respectively. The measured fuel rod centerline temperatures could be related to the local linear rod power at any given time by means of the densities of fission products with different half-lives obtained from fuel rod γ-scans. The fuel temperatures show then already an increase with the burn-up of the FDR-1 which becomes steeper when taking into account the decrease of the EMF measured at irradiated thermocouples taken from the fuel rods. Finally, the determination of effective thermocouple time constants and of fuel rod heat transfer time constants is demonstrated by utilizing the reactor noise to measure the transfer function between neutron flux and fuel temperature signal. (orig.) [de

  16. Intra-temporal facial nerve centerline segmentation for navigated temporal bone surgery

    Science.gov (United States)

    Voormolen, Eduard H. J.; van Stralen, Marijn; Woerdeman, Peter A.; Pluim, Josien P. W.; Noordmans, Herke J.; Regli, Luca; Berkelbach van der Sprenkel, Jan W.; Viergever, Max A.

    2011-03-01

    Approaches through the temporal bone require surgeons to drill away bone to expose a target skull base lesion while evading vital structures contained within it, such as the sigmoid sinus, jugular bulb, and facial nerve. We hypothesize that an augmented neuronavigation system that continuously calculates the distance to these structures and warns if the surgeon drills too close, will aid in making safe surgical approaches. Contemporary image guidance systems are lacking an automated method to segment the inhomogeneous and complexly curved facial nerve. Therefore, we developed a segmentation method to delineate the intra-temporal facial nerve centerline from clinically available temporal bone CT images semi-automatically. Our method requires the user to provide the start- and end-point of the facial nerve in a patient's CT scan, after which it iteratively matches an active appearance model based on the shape and texture of forty facial nerves. Its performance was evaluated on 20 patients by comparison to our gold standard: manually segmented facial nerve centerlines. Our segmentation method delineates facial nerve centerlines with a maximum error along its whole trajectory of 0.40+/-0.20 mm (mean+/-standard deviation). These results demonstrate that our model-based segmentation method can robustly segment facial nerve centerlines. Next, we can investigate whether integration of this automated facial nerve delineation with a distance calculating neuronavigation interface results in a system that can adequately warn surgeons during temporal bone drilling, and effectively diminishes risks of iatrogenic facial nerve palsy.

  17. Fuel-Coolant Interactions: Visualization and Mixing Measurements

    International Nuclear Information System (INIS)

    Loewen, Eric P.; Bonazza, Riccardo; Corradini, Michael L.; Johannesen, Robert E.

    2002-01-01

    Dynamic X-ray imaging of fuel-coolant interactions (FCI), including quantitative measurement of fuel-coolant volume fractions and length scales, has been accomplished with a novel imaging system at the Nuclear Safety Research Center at the University of Wisconsin, Madison. The imaging system consists of visible-light high-speed digital video, low-energy X-ray digital imaging, and high-energy X-ray digital imaging subsystems. The data provide information concerning the melt jet velocity, melt jet configuration, melt volume fractions, void fractions, and spatial and temporal quantification of premixing length scales for a model fuel-coolant system of molten lead poured into a water pool (fuel temperatures 500 to 1000 K; jet diameters 10 to 30 mm; coolant temperatures 20 to 90 deg. C). Overall results indicate that the FCI has three general regions of behavior, with the high fuel-coolant temperature region similar to what might be expected under severe accident conditions. It was observed that the melt jet leading edge has the highest void fraction and readily fragments into discrete masses, which then subsequently subdivide into smaller masses of length scales <10 mm. The intact jet penetrates <3 to 5 jet length/jet diameter before this breakup occurs into discrete masses, which continue to subdivide. Hydrodynamic instabilities can be visually identified at the leading edge and along the jet column with an interfacial region that consists of melt, vapor, and water. This interface region was observed to grow in size as the water pool temperature was increased, indicating mixing enhancement by boiling processes

  18. Laser-Driven Calorimetry Measurements of Petroleum and Biodiesel Fuels.

    Science.gov (United States)

    Presser, Cary; Nazarian, Ashot; Millo, Amit

    2018-02-01

    Thermochemical characteristics were determined for several National Institute of Standards and Technology standard-reference-material petroleum and biodiesel fuels, using a novel laser-heating calorimetry technique. Measurements focused on the sample thermal behavior, specific heat release rate, and total specific heat release. The experimental apparatus consists of a copper sphere-shaped reactor mounted within a chamber, along with laser-beam-steering optical components, gas-supply manifold, and a computer-controlled data-acquisition system. At the center of the reactor, liquid sample is injected onto a copper pan substrate that rests and is in contact with a fine-wire thermocouple. A second thermocouple is in contact with the inner reactor sphere surface. The reactor is heated from opposing sides by a continuous-wave, near-infrared laser to achieve nearly uniform sample temperature. The change in temperature with time (thermogram) is recorded for both thermocouples, and compared to a baseline thermogram (without liquid in the pan). The thermograms are then processed (using an equation for thermal energy conservation) for the thermochemical information of interest. The results indicated that the energy reaching the pan is dominated by radiative heat transfer processes, while the dominant thermal process for the reactor sphere is the stored (internal) thermal energy within the sphere material. Sufficient laser power is necessary to detect the fuel thermal-related characteristics, and the required power can differ from one fuel to another. With sufficient laser power, one can detect the preferential vaporization of the lighter and heavier fuel fractions. The total specific heat release obtained for the different conventional and biodiesel fuels used in this investigation were similar to the expected values available in the literature.

  19. PIV measurement of internal structure of diesel fuel spray

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Z M [Ecotechnology System Lab., Yokohama National Univ. (Japan); Nishino, K [Div. of Artificial Environment and Systems, Yokohama National Univ. (Japan); Mizuno, S [Yokohama National Univ. (Japan); Torii, K [Dept. of Mechanical Engineering and Materials Science, Yokohama National Univ. (Japan)

    2000-12-01

    This paper reports particle image velocimetry (PIV) measurements of diesel fuel spray injected from a single-hole nozzle at injection pressures ranging from 30 to 70 MPa, which are comparable to partial-load operating conditions of commercial diesel engines. The fuel is injected into a non-combusting environment pressurized up to 2.0 MPa. A laser-induced fluorescent (LIF) technique is utilized to visualize internal structures of fuel sprays formed by densely-distributing droplets. A specially designed synchronization system is developed to acquire double-frame spray images at an arbitrary time delay after injection. A direct cross-correlation PIV technique is applied to measure instantaneous droplet velocity distribution. Unique large-scale structures in droplet concentration, called 'branch-like structures' by Azetsu et al. (1990), are observed and shown to be associated with active vortical motions, which appear to be responsible for the mixing between droplets and the surrounding gas. It is found that the droplets tend to move out of the vortical structures and accumulate in the regions of low vorticity. Some other interesting features concerning droplet velocity fields are also presented. (orig.)

  20. Ignition delay time measurements of primary reference fuel blends

    KAUST Repository

    Alabbad, Mohammed

    2017-02-07

    Ignition delay times of four different primary reference fuels (PRF), mixtures of n-heptane and iso-octane, were measured behind reflected shock waves in a high-pressure shock tube facility. The PRFs were formulated to match the RON of two high-octane gasolines (RON 95 and 91) and two prospective low-octane naphtha fuels (RON 80 and 70). Experiments were carried out over a wide range of temperatures (700–1200K), pressures (10, 20, and 40bar) and equivalence ratios (0.5 and 1). Kinetic modeling predictions from four chemical kinetic mechanisms are compared with the experimental data. Ignition delay correlations are developed to reproduce the measured ignition delay times. Brute force sensitivity analyses are carried out to identify reactions that affect ignition delay times at specific temperature, pressure and equivalence ratio. The large experimental data set provided in the current work will serve as a benchmark for the validation of chemical kinetic mechanisms of primary reference fuel blends.

  1. Ignition delay time measurements of primary reference fuel blends

    KAUST Repository

    Alabbad, Mohammed; Javed, Tamour; Khaled, Fathi; Badra, Jihad; Farooq, Aamir

    2017-01-01

    Ignition delay times of four different primary reference fuels (PRF), mixtures of n-heptane and iso-octane, were measured behind reflected shock waves in a high-pressure shock tube facility. The PRFs were formulated to match the RON of two high-octane gasolines (RON 95 and 91) and two prospective low-octane naphtha fuels (RON 80 and 70). Experiments were carried out over a wide range of temperatures (700–1200K), pressures (10, 20, and 40bar) and equivalence ratios (0.5 and 1). Kinetic modeling predictions from four chemical kinetic mechanisms are compared with the experimental data. Ignition delay correlations are developed to reproduce the measured ignition delay times. Brute force sensitivity analyses are carried out to identify reactions that affect ignition delay times at specific temperature, pressure and equivalence ratio. The large experimental data set provided in the current work will serve as a benchmark for the validation of chemical kinetic mechanisms of primary reference fuel blends.

  2. Measuring heat transfer through TR-0 reactor fuel element

    International Nuclear Information System (INIS)

    Nemec, V.; Turzik, Z.; Vitek, M.

    1977-05-01

    The time course of temperatures of the peripheral and the central fuel pins of the TR-O reactor was studied during moderator temperature changes using a model. The formula T=Tsub(e)+(Tsub(o)-Tsub(e)).exp(-t/tsub(e)) applies, where T is the pin temperature, Tsub(o) the initial pin temperature, Tsub(e) is the steady-state bath temperature, tsub(e) the time constant of temperature equilibration and t the time required for a temperature change from value Tsub(o) to T. For the bath level height H=1 m the tsub(e) value for the central pin was determined to be 1.05 hours, for the peripheral pin 0.96 hour; for level height H=2 m the values were 2.1 and 2.12 hours, respectively. The dependence found will allow correcting the experimental results in measurements with heated moderator for fuel temperature changes. (Ha)

  3. An environmental impact measure for nuclear fuel cycle evaluation

    International Nuclear Information System (INIS)

    Ahn, Joonhong

    2004-01-01

    Review of the models and measures for repository performance assessment has revealed that dedicated measures for environmental impacts need to be developed for the purpose of nuclear-fuel-cycle evaluation from the viewpoint of environmental impact minimization. The present study proposes the total toxicity index of released radionuclides that have accumulated in the region exterior to the repository as an environmental impact measure. The measure is quantitatively evaluated by a radionuclide transport model that incorporates the effects of canister-array configuration and the initial mass loading in the waste canister. With the measure, it is demonstrated that the environmental impact of the repository can be effectively reduced by reduction of the initial mass loading and change in the canister-array configuration in the repository. Environmental impacts of the mill tailings and the depleted uranium are as important as those from the high-level radioactive wastes repository. For a fair comparison of various fuel cycles, the sum of these impacts should be compared. (author)

  4. Judgement on the data for fuel assembly outlet temperatures of WWER fuel assemblies in power reactors based on measurements with experimental fuel assemblies

    International Nuclear Information System (INIS)

    Krause, F.

    1986-01-01

    In the period from 1980 to 1985, in the Rheinsberg nuclear power plant experimental fuel assemblies were used on lattices at the periphery of the core. These particular fuel assemblies dispose of an extensive in-core instrumentation with different sensors. Besides this, they are fit out with a device to systematically thottle the coolant flow. The large power gradient present at the core position of the experimental fuel assembly causes a temperature profile along the fuel assemblies which is well provable at the measuring points of the outlet temperature. Along the direction of flow this temperature profile in the coolant degrades only slowly. This effect is to be taken into account when measuring the fuel assembly outlet temperature of WWER fuel assemblies. Besides this, the results of the measurements hinted both at a γ-heating of the temperature measuring points and at tolerances in the calculation of the micro power density distribution. (author)

  5. Hydrocarbon Fuel Thermal Performance Modeling based on Systematic Measurement and Comprehensive Chromatographic Analysis

    Science.gov (United States)

    2016-07-31

    distribution unlimited Hydrocarbon Fuel Thermal Performance Modeling based on Systematic Measurement and Comprehensive Chromatographic Analysis Matthew...vital importance for hydrocarbon -fueled propulsion systems: fuel thermal performance as indicated by physical and chemical effects of cooling passage... analysis . The selection and acquisition of a set of chemically diverse fuels is pivotal for a successful outcome since test method validation and

  6. The measurement of neptunium in fast reactor fuel reprocessing

    International Nuclear Information System (INIS)

    Mair, M.A.; Savage, D.J.; Kyffin, T.W.

    1986-02-01

    Analytical techniques have been developed to measure neptunium in the feed, waste and product streams of a fast reactor fuel reprocessing plant. The estimated level of one microgram per milligram of plutonium in some solutions presented severe separation and measurement problems. An initial separation stage was essential, and both ion exchange and solvent extraction using thenoyltrifluoroacetone were studied. The redox chemistry of neptunium necessary to achieve good separation is considered. Spectrophotometry measurement of the stable neptunium/arsenazo III complex was selected for the final neptunium determination with additional analysis by radiometric methods. Incomplete recovery of neptunium during the separation stages necessitated yield measurements, using either neptunium-237 as an internal standard or the short lived gamma active neptunium-239 isotope as a tracer. The distribution of neptunium between the waste and product streams is discussed, in relation to the chemistry of neptunium in the reprocessing plant. (author)

  7. Measurement and characterization of fission products released from LWR fuel

    International Nuclear Information System (INIS)

    Osborne, M.F.; Collins, J.L.; Lorenz, R.A.; Norwood, K.S.; Strain, R.V.

    1984-01-01

    Samples of commercial LWR fuel have been heated under simulated accident conditions to determine the extent and the chemical forms of fission product release. This project was sponsored by the USNRC under a broad program of reactor safety studies. Of the five tests discussed, the fractional releases of Kr, I, and Cs varied from approx. 2% at 1400 0 C to >50% at 2000 0 C; much smaller fractions of Ru, Ag, Sb, and Te were measured in some tests. The major chemical forms in the effluent appeared to include CsI, CsOH, Sb, Te, and Ag

  8. Measurement and characterization of fission products released from LWR fuel

    International Nuclear Information System (INIS)

    Osborne, M.F.; Collins, J.L.; Lorenz, R.A.; Norwood, K.S.; Strain, R.V.

    1984-01-01

    Samples of commercial LWR fuel have been heated under simulated accident conditions to determine the extent and the chemical forms of fission product release. Of the five tests discussed, the fractional releases of Kr, I, and Cs varied from proportional 2% at 1400 0 C to >50% at 2000 0 C; much smaller fractions of Ru, Ag, Sb, and Te were measured in some tests. The major chemical forms in the effluent appeared to include CsI, CsOH, Sb, Te, and Ag. (orig./HP)

  9. Quality assurance measures for spent fuel shipping and storage containers

    International Nuclear Information System (INIS)

    Droste, B.; Roedel, R.

    1987-01-01

    Quality assurance measures are to be applied in production and operation to ensure the approved fuel-element-container design specifications. The authors concentrate on the official regulations pertaining to the application of a quality assurance system, on the compliance with design specifications ensured by certified manufacturing tests and in-service inspections. For nodular-cast-iron container bodies, the authors demonstrate the procedure by presenting the contents of the materials data sheet characterizing the material, and the production and test sequence plan for container casting. In addition, they state the quality assurance requirements for interim-storage containers which transgress those stipulated for shipping containers. (orig.) [de

  10. Heat transfer and pressure measurements for the SSME fuel turbine

    Science.gov (United States)

    Dunn, Michael G.; Kim, Jungho

    1991-01-01

    A measurement program is underway using the Rocketdyne two-stage Space Shuttle Main Engine (SSME) fuel turbine. The measurements use a very large shock tunnel to produce a short-duration source of heated and pressurized gas which is subsequently passed through the turbine. Within this environment, the turbine is operated at the design values of flow function, stage pressure ratio, stage temperature ratio, and corrected speed. The first stage vane row and the first stage blade row are instrumented in both the spanwise and chordwise directions with pressure transducers and heat flux gages. The specific measurements to be taken include time averaged surface pressure and heat flux distributions on the vane and blade, flow passage static pressure, flow passage total pressure and total temperature distributions, and phase resolved surface pressure and heat flux on the blade.

  11. Evaluation of Instrumentation for Measuring Undissolved Water in Aviation Turbine Fuels per ASTM D3240

    Science.gov (United States)

    2015-11-05

    Undissolved Water in Aviation Turbine Fuels per ASTM D3240 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) Joel Schmitigal... water ) in Aviation Turbine Fuels per ASTM D3240 15. SUBJECT TERMS fuel, JP-8, aviation fuel, contamination, free water , undissolved water , Aqua-Glo 16...Michigan 48397-5000 Evaluation of Instrumentation for Measuring Undissolved Water in Aviation Turbine Fuels per ASTM D3240 Joel Schmitigal Force

  12. Measurement and behaviour of technetium in fast reactor fuel reprocessing

    International Nuclear Information System (INIS)

    Ferguson, C.; Kyffin, T.W.

    1986-02-01

    A method is described for the spectrophotometric measurement of technetium in plant solutions from the reprocessing of fast reactor fuel. The technetium is selectively extracted using tri-iso-octylamine. After back extraction, thiocyanate is added, in the presence of tetrabutyl-ammonium hydroxide, to form the red hexa-thiocyanato anionic complex in a chloroform medium. The concentration of the technetium is then calculated from the spectrophotometric measurement of this complex. This method was applied to bulk samples, collected during a PFR fuel reprocessing campaign, to identify the main routes followed by technetium through the reprocessing plant. In order to understand the probable behaviour of technetium in the process plant streams, an investigation into the influence of plutonium IV nitrate on the extraction of Tc (VII) into 20%v/v tributyl phosphate/odourless kerosene solution from nitric acid solutions, was initiated. The results of this investigation, along with the known distribution coefficient for the extraction of the uranyl/technetium complex U0 2 (N0 3 )(Tc0 4 ).2TBP and the redox chemistry of technetium, are used to predict the probable behaviour of technetium in the process plant streams. This predicted behaviour is compared with the experimental results and reasonable agreement is obtained between experiment and theory, considering the history of the samples analysed. (author)

  13. Measuring the energy security implications of fossil fuel resource concentration

    International Nuclear Information System (INIS)

    Lefevre, Nicolas

    2010-01-01

    Economic assessments of the welfare effects of energy insecurity are typically uncertain and fail to provide clear guidance to policy makers. As a result, governments have had little analytical support to complement expert judgment in the assessment of energy security. This is likely to be inadequate when considering multiple policy goals, and in particular the intersections between energy security and climate change mitigation policies. This paper presents an alternative approach which focuses on gauging the causes of energy insecurity as a way to assist policy making. The paper focuses on the energy security implications of fossil fuel resource concentration and distinguishes between the price and physical availability components of energy insecurity. It defines two separate indexes: the energy security price index (ESPI), based on the measure of market concentration in competitive fossil fuel markets, and the energy security physical availability index (ESPAI), based on the measure of supply flexibility in regulated markets. The paper illustrates the application of ESPI and ESPAI with two case studies-France and the United Kingdom-looking at the evolution of both indexes to 2030.

  14. Measuring the energy security implications of fossil fuel resource concentration

    Energy Technology Data Exchange (ETDEWEB)

    Lefevre, Nicolas [Woodrow Wilson School of Public and International Affairs, Princeton University, New Jersey (United States)

    2010-04-15

    Economic assessments of the welfare effects of energy insecurity are typically uncertain and fail to provide clear guidance to policy makers. As a result, governments have had little analytical support to complement expert judgment in the assessment of energy security. This is likely to be inadequate when considering multiple policy goals, and in particular the intersections between energy security and climate change mitigation policies. This paper presents an alternative approach which focuses on gauging the causes of energy insecurity as a way to assist policy making. The paper focuses on the energy security implications of fossil fuel resource concentration and distinguishes between the price and physical availability components of energy insecurity. It defines two separate indexes: the energy security price index (ESPI), based on the measure of market concentration in competitive fossil fuel markets, and the energy security physical availability index (ESPAI), based on the measure of supply flexibility in regulated markets. The paper illustrates the application of ESPI and ESPAI with two case studies - France and the United Kingdom - looking at the evolution of both indexes to 2030. (author)

  15. Specific features of the WWER Uranium-Gadolinium fuel behavior at BOL

    International Nuclear Information System (INIS)

    Shcheglov, A.; Proselkov, V.; Volkov, B.

    2013-01-01

    The calculated-experimental analysis of the WWER fuel behavior with 5%wt of gadolinium oxide at the beginning of life (BOL) is presented. The results are based on the data on fuel centerline temperature measurements, gas media pressure inside the cladding and fuel elongation obtained during irradiation of the test fuel rods in HBWR (Halden). Computer analysis of experimental data is performed with TOPRA-2, version 2 code. It is shown that specific features of the uranium-gadolinium fuel behavior at the early of life is due to presence of burnable absorber influencing the average linear heat rating, radial power distribution and lower thermal conductivity. In particular, the analysis of “late” relocation effect on the maximum Gd fuel temperature is presented. (authors)

  16. KMRR fuel design

    International Nuclear Information System (INIS)

    Son, D.S.; Sim, B.S.; Kim, T.R.; Hwang, W.; Kim, B.G.; Ku, Y.H.; Lee, C.B.; Lim, I.C.

    1992-06-01

    KMRR fuel rod design criteria on fuel swelling, blistering and oxide spallation have been reexamined. Fuel centerline temperature limit of 250deg C in normal operation condition and fuel swelling limit of 12 % at the end of life have been proposed to prevent fuel failure due to excessive fuel swelling. Fuel temperature limit of 485deg C has been proposed to exclude the possibility of fuel failures during transients or under accident condition. Further analyses are needed to decide the fuel cladding temperature limit to preclude the oxide spallation. Design changes in fuel assembly structure and their effects on related systems have been reviewed from a structural integrity viewpoint. The remained works in fuel mechanical design area have been identified and further efforts of fuel design group will be focused on these aspects. (Author)

  17. Laser induced fluorescence measurements of the mixing of fuel oil with air

    Energy Technology Data Exchange (ETDEWEB)

    Arnold, A; Bombach, R; Hubschmid, W; Kaeppeli, B [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1999-08-01

    We report on measurements of the mixing of fuel oil with air at atmospheric pressure in an industrial premixed gas turbine burner. The concentration of the vaporized fuel oil was measured with laser induced fluorescence. We reason that the fuel oil concentration can be considered with good accuracy as proportional to the fluorescence intensity. (author) 6 fig., 3 refs.

  18. Measurement and analysis of transient vaporization in oxide fuel materials

    International Nuclear Information System (INIS)

    Benson, D.A.; Bergeron, E.G.

    1979-01-01

    This paper describes a series of experiments in which samples are heated to produce high vapor pressure states in times of 10 -6 to 10 -3 seconds. Experimental measurements of vapor pressures over fresh UO 2 from the pulsed electron beam and pulsed reactor heating tests are presented and compared with other high temperature data. The interpretation of the vapor pressure measured in the tests is discussed in detail. Effects of original sample stoichiometry, chemical interactions with the container and non-equilibrium evaporation due to induced temperature gradients are discussed. Special attention is given to dynamic behavior in rapid heating and vaporization of the oxide due to chemical non-equilibrium. Finally, similar projected reactor experiments on irradiated fuel are described and vapor pressure predictions made using available equilibrium models. A discussion of information accessible from such future tests and its importance is presented. (orig.) [de

  19. Measurement and analysis of transient vaporization in oxide fuel materials

    International Nuclear Information System (INIS)

    Gorham-Bergeron, E.; Benson, D.A.

    1978-01-01

    A series of experiments is described in which samples are heated to produce high vapor pressure states in times of 10 -6 to 10 -3 seconds. Experimental measurements of vapor pressures over fresh UO 2 from the pulsed electron beam and pulsed reactor heating tests are presented and compared with other high temperature data. The interpretation of the vapor pressures measured in the tests is discussed in detail. Effects of original sample stoichiometry, chemical interactions with the container and non-equilibrium evaporation due to induced temperature gradients are discussed. Special attention is given to dynamic behavior in rapid heating and vaporization of the oxide due to chemical nonequilibrium. Finally, similar projected reactor experiments on irradiated fuel are described and vapor pressure predictions made using available equilibrium models. A discussion of information accessible from such future tests and its importance is presented

  20. An approach to localize the retinal blood vessels using bit planes and centerline detection.

    Science.gov (United States)

    Fraz, M M; Barman, S A; Remagnino, P; Hoppe, A; Basit, A; Uyyanonvara, B; Rudnicka, A R; Owen, C G

    2012-11-01

    The change in morphology, diameter, branching pattern or tortuosity of retinal blood vessels is an important indicator of various clinical disorders of the eye and the body. This paper reports an automated method for segmentation of blood vessels in retinal images. A unique combination of techniques for vessel centerlines detection and morphological bit plane slicing is presented to extract the blood vessel tree from the retinal images. The centerlines are extracted by using the first order derivative of a Gaussian filter in four orientations and then evaluation of derivative signs and average derivative values is performed. Mathematical morphology has emerged as a proficient technique for quantifying the blood vessels in the retina. The shape and orientation map of blood vessels is obtained by applying a multidirectional morphological top-hat operator with a linear structuring element followed by bit plane slicing of the vessel enhanced grayscale image. The centerlines are combined with these maps to obtain the segmented vessel tree. The methodology is tested on three publicly available databases DRIVE, STARE and MESSIDOR. The results demonstrate that the performance of the proposed algorithm is comparable with state of the art techniques in terms of accuracy, sensitivity and specificity. Copyright © 2011 Elsevier Ireland Ltd. All rights reserved.

  1. A research of road centerline extraction algorithm from high resolution remote sensing images

    Science.gov (United States)

    Zhang, Yushan; Xu, Tingfa

    2017-09-01

    Satellite remote sensing technology has become one of the most effective methods for land surface monitoring in recent years, due to its advantages such as short period, large scale and rich information. Meanwhile, road extraction is an important field in the applications of high resolution remote sensing images. An intelligent and automatic road extraction algorithm with high precision has great significance for transportation, road network updating and urban planning. The fuzzy c-means (FCM) clustering segmentation algorithms have been used in road extraction, but the traditional algorithms did not consider spatial information. An improved fuzzy C-means clustering algorithm combined with spatial information (SFCM) is proposed in this paper, which is proved to be effective for noisy image segmentation. Firstly, the image is segmented using the SFCM. Secondly, the segmentation result is processed by mathematical morphology to remover the joint region. Thirdly, the road centerlines are extracted by morphology thinning and burr trimming. The average integrity of the centerline extraction algorithm is 97.98%, the average accuracy is 95.36% and the average quality is 93.59%. Experimental results show that the proposed method in this paper is effective for road centerline extraction.

  2. Experimental study on centerline velocities of a rectangular capture hood under realistic conditions.

    Science.gov (United States)

    He, Xinjian; Lewis, Braxton V; Guffey, Steven E

    2018-02-01

    Capture hoods are an important component of a local ventilation system designed to reduce exposures to airborne contaminants. The velocity at any point along the centerline of the hood (V x ) is currently estimated using one of many predictive equations developed since the 1930s. It is unproven that those predictive equations for V x are accurate, despite the prodigious number of studies concerning them. Among other issues, almost all experimental verifications were conducted for conditions that were either unrealistically ideal without competing air currents (e.g., zero cross draft) or were not described. This study measured values of V x along the midline using Particle Image Velocimetry (PIV) at distances of 1-14 inches in front of a rectangular capture hood. The experiments were conducted in a large wind tunnel (9' × 12' × 40', H × W × L) using a heated, breathing, anthropomorphically sized manikin. Three 0 degree draft velocities (V draft = 4, 14, and 50 ft/min) were tested, all directed toward the hood face and the back of the manikin (if present). For each value of V draft , the velocity fields were measured in a factorial design with and without the manikin, and with and without a worktable underneath the hood. An ideal condition was represented by a freestanding hood at the 4 fpm draft. Nonideal conditions included the presence of a worktable or manikin, and the combination of table and manikin. Each condition was tested at the three levels of V draft . The experimental results found significant effects (p < 0.001) for V draft , the presence of the manikin, the presence of the worktable, and all combinations of those factors. The effects of the independent variables were most pronounced at distances greater than 10 in (25.4 cm) from the hood face. It is concluded that none of the previously published models accurately predicted V x under the realistic conditions tested in this study. A satisfactory model will have to include terms for V draft and the

  3. Development of examination technique for oxide layer thickness measurement of irradiated fuel rods

    International Nuclear Information System (INIS)

    Koo, D. S.; Park, S. W.; Kim, J. H.; Seo, H. S.; Min, D. K.; Kim, E. K.; Chun, Y. B.; Bang, K. S.

    1999-06-01

    Technique for oxide layer thickness measurement of irradiated fuel rods was developed to measure oxide layer thickness and study characteristic of fuel rods. Oxide layer thickness of irradiated fuels were measured, analyzed. Outer oxide layer thickness of 3 cycle-irradiated fuel rods were 20 - 30 μm, inner oxide layer thickness 0 - 10 μm and inner oxide layer thickness on cracked cladding about 30 μm. Oxide layer thickness of 4 cycle-irradiated fuel rods were about 2 times as thick as those of 1 cycle-irradiated fuel rods. Oxide layer on lower region of irradiated fuel rods was thin and oxide layer from lower region to upper region indicated gradual increase in thickness. Oxide layer thickness from 2500 to 3000 mm showed maximum and oxide layer thickness from 3000 to top region of irradiated fuel rods showed decreasing trend. Inner oxide layer thicknesses of 4 cycle-irradiated fuel rod were about 8 μm at 750 - 3500 mm from the bottom end of fuel rod. Outer oxide layer thickness were about 8 μm at 750 - 1000 mm from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel. Oxide layer thickness technique will apply safety evaluation and study of reactor fuels. (author). 6 refs., 14 figs

  4. Velocity distribution measurement in wire-spaced fuel pin bundle

    International Nuclear Information System (INIS)

    Mizuta, Hiroshi; Ohtake, Toshihide; Uruwashi, Shinichi; Takahashi, Keiichi

    1974-01-01

    Flow distribution measurement was made in the subchannels of a pin bundle in air flow. The present paper is interim because the target of this work is the decision of temperature of the pin surface in contact with wire spacers. The wire-spaced fuel pin bundle used for the experiment consists of 37 simulated fuel pins of stainless steel tubes, 3000 mm in length and 31.6 mm in diameter, which are wound spirally with 6 mm stainless steel wire. The bundle is wrapped with a hexagonal tube, 3500 mm in length and 293 mm in flat-to-flat distance. The bundle is fixed with knock-bar at the entrance of air flow in the hexagonal tube. The pitch of pins in the bundle is 37.6 mm (P/D=1.19) and the wrapping pitch of wire is 1100 mm (H/D=34.8). A pair of arrow-type 5-hole Pitot tubes are used to measure the flow velocity and the direction of air flow in the pin bundle. The measurement of flow distribution was made with the conditions of air flow rate of 0.33 m 3 /sec, air temperature of 45 0 C, and average Reynolds number of 15100 (average air velocity of 20.6 m/sec.). It was found that circular flow existed in the down stream of wire spacers, that axial flow velocity was slower in the subchannels, which contained wire spacers, than in those not affected by the wire, and that the flow angle to the axial velocity at the boundary of subchannels was two thirds smaller than wire wrapping angle. (Tai, I.)

  5. Measuring method for heat-shrinkage of fuel pellet

    International Nuclear Information System (INIS)

    Komono, Akira; Ishizaki, Jin; Inaki, Kiyohiro.

    1997-01-01

    The present invention concerns a method of determining an amount of heat-shrinkage of UR 2 pellets containing gadolinium oxide (Gd 2 O 2 ) based on the difference of the density thereof before and after heating. In a heat shrinkage test of UO 2 pellets containing from 1.0 to 15.0% by weight of gadolinium oxide, the amount of heat-shrinkage is measured under the condition of heat-retaining temperature: from 1700 to 1750degC, temperature elevation time and lowering time: from 90 to 120mins, heat-retaining time: 24hours, inert gas atmosphere, gas pressure: 0.35kg/cm 2 and gas dew point: from -55 to 40degC without changing O/M. This invention has a feature in the use of the inert gas and the elevation of the dew point of the gas. Then, oxygen dissociation phenomenon from crystal lattices of the fuel pellets is suppressed, and normal densification value is shown. Then, fuel pellets of good quality with less fluctuation of the heat-shrinkage can be obtained. (N.H.)

  6. Property measurements and inner state estimation of simulated fuel debris

    Energy Technology Data Exchange (ETDEWEB)

    Hirooka, S.; Kato, M.; Morimoto, K.; Washiya, T. [Japan Atomic Energy Agency, Ibaraki (Japan)

    2014-07-01

    Fuel debris properties and inner state such as temperature profile were evaluated by using analysis of simulated fuel debris manufactured from UO{sub 2} and oxidized zircaloy. The center of the fuel debris was expected to be molten state soon after the melt down accident of LWRs because power density was very high. On the other hand, the surface of the fuel debris was cooled in the water. This large temperature gradient may cause inner stress and consequent cracks were expected. (author)

  7. MEASUREMENTS AND COMPUTATIONS OF FUEL DROPLET TRANSPORT IN TURBULENT FLOWS

    Energy Technology Data Exchange (ETDEWEB)

    Joseph Katz and Omar Knio

    2007-01-10

    The objective of this project is to study the dynamics of fuel droplets in turbulent water flows. The results are essential for development of models capable of predicting the dispersion of slightly light/heavy droplets in isotropic turbulence. Since we presently do not have any experimental data on turbulent diffusion of droplets, existing mixing models have no physical foundations. Such fundamental knowledge is essential for understanding/modeling the environmental problems associated with water-fuel mixing, and/or industrial processes involving mixing of immiscible fluids. The project has had experimental and numerical components: 1. The experimental part of the project has had two components. The first involves measurements of the lift and drag forces acting on a droplet being entrained by a vortex. The experiments and data analysis associated with this phase are still in progress, and the facility, constructed specifically for this project is described in Section 3. In the second and main part, measurements of fuel droplet dispersion rates have been performed in a special facility with controlled isotropic turbulence. As discussed in detail in Section 2, quantifying and modeling the of droplet dispersion rate requires measurements of their three dimensional trajectories in turbulent flows. To obtain the required data, we have introduced a new technique - high-speed, digital Holographic Particle Image Velocimetry (HPIV). The technique, experimental setup and results are presented in Section 2. Further information is available in Gopalan et al. (2005, 2006). 2. The objectives of the numerical part are: (1) to develop a computational code that combines DNS of isotropic turbulence with Lagrangian tracking of particles based on integration of a dynamical equation of motion that accounts for pressure, added mass, lift and drag forces, (2) to perform extensive computations of both buoyant (bubbles) and slightly buoyant (droplets) particles in turbulence conditions

  8. Development of the method for the dimensional measurement of the HANARO nuclear fuel

    International Nuclear Information System (INIS)

    Kim, Tae Yeon; Lee, K. S.; Park, D. G.; Choo, Y. S.; Ahn, S. B.

    1998-06-01

    Dimension of the nuclear fuel is altered in nuclear reactor because of the neutron exposure with high pressure water. If the deformation is overlarge, the severe problem in safety of the nuclear fuel and the reactor come about. Therefore the accurate dimensional data of the nuclear fuel in diameter and length is very important for the design of the nuclear fuel and the estimation of the nuclear safety. Measurement of diameter for the dummy HANARO fuel rod which has not filled with real fuel material was carried out in hot cell. And also the length of the HANARO fuel assembly and the rod are measured. Dimensional measuring method for the HANARO fuel was developed. The test result show our method is good enough to distinguish change in volume with statistical uncertainty of 0.6 %. (author). 2 refs., 7 tabs., 20 figs

  9. Method of performing shutdown reactivity measurements in spent nuclear fuel storage pools

    International Nuclear Information System (INIS)

    Levine, S.H.; Schultz, M.A.; Chang, D.

    1981-01-01

    The objective of this paper is to develop a device to measure the k/infinity/ of a spent fuel assembly used in light water reactors. A subcritical assembly having a cross configuration is designed to allow measurement of the k/sub //infinity/ of a spent fuel assembly by comparing the change in its multiplication with that of a fuel assembly of known k/infinity/. Calculations have been performed using nucleonic codes to develop polynomial equations that relate the k/infinity/ of the spent fuel assembly to measured data. The measurements involve taking count rates with the spent fuel assembly in the center position of the subcritical assembly, and the measured data are the count rate ratio of the spent fuel assembly over the count rate taken with a fuel assembly of known k/infinity/. The polynomial equations are easy to program on a microcomputer, which, together with the subcritical assembly, form the k/infinity/ meter. 9 refs

  10. Measurement of relative permeability of fuel cell diffusion media

    KAUST Repository

    Hussaini, I.S.

    2010-06-01

    Gas diffusion layer (GDL) in PEM fuel cells plays a pivotal role in water management. Modeling of liquid water transport through the GDL relies on knowledge of relative permeability functions in the in-plane and through-plane directions. In the present work, air and water relative permeabilities are experimentally determined as functions of saturation for typical GDL materials such as Toray-060, -090, -120 carbon paper and E-Tek carbon cloth materials in their plain, untreated forms. Saturation is measured using an ex situ gravimetric method. Absolute and relative permeability functions in the two directions of interest are presented and new correlations for in-plane relative permeability of water and air are established. © 2010 Elsevier B.V. All rights reserved.

  11. Parameters affecting level measurement interpretation of nuclear fuel solutions

    International Nuclear Information System (INIS)

    Hunt, B.A.; Landat, D.A.

    1999-01-01

    This paper describes a level measurement technique commonly used in the measurement of radioactive liquids and equipment utilised by the inspectors for safeguards purposes. Some of the influencing parameters affecting the measurement results by this technique are characterised. An essential requisite for successful process operations in chemical facilities involving liquids generally require some physical measurements to be made in-line for both process and quality control in order to achieve the necessary final product specifications . In nuclear fuel reprocessing facilities, the same objectives apply coupled however with an additional requirement of achieving nuclear material accountancy and control. In view of the strategic importance of some of the process vessels in nuclear facilities, accountancy has to be supported by volume and density measurements of low uncertainty. Inspectors therefore require instruments which are at the very least as good as or better than operator's equipment. The classical measurement technique and most widely applied for process liquids in nuclear installations is the bubbler probe or dip-tube technique. Here a regulated flow of air passes through tubes inserted to various depths into the vessel and pressure readings are measured which are a function of the presence of liquid height and density of solution in the tank. These readings, taken together with a pre-determined calibration curve are sufficient for the volume and amount of liquor in a tank to be quantified. All measurement equipment and instrumentation are long distances from the tank environment. The key physical parameter to measure at this location is therefore pressure. Equipment designed developed, commissioned and tested in the tank measurement facilities at Ispra and in nuclear installations in Europe, Japan and the USA, house digital pressure transducer modules with manufacture's declared features of better than 0.01% accuracy and long term stability of 0.01% full

  12. Fission gas and iodine release measured up to 15 GWd/t UO2 burnup

    International Nuclear Information System (INIS)

    Appelhans, A.D.

    1983-01-01

    A summary is presented of the measured release of xenon, krypton and iodine up to 15 GWd/t UO 2 burnup for fuel centerline temperatures ranging from 950 to 1800 K, at average linear heat ratings of 15 to 35 kW/m. The IFA-430 is composed of four 1.28-m-long fuel rods containing 10% enriched UO 2 pellet fuel. Two of the fuel rods are connected, top and bottom, to a gas flow system that permits the fission gases released from the fuel pellets to be swept out of the rods during irradiation and measured via gamma spectrometry. The release/burnup increased significantly between 10 and 15 GWd/t burnup. Fuel temperature did not change. Increased releases were due to physical changes in the fuel-surface area. Changes appeared to be due to higher power operation and burnup

  13. TRIGA fuel element burnup determination by measurement and calculation

    International Nuclear Information System (INIS)

    Zagar, T.; Ravnik, M.; Persic, A.; Jeraj, R.

    2000-01-01

    To estimate the accuracy of the fuel element burnup calculation different factors influencing the calculation were studied. To cover different aspects of burnup calculations, two in-house developed computer codes were used in calculations. The first (TRIGAP) is based on a one-dimensional two-group diffusion approximation, and the second (TRIGLAV) is based on a two-dimensional four-group diffusion equation. Both codes use WIMSD program with different libraries forunit-cell cross section data calculation. The burnup accumulated during the operating history of the TRIGA reactor at Josef Stefan Institute was calculated for all fuel elements. Elements used in the core during this period were standard SS 8.5% fuel elements, standard SS 12% fuel elements and highly enriched FLIP fuel elements. During the considerable period of operational history, FLIP and standard fuel elements were used simultaneously in mixed cores. (authors)

  14. Subcritical Measurements Research Program for Fresh and Spent Materials Test Reactor Fuels

    International Nuclear Information System (INIS)

    Blanchard, A.

    1999-01-01

    'A series of subcritical noise measurements were performed on fresh and spent University of Missouri Research Reactor fuel assemblies. These experimental measurements were performed for the purposes of providing benchmark quality data for validating transport theory computer codes and nuclear cross-section data used to perform criticality safety analyses for highly enriched, uranium-aluminum Material Test Reactor fuel assemblies. A mechanical test rig was designed and built to hold up to four fuel assemblies and neutron detectors in a subcritical array. The rig provided researchers with the ability to evaluate the reactivity effects of variable fuel/detector spacing, fuel rotation, and insertion of metal reflector plates into the lattice.'

  15. Study on Thermal Degradation Characteristics and Regression Rate Measurement of Paraffin-Based Fuel

    Directory of Open Access Journals (Sweden)

    Songqi Hu

    2015-09-01

    Full Text Available Paraffin fuel has been found to have a regression rate that is higher than conventional HTPB (hydroxyl-terminated polybutadiene fuel and, thus, presents itself as an ideal energy source for a hybrid rocket engine. The energy characteristics of paraffin-based fuel and HTPB fuel have been calculated by the method of minimum free energy. The thermal degradation characteristics were measured for paraffin, pretreated paraffin, HTPB and paraffin-based fuel in different working conditions by the using differential scanning calorimetry (DSC and a thermogravimetric analyzer (TGA. The regression rates of paraffin-based fuel and HTPB fuel were tested by a rectangular solid-gas hybrid engine. The research findings showed that: the specific impulse of paraffin-based fuel is almost the same as that of HTPB fuel; the decomposition temperature of pretreated paraffin is higher than that of the unprocessed paraffin, but lower than that of HTPB; with the increase of paraffin, the initial reaction exothermic peak of paraffin-based fuel is reached in advance, and the initial reaction heat release also increases; the regression rate of paraffin-based fuel is higher than the common HTPB fuel under the same conditions; with the increase of oxidizer mass flow rate, the regression rate of solid fuel increases accordingly for the same fuel formulation.

  16. Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad Abir; Fahima Islam; Hyoung Koo Lee; Daniel Wachs

    2014-11-01

    The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the High Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.

  17. Perform Thermodynamics Measurements on Fuel Cycle Case Study Systems

    Energy Technology Data Exchange (ETDEWEB)

    Martin, Leigh R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    This document was prepared to meet FCR&D level 3 milestone M3FT-14IN0304022, “Perform Thermodynamics Measurements on Fuel Cycle Case Study Systems.” This work was carried out under the auspices of the Thermodynamics and Kinetics FCR&D work package. This document reports preliminary work in support of determining the thermodynamic parameters for the ALSEP process. The ALSEP process is a mixed extractant system comprised of a cation exchanger 2-ethylhexyl-phosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) and a neutral solvating extractant N,N,N’,N’-tetraoctyldiglycolamide (TODGA). The extractant combination produces complex organic phase chemistry that is challenging for traditional measurement techniques. To neutralize the complexity, temperature dependent solvent extraction experiments were conducted with neat TODGA and scaled down concentrations of the ALSEP formulation to determine the enthalpies of extraction for the two conditions. A full set of thermodynamic data for Eu, Am, and Cm extraction by TODGA from 3.0 M HNO3 is reported. These data are compared to previous extraction results from a 1.0 M HNO3 aqueous medium, and a short discussion of the mixed HEH[EHP]/TODGA system results is offered.

  18. Manifestations of nonlinearity in fuel center thermocouple steady-state and transient data: implications for data analysis

    International Nuclear Information System (INIS)

    Lanning, D.D.; Barnes, B.O.; Williford, R.E.

    1979-01-01

    The interpretation and verification of fuel centerline thermocouple data are analyzed. Two new concepts are discussed along with their application to in-reactor data from IFA-432, a heavily instrumented six-rod Halden reactor test assembly sponsored by the Nuclear Regulatory Commission. The main ideas presented in this report are that: it is more useful to plot resistance versus power than simply to plot temperature versus power; and the response of the centerline temperature to a linear power decrease is correlated to the rod's current resistance-vs-power behavior. Thus, the resistance-vs-power measurement can be verified by performing a linear power decrease and by plotting the temperature response

  19. Methods for Automating Analysis of Glacier Morphology for Regional Modelling: Centerlines, Extensions, and Elevation Bands

    Science.gov (United States)

    Viger, R. J.; Van Beusekom, A. E.

    2016-12-01

    The treatment of glaciers in modeling requires information about their shape and extent. This presentation discusses new methods and their application in a new glacier-capable variant of the USGS PRMS model, a physically-based, spatially distributed daily time-step model designed to simulate the runoff and evolution of glaciers through time. In addition to developing parameters describing PRMS land surfaces (hydrologic response units, HRUs), several of the analyses and products are likely of interest to cryospheric science community in general. The first method is a (fully automated) variation of logic previously presented in the literature for definition of the glacier centerline. Given that the surface of a glacier might be convex, using traditional topographic analyses based on a DEM to trace a path down the glacier is not reliable. Instead a path is derived based on a cost function. Although only a single path is presented in our results, the method can be easily modified to delineate a branched network of centerlines for each glacier. The second method extends the glacier terminus downslope by an arbitrary distance, according to local surface topography. This product is can be used to explore possible, if unlikely, scenarios under which glacier area grows. More usefully, this method can be used to approximate glacier extents from previous years without needing historical imagery. The final method presents an approach for segmenting the glacier into altitude-based HRUs. Successful integration of this information with traditional approaches for discretizing the non-glacierized portions of a basin requires several additional steps. These include synthesizing the glacier centerline network with one developed with a traditional DEM analysis, ensuring that flow can be routed under and beyond glaciers to a basin outlet. Results are presented based on analysis of the Copper River Basin, Alaska.

  20. Using fine-scale fuel measurements to assess wildland fuels, potential fire behavior and hazard mitigation treatments in the southeastern USA

    Science.gov (United States)

    Roger D. Ottmar; John I. Blake; William T. Crolly

    2012-01-01

    The inherent spatial and temporal heterogeneity of fuel beds in forests of the southeastern United States may require fine scale fuel measurements for providing reliable fire hazard and fuel treatment effectiveness estimates. In a series of five papers, an intensive, fine scale fuel inventory from the Savanna River Site in the southeastern United States is used for...

  1. Simple approximation for estimating centerline gamma absorbed dose rates due to a continuous Gaussian plume

    International Nuclear Information System (INIS)

    Overcamp, T.J.; Fjeld, R.A.

    1987-01-01

    A simple approximation for estimating the centerline gamma absorbed dose rates due to a continuous Gaussian plume was developed. To simplify the integration of the dose integral, this approach makes use of the Gaussian cloud concentration distribution. The solution is expressed in terms of the I1 and I2 integrals which were developed for estimating long-term dose due to a sector-averaged Gaussian plume. Estimates of tissue absorbed dose rates for the new approach and for the uniform cloud model were compared to numerical integration of the dose integral over a Gaussian plume distribution

  2. Thulium oxide fuel characterization study (thulium-170 fueled capsule parametric design)

    Energy Technology Data Exchange (ETDEWEB)

    DesChamps, N.H.

    1968-10-01

    A doubly encapsulated thulia wafer, i.e., individually lined wafers stacked one upon another inside a fuel capsule was studied. The temperature profiles were determined for thulia power densities ranging from 8 to 24 W/cc and fuel capsule surface temperatures ranging from 1000/sup 0/F (538/sup 0/C) to 2000/sup 0/F (1093/sup 0/C). Parametric studies were also carried out on a singly encapsulated configuration in which the thulia wafers were stacked face to face in an infinitely long, lined cylinder. The doubly encapsulated wafer configuration yielded a lower centerline temperature than the singly encapsulated capsule. Only in extreme cases of a large wafer diameter in combination with a high thulia power density did the fuel capsule centerline temperature exceed the thulia melt temperature of 4172/sup 0/F (2300/sup 0/C). Results are also given for the maximum radius attainable without having centerline melting when using a thulia microsphere fuel form.

  3. Biomass burning fuel consumption rates: a field measurement database

    NARCIS (Netherlands)

    van Leeuwen, T.T.; van der Werf, G.R.; Hoffmann, A.A.; Detmers, R.G.; Ruecker, G.; French, N.H.F.; Archibald, S.; Carvalho Jr., J.A.; Cook, G.D.; de Groot, J.W.; Hely, C.; Kasischke, E.S.; Kloster, S.; McCarty, J.L.; Pettinari, M.L.; Savadogo, P.

    2014-01-01

    Landscape fires show large variability in the amount of biomass or fuel consumed per unit area burned. Fuel consumption (FC) depends on the biomass available to burn and the fraction of the biomass that is actually combusted, and can be combined with estimates of area burned to assess emissions.

  4. Concentration measurements of biodiesel in engine oil and in diesel fuel

    International Nuclear Information System (INIS)

    Mäder, A; Eskiner, M; Burger, C; Rossner, M; Krahl, J; Ruck, W

    2012-01-01

    This work comprised a method for concentration measurements of biodiesel in engine oil as well as biodiesel in diesel fuel by a measurement of the permittivity of the mixture at a frequency range from 100 Hz to 20 kHz. For this purpose a special designed measurement cell with high sensitivity was designed. The results for the concentration measurements of biodiesel in the engine oil and diesel fuel shows linearity to the measurement cell signal for the concentration of biodiesel in the engine oil between 0.5% Vol. to 10% Vol. and for biodiesel in the diesel fuel between 0% Vol. to 100% Vol. The method to measure the concentration of biodiesel in the engine oil or the concentration of biodiesel in the diesel fuel is very accurate and low concentration of about 0.5% Vol. biodiesel in engine oil or in diesel fuel can be measured with high accuracy.

  5. Ultrasonic measurement of high burn-up fuel elastic properties

    International Nuclear Information System (INIS)

    Laux, D.; Despaux, G.; Augereau, F.; Attal, J.; Gatt, J.; Basini, V.

    2006-01-01

    The ultrasonic method developed for the evaluation of high burn-up fuel elastic properties is presented hereafter. The objective of the method is to provide data for fuel thermo-mechanical calculation codes in order to improve industrial nuclear fuel and materials or to design new reactor components. The need for data is especially crucial for high burn-up fuel modelling for which the fuel mechanical properties are essential and for which a wide range of experiments in MTR reactors and high burn-up commercial reactor fuel examinations have been included in programmes worldwide. To contribute to the acquisition of this knowledge the LAIN activity is developing in two directions. First one is development of an ultrasonic focused technique adapted to active materials study. This technique was used few years ago in the EdF laboratory in Chinon to assess the ageing of materials under irradiation. It is now used in a hot cell at ITU Karlsruhe to determine the elastic moduli of high burnup fuels from 0 to 110 GWd/tU. Some of this work is presented here. The second on going programme is related to the qualification of acoustic sensors in nuclear environments, which is of a great interest for all the methods, which work, in a hostile nuclear environment

  6. Time-resolved measurements of supersonic fuel sprays using synchrotron x-rays

    International Nuclear Information System (INIS)

    Powell, C.F.; Yue, Y.; Poola, R.; Wang, J.

    2000-11-01

    A time-resolved radiographic technique has been developed for probing the fuel distribution close to the nozzle of a high-pressure single-hole diesel injector. The measurement was made using X-ray absorption of monochromatic synchrotron-generated radiation, allowing quantitative determination of the fuel distribution in this optically impenetrable region with a time resolution of better than 1 μs. These quantitative measurements constitute the most detailed near-nozzle study of a fuel spray to date

  7. Time-resolved measurements of supersonic fuel sprays using synchrotron X-rays.

    Science.gov (United States)

    Powell, C F; Yue, Y; Poola, R; Wang, J

    2000-11-01

    A time-resolved radiographic technique has been developed for probing the fuel distribution close to the nozzle of a high-pressure single-hole diesel injector. The measurement was made using X-ray absorption of monochromatic synchrotron-generated radiation, allowing quantitative determination of the fuel distribution in this optically impenetrable region with a time resolution of better than 1 micros. These quantitative measurements constitute the most detailed near-nozzle study of a fuel spray to date.

  8. In-pile instrumentation improvements for fuel irradiations in test reactor

    International Nuclear Information System (INIS)

    Blanc, J.Y.; Bernard, J.L.; Estrade, J.; Geoffroy, G.

    1996-01-01

    Knowledge of fuel limits and safety margins in normal and off-normal transients in nuclear power plants remains a constant preoccupation for electricity producers and fuel manufacturers. Accurate determination of such limits, through fuel irradiation testing in the OSIRIS reactor at Saclay is closely linked to the reliability of appropriate instrumentation techniques. Two paths are currently followed to obtain short experimental rods: segmented fuel coming directly from power plants, or re-fabrication of rods in hot cells with our FABRICE process. It can be associated with instrumentation such as fuel centerline thermocouple in annular pellets, pressure transducer or fission gas release measurement by gamma-spectrometry using helium sweeping, in analytic experiments. Our present development, to be implemented in 1993, is the the centerline instrumentation of a fuel column with solid pellets. Inserting the thermocouple requires a cold drilling machine, using CO 2 freezing of broken UO 2 (with liquid nitrogen). During the fuel rod irradiation itself, we try to lower the uncertainties associated to power determination, using thermal balance or neutronic calibration, or even gamma spectrometry. A description of the new test train designed for the ISABELLE water loop in OSIRIS is given, with special emphasis on instrumentation: a LVDT for measuring fuel rod elongation and eventual clad failure, and increased number and better localization of thermocouples and SPDN. The third part is devoted to the measurements by optical microdensitometry of neutron radiographs of the fuel pellet dish modification after irradiation. Dishes are generally disappearing through thermal and mechanical deformation of the pellet, and this can eventually be modelized to better understand pellet-cladding mechanical interaction. (author). 3 refs, 5 figs

  9. Fission gas and iodine release measured in IFA-430 up to 15 GWd/t UO2 burnup

    International Nuclear Information System (INIS)

    Appelhans, A.D.; Turnbull, J.A.; White, R.J.

    1983-01-01

    The release of fission products from fuel pellets to the fuel-cladding gap is dependent on the fuel temperature, the power (fission rate) and the burnup (fuel structure). As part of the US Nuclear Regulatory Commission's Fuel Behavior Program, EG and G Idaho, Inc., is conducting fission product release studies in the Heavy Boiling Water Reactor in Halden, Norway. This paper presents a summary of the results up to December, 1982. The data cover fuel centerline temperatures ranging from 700 to 1500 0 C for average linear heat ratings of 16 to 35 kW/m. The measurements have been performed for the period between 4.2 and 14.8 GWd/t UO 2 of burnup of the Instrumented Fuel Assembly 430 (IFA-430). The measurement program has been directed toward quantifying the release of the short-lived radioactive noble gases and iodines

  10. The prospect for fuel ion ratio measurements in ITER by collective Thomson scattering

    DEFF Research Database (Denmark)

    Stejner Pedersen, Morten; Korsholm, Søren Bang; Nielsen, Stefan Kragh

    2012-01-01

    We show that collective Thomson scattering (CTS) holds the potential to become a new diagnostic principle for measurements of the fuel ion ratio, nT/nD, in ITER. Fuel ion ratio measurements will be important for plasma control and machine protection in ITER. Measurements of ion cyclotron structures...... in CTS spectra have been suggested as the basis for a new fuel ion ratio diagnostic which would be well suited for reactor environments and capable of providing spatially resolved measurements in the plasma core. Such measurements were demonstrated in recent experiments in the TEXTOR tokamak. Here we...... conduct a sensitivity study to investigate the potential measurement accuracy of a CTS fuel ion ratio diagnostic on ITER. The study identifies regions of parameter space in which CTS can be expected to provide useful information on plasma composition, and we find that a CTS fuel ion ratio diagnostic could...

  11. Subcritical Noise Analysis Measurements with Fresh and Spent Research Reactor Fuels Elements

    International Nuclear Information System (INIS)

    Valentine, T.E.; Mihalczo, J.T.; Kryter, R.C.; Miller, V.C.

    1999-01-01

    The verification of the subcriticality is of utmost importance for the safe transportation and storage of nuclear reactor fuels. Transportation containers and storage facilities are designed such that nuclear fuels remain in a subcritical state. Such designs often involve excess conservatism because of the lack of relevant experimental data to verify the accuracy of Monte Carlo codes used in nuclear criticality safety analyses. A joint experimental research program between Oak Ridge National Laboratory, Westinghouse Safety Management Solutions, Inc., and the University of Missouri was initiated to obtain measured quantities that could be directly related to the subcriticality of simple arrays of Missouri University Research Reactor (MURR) fuel elements. A series of measurement were performed to assess the reactivity of materials such as BORAL, stainless steel, aluminum, and lead that are typically used in the construction of shipping casks. These materials were positioned between the fuel elements. In addition, a limited number of measurements were performed with configurations of fresh and spent (irradiated) fuel elements to ascertain the reactivity of the spent fuel elements. In these experiments, fresh fuel elements were replaced by spent fuel elements such that the subcritical reactivity change could be measured. The results of these measurements were used by Westinghouse Safety Management Solutions to determine the subcriticality of MURR fuel elements isolated by absorbing materials. The measurements were interpreted using the MCNP-DSP Monte Carlo code to obtain the subcritical neutron multiplication factor k(sub eff), and the bias in K(sub eff) that are used in criticality safety analyses

  12. Measurement of relative permeability of fuel cell diffusion media

    KAUST Repository

    Hussaini, I.S.; Wang, C.Y.

    2010-01-01

    Gas diffusion layer (GDL) in PEM fuel cells plays a pivotal role in water management. Modeling of liquid water transport through the GDL relies on knowledge of relative permeability functions in the in-plane and through-plane directions

  13. Alternative Fuel Vehicles: How Do They Really Measure Up?

    Science.gov (United States)

    similar to that of comparable gasoline vehicles. The publication is available in .pdf format through DOE's Alternative Fuels Data Center home page on the World Wide Web. For a hard copy, call the National Alternative

  14. In Vivo Calibration Measurements for Mixed Oxide Nuclear Reactor Fuel

    National Research Council Canada - National Science Library

    Shaw, Daniel

    2003-01-01

    .... Thus, the final isotopic composition of MOX depends upon the origin of the plutonium, how well chemical and radiological impurities are removed from the plutonium prior to fuel fabrication, the time...

  15. Technical measurement of small fission gas inventory in fuel rod with laser puncturing system

    International Nuclear Information System (INIS)

    Kim, Hee Moon; Kim, Sung Ryul; Lee, Byoung Oon; Yang, Yong Sik; Baek, Sang Ryul; Song, Ung Sup

    2012-01-01

    The fission gas release cause degradation of fuel rod. It influences fuel temperature and internal pressure due to low thermal conductivity. Therefore, fission gas released to internal void of fuel rod must be measured with burnup. To measure amount of fission gas, fuel rod must be punctured by a steel needle in a closed chamber. Ideal gas law(PV=nRT) is applied to obtain atomic concentration(mole). Steel needle type is good for large amount of fission gas such as commercial spent fuel rod. But, some cases with small fuel rig in research reactor for R/D program are not available to use needle type because of large chamber volume. The laser puncturing technique was developed to solve measurement of small amount of fission gas. This system was very rare equipment in other countries. Fine pressure gage and strong vacuum system were installed, and the chamber volume was reduced at least. Fiber laser was used for easy operation

  16. Development of CANDU fuel in-bay inspection and dimensional measurement system

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Chang Keun; Cho, Moon Sung; Suk, Ho Chun; Koo, Dae Seo; Park, Kwang June; Jun, Ji Su; Jung, Jong Yeob

    2004-09-01

    In this report, we describe a spent fuel inspection and dimensional measuring system for CANDU fuel bundles in atmosphere or water. The submissible camera with radiation resistance was used to inspect a spent fuel in water. The design criteria of the dimensional measuring system was set up {+-}0.01mm(10{mu}m) for measuring accuracy. The LVDT (Linear Variable Differential Transformer) was used as measuring sensor in this dimensional measuring system. An LVDT calibration equipment was developed in order to satisfy the required accuracy of the system. Also, aluminum master and CANDU Master with same dimension of fuel bundle was made to calibrate the mechanical error of the dimensional measuring system. The accuracy of the fuel inspection system was examined. The results show that the accuracy in dimensional measurement of fuel rod profile and bearing pad profile, diameter of fuel bundle, fuel rod length, and end plate profile using standard test equipment satisfies the design criteria, i.e., maximum measurement error of {+-}0.01mm(10{mu}m)

  17. Measurement and evaluation of fuels and technologies for passenger rail service in North Carolina.

    Science.gov (United States)

    2012-08-01

    The purpose of this project is to measure a baseline for fuel use and emission rates on the rebuilt or replaced engines on each locomotive in the NCDOT Rail Division fleet, using ultra-low sulfur diesel (ULSD) fuel; measure real-world, in-use over...

  18. Remotely sensed measurements of forest structure and fuel loads in the Pinelands of New Jersey

    Science.gov (United States)

    Nicholas Skowronski; Kenneth Clark; Ross Nelson; John Hom; Matt Patterson

    2007-01-01

    We used a single-beam, first return profiling LIDAR (Light Detection and Ranging) measurements of canopy height, intensive biometric measurements in plots, and Forest Inventory and Analysis (FIA) data to quantify forest structure and ladder fuels (defined as vertical fuel continuity between the understory and canopy) in the New Jersey Pinelands. The LIDAR data were...

  19. Estimation of centerline temperature of the waste form for the rare earth waste generated from pyrochemical process

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jung-Hoon, E-mail: mrchoijh@kaeri.re.kr; Eun, Hee-Chul; Lee, Tae-Kyo; Lee, Ki-Rak; Han, Seung-Youb; Jeon, Min-Ku; Park, Hwan-Seo; Ahn, Do-Hee

    2017-01-15

    Estimation of centerline temperature of nuclear glass waste form for each waste stream is very essential in the period of storage because the centerline temperature being over its glass transition temperature results in the increase of leaching rate of radioactive nuclides due to the devitrification of glass waste form. Here, to verify the effects of waste form diameter and transuranic element content in the rare earth waste on the centerline temperature of the waste form, the surrogate rare earth glass waste generated from pyrochemical process was immobilized with SiO{sub 2}−Al{sub 2}O{sub 3}−B{sub 2}O{sub 3} glass frit system, and thermal properties of the rare earth glass waste form were determined by thermomechanical analysis and thermal conductivity analysis. The estimation of centerline temperature was carried out using the experimental thermal data and steady-state conduction equation in a long and solid cylinder type waste form. It was revealed that thermal stability of waste form in case of 0.3 m diameter was not affected by the TRU content even in the case of 80% TRU recovery ratio in the electrowinning process, meaning that the waste form of 0.3 m diameter is thermally stable due to the low centerline temperature relative to its glass transition temperature of the rare earth glass waste form.

  20. An improved characterization method for international accountancy measurements of fresh and irradiated mixed oxide (MOX) fuel: helping achieve continual monitoring and safeguards through the fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Evans, Louise G [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Tobin, S. J. [Los Alamos National Laboratory; Menlove, H. O. [Los Alamos National Laboratory; Schear, M. A. [Los Alamos National Laboratory; Worrall, Andrew [U.K. NNL

    2011-01-13

    Nuclear fuel accountancy measurements are conducted at several points through the nuclear fuel cycle to ensure continuity of knowledge (CofK) of special nuclear material (SNM). Non-destructive assay (NDA) measurements are performed on fresh fuel (prior to irradiation in a reactor) and spent nuclear fuel (SNF) post-irradiation. We have developed a fuel assembly characterization system, based on the novel concept of 'neutron fingerprinting' with multiplicity signatures to ensure detailed CofK of nuclear fuel through the entire fuel cycle. The neutron fingerprint in this case is determined by the measurement of the various correlated neutron signatures, specific to fuel isotopic composition, and therefore offers greater sensitivity to variations in fissile content among fuel assemblies than other techniques such as gross neutron counting. This neutron fingerprint could be measured at the point of fuel dispatch (e.g. from a fuel fabrication plant prior to irradiation, or from a reactor site post-irradiation), monitored during transportation of the fuel assembly, and measured at a subsequent receiving site (e.g. at the reactor site prior to irradiation, or reprocessing facility post-irradiation); this would confirm that no unexpected changes to the fuel composition or amount have taken place during transportation and/ or reactor operations. Changes may indicate an attempt to divert material for example. Here, we present the current state of the practice of fuel measurements for both fresh mixed oxide (MOX) fuel and SNF (both MOX and uranium dioxide). This is presented in the framework of international safeguards perspectives from the US and UK. We also postulate as to how the neutron fingerprinting concept could lead to improved fuel characterization (both fresh MOX and SNF) resulting in: (a) assured CofK of fuel across the nuclear fuel cycle, (b) improved detection of SNM diversion, and (c) greater confidence in safeguards of SNF transportation.

  1. Measuring method for effective neutron multiplication factor upon containing irradiated fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Makoto; Mitsuhashi, Ishi; Sasaki, Tomoharu.

    1993-01-01

    A portion of irradiated fuel assemblies at a place where a reactivity effect is high, that is, at a place where neutron importance is high is replaced with standard fuel assemblies having a known composition to measure neutron fluxes at each of the places. An effective composition at the periphery of the standard fuel assemblies is determined by utilizing a calibration curve determined separately based on the composition and neutron flux values of the standard assemblies. By using the calibration curve determined separately based on this composition and the known composition of the standard fuel assemblies, an effective neutron multiplication factor for the fuel containing portion containing the irradiated fuel assemblies is recognized. Then, subcriticality is ensured and critical safety upon containing the fuel assemblies can be secured quantitatively. (N.H.)

  2. Temperature and neutron dose rate measurements at a spent fuel shipping cask

    International Nuclear Information System (INIS)

    Krause, F.

    1982-01-01

    Apart from some other requirements, spent fuel shipping casks have to ensure sufficient heat removal and radiation shielding. Results of temperature and neutron dose rate measurements at a spent fuel shipping cask are presented for different loading and heat removal by air. The measurements show that in shipping higher burnup fuel assemblies neutron radiation has to be taken into account when estimating the shielding of the shipping cask. On the other hand, unallowable high temperatures have been observed neither at the fuel assemblies nor at the shipping cask for a maximum heat output of Q <= 12 kW. (author)

  3. Synergistic smart fuel for in-pile nuclear reactor measurements

    Energy Technology Data Exchange (ETDEWEB)

    Smith, J.A.; Kotter, D.K. [Idaho National Laboratories, Idaho Falls (United States); Ali, R.A.; Garrett, S.L. [Penn State University, University Park, State College, PA 16801 (United States)

    2013-07-01

    The thermo-acoustic fuel rod sensor developed in this research has demonstrated a novel technique for monitoring the temperature within the core of a nuclear reactor or the temperature of the surrounding heat-transfer fluid. It uses the heat from the nuclear fuel to generate sustained acoustic oscillations whose frequency will be indicative of the temperature. Converting a nuclear fuel rod into this type of thermo-acoustic sensor simply requires the insertion of a porous material (stack). This sensor has demonstrated a synergy with the elevated temperatures that exist within the nuclear reactor using materials that have only minimal susceptibility to high-energy particle fluxes. When the sensor is in operation, the sound waves radiated from the fuel rod resonator will propagate through the surrounding cooling fluid. The frequency of these oscillations is directly correlated with an effective temperature within the fuel rod resonator. This device is self-powered and is operational even in case of total loss of power of the reactor.

  4. Robuste Verzweigungserkennung von Gefäßen in CTA-Datensätzen zur modellbasierten Extraktion der Centerline

    Science.gov (United States)

    Beck, Thomas; Fritz, Dominik; Biermann, Christina; Dillmann, Rüdiger

    Bei der Befundung und Visualisierung von Blutgefäßen ist deren Centerline von zentraler Bedeutung. Die Unterscheidung zwischen unverzweigten Abschnitten des Gefäßes und Verzweigungsbereichen ermöglicht den Einsatz spezialisierter und sehr effizienter Algorithmen zur modellbasierten Extraktion der Centerline. In diesem Artikel wird ein robustes Verfahren zur Verzweigungserkennung vorgestellt. Das Verfahren beruht auf einem Front-Propagation-Ansatz mit dynamisch angepassten Schwellwerten und einer anschließenden Clusteranalyse. Die vorgestellte Methode zur Verzweigungserkennung wurde als Komponente einer Architektur zur Extraktion der Centerline auf handannotierten Datensätzen getestet. Erste Ergebnisse sind sehr vielversprechend und ermöglichen auch bei pathologischen Gefäßen eine robuste Detektion von Gefäßverzweigungen.

  5. Measurement of dynamic interaction between a vibrating fuel element and its support

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, N.J.; Tromp, J.H.; Smith, B.A.W. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada). Chalk River Labs.

    1996-12-01

    Flow-induced vibration of CANDU{reg_sign} fuel can result in fretting damage of the fuel and its support. A WOrk-Rate Measuring Station (WORMS) was developed to measure the relative motion and contact forces between a vibrating fuel element and its support. The fixture consists of a small piece of support structure mounted on a micrometer stage. This arrangement permits position of the support relative to the fuel element to be controlled to within {+-} {micro}m. A piezoelectric triaxial load washer is positioned between the support and micrometer stage to measure contact forces, and a pair of miniature eddy-current displacement probes are mounted on the stage to measure fuel element-to-support relative motion. WORMS has been utilized to measure dynamic contact forces, relative displacements and work-rates between a vibrating fuel element and its support. For these tests, the fuel element was excited with broadband random force excitation to simulate flow-induced vibration due to axial flow. The relationship between fuel element-to-support gap or preload (i.e., interference or negative gap) and dynamic interaction (i.e., relative motion, contact forces and work-rates) was derived. These measurements confirmed numerical simulations of in-reactor interaction predicted earlier using the VIBIC code.

  6. In-pile measurement of the thermal conductivity of irradiated metallic fuel

    International Nuclear Information System (INIS)

    Bauer, T.H.; Holland, J.W.

    1995-01-01

    Transient test data and posttest measurements from recent in-pile overpower transient experiments are used for an in situ determination of metallic fuel thermal conductivity. For test pins that undergo melting but remain intact, a technique is described that relates fuel thermal conductivity to peak pin power during the transient and a posttest measured melt radius. Conductivity estimates and their uncertainty are made for a database of four irradiated Integral Fast Reactor-type metal fuel pins of relatively low burnup (<3 at.%). In the assessment of results, averages and trends of measured fuel thermal conductivity are correlated to local burnup. Emphasis is placed on the changes of conductivity that take place with burnup-induced swelling and sodium logging. Measurements are used to validate simple empirically based analytical models that describe thermal conductivity of porous media and that are recommended for general thermal analyses of irradiated metallic fuel

  7. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumentation and measurement techniques in fuel fabrication facilities

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-01-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period

  8. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumentation and measurement techniques in fuel fabrication facilities

    Energy Technology Data Exchange (ETDEWEB)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-01-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period.

  9. Development of nuclear fuel materials for research reactor

    International Nuclear Information System (INIS)

    Kim, Chang Kyu; Park, H. D.; Kim, K. H.; Lee, J. T.; Ryu, W. S.; Hwang, W.; Kim, H. N.; Kim, H. I.; Kwon, H. I.; Park, C.; Lee, B. C.; Park, J. M.; Lee, C. S.; Chae, H. T.; Im, N. J.; Cho, M. S.; Im, I. C.; Nam, C.; Lee, D. B.; Goh, Y. M.; Kim, J. D.; Ahn, H. S.; Woo, Y. M.; Chang, S. J.; Cho, H. D.

    1997-09-01

    This project has aimed at the development of U 3 Si dispersion fuel for the localization of HANARO fuel and the application of atomization process to advanced RERTR fuel development. The design criteria were established through the modified computer codes. Design documents were prepared and issued. The acceptable co-extrusion cladding was achieved. The electron beam welding technology has been developed and the sealing of the end plug and cladding was accomplished without defects. The atomization fuel meats have about 200% higher elongation and about 20% higher than comminution fuel meats. The thermal compatibility test showed that atomization fuel have about 30% higher stability that the comminution fuel. The pressure drops of 18 rods fuel assembly and 36 rods fuel assembly were measured to have 213 kPa and 205 kPa respectively. Apparent wear was not found in endurance test. The irradiation fuel was designed and fabricated by using low enriched uranium metal following the developed Q/A system. The safety analysis of irradiation fuel assembly was performed through linear power calculation by using MCNP4A code and centerline temperature calculation by using DIFAIR code. The quality assurance system has been established. The quality inspection technologies were developed. By acquiring the license, low enriched uranium of 100 kg as well as depleted uranium can be used. U 3 Si 2 -Al fuel swelled less than comminution fuel irrespective of temperature and fuel fraction in a compatibility test. The atomized U-10wt.%Mo powder were found to have gamma phase of isotropic structure. Gamma structure remained with a little swelling without any structure change at 400 deg C for 100 hours. Irradiation miniplate and test rig were designed preliminary manufactured. Thermal hydraulic and linear power calculations were performed by using PLTEMP and MCNP4A computer codes respectively. The hydraulic test showed that the pressure drop met the HANARO requirement. The vibration

  10. gamma-ray spectra measurements for long cooled MOX spent fuels

    International Nuclear Information System (INIS)

    Murakami, Kiyonobu; Kobayashi, Iwao

    1993-09-01

    Gamma-ray spectra of spent fuels have important informations in the estimation of burnup rate, concentration of fission products, cooling time and etc. which are required in the fuel loading control of reactors and special nuclear materials accountancy from the view point of safe guard. Although, some available data are given about uranium dioxide fuels, few data are given about uranium and plutonium dioxide mixtures (MOX fuels). Especially, there is few data about MOX fuels which are irradiated in thermal reactors and cooled more than ten years. Gamma-ray spectra are measured for PuO 2 -UO 2 fuel rods (IFA-159, IFA-160) which are irradiated at HBWR in Norway up to 9,420 and 5,340MWd/t respectively. Gamma-ray spectra had been measured about the two fuels ten years ago at the spent fuel pond of Japan Demonstration Reactor (JPDR). The objectives of this measurement is to know how decayed the gamma-ray spectra in these ten years and some fission products are there which are effective to estimate burnup rate of spent MOX fuels. (author)

  11. Contact-type displacement measuring mechanism for fuel assembly in reactor

    International Nuclear Information System (INIS)

    Yokota, Yoshio; Ko, Kuniaki.

    1995-01-01

    The measuring mechanism of the present invention, which is used in a lmfbr type reactor, is suspended by a gripper of a fuel handing machine, and it comprises a combination of a displacement amount measuring jig allowed to be inserted into a handling head of a fuel assembly and a displacement amount measuring ring disposed at the lower portion in the handling head. The displacement amount measuring jig has a structure comprising a releasable handle and a columnar or cylindrical measuring portion allowable to be inserted into the handling head formed at the lower portion of the handle, which are connected with each other. When an interference (contact) occurred between the displacement amount measuring jig and the stepwise displacement amount measuring ring during the measurement, change of load and a phenomenon that the fuel handing machine can not be lowered are recognized, so that core displacement amount can be recognized based on the stroke of the gripper portion. Then, remote measurement is possible for displacement and deformation of the fuel assembly in the reactor container, and the measurement can be conducted by the same procedures and in the same period of time as in a case of ordinary fuel exchange operation. A flow channel for coolants passing through the fuel assembly can be ensured, thereby enabling to measure the amount of core displacement which is closer to an actual value in the reactor. (N.H.)

  12. Measuring fuel moisture content in Alaska: standard methods and procedures.

    Science.gov (United States)

    Rodney A. Norum; Melanie. Miller

    1984-01-01

    Methods and procedures are given for collecting and processing living and dead plant materials for the purpose of determining their water content. Wild-land fuels in Alaska are emphasized, but the methodology is applicable elsewhere. Guides are given for determining the number of samples needed to attain a chosen precision. Detailed procedures are presented for...

  13. Clean Cities: AFLEET Measures Impacts of Vehicles and Fuels

    Energy Technology Data Exchange (ETDEWEB)

    None

    2016-12-01

    AFLEET is a free tool from the U.S. Department of Energy (DOE) that fleet managers can use to quantify the environmental and economic impacts of new fuels and vehicle technologies. The AFLEET factsheet explains how the tool works and how to access it.

  14. Automatic Centerline Extraction of Coverd Roads by Surrounding Objects from High Resolution Satellite Images

    Science.gov (United States)

    Kamangir, H.; Momeni, M.; Satari, M.

    2017-09-01

    This paper presents an automatic method to extract road centerline networks from high and very high resolution satellite images. The present paper addresses the automated extraction roads covered with multiple natural and artificial objects such as trees, vehicles and either shadows of buildings or trees. In order to have a precise road extraction, this method implements three stages including: classification of images based on maximum likelihood algorithm to categorize images into interested classes, modification process on classified images by connected component and morphological operators to extract pixels of desired objects by removing undesirable pixels of each class, and finally line extraction based on RANSAC algorithm. In order to evaluate performance of the proposed method, the generated results are compared with ground truth road map as a reference. The evaluation performance of the proposed method using representative test images show completeness values ranging between 77% and 93%.

  15. MTR fuel element burn-up measurements by the reactivity method

    International Nuclear Information System (INIS)

    Zuniga, A.; Cuya, T.R.; Ravnik, M.

    2003-01-01

    Fuel element burn-up was measured by the reactivity method in the 10 MW Peruvian MTR reactor RP-10. The main purpose of the experiment was testing the reactivity method for an MTR reactor as the reactivity method was originally developed for TRIGA reactors. The reactivity worth of each measured fuel element was measured in its original core position in order to measure the burn-up of the fuel elements that were part of the experimental core. The burn-up of each measured fuel element was derived by interpolating its reactivity worth from the reactivity worth of two reference fuel elements of known burn-up, whose reactivity worth was measured in the position of the measured fuel element. The accuracy of the method was improved by separating the reactivity effect of burn-up from the effect of the position in the core. The results of the experiment showed that the modified reactivity method for fuel element burn-up determination could be applied also to MTR reactors. (orig.)

  16. Measurement of the Velocity and Pressure Drop in a Tubular Type Fuel

    International Nuclear Information System (INIS)

    Jonghark Park; Heetaek Chae; Cheol Park; Heonil Kim

    2006-01-01

    We have developed a tubular type fuel assembly design as one of candidates for fuel to be used in the Advanced HANARO Reactor (AHR). The tubular type fuel has several merits over a rod type fuel with respect to the thermal-hydraulic and structural safety; the larger ratio of surface area to volume makes the surface temperature of a fuel element become lower, and curved plate is stronger against longitudinal bending and vibration. In the other side, a disadvantage is expected such that the flow velocity can be distributed unevenly channel by channel because the flow channels are isolated from each other in a tubular type fuel assembly. In addition to the design development, we also investigated the flow characteristics of the tubular fuel experimentally. To examine the flow velocity distribution and pressure drop, we made an experiment facility and a mockup of the tubular fuel assembly. The fuel assembly consists of 6 concentric fuel tubes so that 7 layers are made between fuel tubes. Since each layer is divided into three sections by stiffeners, 21 isolated flow channels are made in total. We employed pitot-tubes to measure the coolant velocity in each channel. The maximum velocity was measured as large as about 28% of the average velocity. It was observed in the innermost channel contrarily to the expectation from the hydraulic diameter. A change in the total flow rate did not affect the flow distribution. Meanwhile, the pressure drop was measured as about 70% of the drop in the rod type fuel assembly in use in HANARO. (authors)

  17. Threshold burnup for recrystallization and model for rim porosity in the high burnup UO2 fuel

    International Nuclear Information System (INIS)

    Lee, Byung Ho; Koo, Yang Hyun; Sohn, Dong Seong

    1998-01-01

    Applicability of the threshold burnup for rim formation was investigated as a function of temperature by Rest's model. The threshold burnup was the lowest in the intermediate temperature region, while on the other temperature regions the threshold burnup is higher. The rim porosity was predicted by the van der Waals equation based of the rim pore radius of 0.75μm and the overpressurization model on rim pores. The calculated centerline temperature is in good agreement with the measured temperature. However, more efforts seem to be necessary for the mechanistic model of the rim effect including rim growth with the fuel burnup

  18. Linear variable differential transformer and its uses for in-core fuel rod behavior measurements

    International Nuclear Information System (INIS)

    Wolf, J.R.

    1979-01-01

    The linear variable differential transformer (LVDT) is an electromechanical transducer which produces an ac voltage proportional to the displacement of a movable ferromagnetic core. When the core is connected to the cladding of a nuclear fuel rod, it is capable of producing extremely accurate measurements of fuel rod elongation caused by thermal expansion. The LVDT is used in the Thermal Fuels Behavior Program at the U.S. Idaho National Engineering Laboratory (INEL) for measurements of nuclear fuel rod elongation and as an indication of critical heat flux and the occurrence of departure from nucleate boiling. These types of measurements provide important information about the behavior of nuclear fuel rods under normal and abnormal operating conditions. The objective of the paper is to provide a complete account of recent advances made in LVDT design and experimental data from in-core nuclear reactor tests which use the LVDT

  19. Measurements relating fire radiative energy density and surface fuel consumption - RxCADRE 2011 and 2012

    Science.gov (United States)

    Andrew T. Hudak; Matthew B. Dickinson; Benjamin C. Bright; Robert L. Kremens; E. Louise Loudermilk; Joseph J. O' Brien; Benjamin S. Hornsby; Roger D. Ottmar

    2016-01-01

    Small-scale experiments have demonstrated that fire radiative energy is linearly related to fuel combusted but such a relationship has not been shown at the landscape level of prescribed fires. This paper presents field and remotely sensed measures of pre-fire fuel loads, consumption, fire radiative energy density (FRED) and fire radiative power flux density (FRFD),...

  20. Reactivity Measurements On Burnt And Reference Fuel Samples In LWR-PROTEUS Phase II

    International Nuclear Information System (INIS)

    Murphy, M.; Jatuff, F.; Grimm, P.; Seiler, R.; Luethi, A.; Van Geemert, R.; Brogli, R.; Chawla, R.; Meier, G.; Berger, H.-D.

    2003-01-01

    During the year 2002, the PROTEUS research reactor was used to make a series of reactivity measurements on Pressurised Water Reactor (PWR) burnt fuel samples, and on a series of specially prepared standards. These investigations have been made in two different neutron spectra. In addition, the intrinsic neutron emissions of the burnt fuel samples have been determined. (author)

  1. Comparison of Global Sizing Velocimetry and Phase Doppler Anemometry measurements of alternative jet fuel sprays

    Science.gov (United States)

    Sadr, Reza; Kannaiyan, Kumaran

    2013-11-01

    Atomization plays a crucial precursor role in liquid fuel combustion that directly affects the evaporation, mixing, and emission levels. Laser diagnostic techniques are often used to study the spray characteristics of liquid fuels. The objective of this work is to compare the spray measurements of Gas-to Liquid (GTL) jet fuels obtained using Global Sizing Velocimetry (GSV) and Phase Doppler Anemometry (PDA) techniques at global and local levels, respectively. The chemical and physical properties of GTL fuels are different from conventional jet fuels, owing to the difference in their production methodology. In this work, the experimental facility, the measurement techniques, and spray characteristics of two different GTL fuels are discussed and compared with those of Jet A-1 fuel. Results clearly demonstrate that although the global measurement gives an overall picture of the spray, fine details are obtained only through local measurements and complement in gaining more inferences into the spray characteristics. The results also show a close similarity in spray characteristics between GTL and Jet A-1 fuels. Funded by Qatar Science and Technology Park.

  2. Aluminum hydroxide coating thickness measurements and brushing tests on K West Basin fuel elements

    International Nuclear Information System (INIS)

    Pitner, A.L.

    1998-01-01

    Aluminum hydroxide coating thicknesses were measured on fuel elements stored in aluminum canisters in K West Basin using specially developed eddy current probes . The results were used to estimate coating inventories for MCO fuel,loading. Brushing tests successfully demonstrated the ability to remove the coating if deemed necessary prior to MCO loading

  3. Accuracy of dimension measurements from neutron radiographs of nuclear fuel pins

    International Nuclear Information System (INIS)

    Domanus, J.C.

    1976-01-01

    A comparison is given of accuracies obtained with measuring the dimensions (pellet diameter and fuel-clad gap) from neutron and X-ray radiographs of a calibrated nuclear fuel pin performed with a projection microscope, microdensitometers and a video micrometer

  4. Use of multi-functional flexible micro-sensors for in situ measurement of temperature, voltage and fuel flow in a proton exchange membrane fuel cell.

    Science.gov (United States)

    Lee, Chi-Yuan; Chan, Pin-Cheng; Lee, Chung-Ju

    2010-01-01

    Temperature, voltage and fuel flow distribution all contribute considerably to fuel cell performance. Conventional methods cannot accurately determine parameter changes inside a fuel cell. This investigation developed flexible and multi-functional micro sensors on a 40 μm-thick stainless steel foil substrate by using micro-electro-mechanical systems (MEMS) and embedded them in a proton exchange membrane fuel cell (PEMFC) to measure the temperature, voltage and flow. Users can monitor and control in situ the temperature, voltage and fuel flow distribution in the cell. Thereby, both fuel cell performance and lifetime can be increased.

  5. Construction and tests of a gamma device for experimental measurements of burnup of nuclear reactor fuel

    International Nuclear Information System (INIS)

    Brandao Junior, F.A.

    1982-01-01

    The gamma-scanning method is an important tool for the measurement of burnup of nuclear reactor fuel. The adequate knowledge of burnup allows for a better inventory of 'sensitive' fissile materials, better fuel management and provides insight on fuel behaviour and safety margins. This paper is related to the description, construction and operation of a first gamma scanning device, tested by irradiation of prototype PWR fuel pins, 14 cm long, in a Triga Mark-I reactor at very low power. Despite the limitations imposed by the low burnup, the experiment permitted a good checking of the main physical concepts and devices involved in the method. (Author) [pt

  6. Temperature measurements of the aluminium claddings of fuel elements in nuclear reactor

    International Nuclear Information System (INIS)

    Chen Daolong

    1986-01-01

    A method for embedding the sheathed thermocouples in the aluminium claddings of some fuel elements of experimental reactors by ultrasonic welding technique is described. The measurement results of the cladding temperature of fuel elements in reactors are given. By means of this method, the joint between the sheathed thermocouples and the cladding of fuel elements can be made very tight, there are no bulges on the cladding surfaces, and the sheathed thermocouples are embedded strongly and reliably. Therefore an essential means is provided for acquiring the stable and dynamic state data of the cladding temperature of in-core fuel elements

  7. The development of the neutron flux measurement technology using SPNDs during nuclear fuel irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B. G.; Kang, Y. H.; Cho, M. S.; Joo, K. N.; Choi, M. H.; Park, S. J.; Shin, Y. T.; Oh, J. M.; Kim, Y. J

    2004-03-01

    As a part of the development of instrumentation technologies for a nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), a study is performed to measure and evaluate the neutron flux at the same position as the nuclear fuel during irradiation test using the SPND(Self Powered Neutron Detector). To perform this study, rhodium type SPNDs and amplifier are selected suitable to irradiation test, and the selected SPNDs are installed in instrumented fuel capsule(02F-11K). The irradiation test using a instrumented fuel capsule are performed in the OR5 vertical hole of HANARO for about 54 days, and SPND output signals are acquired successfully during irradiation test. Acquired SPND signals are analyzed and evaluated as a reliable data by COSMOS Code. This will be utilized for the fuel related research together with fuel center temperature and reactor operation data.

  8. Fuel Rod Vibration Measurement Method using a Flap and its Verification

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Joo Young; Park, Nam Gyu; Suh, Jung Min; Jeon, Kyeong Lak [KEPCO NF Co., Daejeon (Korea, Republic of)

    2011-10-15

    Flow-induced vibration is a critical factor for the mechanical integrity of a fuel rod. This vibration can cause leaked fuel through the mechanism, such as grid to rod fretting. To minimize the failures caused by flow-induced vibration, a robust design is needed which takes into account vibrational characteristics. That is, the spacer grid design should be developed to avoid any excessive vibration. On the one hand, if fuel rod vibration can be measured, an estimation of the excitation forces, which are a critical cause of rod failure, should be possible. Therefore, by applying an external force, flow-induced vibration can be roughly estimated when the fuel rod vibration model is used. KEPCO Nuclear Fuel developed the test loop to research flow-induced vibration as shown in Fig.1. The investigation flow-induced vibration (INFINIT) - the test facility - can measure the grid strap vibration and pressure drop of a 5x5 small scale fuel bundle. Basically, using a Laser Doppler Vibrometer (LDV), the vibration of a structure immersed in high speed fluid can be measured. Grid strap vibration is easily measured using an LDV. However, it is quite difficult to measure fuel rod vibration because of the round surface shape of the rods. In addition, measuring current method using the LDV, it was only possible to directly measure fuel rod vibration at the first row of the bundle as the rods behind the first row are obscured. To solve this problem, a thin flap, as shown in Fig. 2(a) can be used as a reflecting target, gaining access to rods within the bundle. The flap is attached to the fuel rod, as in Fig. 2(b). As a result, most of the inner rod vibration can be measured. Before using a flap to measure fuel rod vibration, a verification process was needed to show whether the LDV signal from the flap vibration provided equivalent and reliable signals. Therefore, impact testing was carried out on the fuel rod using a flap. The LDV signals were then compared with accelerometer

  9. A STUDY OF THE DISCREPANCY BETWEEN FEDERAL AND STATE MEASUREMENTS OF ON-HIGHWAY FUEL CONSUMPTION

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, HL

    2003-08-11

    Annual highway fuel taxes are collected by the Treasury Department and placed in the Highway Trust Fund (HTF). There is, however, no direct connection between the taxes collected by the Treasury Department and the gallons of on-highway fuel use, which can lead to a discrepancy between these totals. This study was conducted to determine how much of a discrepancy exists between the total fuel usages estimated based on highway revenue funds as reported by the Treasury Department and the total fuel usages used in the apportionment of the HTF to the States. The analysis was conducted using data from Highway Statistics Tables MF-27 and FE-9 for the years 1991-2001. It was found that the overall discrepancy is relatively small, mostly within 5% difference. The amount of the discrepancy varies from year to year and varies among the three fuel types (gasoline, gasohol, special fuels). Several potential explanations for these discrepancies were identified, including issues on data, tax measurement, gallon measurement, HTF receipts, and timing. Data anomalies caused by outside forces, such as deferment of tax payments from one fiscal year to the next, can skew fuel tax data. Fuel tax evasion can lead to differences between actual fuel use and fuel taxes collected. Furthermore, differences in data collection and reporting among States can impact fuel use data. Refunds, credits, and transfers from the HTF can impact the total fuel tax receipt data. Timing issues, such as calendar year vs. fiscal year, can also cause some discrepancy between the two data sources.

  10. Improved HOR fuel management by flux measurement data feedback

    Energy Technology Data Exchange (ETDEWEB)

    Serov, I.V.; Leege, P.F.A. de; Hoogenboom, J.E.; Gibcus, H.P.M. [Delft University of Technology, Reactor Physics Dep., Interfaculty Reactor Inst., Delft (Netherlands)

    1997-07-01

    Flux distribution in a nuclear reactor can be obtained by utilizing different calculational and experimental methods. The obtained flux distributions are associated with uncertainties and therefore always differ from each other. By combining information from the calculation and experiment using the confluence method, it is possible to obtain a more reliable estimate of the flux distribution than exhibited by the calculation or experiment separately. As a feedback, the fuel burnup distribution, which is used as initial data to the calculation can be improved as well. The confluence method is applied to improvement of the burnup distribution estimates for the HOR research reactor of the Delft University of Technology. An integrated code system CONHOR is developed to match the CITATION results of in-core foil activation rate calculations with in-core experimental data through confluence. The system forms the basis for the advanced fuel management of the reactor. (author)

  11. Improved HOR fuel management by flux measurement data feedback

    Energy Technology Data Exchange (ETDEWEB)

    Serov, I.V.; Leege, P.F.A. de; Hoogenboom, J.E.; Gibcus, H.P.M. [Delft University of Technology, Reactor Physics Dep., Interfaculty Reactor Inst., Delft (Netherlands)

    1997-07-01

    Flux distribution in a nuclear reactor can be obtained by utilizing different calculational and experimental methods. The obtained flux distributions are associated with uncertainties and therefore always differ from each other. By combining information from the calculation and experiment using the confluence method, it is possible to obtain a more reliable estimate of the flux distribution than exhibited by the calculation or experiment separately. As a feedback, the fuel burnup distribution, which is used as initial data to the calculation can be improved as well. The confluence method is applied to improvement of the burnup distribution estimates for the HOR research reactor of the Delft University of Technology. An integrated code system CONHOR is developed to match the CITATION results of in-core foil activation rate calculations with in-core experimental data through confluence. The system forms the basis for the advanced fuel management of the reactor. (author) 1 fig., 8 refs.

  12. Improved HOR fuel management by flux measurement data feedback

    International Nuclear Information System (INIS)

    Serov, I.V.; Leege, P.F.A. de; Hoogenboom, J.E.; Gibcus, H.P.M.

    1997-01-01

    Flux distribution in a nuclear reactor can be obtained by utilizing different calculational and experimental methods. The obtained flux distributions are associated with uncertainties and therefore always differ from each other. By combining information from the calculation and experiment using the confluence method, it is possible to obtain a more reliable estimate of the flux distribution than exhibited by the calculation or experiment separately. As a feedback, the fuel burnup distribution, which is used as initial data to the calculation can be improved as well. The confluence method is applied to improvement of the burnup distribution estimates for the HOR research reactor of the Delft University of Technology. An integrated code system CONHOR is developed to match the CITATION results of in-core foil activation rate calculations with in-core experimental data through confluence. The system forms the basis for the advanced fuel management of the reactor. (author)

  13. Measurement of fission product release during LWR fuel failure

    International Nuclear Information System (INIS)

    Osetek, D.J.; King, J.J.

    1979-01-01

    The PBF is a specialized test reactor consisting of an annular core and a central test space 21 cm in diameter and 91 cm high. A test loop circulates coolant through the central experimental section at typical power reactor conditions. Light-water-reactor-type fuel rods are exposed to power bursts simulating reactivity insertion transients, and to power-cooling-mismatch conditions during which the rods are allowed to operate in film boiling. Fission product concentrations in the test loop coolant are continuously monitored during these transients by a Ge(Li) detector based gamma spectrometer. Automatic batch processing of pulse height spectra results in a list of radionuclide concentrations present in the loop coolant as a function of time during the test. Fission product behavior is then correlated to test parameters and posttest examination of the fuel rods. Data are presented from Test PCM-1

  14. Policy measures to increase the competitiveness of biodiesel fuel

    International Nuclear Information System (INIS)

    Assink, R.A.J.; Kerkhof, F.P.J.M.; Das, A.

    1993-01-01

    As a transport fuel of agricultural origin, biodiesel, which may be produced by means of a simple process from any available vegetable oil, is gathering more and more interest. Biodiesel is a mixture of methylesters of linear carbonic acids, which may be combusted in existing diesel engines. In this article the cultivation of the agricultural raw material and the fuel characteristics of biodiesel from rapeseed oil are elucidated. Also attention is paid to technological backgrounds and economical aspects of biodiesel production. At a rapeseed oil price of 750 Dutch guilders per ton, the cost price of biodiesel is 0.90 Dutch guilders per liter. Commercial demand can be created at an 85% reduction of the usual excises and levies. 9 figs., 4 tabs., 9 refs

  15. MOX fuel irradiation behavior in steady state (irradiation test in HBWR)

    Energy Technology Data Exchange (ETDEWEB)

    Kohno, S; Kamimura, K [Power Reactor and Nuclear Fuel Development Corp., Naka, Ibaraki (Japan)

    1997-08-01

    Two rigs of plutonium-uranium oxide (MOX) fuel rods have been irradiated in Halden boiling water reactor (HBWR) to investigate high burnup MOX fuel behavior for thermal reactor. The objective of irradiation tests is to investigate fuel behavior as influenced by pellet shape, pellet surface treatment, pellet-cladding gap size and MOX fuel powder preparations process. The two rigs have instrumentations for in-pile measurements of the fuel center-line temperature, plenum pressure, cladding elongation and fuel stack length change. The data, taken through in-operation instrumentation, have been analysed and compared with those from post-irradiation examination. The following observations are made: 1) PNC MOX fuels have achieved high burn-up as 59GWd/tMOX (67GWd/tM) at pellet peak without failure; 2) there was no significant difference in fission gas release fraction between PNC MOX fuels and UO{sub 2} fuels; 3) fission gas release from the co-converted fuel was lower than that from the mechanically blended fuel; 4) gap conductance was evaluated to decrease gradually with burn-up and to get stable in high burn-up region. 5) no evident difference of onset LHR for PCMI in experimental parameters (pellet shape and pellet-cladding gap size) was observed, but it decreased with burn-up. (author). 13 refs, 15 figs, 3 tabs.

  16. Development method for measuring thickness of nuclei and coating of fuel plates

    International Nuclear Information System (INIS)

    Borges Junior, Reinaldo

    2013-01-01

    One of the most important components of a nuclear reactor is the Nuclear Fuel. Currently, the most advanced commercial fuel, whose applicability in Brazilian reactors has been developed by IPEN since 1985, is the silicide U 3 Si 2 . This is formed by fuel plates with nuclei dispersion (where the fissile material (U 3 Si 2 ) is homogeneously dispersed in a matrix of aluminum) coated aluminum. This fuel is produced in Brazil with developed technology, the result of the efforts made by the group of manufacturing nuclear fuel (CCN - Center of Nuclear Fuel) of IPEN. Considering the necessity of increasing the power of the IEA- R1 and Brazilian Multipurpose Reactor Building (RMB), for the production of radioisotopes - mainly for the area of medicine - there will be significant increase in the production of nuclear fuel at IPEN. Given this situation, if necessary, make the development of more modern and automated classification techniques. Aiming at this goal, this work developed a new computational method for measuring thickness of core and cladding of fuel plates, which are able to perform such measurements in less time and with more meaningful statistical data when compared with the current method of measurement. (author)

  17. Burn-up measurements of spent fuel using gamma spectrometry technique

    International Nuclear Information System (INIS)

    Pereda, C.; Henriquez, C.; Klein, J.; Medel, J.

    2005-01-01

    Burn-up results obtained for HEU (45% of 235 U) fuel assemblies of the RECH-1 Research Reactor using gamma spectrometry technique are presented. The spectra were got from an in-pool facility built in the reactor to be mainly used to measure the burnup of irradiated fuel assemblies with short cooling time, where 95 Zr is being evaluated as possible fission monitor. A program to measure all spent fuel assemblies of the RECH-1 reactor was initiated in the frame of the Regional Project RLA/4/018: 'Management of Spent Fuel from Research Reactors'. The results presented here were obtained from HEU spent fuel assemblies with cooling time greater than 100 days and 137 Cs was used as fission monitor. The efficiency of the in-pool system was determined using a slightly burnt experimental fuel assembly, which has one fuel plate (one of the outer plates) and the rest are dummy plates. An average burn-up of 2.8% of 235 U was previously measured for the experimental fuel assembly utilizing a facility installed in a hot cell and 137 Cs was used as monitor. (author)

  18. Experimental Measurement and Numerical Modeling of the Effective Thermal Conductivity of TRISO Fuel Compacts

    International Nuclear Information System (INIS)

    Folsom, Charles

    2015-01-01

    Accurate modeling capability of thermal conductivity of tristructural-isotropic (TRISO) fuel compacts is important to fuel performance modeling and safety of Generation IV reactors. To date, the effective thermal conductivity (ETC) of tristructural-isotropic (TRISO) fuel compacts has not been measured directly. The composite fuel is a complicated structure comprised of layered particles in a graphite matrix. In this work, finite element modeling is used to validate an analytic ETC model for application to the composite fuel material for particle-volume fractions up to 40%. The effect of each individual layer of a TRISO particle is analyzed showing that the overall ETC of the compact is most sensitive to the outer layer constituent. In conjunction with the modeling results, the thermal conductivity of matrix-graphite compacts and the ETC of surrogate TRISO fuel compacts have been successfully measured using a previously developed measurement system. The ETC of the surrogate fuel compacts varies between 50-30 W m -1 K -1 over a temperature range of 50-600°C. As a result of the numerical modeling and experimental measurements of the fuel compacts, a new model and approach for analyzing the effect of compact constituent materials on ETC is proposed that can estimate the fuel compact ETC with approximately 15-20% more accuracy than the old method. Using the ETC model with measured thermal conductivity of the graphite matrix-only material indicate that, in the composite form, the matrix material has a much greater thermal conductivity, which is attributed to the high anisotropy of graphite thermal conductivity. Therefore, simpler measurements of individual TRISO compact constituents combined with an analytic ETC model, will not provide accurate predictions of overall ETC of the compacts emphasizing the need for measurements of composite, surrogate compacts.

  19. Field Measurement and Calibration of HDM-4 Fuel Consumption Model on Interstate Highway in Florida

    Directory of Open Access Journals (Sweden)

    Xin Jiao

    2015-03-01

    Full Text Available Fuel consumptions are measured by operating passenger car and tractor-trailer on two interstate roadway sites in Florida. Each site contains flexible pavement and rigid pavement with similar pavement, traffic and environmental condition. Field test reveals that the average fuel consumption differences between vehicle operating on flexible pavement and rigid pavement at given test condition are 4.04% for tractor-trailer and 2.50% for passenger car, with a fuel saving on rigid pavement. The fuel consumption differences are found statistically significant at 95% confidence level for both vehicle types. Test data are then used to calibrate the Highway Development and Management IV (HDM-4 fuel consumption model and model coefficients are obtained for three sets of observations. Field measurement and prediction by calibrated model shows generally good agreement. Nevertheless, verification and adjustment with more experiment or data sources would be expected in future studies.

  20. Measurements of fuel temperature coefficient of reactivity on a commercial AGR

    International Nuclear Information System (INIS)

    Telford, A.; Bridge, M.J.

    1978-01-01

    Tests have been carried out on the commercial AGR at Hikley Point to determine the fuel temperature coefficient of reactivity, an important safety related parameter. Reactor neutron flux was measured during transients induced by movement of a bank of control rods from one steady position to another. An inverse kinetics analysis was applied to the measured flux to determine the change which occured in core reactivity as the fuel temperature changed. The variation of mean fuel temperature was deduced from the flux transient by means of a nine-plane thermal hydraulics representation of the AGR fuel channel. Results so far obtained confirm the predicted variation of fuel temperature coefficient with butn-up. (author)

  1. Technique of measuring and evaluating the R3E dia. 112 experimental fuel element

    International Nuclear Information System (INIS)

    Stanc, S.; Badiar, S.

    1974-10-01

    The charging is considered of the R3E experimental fuel element having a diameter of 112 mm which will serve the simultaneous temperature measurement on 15 sites and neutron flux measurement on 7 sites aimed at verifying the improved R3-type fuel element with extended cladding and a new toroidal spacing. The temperature will be measured using jacket thermocouples while neutron flux will be measured by self-powered detectors. The methods are described of checking the measuring chain by determining the loop resistance of the thermocouples and SPN detectors in order to verify the correctness of the measured data. (J.B.)

  2. Measuring the linear heat generation rate of a nuclear reactor fuel pin

    International Nuclear Information System (INIS)

    Smith, R.D.

    1981-01-01

    A miniature gamma thermometer is described which is capable of travelling through bores distributed in an array through a nuclear reactor core and measure the linear heat generation rate of the fuel pins. (U.K.)

  3. Measuring method for amount of fissionable gas in spent fuel pellet

    International Nuclear Information System (INIS)

    Kashibe, Shinji.

    1992-01-01

    The method of the present invention separately measures the amount of both of a fission product (FP) gas accumulated in bubbles at the crystal grain boundary of spent fuel pellets and an FP gas accumulated in the crystal grains. That is, in a radial position of the spent fuel pellet, a microfine region is mechanically destroyed. The amount of the FP gas released by the destruction from the crystal grains is measured by using a mass analyzer. Then, when the destroyed pieces formed by the destruction are recovered and dissolved, FP gas accumulated in the crystal grains of the pellet is released. The amount released is measured by the mass analyzer. With such procedures, the amount of FP gas accumulated in the bubbles at the crystal grain boundary and in the crystal grains at the radial position of the spent fuel pellet can be measured discriminately. Accordingly, the integrity of the fuel pellet can be recognized appropriately. (I.S.)

  4. Principles of fuel ion ratio measurements in fusion plasmas by collective Thomson scattering

    DEFF Research Database (Denmark)

    Stejner Pedersen, Morten; Nielsen, Stefan Kragh; Bindslev, Henrik

    2011-01-01

    ratio. Measurements of the fuel ion ratio will be important for plasma control and machine protection in future experiments with burning fusion plasmas. Here we examine the theoretical basis for fuel ion ratio measurements by CTS. We show that the sensitivity to plasma composition is enhanced......For certain scattering geometries collective Thomson scattering (CTS) measurements are sensitive to the composition of magnetically confined fusion plasmas. CTS therefore holds the potential to become a new diagnostic for measurements of the fuel ion ratio—i.e. the tritium to deuterium density...... by the signatures of ion cyclotron motion and ion Bernstein waves which appear for scattering geometries with resolved wave vectors near perpendicular to the magnetic field. We investigate the origin and properties of these features in CTS spectra and give estimates of their relative importance for fuel ion ratio...

  5. The underwater coincidence counter for plutonium measurements in mixed-oxide fuel assemblies manual

    International Nuclear Information System (INIS)

    Eccleston, G.W.; Menlove, H.O.; Abhold, M.; Baker, M.; Pecos, J.

    1999-01-01

    This manual describes the Underwater Coincidence Counter (UWCC) that has been designed for the measurement of plutonium in mixed-oxide (MOX) fuel assemblies prior to irradiation. The UWCC uses high-efficiency 3 He neutron detectors to measure the spontaneous-fission and induced-fission rates in the fuel assembly. Measurements can be made on MOX fuel assemblies in air or underwater. The neutron counting rate is analyzed for singles, doubles, and triples time correlations to determine the 240 Pu effective mass per unit length of the fuel assembly. The system can verify the plutonium loading per unit length to a precision of less than 1% in a measurement time of 2 to 3 minutes. System design, components, performance tests, and operational characteristics are described in this manual

  6. Measurement of gamma ray from fuel of high temperature engineering test reactor. Method of measurement and results

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto, Nozomu; Nojiri, Naoki; Takada, Eiji [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2001-02-01

    To obtain information in the HTTR core directly, gamma ray from fuel blocks was measured when fuel blocks were discharged from the core and reloaded to the core. Gamma ray was measured using GM detector, CZT semiconductor detector installed in a door valve and area monitors installed in a stand pipe compartment. The measurement was carried out for 20 fuel blocks in 4 columns considering the symmetry of uranium enrichment distribution in the core. Relative axial distribution in each column obtained by the GM detector and CZT detector agreed with calculated results. However, calculation values showed higher values than measured values in upper region of the core, lower those in lower region of the core. The axial distributions were also evaluated by the area monitors. The measured values agreed with calculated values. It became clear that it was possible to obtain the data inside the core by this method. (author)

  7. Measuring the plutonium distribution in fuel elements by the gamma scanning method

    International Nuclear Information System (INIS)

    Gorobets, A.K.; Leshchenko, Yu.I.; Semenov, A.L.

    1982-01-01

    An on-line system designed for measuring Pu distribution in the length of fresh fuel elements with vibrocompacted UO 2 -PuO 2 fuel rods by the γ-scanning method is described. An algorithm for measurement result processing and the procedure of determination of calibration parameters necessary for the valid signal separat.ion by means of a two-channel analyzer and for evaluation of the self-absorption effect are considered. The device scanning unit consists of two NaI(Tl) detectors simultaneously detecting γ-radiation from the opposite sides of a measured fuel rod section. The cesium source with Esub(γ)=660 keV is used for fuel scanning. On the base of the analysis of the results obtained when studying the BOR-60 experimental fuel elements with fuel rods of 400 mm long by means of the described device clusion is made that fuel element scanning during 20 min (scanning step is 4 mm, measuring time at each step is 10 s) makes it possible to determine Pu distribution with the error less than +-4% at the confidence probability of 0.68

  8. Development of an instantaneous local fuel-concentration measurement probe: an engine application

    Science.gov (United States)

    Guibert, P.; Boutar, Z.; Lemoyne, L.

    2003-11-01

    This work presents a new tool which can deliver instantaneous local measurements of fuel concentration in an engine cylinder with a high temporal resolution, particularly during compression strokes. Fuel concentration is represented by means of equivalence fuel-air ratio, i.e. the real engine mass ratio of fuel to air divided by the same ratio in ideal stoichiometry conditions. Controlling the mixture configuration for any strategy in a spark ignition engine and for auto-ignition combustion has a dominant effect on the subsequent processes of ignition, flame propagation and auto-ignition combustion progression, pollutant formation under lean or even stoichiometric operating conditions. It is extremely difficult, under a transient operation, to control the equivalence air/fuel ratio precisely at a required value and at the right time. This requires the development of a highly accurate equivalence air/fuel ratio control system and a tool to measure using crank angle (CA) resolution. Although non-intrusive laser techniques have considerable advantages, they are most of the time inappropriate due to their optical inaccessibility or the complex experimental set-up involved. Therefore, as a response to the demand for a relatively simple fuel-concentration measurement system a probe is presented that replaces a spark plug and allows the engine to run completely normally. The probe is based on hot-wire like apparatus, but involves catalytic oxidation at the wire surface. The development, characteristics and calibration of the probe are presented followed by applications to in-cylinder engine measurements.

  9. Experimental results on the MOX fuel. Study of the calculus/measures divergences

    International Nuclear Information System (INIS)

    Martin, S.

    1997-01-01

    For each nuclear plant unit restart, all safety criterion have to be respected. Various parameters as boron concentration, temperature coefficient, worth or power and activity distributions related to fuel assemblies, have to be calculated. To compute these parameters Framatome uses the neutronic channel Science. Before the validation they are compare to experimental measures. For UO 2 fuel the divergence calculus/measures are correct. But for MOX fuels the divergence worsening. This paper discusses tis divergence and research the origin. (A.L.B.)

  10. Measurement of enriched uranium and uranium-aluminum fuel materials with the AWCC

    International Nuclear Information System (INIS)

    Krick, M.S.; Menlove, H.O.; Zick, J.; Ikonomou, P.

    1985-05-01

    The active well coincidence counter (AWCC) was calibrated at the Chalk River Nuclear Laboratories (CRNL) for the assay of 93%-enriched fuel materials in three categories: (1) uranium-aluminum billets, (2) uranium-aluminum fuel elements, and (3) uranium metal pieces. The AWCC was a standard instrument supplied to the International Atomic Energy Agency under the International Safeguards Project Office Task A.51. Excellent agreement was obtained between the CRNL measurements and previous Los Alamos National Laboratory measurements on similar mockup fuel material. Calibration curves were obtained for each sample category. 2 refs., 8 figs., 15 tabs

  11. The out-of-pile test for internal pressure measurement of nuclear fuel rod using LVDT

    Energy Technology Data Exchange (ETDEWEB)

    Son, J. M.; Kim, B. K.; Kim, D. S.; Joo, K. N.; Park, S. J.; Kang, Y. H.; Kim, Y. K.; Yeum, K. I. [KAERI, Taejon (Korea, Republic of)

    2002-05-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), the internal pressure measurement technique of the nuclear fuel rod is being developed using LVDT(Linear Variable Differential Transformer). The objectives of this test were to understand the LVDT's characteristics and to study its application techniques for fuel irradiation technology. It will be required to analyze the acquired internal pressure of fuel rod during fuel irradiation test in HANARO. Therefore, the out of pile test system for pressure measurement was developed, and the test with the LVDT at room temperature were performed. This test were implemented in 1 kg/cm{sup 2} increment from 1 kg/cm{sup 2} to 30 kg/cm{sup 2}, and repeated 6 times at same condition. The LVDT's sensitivities were obtained by following two ways, the one by test and the other by calculation from characteristics data. These two sensitivities were compared and analyzed. The calculation method for internal pressure of nuclear fuel rod at specified temperature was also established. The results of the out-of-pile test will be used to predict accurately the internal pressure of fuel rod during irradiation test. And, the well qualified out-of-pile tests are needed to understand the LVDT's detail characteristics at high temperature for the detail design of the fuel irradiation capsule.

  12. The out-of-pile test for internal pressure measurement of nuclear fuel rod using LVDT

    Energy Technology Data Exchange (ETDEWEB)

    Min, Sohn Jae; Kang, Y. H.; Kim, B. G. [and others

    2001-11-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO, the internal pressure measurement technique of the nuclear fuel rod is being developed using LVDT. The objectives of this test were to understand the LVDT's characteristics and to study its application techniques for fuel irradiation technology. It will be required to analyze the acquired internal pressure of fuel rod during fuel irradiation test in HANARO. The out-of-pile test system for pressure measurement was developed, and the test with the LVDT at room temperature(19 .deg. C) were performed. A out-of-pile test were implemented in 1 kg/cm{sup 2} increment from 1 kg/cm{sup 2} to 30 kg/cm{sup 2} and repeated 6 times at each condition. The LVDT's sensitivities were obtained by following two ways, the one by test and the other by calculation from characteristics data. These two sensitivities were compared and analyzed. The calculation method for internal pressure of nuclear fuel rod at specified temperature was also established. This report describes the system configuration, the out-of-pile test procedures, and the results. The results of the out-of-pile test will be used to predict accurately the internal pressure of fuel rod during irradiation test. And, the well qualified out-of-pile tests are needed to understand the LVDT's detail characteristics for the detail design of the fuel irradiation capsule.

  13. Improved measurement of aluminum in irradiated fuel reprocessed at the Savannah River Site

    International Nuclear Information System (INIS)

    Maxwell, S.L. III.

    1991-01-01

    At the Savannah River Site (SRS), irradiated fuel from research reactor operators or their contract fuel service companies is reprocessed in the H-Canyon Separations Facility. Final processing costs are based on analytical measurements of the amount of total metal dissolved. Shipper estimates for uranium and uranium-235 and measured values at SRS have historically agreed very well. There have occasionally been significant differences between shipper estimates for aluminum and the aluminum content determined at SRS. To minimize analytical error that might contribute to poor shipper-receiver agreement for the reprocessing of off-site fuel, a new analytical method to measure aluminum was developed by SRS Analytical Laboratories at the Central Laboratory Facilities. An EDTA (ethylenediaminetetraacetic acid) titration method, subject to dissolver matrix interferences, was previously used at SRS to measure aluminum in H-Canyon dissolver during the reprocessing of offsite fuel. The new method combines rapid ion exchange technology with direct current argon plasma spectrometry to enhance the reliability of aluminum measurements for off-site fuel. The technique rapidly removes spectral interferences such as uranium and significantly lowers gamma levels due to fission products. Aluminium is separated quantitatively by using an anion exchange technique that employs oxalate complexing, small particle size resin and rapid flow rates. The new method, which has eliminated matrix interference problems with these analyses and improved the quality of aluminum measurements, has improved the overall agreement between shipper-receiver values for offsite fuel processed SRS

  14. Synergistic Smart Fuel For In-pile Nuclear Reactor Measurements

    Energy Technology Data Exchange (ETDEWEB)

    James A. Smith; Dale K. Kotter; Randall A. Ali; Steven L . Garrett

    2013-10-01

    In March 2011, an earthquake of magnitude 9.0 on the Richter scale struck Japan with its epicenter on the northeast coast, near the Tohoku region. In addition to the immense physical destruction and casualties across the country, several nuclear power plants (NPP) were affected. It was the Fukushima Daiichi NPP that experienced the most severe and irreversible damage. The earthquake brought the reactors at Fukushima to an automatic shutdown and because the power transmission lines were damaged, emergency diesel generators (EDGs) were activated to ensure that there was continued cooling of the reactors and spent fuel pools. The situation was being successfully managed until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to the reactors.2 At this point, the situation became critical. There was a loss of the sensors and instrumentation within the reactor that could have provided valuable information to guide the operators to make informed decisions and avoid the unfortunate events that followed. In the light of these events, we have developed and tested a potential self-powered thermoacoustic system, which will have the ability to serve as a temperature sensor and can transmit data independently of electronic networks. Such a device is synergistic with the harsh environment of the nuclear reactor as it utilizes the heat from the nuclear fuel to provide the input power.

  15. The Measurement of Fuel-Air Ratio by Analysis for the Oxidized Exhaust Gas

    Science.gov (United States)

    Gerrish, Harold C.; Meem, J. Lawrence, Jr.

    1943-01-01

    An investigation was made to determine a method of measuring fuel-air ratio that could be used for test purposes in flight and for checking conventional equipment in the laboratory. Two single-cylinder test engines equipped with typical commercial engine cylinders were used. The fuel-air ratio of the mixture delivered to the engines was determined by direct measurement of the quantity of air and of fuel supplied and also by analysis of the oxidized exhaust gas and of the normal exhaust gas. Five fuels were used: gasoline that complied with Army-Navy fuel Specification No. AN-VV-F-781 and four mixtures of this gasoline with toluene, benzene, and xylene. The method of determining the fuel-air ratio described in this report involves the measurement of the carbon-dioxide content of the oxidized exhaust gas and the use of graphs for the presented equation. This method is considered useful in aircraft, in the field, or in the laboratory for a range of fuel-air ratios from 0.047 to 0.124.

  16. The Measurement of Fuel-air Ratio by Analysis of the Oxidized Exhaust Gas

    Science.gov (United States)

    Memm, J. Lawrence, Jr.

    1943-01-01

    An investigation was made to determine a method of measuring fuel-air ratio that could be used for test purposes in flight and for checking conventional equipment in the laboratory. Two single-cylinder test engines equipped with typical commercial engine cylinders were used. The fuel-air ratio of the mixture delivered to the engines was determined by direct measurement of the quantity of air and of fuel supplied and also by analysis of the oxidized exhaust gas and of the normal exhaust gas. Five fuels were used: gasoline that complied with Army-Navy Fuel Specification, No. AN-VV-F-781 and four mixtures of this gasoline with toluene, benzene, and xylene. The method of determining the fuel-air ratio described in this report involves the measurement of the carbon-dioxide content of the oxidized exhaust gas and the use of graphs or the presented equation. This method is considered useful in aircraft, in the field, or in the laboratory for a range of fuel-air ratios from 0.047 to 0.124

  17. Measurements of decay heat and gamma-ray intensity of spent LWR fuel assemblies

    International Nuclear Information System (INIS)

    Vogt, J.; Agrenius, L.; Jansson, P.; Baecklin, A.; Haakansson, A.; Jacobsson, S.

    1999-01-01

    Calorimetric measurements of the decay heat of a number of BWR and PWR fuel assemblies have been performed in the pools at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel, CLAB. Gamma-ray measurements, using high-resolution gamma-ray spectroscopy (HRGS), have been carried out on the same fuel assemblies in order to test if it is possible to find a simple and accurate correlation between the 137 CS -intensity and the decay heat for fuel with a cooling time longer than 10-12 years. The results up to now are very promising and may ultimately lead to a qualified method for quick and accurate determination of the decay heat of old fuel by gamma-ray measurements. By means of the gamma spectrum the operator declared data on burnup, cooling time and initial enrichment can be verified as well. CLAB provides a unique opportunity in the world to follow up the decay heat of individual fuel assemblies during several decades to come. The results will be applicable for design and operation of facilities for wet and dry interim storage and subsequent encapsulation for final disposal of the fuel. (author)

  18. Measurement station for interim inspections of Lightbridge metallic fuel rods at the Halden Boiling Water Reactor

    Science.gov (United States)

    Hartmann, C.; Totemeier, A.; Holcombe, S.; Liverud, J.; Limi, M.; Hansen, J. E.; Navestad, E. AB(; )

    2018-01-01

    Lightbridge Corporation has developed a new Uranium-Zirconium based metallic fuel. The fuel rods aremanufactured via a co-extrusion process, and are characterized by their multi-lobed (cruciform-shaped) cross section. The fuel rods are also helically-twisted in the axial direction. Two experimental fuel assemblies, each containing four Lightbridge fuel rods, are scheduled to be irradiated in the Halden Boiling Water Reactor (HBWR) starting in 2018. In addition to on-line monitoring of fuel rod elongation and critical assembly conditions (e.g. power, flow rates, coolant temperatures, etc.) during the irradiation, several key parameters of the fuel will be measured out-of-core during interim inspections. An inspection measurement station for use in the irradiated fuel handling compartment at the HBWR has therefore been developed for this purpose. The multi-lobed cladding cross section combined with the spiral shape of the Lightbridge metallic fuel rods requires a high-precision guiding system to ensure good position repeatability combined with low-friction guiding. The measurement station is equipped with a combination of instruments and equipment supplied from third-party vendors and instruments and equipment developed at Institute for Energy Technology (IFE). Two sets of floating linear voltage differential transformer (LVDT) pairs are used to measure swelling and diameter changes between the lobes and the valleys over the length of the fuel rods. Eddy current probes are used to measure the thickness of oxide layers in the valleys and on the lobe tips and also to detect possible surface cracks/pores. The measurement station also accommodates gamma scans. Additionally, an eddy-current probe has been developed at IFE specifically to detect potential gaps or discontinuities in the bonding layer between the metallic fuel and the Zirconium alloy cladding. Potential gaps in the bonding layer will be hidden behind a 0.5-1.0 mm thick cladding wall. It has therefore been

  19. Validating criticality calculations for spent fuel with 252Cf-source-driven noise measurements

    International Nuclear Information System (INIS)

    Mihalczo, J.T.; Krass, A.W.; Valentine, T.E.

    1992-01-01

    The 252 Cf-Source-driven noise analysis method can be used for measuring the subcritical neutron multiplication factor k of arrays of spent light water reactor (LWR) fuel. This type of measurement provides a parameter that is directly related to the criticality state of arrays of LWR fuel. Measurements of this parameter can verify the criticality safety margins of spent LWR fuel configurations and thus could be a means of obtaining the information to justify burnup credit for spent LWR transportation/storage casks. The practicality of a measurement depends on the ability to install the hardware required to perform the measurement. Source chambers containing the 252 Cf at the required source intensity for this application have been constructed and have operated successfully for ∼10 years and can be fabricated to fit into control rod guide tubes of PWR fuel elements. Fission counters especially developed for spent-fuel measurements are available that would allow measurements of a special 3 x 3 spent fuel array and a typical burnup credit rail cask with spent fuel in unborated water. Adding a moderator around these fission counters would allow measurements with the typical burnup credit rail cask with borated water and the special 3 x 3 array with borated water. The recent work of Ficaro on modifying the KENO Va code to calculate by the Monte Carlo method the time sequences of pulses at two detectors near a fissile assembly from the fission chain multiplication process, initiated by a 252 Cf source in the assembly allows a direct computer calculation of the noise analysis data from this measurement method

  20. Testing and preformance measurement of straight vegetable oils as an alternative fuel for diesel engines

    Science.gov (United States)

    Lakshminarayanan, Arunachalam

    Rising fuel prices, growing energy demand, concerns over domestic energy security and global warming from greenhouse gas emissions have triggered the global interest in bio-energy and bio-fuel crop development. Backlash from these concerns can result in supply shocks of traditional fossil fuels and create immense economic pressure. It is thus widely argued that bio-fuels would particularly benefit developing countries by off-setting their dependencies on imported petroleum. Domestically, the transportation sector accounts for almost 40% of liquid fuel consumption, while on-farm application like tractors and combines for agricultural purposes uses close to an additional 18%. It is estimated that 40% of the farm budget can be attributed to the fuel costs. With the cost of diesel continuously rising, farmers are now looking at using Straight Vegetable Oil (SVO) as an alternative fuel by producing their own fuel crops. This study evaluates conventional diesel compared to the use of SVO like Camelina, Canola and Juncea grown on local farms in Colorado for their performance and emissions on a John Deere 4045 Tier-II engine. Additionally, physical properties like density and viscosity, metal/mineral content, and cold flow properties like CFPP and CP of these oils were measured using ASTM standards and compared to diesel. It was found that SVOs did not show significant differences compared to diesel fuel with regards to engine emissions, but did show an increase in thermal efficiency. Therefore, this study supports the continued development of SVO production as a viable alternative to diesel fuels, particularly for on-farm applications. The need for providing and developing a sustainable, economic and environmental friendly fuel alternative has taken an aggressive push which will require a strong multidisciplinary education in the field of bio-energy. Commercial bio-energy development has the potential to not only alleviate the energy concerns, but also to give renewed

  1. Reactivity change measurements on plutonium-uranium fuel elements in hector experimental techniques and results

    International Nuclear Information System (INIS)

    Tattersall, R.B.; Small, V.G.; MacBean, I.J.; Howe, W.D.

    1964-08-01

    The techniques used in making reactivity change measurements on HECTOR are described and discussed. Pile period measurements were used in the majority of oases, though the pile oscillator technique was used occasionally. These two methods are compared. Flux determinations were made in the vicinity of the fuel element samples using manganese foils, and the techniques used are described and an error assessment made. Results of both reactivity change and flux measurements on 1.2 in. diameter uranium and plutonium-uranium alloy fuel elements are presented, these measurements being carried out in a variety of graphite moderated lattices at temperatures up to 450 deg. C. (author)

  2. Experimental verification of neutron emission method for measuring of fissile material content in spent fuel

    International Nuclear Information System (INIS)

    Abou-Zaid, A.A.; Pytel, K.

    1999-01-01

    A non-destructive method of measurement of fissile nuclides content remained in spent fuel from research reactor is presented. The method, called the neutron emission one, is based on counting of fission neutrons emitted from fissile isotopes: 235 U, 239 Pu, 241 Pu. Fissions are induced mainly by neutrons supplied by the external neutron source. Another effects contribute also to the measured neutron population, e. g. source neutrons from penetrating the fuel without being captured and scattered, neutrons (α,n) reactions and from spontaneous fissions of actinides. Complexity of phenomena occurring within the measurement facility required the detailed numerical simulation and experimental studies prior design of ultimate measurement stand. In the previous paper, the results of Monte Carlo simulation on optimisation of measuring stand for neutron emission method were presented. On the basis of those results, the experimental stand for Maria reactor fuel investigation has been designed and manufactured. The present paper, being the continuation of previous one, contains the description of experimental facility and the results of measurements for the fresh fuel (without burnup) and the fuel mock-up (without fissile materials). Although some discrepancies were found between Monte Carlo and experimental results, the main conclusions concerning the optimal geometry of measuring facility have been confirmed. (author)

  3. Recent advances in hardware and software are to improve spent fuel measurements

    International Nuclear Information System (INIS)

    Staples, P.; Beddingfield, D.H.; Lestone, J.P.; Pelowitz, D.G.; Bytchkov, M.; Starovich, Z.; Harizanov, I.; Luna-Vellejo, J.; Lavender, C.

    2001-01-01

    Vast quantities of spent fuel are available for safeguard measurements, primarily in Commonwealth of Independent States (CIS) of the former Soviet Union. This spent fuel, much of which consists of long-cooling-time material, is going to become less unique in the world safeguards arena as reprocessing projects or permanent repositories continue to be delayed or postponed. The long cooling time of many of the spent fuel assemblies being prepared for intermediate term storage in the CIS countries promotes the possibility of increased accuracy in spent fuel assays. This improvement is made possible through the process of decay of the Curium isotopes and of fission products. An important point to consider for the future that could advance safeguards measurements for reverification and inspection would be to determine what safeguards requirements should be imposed upon this 'new' class of spent fuel, Improvements in measurement capability will obviously affect the safeguards requirements. What most significantly enables this progress in spent fuel measurements is the improvement in computer processing power and software enhancements leading to user-friendly Graphical User Interfaces (GUT's). The software used for these projects significantly reduces the IAEA inspector's time expenditure for both learning and operating computer and data acquisition systems, At the same time, by standardizing the spent fuel measurements, it is possible to increase reproducibility and reliability of the measurement data. Hardware systems will be described which take advantage of the increased computer control available to enable more complex measurement scenarios. A specific example of this is the active regulation of a spent fuel neutron coincident counter's 3 He tubes high voltage, and subsequent scaling of measurement results to maintain a calibration for direct assay of the plutonium content of Fast Breeder Reactor spent fuel. The plutonium content has been successfully determined for

  4. In-core fuel element temperature and flow measurment of HFETR

    International Nuclear Information System (INIS)

    Chen Daolong; Jiang Pei

    1988-02-01

    The HFETR in-core fuel element temperature-flow measurement facility and its measurement system are expounded. The applications of the instrumented fuel element to stationary and transient states measurements during the lift of power, the operation test of all lifetime at first load, and the deepening burn-up test at second load are described. The method of determination of the hot point temperature under the fin is discussed. The error analysis is made. The fuel element out-of-pile water deprivation test is described. The development of this measurement facility and succesful application have made important contribution to high power and deep burn-up safe operation at two load, in-core fuel element irradiation, and varied investigation of HFETR. After operation at two loads, the integrated power of this instrumented fuel element arrives at 90.88 MWd, its maximum point burn-up is about 64.9%, so that the economy of fuel use of HFETR is raised very much

  5. Hydrodynamic mobility of a sphere moving on the centerline of an elastic tube

    Science.gov (United States)

    Daddi-Moussa-Ider, Abdallah; Lisicki, Maciej; Gekle, Stephan

    2017-11-01

    Elastic channels are an important component of many soft matter systems, in which hydrodynamic interactions with confining membranes determine the behavior of particles in flow. In this work, we derive analytical expressions for Green's functions associated with a point-force (Stokeslet) directed parallel or perpendicular to the axis of an elastic cylindrical channel exhibiting resistance against shear and bending. We then compute the leading order self- and pair mobility functions of particles on the cylinder axis, finding that the mobilities are primarily determined by membrane shear and that bending does not play a significant role. In the quasi-steady limit of vanishing frequency, the particle self- and pair mobilities near a no-slip hard cylinder are recovered only if the membrane possesses a non-vanishing shear rigidity. We further compute the membrane deformation, finding that deformation is generally more pronounced in the axial (radial) directions, for the motion along (perpendicular to) the cylinder centerline, respectively. Our analytical calculations for Green's functions in an elastic cylinder can serve as a fundamental building block for future studies and are verified by fully resolved boundary integral simulations where very good agreement is obtained.

  6. Characterization of projected DWPF glasses heat treated to simulate canister centerline cooling

    International Nuclear Information System (INIS)

    Marra, S.L.; Jantzen, C.M.

    1992-05-01

    Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. The glass will be produced and poured into stainless steel canisters in the Defense Waste Processing Facility (DWPF). Eventually these canistered waste forms will be sent to a geologic repository for final disposal. In order to assure acceptability by the repository, the Department of Energy has defined requirements which DWPF canistered waste forms must meet. These requirements are the Waste Acceptance Preliminary Specifications (WAPS). The WAPS require DWPF to identify the crystalline phases expected to be present in the final glass product. Knowledge of the thermal history of the borosilicate glass during filling and cooldown of the canister is necessary to determine the amount and type of crystalline phases present in the final glass product. Glass samples of seven projected DWPF compositions were cooled following the same temperature profile as that of glass at the centerline of the full-scale DWPF canister. The glasses were characterized by x-ray diffraction and scanning electron microscopy to identify the crystalline phases present The volume percents of each crystalline phase present were determined by quantitative x-ray diffraction. The Product Consistency Test (PCI) was used to determine the durability of the heat-treated glasses

  7. Spent-fuel composition: a comparison of predicted and measured data

    International Nuclear Information System (INIS)

    Thomas, C.C. Jr.; Cobb, D.D.; Ostenak, C.A.

    1981-03-01

    The uncertainty in predictions of the nuclear materials content of spent light-water reactor fuel was investigated to obtain guidelines for nondestructive spent-fuel verification and assay. Values predicted by the reactor operator were compared with measured values from fuel reprocessors for six reactors (three PWR and three BWR). The study indicates that total uranium, total plutonium, fissile uranium, fissile plutonium, and total fissile content can be predicted with biases ranging from 1 to 6% and variabilities (1-sigma) ranging from 2 to 7%. The higher values generally are associated with BWRs. Based on the results of this study, nondestructive assay measurements that are accurate and precise to 5 to 10% (1sigma) or better should be useful for quantitative analyses of typical spent fuel

  8. Development of Vision System for Dimensional Measurement for Irradiated Fuel Assembly

    International Nuclear Information System (INIS)

    Shin, Jungcheol; Kwon, Yongbock; Park, Jongyoul; Woo, Sangkyun; Kim, Yonghwan; Jang, Youngki; Choi, Joonhyung; Lee, Kyuseog

    2006-01-01

    In order to develop an advanced nuclear fuel, a series of pool side examination (PSE) is performed to confirm in-pile behavior of the fuel for commercial production. For this purpose, a vision system was developed to measure for mechanical integrity, such as assembly bowing, twist and growth, of the loaded lead test assembly. Using this vision system, three(3) times of PSE were carried out at Uljin Unit 3 and Kori Unit 2 for the advanced fuels, PLUS7 TM and 16ACE7 TM , developed by KNFC. Among the main characteristics of the vision system is very simple structure and measuring principal. This feature enables the equipment installation and inspection time to reduce largely, and leads the PSE can be finished without disturbance on the fuel loading and unloading activities during utility overhaul periods. And another feature is high accuracy and repeatability achieved by this vision system

  9. Application of reactivity method to MTR fuel burn-up measurement

    International Nuclear Information System (INIS)

    Zuniga, A.; Ravnik, M.; Cuya, R.

    2001-01-01

    Fuel element burn-up has been measured for the first time by reactivity method in a MTR reactor. The measurement was performed in RP-10 reactor of Peruvian Institute for Nuclear Energy (IPEN) in Lima. It is a pool type 10MW material testing reactor using standard 20% enriched uranium plate type fuel elements. A fresh element and an element with well defined burn-up were selected as reference elements. Several elements in the core were selected for burn-up measurement. Each of them was replaced in its original position by both reference elements. Change in excess reactivity was measured using control rod calibration curve. The burn-up reactivity worth of fuel elements was plotted as a function of their calculated burnup. Corrected burn-up values of the measured fuel elements were calculated using the fitting function at experimental reactivity for all elements. Good agreement between measured and calculated burn-up values was observed indicating that the reactivity method can be successfully applied also to MTR fuel element burn-up determination.(author)

  10. Measurement of plutonium in spent nuclear fuel by self-induced x-ray fluorescence

    Energy Technology Data Exchange (ETDEWEB)

    Hoover, Andrew S [Los Alamos National Laboratory; Rudy, Cliff R [Los Alamos National Laboratory; Tobin, Steve J [Los Alamos National Laboratory; Charlton, William S [Los Alamos National Laboratory; Stafford, A [TEXAS A& M; Strohmeyer, D [TEXAS A& M; Saavadra, S [ORNL

    2009-01-01

    Direct measurement of the plutonium content in spent nuclear fuel is a challenging problem in non-destructive assay. The very high gamma-ray flux from fission product isotopes overwhelms the weaker gamma-ray emissions from plutonium and uranium, making passive gamma-ray measurements impossible. However, the intense fission product radiation is effective at exciting plutonium and uranium atoms, resulting in subsequent fluorescence X-ray emission. K-shell X-rays in the 100 keV energy range can escape the fuel and cladding, providing a direct signal from uranium and plutonium that can be measured with a standard germanium detector. The measured plutonium to uranium elemental ratio can be used to compute the plutonium content of the fuel. The technique can potentially provide a passive, non-destructive assay tool for determining plutonium content in spent fuel. In this paper, we discuss recent non-destructive measurements of plutonium X-ray fluorescence (XRF) signatures from pressurized water reactor spent fuel rods. We also discuss how emerging new technologies, like very high energy resolution microcalorimeter detectors, might be applied to XRF measurements.

  11. LWR spent-fuel radiochemical measurements and comparison with ORIGEN2 predictions

    International Nuclear Information System (INIS)

    Blahnik, D.E.; Jenquin, U.P.; Guenther, R.J.

    1988-01-01

    The Materials Characterization Center (MCC) at Pacific Northwest Laboratory is responsible for providing characterized spent fuel, designated approved testing material (ATMs) for subsequent use in the investigation of nuclear waste disposal forms by the US Department of Energy geologic repository project. The ATMs are selected to assure that test material is available that has a representative range of characteristics important to spent-fuel behavior in a geologic repository. Burnup and fission gas release were the primary criteria for selecting the ATMs. The five spent-fuel ATMs (ATM-101, -103, -104, -105, and -106) currently being characterized by the MCC have rod average burnups ranging from 20 to 43 MWd/kg M, fission gas releases ranging from 0.2 to 11.2%, and are from both boiling water reactors and pressurized water reactors. Radiochemical analyses of the fuel included measurements of 148 Nd (for fuel burnup), the isotopes of uranium and plutonium, and nuclides of importance to repository performance. Cladding samples were analyzed for 14 C. The measured values of selected nuclides were compared with values obtained from calculations with the ORIGEN2 code that was used to predict isotopic quantities for all of the ATMS. Ratios of the ORIGEN2 calculated values to the measured values for ATM-103 and ATM-106 fuel are given

  12. Flow measurement by Laser Doppler Anemometry in a nuclear fuel assembly

    International Nuclear Information System (INIS)

    Kehoe, A.

    1984-12-01

    Development of a Laser Doppler Anemometer measurement system and its operation are examined in this research. The system is designed for flow measurement in laboratory models of nuclear fuel assemblies. Use of the system is demonstrated by measuring turbulent velocity profiles in the laboratory model at full scale reactor flow rates. The reactors at the Savanah River Plant (SRP) are heavy water moderated and operate at low temperatures and pressures. Reactor power is currently limited by the temperature of the water in the nuclear fuel assembly. These temperature limits are conservatively calculated without allowing for any turbulent mixing. This research incorporates the design, fabriction and operation of a plexiglas model fuel assembly for the purpose of making turbulent velocity measurement via a Laser Doppler Anemometer System

  13. Analysis of systematic error deviation of water temperature measurement at the fuel channel outlet of the reactor Maria

    International Nuclear Information System (INIS)

    Bykowski, W.

    2000-01-01

    The reactor Maria has two primary cooling circuits; fuel channels cooling circuit and reactor pool cooling circuit. Fuel elements are placed inside the fuel channels which are parallely linked in parallel, between the collectors. In the course of reactor operation the following measurements are performed: continuous measurement of water temperature at the fuel channels inlet, continuous measurement of water temperature at the outlet of each fuel channel and continuous measurement of water flow rate through each fuel channel. Based on those thermal-hydraulic parameters the instantaneous thermal power generated in each fuel channel is determined and by use of that value the thermal balance and the degree of fuel burnup is assessed. The work contains an analysis concerning estimate of the systematic error of temperature measurement at outlet of each fuel channel and so the erroneous assessment of thermal power extracted in each fuel channel and the burnup degree for the individual fuel element. The results of measurements of separate factors of deviations for the fuel channels are enclosed. (author)

  14. Fuel film thickness measurements using refractive index matching in a stratified-charge SI engine operated on E30 and alkylate fuels

    Science.gov (United States)

    Ding, Carl-Philipp; Sjöberg, Magnus; Vuilleumier, David; Reuss, David L.; He, Xu; Böhm, Benjamin

    2018-03-01

    This study shows fuel film measurements in a spark-ignited direct injection engine using refractive index matching (RIM). The RIM technique is applied to measure the fuel impingement of a high research octane number gasoline fuel with 30 vol% ethanol content at two intake pressures and coolant temperatures. Measurements are conducted for an alkylate fuel at one operating case, as well. It is shown that the fuel volume on the piston surface increases for lower intake pressure and lower coolant temperature and that the alkylate fuel shows very little spray impingement. The fuel films can be linked to increased soot emissions. A detailed description of the calibration technique is provided and measurement uncertainties are discussed. The dependency of the RIM signal on refractive index changes is measured. The RIM technique provides quantitative film thickness measurements up to 0.9 µm in this engine. For thicker films, semi-quantitative results of film thickness can be utilized to study the distribution of impinged fuel.

  15. Enrichment measurement in TRIGA type fuels; Medicion de enriquecimiento en combustibles tipo Triga

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F.; Mazon R, R. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2001-05-15

    The Department of Energy of the United States of North America, through the program 'Idaho Operations Nuclear Spent Fuel Program' of the Idaho National Engineering and Environmental Laboratory (INEEL), in Idaho Falls; Idaho USA, hires to Global Technologies Inc. (GTI) to develop a prototype device of detection enrichment uranium (DEU Detection of Enrichment of Uranium) to determine quantitatively the enrichment in remainder U-235 in a TRIGA fuel element at the end of it useful life. The characteristics of the prototype developed by GTI are the following ones: It allows to carry out no-destructive measurements of TRIGA type fuel. Easily transportable due to that reduced of it size. The determination of the enrichment (in grams of U-235) it is obtained with a precision of 5%. The National Institute of Nuclear Research (ININ), in its facilities of the Nuclear Center of Mexico, it has TRIGA type fuel of high and low enrichment (standard and FLIP) fresh and with burnt, it also has the infrastructure (hot cells, armor-plating of transport, etc) and qualified personnel to carry out the necessary maneuvers to prove the operation of the DEU prototype. For this its would be used standard type fuel elements and FLIP, so much fresh as with certain burnt one. In the case of the fresh fuels the measurement doesn't represent any risk, the fuels before and after the measurement its don't contain a quantity of fission products that its represent a radiological risk in its manipulation; but in the case of the fuels with burnt the handling of the same ones represents an important radiological risk reason why for its manipulation it was used the transport armor-plating and the hot cells. (Author)

  16. Fusion of Pixel-based and Object-based Features for Road Centerline Extraction from High-resolution Satellite Imagery

    Directory of Open Access Journals (Sweden)

    CAO Yungang

    2016-10-01

    Full Text Available A novel approach for road centerline extraction from high spatial resolution satellite imagery is proposed by fusing both pixel-based and object-based features. Firstly, texture and shape features are extracted at the pixel level, and spectral features are extracted at the object level based on multi-scale image segmentation maps. Then, extracted multiple features are utilized in the fusion framework of Dempster-Shafer evidence theory to roughly identify the road network regions. Finally, an automatic noise removing algorithm combined with the tensor voting strategy is presented to accurately extract the road centerline. Experimental results using high-resolution satellite imageries with different scenes and spatial resolutions showed that the proposed approach compared favorably with the traditional methods, particularly in the aspect of eliminating the salt noise and conglutination phenomenon.

  17. Fuel Performance Modeling of U-Mo Dispersion Fuel: The thermal conductivity of the interaction layers of the irradiated U-Mo dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mistarhi, Qusai M.; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    U-Mo/Al dispersion fuel performed well at a low burn-up. However, higher burn-up and higher fission rate irradiation testing showed enhanced fuel meat swelling which was caused by high interaction layer growth and pore formation. The performance of the dispersion type fuel in the irradiation and un-irradiation environment is very important. During the fabrication of the dispersion type fuel an Interaction Layer (IL) is formed due to the inter-diffusion between the U-Mo fuel particles and the Al matrix which is an intermetallic compound (U,Mo)Alx. During irradiation, the IL becomes amorphous causing a further decrease in the thermal conductivity and an increase in the centerline temperature of the fuel meat. Several analytical models and numerical methods were developed to study the performance of the unirradiated U-Mo/Al dispersion fuel. Two analytical models were developed to study the performance of the irradiated U-Mo/Al dispersion fuel. In these models, the thermal conductivity of the IL was assumed to be constant. The properties of the irradiated U-Mo dispersion fuel have been investigated recently by Huber et al. The objective of this study is to develop a correlation for IL thermal conductivity during irradiation as a function of the temperature and fission density from the experimentally measured thermal conductivity of the irradiated U-Mo/Al dispersion fuel. The thermal conductivity of IL during irradiation was calculated from the experimentally measured data and a correlation was developed from the thermal conductivity of IL as a function of T and fission density.

  18. Current density distribution mapping in PEM fuel cells as an instrument for operational measurements

    Energy Technology Data Exchange (ETDEWEB)

    Geske, M.; Heuer, M.; Heideck, G.; Styczynski, Z. A. [Otto-von-Guericke University Magdeburg, Chair Electric Power Networks and Renewable Energy Sources, Magdeburg (Germany)

    2010-07-01

    A newly developed measurement system for current density distribution mapping has enabled a new approach for operational measurements in proton exchange membrane fuel cells (PEMFC). Taking into account previously constructed measurement systems, a method based on a multi layer printed circuit board was chosen for the development of the new system. This type of system consists of a sensor, a special electronic device and the control and visualization PC. For the acquisition of the current density distribution values, a sensor device was designed and installed within a multilayer printed circuit board with integrated shunt resistors. Varying shunt values can be taken into consideration with a newly developed and evaluated calibration method. The sensor device was integrated in a PEM fuel cell stack to prove the functionality of the whole measurement system. A software application was implemented to visualize and save the measurement values. Its functionality was verified by operational measurements within a PEMFC system. Measurement accuracy and possible negative reactions of the sensor device during PEMFC operation are discussed in detail in this paper. The developed system enables operational measurements for different operating phases of PEM fuel cells. Additionally, this can be seen as a basis for new opportunities of optimization for fuel cell design and operation modes. (author)

  19. Current Density Distribution Mapping in PEM Fuel Cells as An Instrument for Operational Measurements

    Directory of Open Access Journals (Sweden)

    Martin Geske

    2010-04-01

    Full Text Available A newly developed measurement system for current density distribution mapping has enabled a new approach for operational measurements in proton exchange membrane fuel cells (PEMFC. Taking into account previously constructed measurement systems, a method based on a multi layer printed circuit board was chosen for the development of the new system. This type of system consists of a sensor, a special electronic device and the control and visualization PC. For the acquisition of the current density distribution values, a sensor device was designed and installed within a multilayer printed circuit board with integrated shunt resistors. Varying shunt values can be taken into consideration with a newly developed and evaluated calibration method. The sensor device was integrated in a PEM fuel cell stack to prove the functionality of the whole measurement system. A software application was implemented to visualize and save the measurement values. Its functionality was verified by operational measurements within a PEMFC system. Measurement accuracy and possible negative reactions of the sensor device during PEMFC operation are discussed in detail in this paper. The developed system enables operational measurements for different operating phases of PEM fuel cells. Additionally, this can be seen as a basis for new opportunities of optimization for fuel cell design and operation modes.

  20. Measurement of the heavy water level in the fuel channels of the RA reactor - Annex 11

    International Nuclear Information System (INIS)

    Nikolic, M.

    1964-01-01

    The objective of measuring the heavy water level in the reactor channels was to verify experimentally the possibilities of reactor cooling with parallel operation of heavy water pumps od 1500 rotations/min at nominal power of 6.5 MW. Measurements were done in 2 periphery and 2 central fuel channels with pumps speed 1500, 1800 and 3000 rotations/min by a contact probe with electric resistance measuring device. precision of the measurement was ±1 cm

  1. Electronic temperature control and measurements reactor fuel rig circuits

    Energy Technology Data Exchange (ETDEWEB)

    Glowacki, S W

    1980-01-01

    The electronic circuits of two digital temperature meters developed for the thermocouple of Ni-NiCr type are described. The output thermocouple signal as converted by means of voltage-to-freguency converter. The frequency is measured by a digital scaler controled by quartz generator signals. One of the described meter is coupled with digital temperature controler which drives the power stage of the reactor rig heater. The internal rig temperature is measured by the thermocouple providing the input signal to the mentioned voltage-to-frequency converter, that means the circuits work in the negative feedback loop. The converter frequency-to-voltage ratio is automatically adjusted to match to thermocouple sensitivity changes in the course of the temperature variations. The accuracy of measuring system is of order of +- 1degC for thermocouple temperature changes from 523 K up to 973 K (50degC up to 700degC).

  2. The electronic temperature control and measurements reactor fuel rig circuits

    International Nuclear Information System (INIS)

    Glowacki, S.W.

    1980-01-01

    The electronic circuits of two digital temperature meters developed for the thermocouple of Ni-NiCr type are described. The output thermocouple signal as converted by means of voltage-to-freguency converter. The frequency is measured by a digital scaler controled by quartz generator signals. One of the described meter is coupled with digital temperature controler which drives the power stage of the reactor rig heater. The internal rig temperature is measured by the thermocouple providing the input signal to the mentioned voltage-to-frequency converter, that means the circuits work in the negative feedback loop. The converter frequency-to-voltage ratio is automatically adjusted to match to thermocouple sensitivity changes in the course of the temperature variations. The accuracy of measuring system is of order of +- 1degC for thermocouple temperature changes from 523 K up to 973 K (50degC up to 700degC). (author)

  3. Nuclear reactor fuel assembly grid measuring method and device

    International Nuclear Information System (INIS)

    Fink, D.J.; Cooper, F.W. Jr.

    1987-01-01

    A device is described for remotely measuring a dimension of a workpiece, comprising: (a) first means for measuring the dimension of the workpiece; (b) second means for limiting the force exerted by the first means against the workpiece; (c) third means connected to the first means for moving the first means in the X, Y, Z axes simultaneously relative to the workpiece; (d) fourth means for limiting the force exerted by the third means moving the first means in the X, Y, Z axes; and (e) fifth means remote from the workpiece for monitoring and controlling the movement of the first means relative to the workpiece

  4. Fuel cycle covariance of plutonium and americium separations to repository capacity using information theoretic measures

    International Nuclear Information System (INIS)

    Scopatz, Anthony; Schneider, Erich; Li, Jun; Yim, Man-Sung

    2011-01-01

    A light water reactor, fast reactor symbiotic fuel cycle scenario was modeled and parameterized based on thirty independent inputs. Simultaneously and stochastically choosing different values for each of these inputs and performing the associated fuel cycle mass-balance calculation, the fuel cycle itself underwent Monte Carlo simulation. A novel information theoretic metric is postulated as a measure of system-wide covariance. This metric is the coefficient of variation of the set of uncertainty coefficients generated from 2D slices of a 3D contingency table. It is then applied to the fuel cycle, taking fast reactor used fuel plutonium and americium separations as independent variables and the capacity of a fully-loaded tuff repository as the response. This set of parameters is known from prior studies to have a strong covariance. When measured with all 435 other input parameters possible, the fast reactor plutonium and americium separations pair was found to be ranked the second most covariant. This verifies that the coefficient of variation metric captures the desired sensitivity of sensitivity effects in the nuclear fuel cycle. (author)

  5. Measurement techniques in dry-powdered processing of spent nuclear fuels

    International Nuclear Information System (INIS)

    Bowers, D. L.; Hong, J.-S.; Kim, H.-D.; Persiani, P. J.; Wolf, S. F.

    1999-01-01

    High-performance liquid chromatography (HPLC) with inductively coupled plasma mass spectrometry (ICPMS) detection, α-spectrometry (α-S), and γ-spectrometry (γ-S) were used for the determination of nuclide content in five samples excised from a high-burnup fuel rod taken from a pressurized water reactor (PWR). The samples were prepared for analysis by dissolution of dry-powdered samples. The measurement techniques required no separation of the plutonium, uranium, and fission products. The sample preparation and analysis techniques showed promise for in-line analysis of highly-irradiated spent fuels in a dry-powdered process. The analytical results allowed the determination of fuel burnup based on 148 Nd, Pu, and U content. A goal of this effort is to develop the HPLC-ICPMS method for direct fissile material accountancy in the dry-powdered processing of spent nuclear fuel

  6. Feasibility of subcriticality and NDA measurements for spent fuel by frequency analysis techniques with 252Cf

    International Nuclear Information System (INIS)

    Mihalczo, J.T.; Valentine, T.E.; Mattingly, J.K.

    1996-01-01

    The 252 Cf-source-driven frequency analysis method can be used for measuring the subcritical neutron multiplication factor of arrays of LWR fuel and as little as a single PWR fuel assembly. These measurements can be used to verify the criticality safety margins of spent LWR fuel configurations and thus could be a means of obtaining the information to justify burnup credit for spent LWR transportation/storage casks. In addition, the data can be used to validate calculational methods for criticality safety. These measurements provide parameters that have a higher sensitivity to changes in fissile mass than neutron multiplication factor and thus serve as a better test of calculational methods. The analysis have also shown that measurement of the cross power spectral density (CPSD) between detectors on one side of a single fuel assembly and an internal or external 252 Cf source driving the fission chain multiplication process can be used for nondestructive assay of fissile mass along the length of the assembly. This CPSD is a smooth function of fissile mass and does not depend on the varying inherent source in the fuel assembly and thus is ideal for fissile mass assay

  7. Measurement of gap and grain-boundary inventories of 129I in used CANDU fuels

    International Nuclear Information System (INIS)

    Stroes-Gascoyne, S.; Moir, D.L.; Kolar, M.; Porth, R.J.; McConnell, J.L.; Kerr, A.H.

    1995-01-01

    Combined gap and grain-boundary inventories of 129 I in 14 used CANDU fuel elements were measured by crushing and simultaneously leaching fuel segments for 4 h in a solution containing KI carrier. From analogy with previous work a near one-to-one correlation was anticipated between the amount of stable Xe and the amount of 128 I in the combined gap and grain-boundary regions of the fuel. However, the results showed that such a correlation was only apparent for low linear power rating (LLPR) fuels with an average linear power rating of 44 kW/m), the 129 I values were considerably smaller than expected. The combined gap and grain-boundary inventories of 129 I in the 14 fuels tested varied from 1.8 to 11.0%, with an average value of 3.6 ± 2.4% which suggests that the average value of 8.1 ± 1% used in safety assessment calculations overestimates the instant release fraction for 129 I. Segments of used CANDU fuels were leached for 92 d (samples taken at 5, 28 and 92 d) to determine the kinetics of 129 I release. Results could be fitted tentatively to half-order reaction kinetics, implying that 129 I release is a diffusion-controlled process for LLPR fuels, and also for HLPR fuels, once the gap inventory has been leached. However, more data are needed over longer leaching periods to gain more understanding of the processes that control grain-boundary release of 129 I from used CANDU fuel

  8. Biomass burning fuel consumption rates: a field measurement database

    CSIR Research Space (South Africa)

    Van Leeuwen, TT

    2014-01-01

    Full Text Available residue (n = 4, FC = 6.5), chaparral (n = 2, FC = 32), tropical peatland (n = 4, FC = 314), boreal peatland (n = 2, FC = 42), and tundra (n = 1, FC = 40). Within biomes the regional variability in the number of measurements was sometimes large, with e...

  9. Measurement of Weight of Kernels in a Simulated Cylindrical Fuel Compact for HTGR

    International Nuclear Information System (INIS)

    Kim, Woong Ki; Lee, Young Woo; Kim, Young Min; Kim, Yeon Ku; Eom, Sung Ho; Jeong, Kyung Chai; Cho, Moon Sung; Cho, Hyo Jin; Kim, Joo Hee

    2011-01-01

    The TRISO-coated fuel particle for the high temperature gas-cooled reactor (HTGR) is composed of a nuclear fuel kernel and outer coating layers. The coated particles are mixed with graphite matrix to make HTGR fuel element. The weight of fuel kernels in an element is generally measured by the chemical analysis or a gamma-ray spectrometer. Although it is accurate to measure the weight of kernels by the chemical analysis, the samples used in the analysis cannot be put again in the fabrication process. Furthermore, radioactive wastes are generated during the inspection procedure. The gamma-ray spectrometer requires an elaborate reference sample to reduce measurement errors induced from the different geometric shape of test sample from that of reference sample. X-ray computed tomography (CT) is an alternative to measure the weight of kernels in a compact nondestructively. In this study, X-ray CT is applied to measure the weight of kernels in a cylindrical compact containing simulated TRISO-coated particles with ZrO 2 kernels. The volume of kernels as well as the number of kernels in the simulated compact is measured from the 3-D density information. The weight of kernels was calculated from the volume of kernels or the number of kernels. Also, the weight of kernels was measured by extracting the kernels from a compact to review the result of the X-ray CT application

  10. Dry, portable calorimeter for nondestructive measurement of the activity of nuclear fuel

    Science.gov (United States)

    Beyer, Norman S.; Lewis, Robert N.; Perry, Ronald B.

    1976-01-01

    The activity of a quantity of heat-producing nuclear fuel is measured rapidly, accurately and nondestructively by a portable dry calorimeter comprising a preheater, an array of temperature-controlled structures comprising a thermally guarded temperature-controlled oven, and a calculation and control unit. The difference between the amounts of electric power required to maintain the oven temperature with and without nuclear fuel in the oven is measured to determine the power produced by radioactive disintegration and hence the activity of the fuel. A portion of the electronic control system is designed to terminate a continuing sequence of measurements when the standard deviation of the variations of the amount of electric power required to maintain oven temperature is within a predetermined value.

  11. Temperature measurement in cans of fuel rods and fuel rod simulators

    International Nuclear Information System (INIS)

    Tschoeke, H.; Moeller, R.

    1977-01-01

    On the sodium-cooled 19-rod cluster model for the SNR 300 the can wall temperature distributions of the non-uniformly cooled rods were measured with thermocouples mounted in outer grooves in the peripheral zone, permitting, in connection with Ni solder, a practically undisturbed measurement. For a more exact determination of the local surface temperature a calibration method, the so-called double-wall method, was developed and applied. The description of this calibration method and the experimental results achieved until now are presented. (orig./RW) [de

  12. Outlet temperature measurement correction of Gd fuel assemblies at Dukovany NPP

    International Nuclear Information System (INIS)

    Jurickova, M.

    2008-01-01

    In year 2006 we started data processing from the Dukovany NPP operating history database that contained data from the old measurement system VK3 and the new Scorpio-VVER. The work has been done in cooperation with the reactor physicists at Dukovany NPP. Obtained data from database were compared with calculated parameters from 3D diffusion macrocode Mobydick. During the data processing it was found that the Gd fuel assemblies have different time plot of measured assembly outlet temperature compared to the non-Gd fuel assemblies. Experimental studies in RRC KI found that there is insufficient coolant mixing in the region from the fuel bundle to the fuel assembly thermocouple. Due to this fact the thermocouple measure temperature is systematically higher than real temperature. There are two methods to solve this problem. The first method analyses the flow and heat transfer in the region from the fuel bundle to the fuel assembly thermocouple - this method is developed in Skoda JS. The second method statistically studies differences between the measured and calculated temperature by the Mobydick code using the operational history database. Our study is focused on the second method. Several calculation methods for the correction of measured assembly outlet temperature were developed. All correction methods were applied to the measured temperatures from the Dukovany NPP operating history database and the methods were mutually compared. In near future it is planned to compare results of our chosen correction method with modeling method, which is developing in Skoda JS and it is planned to validate both of them. Consequently, the one of these correction methods will be implemented in the modernized Scorpio-VVER for Dukovany NPP. (author)

  13. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE and AFTER IRRADIATION

    International Nuclear Information System (INIS)

    SCHWINKENDORF, K.N.

    2006-01-01

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements. The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprising two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with ''green'' (fresh) fuel and one with spent fuel. Both the green and spent fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k eff = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, 3 green fuel and 4 spent fuel loading configurations were considered to serve as benchmark models. However, shortcomings in experimental data failed to meet the high standards for a benchmark problem. Nevertheless, the data provided by these subcritical measurements can supply useful

  14. Measurements of nuclear properties of shielding materials and nuclear fuels

    International Nuclear Information System (INIS)

    Todorovic, J.

    1961-12-01

    The objective of this study was determine total cross sections in resonant, thermal and sub thermal neutron energy range, i.e. from few tens of electron volts to 10 -4 . First part of this report contains theoretical results which enable obtaining parameters of nuclides resonant level from experimental data. in some cases it is possible to calculate total scattering cross section, coherent and incoherent cross section. Second part covers the description of transmission method for measuring the total cross section

  15. Combustion and emission formation in a biomass fueled grate furnace - measurements and modelling

    International Nuclear Information System (INIS)

    Lindsjoe, H.

    1997-06-01

    A study of turbulent combustion with special emphasis on the formation of nitrous oxide emissions in a biomass fueled grate furnace has been conducted with the aid of measurements, literature studies and CFD-computations. The literature study covers nitrous oxide formation and the pyrolysis, gasification and combustion of biomass fuel. The measurements were conducted inside the furnace and at the outlet, and temperature and some major species were measured. A tool for the treatment of the bed processes (pyrolysis, gasification and combustion) has been developed. The measurements show significantly higher concentrations of oxygen above the fuel bed than expected. The gas production in the bed was shown to be very unevenly distributed over the width of the furnace. The measured temperatures were relatively low and in the same order as reported from other, similar measurements. The computational results are in good quantitative agreement with the measurements, even for the nitrous oxide emissions. It was necessary to include tar as one of the combustible species to achieve reasonable results. The computations point out that the fuel-NO mechanism is the most important reaction path for the formation of nitrous oxide in biomass combustion in grate furnaces. The thermal NO mechanism is responsible for less than 10% of the total amount of NO-emissions. Although the results are quantitatively in good agreement with the measurements, a sensitivity study showed that the fuel-NO model did not respond to changes in the distribution of secondary air as the measurements indicate. The results from this work have lead to some guidelines on how the furnace should be operated to achieve minimum NO-emissions. Some proposals of smaller changes in the construction are also given. 33 refs, 37 figs, 7 tabs

  16. Measurement and analysis of vibrational behaviour of an SNR-fuel element in sodium flow

    International Nuclear Information System (INIS)

    Hess, B.F.H.; Ruppert, E.; Schmidt, H.; Vinzens, K.

    1975-01-01

    Within the framework of SNR-300 fuel element development programme a complete full size fuel element dummy has been tested thoroughly for nearly 3000 hours at 650 0 C system temperature in the AKB sodium loop at Interatom, Bensberg. Investigations of the hydraulic characteristics by measurements of specific pressure losses, flow velocities, leakage flow through the piston rings and investigations of its vibrational behaviour were part of this endurance test at elevated temperatures. The pressure drop versus flow and the leakage measurement are mentioned briefly to confirm the correctness of the test hydraulics. The vibrational behaviour of the element and the approach to analysis is the main object of this report. (Auth.)

  17. Accuracy of dimension measurements from neutron radiographs of nuclear fuel pins

    International Nuclear Information System (INIS)

    Domanus, J. C.

    1976-03-01

    A review of different methods used for dimension measurements from neutron radiographs. The results are presented of an investigation performed using unirradiated fuel pins with calibrated UO 2 pellet-diameters and fuel-to-clad gaps. A projection microscope, three types of travelling microdensitometers and an electronic image analyzer were used to measure diameters and gaps from neutron radiographs produced at Risoe and Studsvik (Sweden) using different brands of X-ray films and transfer technique with 0.1 mm Dy foil. (author)

  18. Burnup measurements on spent fuel elements of the RP-10 research reactor

    International Nuclear Information System (INIS)

    Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro

    2011-01-01

    This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using 137 Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

  19. Burnup measurements on spent fuel elements of the RP-10 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro, E-mail: mvela@ipen.gob.p [Instituto Peruano de Energia Nuclear (IPEN/Peru), Lima (Peru). Grupo de Calculo, Analisis y Seguridad de Reactores; Terremoto, Luis Antonio Albiac, E-mail: laaterre@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using {sup 137}Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

  20. Tuneable diode laser gas analyser for methane measurements on a large scale solid oxide fuel cell

    Science.gov (United States)

    Lengden, Michael; Cunningham, Robert; Johnstone, Walter

    2011-10-01

    A new in-line, real time gas analyser is described that uses tuneable diode laser spectroscopy (TDLS) for the measurement of methane in solid oxide fuel cells. The sensor has been tested on an operating solid oxide fuel cell (SOFC) in order to prove the fast response and accuracy of the technology as compared to a gas chromatograph. The advantages of using a TDLS system for process control in a large-scale, distributed power SOFC unit are described. In future work, the addition of new laser sources and wavelength modulation will allow the simultaneous measurement of methane, water vapour, carbon-dioxide and carbon-monoxide concentrations.

  1. A technique of melting temperature measurement and its application for irradiated high-burnup MOX fuels

    International Nuclear Information System (INIS)

    Namekawa, Takashi; Hirosawa, Takashi

    1999-01-01

    A melting temperature measurement technique for irradiated oxide fuels is described. In this technique, the melting temperature was determined from a thermal arrest on a heating curve of the specimen which was enclosed in a tungsten capsule to maintain constant chemical composition of the specimen during measurement. The measurement apparatus was installed in an alpha-tight steel box within a gamma-shielding cell and operated by remote handling. The temperature of the specimen was measured with a two-color pyrometer sighted on a black-body well at the bottom of the tungsten capsule. The diameter of the black-body well was optimized so that the uncertainties of measurement were reduced. To calibrate the measured temperature, two reference melting temperature materials, tantalum and molybdenum, were encapsulated and run before and after every oxide fuel test. The melting temperature data on fast reactor mixed oxide fuels irradiated up to 124 GWd/t were obtained. In addition, simulated high-burnup mixed oxide fuel up to 250 GWd/t by adding non-radioactive soluble fission products was examined. These data shows that the melting temperature decrease with increasing burnup and saturated at high burnup region. (author)

  2. Procedure and apparatus for measuring the radial gap between fuel and surrounding cladding in a fuel rod for a nuclear reactor

    International Nuclear Information System (INIS)

    Olshausen, K.D.

    1976-01-01

    A device is described for measuring non-destructively the annular fuel-cladding gap in an irradiated or fresh fuel rod. The principle applied is that a force is applied to an arm which presses the cladding diametrically, thus deforming it until it touches the fuel pellet. By presenting the values of the force applied and the deformation produced on an XY recorder, the width of the gap is obtained. Alternatively the gap width may be obtained digitally. Since the gap is so small that the deformation is within the elastic range, the fuel rod may be reloaded in the reactor for further irradiation. (JIW)

  3. Notification determining technical details concerning measures for transportation of nuclear fuel materials

    International Nuclear Information System (INIS)

    1977-01-01

    These provisions are established on the basis of and to enforce ''The regulation for installation and operation of reactor'', ''The regulation concerning the fabricating business of nuclear fuel'' and ''The regulations concerning the reprocessing business of spent fuel''. The terms used hereinafter are according to those used in such regulations. The limit of radioactivity concentration of things contaminated by the nuclear fuel materials which are not required to be enclosed in vessels is defined in the lists attached. In the applications for the approval of the measures concerning the transport of things remarkably difficult to be enclosed in vessels, the name and the address of the applicant, the kind, quantity, form and constitution of the thing contaminated by the nuclear fuel materials to be transported, the date and route of the transport and the measures for the prevention of injuries during the transport must be written. The limit of quantity of nuclear fuel materials classifying the performance of vessels is defined respectively in the lists attached. The radiation dose rates provided for by the Director General of the Science and Technology Agency concerning transported things and transporting apparatuses are 200 millirem per hour on the surfaces of such things and containers. The nuclear fission materials specified, for which the measures for the prevention of criticality are especially required, include uranium 233, uranium 235, plutonium 238, plutonium 239, plutonium 241, and the chemical compounds of such substances, and the nuclear fuel materials containing one or two and more of such substances, excluding the nuclear fuel materials with less than 15 grams of such uranium and plutonium. (Okada, K.)

  4. Temperature sensitivity study of eddy current and digital gauge probes for nuclear fuel rod oxide measurement

    Science.gov (United States)

    Beck, Faith R.; Lind, R. Paul; Smith, James A.

    2018-04-01

    Novel fuels are part of the nationwide effort to reduce the enrichment of Uranium for energy production. Performance of such fuels is determined by irradiating their surfaces. To test irradiated samples, the instrumentation must operate remotely. The plate checker used in this experiment at Idaho National Lab (INL) performs non-destructive testing on fuel rod and plate geometries with two different types of sensors: eddy current and digital thickness gauges. The sensors measure oxide growth and total sample thickness on research fuels, respectively. Sensor measurement accuracy is crucial because even 10 microns of error is significant when determining the viability of an experimental fuel. One parameter known to affect the eddy current and thickness gauge sensors is temperature. Since both sensor accuracies depend on the ambient temperature of the system, the plate checker has been characterized for these sensitivities. The manufacturer of the digital gauge probes has noted a rather large coefficient of thermal expansion for their linear scale. It should also be noted that the accuracy of the digital gauge probes are specified at 20°C, which is approximately 7°C cooler than the average hot-cell temperature. In this work, the effect of temperature on the eddy current and digital gauge probes is studied, and thickness measurements are given as empirical functions of temperature.

  5. Reactivity and neutron emission measurements of highly burnt PWR fuel rod samples

    International Nuclear Information System (INIS)

    Murphy, M.F.; Jatuff, F.; Grimm, P.; Seiler, R.; Brogli, R.; Meier, G.; Berger, H.-D.; Chawla, R.

    2006-01-01

    Fuel rods with burnup values beyond 50 GWd/t are characterised by relatively large amounts of fission products and a high abundance of major and minor actinides. Of particular interest is the change in the reactivity of the fuel as a function of burnup and the capability of modern codes to predict this change. In addition, the neutron emission from burnt fuel has important implications for the design of transport and storage facilities. Measurements have been made of the reactivity effects and the neutron emission rates of highly burnt uranium oxide and mixed oxide fuel rod samples coming from a pressurised water reactor (PWR). The reactivity measurements have been made in a PWR lattice in the PROTEUS zero-energy reactor moderated in turn with: water, a water and heavy water mixture and water containing boron. A combined transport flask and sample changer was used to insert the 400 mm long burnt fuel rod segments into the reactor. Both control rod compensation and reactor period methods were used to determine the reactivities of the samples. For the range of burnup values investigated, an interesting exponential relationship has been found between the neutron emission rate and the measured reactivity

  6. Observer-Based Fuel Control Using Oxygen Measurement

    DEFF Research Database (Denmark)

    Andersen, Palle; Bendtsen, Jan Dimon; Mortensen, Jan Henrik

    is constructed and validated against data obtained at the plant. A Kalman filter based on measurements of combustion air flow led into the furnace and oxygen concentration in the flue gas is designed to estimate the actual coal flow. With this estimate, it becomes possible to close an inner loop around the coal......This report describes an attempt to improve the existing control af coal mills used at the Danish power plant Nordjyllandsværket Unit 3. The coal mills are not equipped with coal flow sensors; thus an observer-based approach is investigated. A nonlinear differential equation model of the boiler...

  7. Novel method for the measurement of liquid film thickness during fuel spray impingement on surfaces.

    Science.gov (United States)

    Henkel, S; Beyrau, F; Hardalupas, Y; Taylor, A M K P

    2016-02-08

    This paper describes the development and application of a novel optical technique for the measurement of liquid film thickness formed on surfaces during the impingement of automotive fuel sprays. The technique makes use of the change of the light scattering characteristics of a metal surface with known roughness, when liquid is deposited. Important advantages of the technique over previously established methods are the ability to measure the time-dependent spatial distribution of the liquid film without a need to add a fluorescent tracer to the liquid, while the measurement principle is not influenced by changes of the pressure and temperature of the liquid or the surrounding gas phase. Also, there is no need for non-fluorescing surrogate fuels. However, an in situ calibration of the dependence of signal intensity on liquid film thickness is required. The developed method can be applied to measure the time-dependent and two-dimensional distribution of the liquid fuel film thickness on the piston or the liner of gasoline direct injection (GDI) engines. The applicability of this technique was evaluated with impinging sprays of several linear alkanes and alcohols with different thermo-physical properties. The surface temperature of the impingement plate was controlled to simulate the range of piston surface temperatures inside a GDI engine. Two sets of liquid film thickness measurements were obtained. During the first set, the surface temperature of the plate was kept constant, while the spray of different fuels interacted with the surface. In the second set, the plate temperature was adjusted to match the boiling temperature of each fuel. In this way, the influence of the surface temperature on the liquid film created by the spray of different fuels and their evaporation characteristics could be demonstrated.

  8. Impact of Alternative Jet Fuels on Engine Exhaust Composition During the 2015 ECLIF Ground-Based Measurements Campaign.

    Science.gov (United States)

    Schripp, Tobias; Anderson, Bruce; Crosbie, Ewan C; Moore, Richard H; Herrmann, Friederike; Oßwald, Patrick; Wahl, Claus; Kapernaum, Manfred; Köhler, Markus; Le Clercq, Patrick; Rauch, Bastian; Eichler, Philipp; Mikoviny, Tomas; Wisthaler, Armin

    2018-04-17

    The application of fuels from renewable sources ("alternative fuels") in aviation is important for the reduction of anthropogenic carbon dioxide emissions, but may also attribute to reduced release of particles from jet engines. The present experiment describes ground-based measurements in the framework of the ECLIF (Emission and Climate Impact of Alternative Fuels) campaign using an Airbus A320 (V2527-A5 engines) burning six fuels of chemically different composition. Two reference Jet A-1 with slightly different chemical parameters were applied and further used in combination with a Fischer-Tropsch synthetic paraffinic kerosene (FT-SPK) to prepare three semi synthetic jet fuels (SSJF) of different aromatic content. In addition, one commercially available fully synthetic jet fuel (FSJF) featured the lowest aromatic content of the fuel selection. Neither the release of nitrogen oxide or carbon monoxide was significantly affected by the different fuel composition. The measured particle emission indices showed a reduction up to 50% (number) and 70% (mass) for two alternative jet fuels (FSJF, SSJF2) at low power settings in comparison to the reference fuels. The reduction is less pronounced at higher operating conditions but the release of particle number and particle mass is still significantly lower for the alternative fuels than for both reference fuels. The observed correlation between emitted particle mass and fuel aromatics is not strict. Here, the H/C ratio is a better indicator for soot emission.

  9. Performance of an accountability measurement system at an operating fuel reprocessing facility

    International Nuclear Information System (INIS)

    Wade, M.A.; Spraktes, F.W.; Hand, R.L.; Baldwin, J.M.; Filby, E.E.; Lewis, L.C.

    1978-01-01

    The ICPP has been engaged for 25 years in the recovery of uranium from spent reactor fuels. In concert with the reprocessing activity, an accountability measurements system has been operated throughout the history of the ICPP. The structure and functions of the accountability measurements system are presented. Its performance is evaluated in order to illustrate the relation of analytical methodology to the overall measurements system. 6 figures, 5 tables

  10. Measurements of temperature effects in tubular loose coated particle fuel at AEE Winfrith

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, A J

    1971-07-01

    A series of measurements of reaction rates and reactivity changes were made in 1969 in the central region of HECTOR which can be heated to 430 deg C. These measurements were complementary to the NESTOR measurements of cold reaction rates reported at the ninth DRAGON Countries Physics Meeting and made use of fuels which were also studied in NESTOR. A final evaluation of all cores is not yet complete, so this report is an interim statement of the current position.

  11. International auspices for the storage of spent nuclear fuel as a nonproliferation measure

    International Nuclear Information System (INIS)

    O'Brien, J.N.

    1981-01-01

    The maintenance of spent nuclear fuel from power reactors will pose problems regardless of how or when the debate over reprocessing is resolved. At present, many reactor sites contain significant buildups of spent fuel stored in holding pools, and no measure short of shutting down reactors with no remaining storage capacity will alleviate the need for away-from-reactor storage. Although the federal government has committed itself to dealing with the spent fuel problem, no solution has been reached, largely because of a debate over differing projections of storage capacity requirements. Proliferation of weapons grade nuclear material in many nations presents another pressing issue. If nations with small nuclear programs are forced to deal with their own spent fuel accumulations, they will either have to reprocess it indigenously or contract to have it reprocessed in a foreign reprocessing plant. In either case, these nations may eventually possess sufficient resources to assemble a nuclear weapon. The problem of spent fuel management demands real global solutions, and further delay in solving the problem of spent nuclear fuel accumulation, both nationally and globally, can benefit only a small class of elected officials in the short term and may inflict substantial costs on the American public, and possibly the world

  12. Recent standardisation work in Sweden related to measurement of biomass fuel quality

    Energy Technology Data Exchange (ETDEWEB)

    Maansson, Margret [Swedish National Testing and Research Inst., Boraas (Sweden)

    1998-06-01

    Work on Swedish standards for peat and biofuels started close to fifteen years ago. The same technical committee that has the responsibility for peat and solid biofuels is also handling the standardisation work on solid mineral fuels. Its counterpart within the ISO is TC 27 Solid mineral fuels. A number of the Swedish analysis standards are structured such that they define methods for all of the solid fuels in the same standard, with specific requirements for the type of fuel if necessary. By now, twenty Swedish biomass standards have been prepared and adopted, half of them already revised at least once. There are dedicated biofuel standards for terminology, sampling and sample preparation and for determination of parameters such as moisture, ash, size distribution, bulk density and mechanical strength. Solid fuels standards that include biomass and peat in their range of application exist for the determination of volatile matter, sulfur chlorine and calorific value. Solid fuel ash methods have been specifically developed for the determination of unburned material and sulfur content. At the present time, standard methods are being defined for the determination of total amounts of heavy metals in ash, and also methods for measuring the availability (leaching properties) of certain elements in ash, in particular ash from combustion of biomass. Ash methods are of interest because of the focus on the possibilities of returning biomass-origin ash to forest soil as a fertilizer and also to prevent depletion of trace elements caused by the increase in the utilisation of the forest growth

  13. FITTING A THREE DIMENSIONAL PEM FUEL CELL MODEL TO MEASUREMENTS BY TUNING THE POROSITY AND

    DEFF Research Database (Denmark)

    Bang, Mads; Odgaard, Madeleine; Condra, Thomas Joseph

    2004-01-01

    the distribution of current density and further how thisaffects the polarization curve.The porosity and conductivity of the catalyst layer are some ofthe most difficult parameters to measure, estimate and especiallycontrol. Yet the proposed model shows how these two parameterscan have significant influence...... on the performance of the fuel cell.The two parameters are shown to be key elements in adjusting thethree-dimensional model to fit measured polarization curves.Results from the proposed model are compared to single cellmeasurements on a test MEA from IRD Fuel Cells.......A three-dimensional, computational fluid dynamics (CFD) model of a PEM fuel cell is presented. The model consists ofstraight channels, porous gas diffusion layers, porous catalystlayers and a membrane. In this computational domain, most ofthe transport phenomena which govern the performance of the...

  14. Automatic Gamma-Scanning System for Measurement of Residual Heat in Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Osifo, Otasowie

    2007-03-01

    In Sweden, spent nuclear fuel will be encapsulated and placed in a deep geological repository. In this procedure, reliable and accurate spent fuel data such as discharge burnup, cooling time and residual heat must be available. The gamma scanning method was proposed in earlier work as a fast and reliable method for the experimental determination of such spent fuel data. This thesis is focused on the recent achievements in the development of a pilot gamma scanning system and its application in measuring spent fuel residual heat. The achievements include the development of dedicated spectroscopic data-acquisition and analysis software and the use of a specially designed calorimeter for calibrating the gamma scanning system. The pilot system is described, including an evaluation of the performance of the spectrum analysis software. Also described are the gamma-scanning measurements on 31 spent PWR fuel assemblies performed using the pilot system. The results obtained for the determination of residual heat are presented, showing an agreement of (2-3) % with both calorimetric and calculated data. In addition, the ability to verify declared data such as discharge burnup and cooling time is demonstrated

  15. Fuel-clad heat transfer coefficient of a defected fuel rod

    International Nuclear Information System (INIS)

    Bruet, M.; Stora, J.P.

    1976-01-01

    A special rod has been built with a stack of UO 2 pellets inside a thick zircaloy clad. The atmosphere inside the fuel rod can be changed and particularly the introduction of water is possible. The capsule was inserted in the Siloe pool reactor in a special device equipped with a neutron flux monitor. The fuel centerline temperature and the temperature at a certain radius of the clad were recorded by two thermocouples. The temperature profiles in the fuel and in the cladding have been calculated and then the heat transfer coefficient. In order to check the proper functioning of the device, two runs were successively achieved with a helium atmosphere. Then the helium atmosphere inside the fuel rod was removed and replaced by water. The heat transfer coefficients derived from the measurements at low power level are in agreement with the values given by the model based on thermal conductivity. However, for higher power levels, the heat transfer coefficients become higher than those based on the calculated gap

  16. An Electronic Measurement Instrumentation of the Impedance of a Loaded Fuel Cell or Battery.

    Science.gov (United States)

    Aglzim, El-Hassane; Rouane, Amar; El-Moznine, Reddad

    2007-10-17

    In this paper we present an inexpensive electronic measurement instrumentationdeveloped in our laboratory, to measure and plot the impedance of a loaded fuel cell orbattery. Impedance measurements were taken by using the load modulation method. Thisinstrumentation has been developed around a VXI system stand which controls electroniccards. Software under Hpvee ® was developed for automatic measurements and the layout ofthe impedance of the fuel cell on load. The measurement environment, like the ambienttemperature, the fuel cell temperature, the level of the hydrogen, etc..., were taken withseveral sensors that enable us to control the measurement. To filter the noise and theinfluence of the 50Hz, we have implemented a synchronous detection which filters in a verynarrow way around the useful signal. The theoretical result obtained by a simulation underPspice ® of the method used consolidates the choice of this method and the possibility ofobtaining correct and exploitable results. The experimental results are preliminary results ona 12V vehicle battery, having an inrush current of 330A and a capacity of 40Ah (impedancemeasurements on a fuel cell are in progress, and will be the subject of a forthcoming paper).The results were plotted at various nominal voltages of the battery (12.7V, 10V, 8V and 5V)and with two imposed currents (0.6A and 4A). The Nyquist diagram resulting from theexperimental data enable us to show an influence of the load of the battery on its internalimpedance. The similitude in the graph form and in order of magnitude of the valuesobtained (both theoretical and practical) enables us to validate our electronic measurementinstrumentation. One of the future uses for this instrumentation is to integrate it with several control sensors, on a vehicle as an embedded system to monitor the degradation of fuel cell membranes.

  17. An Electronic Measurement Instrumentation of the Impedance of a Loaded Fuel Cell or Battery

    Directory of Open Access Journals (Sweden)

    Reddad El-Moznine

    2007-10-01

    Full Text Available In this paper we present an inexpensive electronic measurement instrumentationdeveloped in our laboratory, to measure and plot the impedance of a loaded fuel cell orbattery. Impedance measurements were taken by using the load modulation method. Thisinstrumentation has been developed around a VXI system stand which controls electroniccards. Software under Hpvee® was developed for automatic measurements and the layout ofthe impedance of the fuel cell on load. The measurement environment, like the ambienttemperature, the fuel cell temperature, the level of the hydrogen, etc..., were taken withseveral sensors that enable us to control the measurement. To filter the noise and theinfluence of the 50Hz, we have implemented a synchronous detection which filters in a verynarrow way around the useful signal. The theoretical result obtained by a simulation underPspice® of the method used consolidates the choice of this method and the possibility ofobtaining correct and exploitable results. The experimental results are preliminary results ona 12V vehicle battery, having an inrush current of 330A and a capacity of 40Ah (impedancemeasurements on a fuel cell are in progress, and will be the subject of a forthcoming paper.The results were plotted at various nominal voltages of the battery (12.7V, 10V, 8V and 5Vand with two imposed currents (0.6A and 4A. The Nyquist diagram resulting from theexperimental data enable us to show an influence of the load of the battery on its internalimpedance. The similitude in the graph form and in order of magnitude of the valuesobtained (both theoretical and practical enables us to validate our electronic measurementinstrumentation. One of the future uses for this instrumentation is to integrate it with several control sensors, on a vehicle as an embedded system to monitor the degradation of fuel cell membranes.

  18. A review on the welding technology for the sealing of irradiation test fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. W.; Kang, Y. H.; Kim, B. G.; Joo, K. N.; Oh, J. M.; Park, S. J.; Shin, Y. T

    2000-02-01

    For the irradiation test of nuclear fuel in a research reactor, the fuel manufacturing technology should be developed in advance. Highly radioactive fission products are produced and can be released from the fuel materials during irradiation. Therefore, The sealing of the test is one of the most important procedure among the test fuel manufacturing processes, considering its impacts on the safety of a reactor operation.many welding techniques such as TIG, EBW, LBW, upset butt welding and flash welding are applied in sealing the end of fuel elements. These welding techniques are adopted in conjunction with the weld material, weldability, weld joint design and cost effectiveness. For fuel irradiation test, the centerline temperature of fuel pellets is one of the important item to be measured. For this, a thermocouple is installed into the center of the fuel pellet. The sealing of the penetration hole of the thermocouple sheath should be conducted and the hole should be perfectly sealed using the dissimilar metal joining technique. For this purpose, the dissimilar metal welding between zircaloy-4 and Inconel or stainless steel is needed to be developed. This report describes the techniques sealing the end cap and the penetration of a thermocouple sheath by welding. (author)

  19. Measurement of thermal neutron distribution of flux in the fuel element cluster - Progress report

    International Nuclear Information System (INIS)

    Krcevinac, S.B.; Takac, S.M.

    1966-12-01

    Relative distribution of thermal neutron flux in the fuel element cluster (19 UO 2 rods with Zr-II cladding, D 2 O moderator and coolant) was measured by newly developed cell perturbation method. The obtained values of mean density ratios are compared to the results obtained by TER-I code using Amouyal - Benoist model [sr

  20. Measurement of burnup in FBR MOX fuel irradiated to high burnup

    International Nuclear Information System (INIS)

    Koyama, Shin-ichi; Osaka, Masahiko; Sekine, Takashi; Morozumi, Katsufumi; Namekawa, Takashi; Itoh, Masahiko

    2003-01-01

    The burnup of fuel pins in the subassemblies irradiated at the range from 0.003 to 13.28% FIMA in the JOYO MK-II core were measured by the isotope dilution analysis. For the measurement, 75 and 51 specimens were taken from the fuel pins of driver fuel and irradiation test subassemblies, respectively. The data of burnup could be obtained within an experimental error of 4%, and were compared with the ones calculated by 3-dimensional neutron diffusion codes MAGI and ESPRIT-J, which are used for JOYO core management system. Both data of burnup almost agree with each other within an error of 5%. For the fuel pins loaded at the outer region of the subassembly in the 4th row, which was adjacent to reflectors, however, some of the calculation results were 15% less at most than the measured values. It is suggested from the calculation by a Monte Carlo code MCNP-4A that this difference between the calculated and the measured data attribute from the softening of neutron flux in the region adjacent to the reflector. (author)

  1. Magnet Fiducialization with Coordinate Measuring Machines

    International Nuclear Information System (INIS)

    Friedsam, H.; Oren, W.; Pietryka, M.; SLAC

    2005-01-01

    One of the fundamental alignment problems encountered when building a particle accelerator is the transfer of a component's magnetic centerline position to external fiducials. This operation, dubbed fiducialization, is critical because it can contribute significantly to the alignment error budget. The fiducialization process requires two measurements: (1) from magnetic centerline to mechanical centerline, and (2) from mechanical centerline to external fiducials. This paper will focus on methods for observing the second measurement. Two Stanford Linear Collider (SLC) examples are presented. The object of magnet fiducialization is to relate the magnet-defined beamline position to exterior reference surfaces. To be useful for later component alignment, this relationship must be established in a manner consistent with overall positioning tolerances. The error budget for the SLC's ± 100 (micro)m component to component alignment tolerance is as follows: magnetic centerline to mechanical centerline--σ = ±30 (micro)m; mechanical centerline to fiducial marks--σ = ±50 (micro)m; and fiducial marks to adjacent components--σ = ±80 (micro)m; the TOTAL σ = ±100 (micro)m. The offset between the mechanical and magnetic centerlines of well-known magnets is generally smaller than the ±30 (micro)m measurement tolerance. It is commonly assumed to be zero without measurement. When this tiny value must be measured, extreme care is necessary to avoid obscuring the offset with measurement tool registration errors. In contrast, the mechanical centerline to fiducial measurement must be performed on every magnet. The 50 (micro)m tolerance for this operation is only slightly larger and pushes conventional surveying technology to its limit

  2. Magnet Fiducialization with Coordinate Measuring Machines

    Energy Technology Data Exchange (ETDEWEB)

    Friedsam, H.; Oren, W.; Pietryka, M.; /SLAC

    2005-08-12

    One of the fundamental alignment problems encountered when building a particle accelerator is the transfer of a component's magnetic centerline position to external fiducials. This operation, dubbed fiducialization, is critical because it can contribute significantly to the alignment error budget. The fiducialization process requires two measurements: (1) from magnetic centerline to mechanical centerline, and (2) from mechanical centerline to external fiducials. This paper will focus on methods for observing the second measurement. Two Stanford Linear Collider (SLC) examples are presented. The object of magnet fiducialization is to relate the magnet-defined beamline position to exterior reference surfaces. To be useful for later component alignment, this relationship must be established in a manner consistent with overall positioning tolerances. The error budget for the SLC's {+-} 100 {micro}m component to component alignment tolerance is as follows: magnetic centerline to mechanical centerline--{sigma} = {+-}30 {micro}m; mechanical centerline to fiducial marks--{sigma} = {+-}50 {micro}m; and fiducial marks to adjacent components--{sigma} = {+-}80 {micro}m; the TOTAL {sigma} = {+-}100 {micro}m. The offset between the mechanical and magnetic centerlines of well-known magnets is generally smaller than the {+-}30 {micro}m measurement tolerance. It is commonly assumed to be zero without measurement. When this tiny value must be measured, extreme care is necessary to avoid obscuring the offset with measurement tool registration errors. In contrast, the mechanical centerline to fiducial measurement must be performed on every magnet. The 50 {micro}m tolerance for this operation is only slightly larger and pushes conventional surveying technology to its limit.

  3. The Fuel Performance Analysis of LWR Fuel containing High Thermal Conductivity Reinforcements

    International Nuclear Information System (INIS)

    Kim, Seung Su; Ryu, Ho Jin

    2015-01-01

    The thermal conductivity of fuel affects many performance parameters including the fuel centerline temperature, fission gas release and internal pressure. In addition, enhanced safety margin of fuel might be expected when the thermal conductivity of fuel is improved by the addition of high thermal conductivity reinforcements. Therefore, the effects of thermal conductivity enhancement on the fuel performance of reinforced UO2 fuel with high thermal conductivity compounds should be analyzed. In this study, we analyzed the fuel performance of modified UO2 fuel with high thermal conductivity reinforcements by using the FRAPCON-3.5 code. The fissile density and mechanical properties of the modified fuel are considered the same with the standard UO2 fuel. The fuel performance of modified UO2 with high thermal conductivity reinforcements were analyzed by using the FRAPCON-3.5 code. The thermal conductivity enhancement factors of the modified fuels were obtained from the Maxwell model considering the volume fraction of reinforcements

  4. Acoustic sensor for in-pile fuel rod fission gas release measurement

    International Nuclear Information System (INIS)

    Fourmentel, D.; Villard, J. F.; Ferrandis, J. Y.; Augereau, F.; Rosenkrantz, E.; Dierckx, M.

    2009-01-01

    We have developed a specific acoustic sensor to improve the knowledge of fission gas release in Pressurized Water Reactor (PWR) fuel rods when irradiated in materials testing reactors. In order to perform experimental programs related to the study of the fission gas release kinetics, the CEA (French Nuclear Energy Commission) acquired the ability to equip a pre-irradiated PWR fuel rod with three sensors, allowing the simultaneous on-line measurements of the following parameters: - fuel temperature with a centre-line thermocouple type C, - internal pressure with a specific counter-pressure sensor, - fraction of fission gas released in the fuel rod with an innovative acoustic sensor. The third detector is the subject of this paper. This original acoustic sensor has been designed to measure the molar mass and pressure of the gas contained in the fuel rod plenum. For in-pile instrumentation, the fraction of fission gas, such as Krypton and Xenon, in Helium, can be deduced online from this measurement. The principle of this acoustical sensor is the following: a piezoelectric transducer generates acoustic waves in a cavity connected to the fuel rod plenum. The acoustic waves are propagated and reflected in this cavity and then detected by the transducer. The data processing of the signal gives the velocity of the acoustic waves and their amplitude, which can be related respectively to the molar mass and to the pressure of the gas. The piezoelectric material of this sensor has been qualified in nuclear conditions (gamma and neutron radiations). The complete sensor has also been specifically designed to be implemented in materials testing reactors conditions. For this purpose some technical points have been studied in details: - fixing of the piezoelectric sample in a reliable way with a suitable signal transmission, - size of the gas cavity to avoid any perturbation of the acoustic waves, - miniaturization of the sensor because of narrow in-pile experimental devices

  5. In-reactor measurements of thermo mechanical behaviour and fission gas release of water reactor fuel

    International Nuclear Information System (INIS)

    Kolstad, E.; Vitanza, C.

    1983-01-01

    the fuel performance during and after a power ramp can be investigated by direct in-pile measurements related to the thermal, mechanical and fission gas release behaviour. The thermal response is examined by thermocouples placed at the centre of the fuel. Such measurements allow the determination of thermal feedback effects induced by the simultaneous liberation of fission gases. The thermal feedback effect is also being separately studied out-of-pile in a specially designed rod where the fission gas release is simulated by injecting xenon in known quantities at different axial positions within the rod. Investigations on the mechanical behaviour are based on axial and diametral cladding deformation measurements. This enables the determination of the amount of local cladding strain and ridging during ramping, the extent of relaxation during the holding time and the amount of residual (plastic) deformation. Gap width measurements are also performed in operating fuel rods using a cladding deflection technique. Fission gas release data are obtained, besides from post-irradiation puncturing, by continuous measurements of the rod internal pressure. This type of measurement leads to the description of the kinetics of the fission gas release process at different powers. The data tend to indicate that the time-dependent release can be reasonably well described by simple diffusion. The paper describes measuring techniques developed and currently in use in Halden, and presents and discusses selected experimental results obtained during various power ramps and transients. (author)

  6. Using fine-scale fuel measurements to assess wildland fuels, potential fire behavior and hazard mitigation treatments in the southeastern USA

    International Nuclear Information System (INIS)

    Ottmar, Roger D.; Blake, John I.; Crolly, William T.

    2012-01-01

    The inherent spatial and temporal heterogeneity of fuelbeds in forests of the southeastern United States may require fine scale fuel measurements for providing reliable fire hazard and fuel treatment effectiveness estimates. In a series of five papers, an intensive, fine scale fuel inventory from the Savanna River Site in the southeastern United States is used for building fuelbeds and mapping fire behavior potential, evaluating fuel treatment options for effectiveness, and providing a comparative analysis of landscape modeled fire behavior using three different data sources including the Fuel Characteristic Classification System, LANDFIRE, and the Southern Wildfire Risk Assessment. The research demonstrates that fine scale fuel measurements associated with fuel inventories repeated over time can be used to assess broad scale wildland fire potential and hazard mitigation treatment effectiveness in the southeastern USA and similar fire prone regions. Additional investigations will be needed to modify and improve these processes and capture the true potential of these fine scale data sets for fire and fuel management planning.

  7. Device for measuring the local power in nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Thor, E.R.; Korpaas, Harald; Leyse, R.H.; Smith, R.D.

    1977-01-01

    The purpose of the invention is to make possible the fabrication of a gamma ray thermometer, capable of being accurately calibrated, before being mounted in the core of the reactor, by means of a system simulating the power released by the fuel. As already indicated, the principle used for performing the measurement with a gamma ray thermometer consists, schematically, in establishing a constant and given conductance path in the mass of an absorbent body and, after calibration, deducting from the temperature difference recorded at the ends of this path the value of the calorific power absorbed by the body and therefore generated by the surrounding nuclear fuel [fr

  8. EURATOM work on standard defects and dimensional measurements in neutron radiography of nuclear fuel elements

    International Nuclear Information System (INIS)

    Domanus, J.C.

    1981-10-01

    In 1979 a working group on neutron radiography was formed at Euratom. The purpose of this group is the standardization of neutron radiographic methods in the field of nuclear fuel. First priority was given to the development of image quality indicators and standard objects for the determination of accuracy of dimensional measurements from neutron radiographs. For that purpose beam purity and sensitivity indicators as well as a calibration fuel pin were designed and fabricated at Risoe. All the Euratom neutron radiography centers have recieved the above items for comparative neutron radiography. The measuring results obtained, using various measuring apparatus, will form the basis to formulate conclusions about the best measuring methods and instruments to be used in that field. (author)

  9. Spent fuel measurements. passive neutron albedo reactivity (PNAR) and photon signatures

    Energy Technology Data Exchange (ETDEWEB)

    Eigenbrodt, Julia [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Menlove, Howard Olsen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-29

    The International Atomic Energy Agency’s (IAEA) safeguards technical objective is the timely detection of a diversion of a significant quantity of nuclear material from peaceful activities to the manufacture of nuclear weapons or of other nuclear explosive devices or for purposes unknown, and deterrence of such diversion by the risk of early detection. An important IAEA task towards meeting this objective is the ability to accurately and reliably measure spent nuclear fuel (SNF) to verify reactor operating parameters and verify that the fuel has not been removed from reactors or SNF storage facilities. This dissertation analyzes a method to improve the state-of-the-art of nuclear material safeguards measurements using two combined measurement techniques: passive neutron albedo reactivity (PNAR) and passive spectral photon measurements.

  10. Prototype vibration measurement program for reactor internals (177-fuel assembly plant). Supplement 1

    International Nuclear Information System (INIS)

    Simonis, J.C.; Post, R.C.; Thoren, D.E.

    1976-08-01

    The surveillance specimen holder tubes installed in the Babcock and Wilcox 177-fuel assembly plants have been redesigned. The structural adequacy of this design has been verified through extensive analysis. The design adequacy will be further confirmed by measuring the vibrational response of the surveillance specimen holder tube during normal and transient flow operation. This report describes the vibration measurement program that will be conducted at Toledo Edison's Davis Besse 1 site

  11. Calculation of the internal pressure of fuel rod from measurements of krypton-85 at its plenum

    International Nuclear Information System (INIS)

    Arana, I.; Doncel, N.; Casado, C.

    2012-01-01

    ENUSA carried out numerous campaigns of measurement internal pressure of fuel rod irradiated. All of them have been performed of form destructively in a hot cell laboratory which implies a time high to obtain results and a high economic cost to obtain a single data by rod, representative of the end of the irradiation. The objective of the project is to develop a non-destructive measurement and a methodology for reliable calculation that eliminates these problems.

  12. Comparison of measured and calculated burn-up of AVR-Fuel-Elements

    Energy Technology Data Exchange (ETDEWEB)

    Wagemann, R.

    1974-03-15

    Burn-up comparisons are made for small batches of three types of AVR fuel elements using a coupled EREBUS-MUPO neutronic analysis compared against test results from both nondestructive gamma-ray measurements of cesium-137 activity and destructive mass spectrometry measurements of the ratio of U-233 to U-235. The comparisons are relatively good for average burn-up and reasonably good for burn-up distributions.

  13. System for uranium superficial density measurement in U3Si2 MTR fuel plates using radiography

    International Nuclear Information System (INIS)

    Hey, Martin A.; Gomez Marlasca, Fernando

    2003-01-01

    The paper describes a method for measuring uranium superficial density in high density uranium silicide (U 3 Si 2 ) MTR fuel plates, through the use of industrial radiography, a set of patterns built for this purpose, a transmission optical densitometer, and a quantitative model of analysis and measurement. Our choice for this particular method responds to its high accuracy, low cost and easy implementation according to the standing quality control systems. (author)

  14. New development in nondestructive measurement and verification of irradiated LWR fuels

    International Nuclear Information System (INIS)

    Lee, D.M.; Phillips, J.R.; Halbig, J.K.; Hsue, S.T.; Lindquist, L.O.; Ortega, E.M.; Caine, J.C.; Swansen, J.; Kaieda, K.; Dermendjiev, E.

    1979-01-01

    Nondestructive techniques for characterizing irradiated LWR fuel assemblies are discussed. This includes detection systems that measure the axial activity profile, neutron yield and gamma yield. A multi-element profile monitor has been developed that offers a significant improvement in speed and complexity over existing mechanical scanning systems. New portable detectors and electronics, applicable to safeguard inspection, are presented and results of gamma-ray and neutron measurements at commercial reactor facilities are given

  15. Density and Viscosity Measurement of Diesel Fuels at Combined High Pressure and Elevated Temperature

    Directory of Open Access Journals (Sweden)

    Carl Schaschke

    2013-07-01

    Full Text Available We report the measurement of the viscosity and density of various diesel fuels, obtained from British refineries, at elevated pressures up to 500 MPa and temperatures in the range 298 K to 373 K. The measurement and prediction procedures of fluid properties under high pressure conditions is of increasing interest in many processes and systems including enhanced oil recovery, automotive engine fuel injection, braking, and hydraulic systems. Accurate data and understanding of the fluid characteristic in terms of pressure, volume and temperature is required particularly where the fluid is composed of a complex mixture or blend of aliphatic or aromatic hydrocarbons. In this study, high pressure viscosity data was obtained using a thermostatically-controlled falling sinker-type high pressure viscometer to provide reproducible and reliable viscosity data based on terminal velocity sinker fall times. This was supported with density measurements using a micro-pVT device. Both high-pressure devices were additionally capable of illustrating the freezing points of the hydrocarbon mixtures. This work has, thus, provided data that can extend the application of mixtures of commercially available fuels and to test the validity of available predictive density and viscosity models. This included a Tait-style equation for fluid compressibility prediction. For complex diesel fuel compositions, which have many unidentified components, the approach illustrates the need to apply appropriate correlations, which require accurate knowledge or prediction of thermodynamic properties.

  16. Ignition delay measurements of light naphtha: A fully blended low octane fuel

    KAUST Repository

    Javed, Tamour

    2016-06-15

    Light naphtha is a fully blended, low-octane (RON. = 64.5, MON. = 63.5), highly paraffinic (>. 90% paraffinic content) fuel, and is one of the first distillates obtained during the crude oil refining process. Light naphtha is an attractive low-cost fuel candidate for advanced low-temperature compression ignition engines where autoignition is the primary control mechanism. We measured ignition delay times for light naphtha in a shock tube and a rapid compression machine (RCM) over a broad range of temperatures (640-1250. K), pressures (20 and 40. bar) and equivalence ratios (0.5, 1 and 2). Ignition delay times were modeled using a two-component primary reference fuel (PRF) surrogate and a multi-component surrogate. Both surrogates adequately captured the measured ignition delay times of light naphtha under shock tube conditions. However, for low-temperature RCM conditions, simulations with the multi-component surrogate showed better agreement with experimental data. These simulated surrogate trends were confirmed by measuring the ignition delay times of the PRF and multi-component surrogates in the RCM at . P = 20. bar, . ϕ = 2. Detailed kinetic analyses were undertaken to ascertain the dependence of the surrogates\\' reactivity on their chemical composition. To the best of our knowledge, this is the first fundamental autoignition study on the reactivity of a low-octane fully blended fuel and the use of a suitably formulated multi-component surrogate to model its behavior.

  17. Deposit Probe Measurements in Danish Grate and Pulverized Fuel Biomass Power Boilers

    DEFF Research Database (Denmark)

    Hansen, Stine Broholm; Jensen, Peter Arendt; Jappe Frandsen, Flemming

    2012-01-01

    . Corresponding samples of fuels, ash deposits and fly ash have provided information on the transformation of inorganics in the boiler. Generally, grate fired boilers provide a fly ash containing high contents of K, Cl and S compared to the fuel ash, while suspension fired boilers fly ash has a composition nearly...... similar to the fuel ash. Inner most biomass deposits are always salt-rich, while thicker deposit layers also contain some Si and Ca. Deposit probe formation rate measurements have been performed in different ways on several boilers. Grate and suspension fired boilers seems to cause similar deposit...... formation rates. Suspension fired boilers generate more fly ash, while grate boilers form a fly ash with a higher fraction of melt formation (and thereby a higher sticking probability) at similar temperatures. For suspension fired units it is observed that wood with a lower ash content than straw gives rise...

  18. Measuring the environmental benefits of hydrogen transportation fuel cycles under uncertainty about external costs

    International Nuclear Information System (INIS)

    Chernyavs'ka, Liliya; Gulli, Francesco

    2010-01-01

    In this paper, we attempt to measure the environmental benefits of hydrogen deployment in the transportation sector. We compare the hydrogen pathways to the conventional transportation fuel cycles in terms of external costs, estimated using the results of the most accurate methodologies available in this field. The central values of performed analysis bring us ambiguous results. The external cost of the best conventional solution ('oil to diesel hybrid internal-combustion engine') in some cases is just higher and in others just lower than that of the best fossil fuel to hydrogen solution ('natural gas to hydrogen fuel cell'). Nevertheless, by accounting for the uncertainty about external costs, we are able to remove this ambiguity highlighting that the hydrogen pathway provides significant environmental benefits ,especially in densely populated areas, assuming 100% city driving.

  19. Burn-up measurements on nuclear reactor fuels using high performance liquid chromatography

    International Nuclear Information System (INIS)

    Sivaraman, N.; Subramaniam, S.; Srinivasan, T.G.; Vasudeva Rao, P.R.

    2002-01-01

    Burn-up measurements on thermal as well as fast reactor fuels were carried out using high performance liquid chromatography (HPLC). A column chromatographic technique using di-(2-ethylhexyl) phosphoric acid (HDEHP) coated column was employed for the isolation of lanthanides from uranium, plutonium and other fission products. Ion-pair HPLC was used for the separation of individual lanthanides. The atom percent fissions were calculated from the concentrations of the lanthanide (neodymium in the case of thermal reactor and lanthanum for the fast reactor fuels) and from uranium and plutonium contents of the dissolver solutions. The HPLC method was also used for determining the fractional fissions from uranium and plutonium for the thermal reactor fuel. (author)

  20. A fuel performance code TRUST VIc and its validation

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, M; Kogai, T [Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki (Japan)

    1997-08-01

    This paper describes a fuel performance code TRUST V1c developed to analyze thermal and mechanical behavior of LWR fuel rod. Submodels in the code include FP gas models depicting gaseous swelling, gas release from pellet and axial gas mixing. The code has FEM-based structure to handle interaction between thermal and mechanical submodels brought by the gas models. The code is validated against irradiation data of fuel centerline temperature, FGR, pellet porosity and cladding deformation. (author). 9 refs, 8 figs.

  1. A fuel performance code TRUST VIc and its validation

    International Nuclear Information System (INIS)

    Ishida, M.; Kogai, T.

    1997-01-01

    This paper describes a fuel performance code TRUST V1c developed to analyze thermal and mechanical behavior of LWR fuel rod. Submodels in the code include FP gas models depicting gaseous swelling, gas release from pellet and axial gas mixing. The code has FEM-based structure to handle interaction between thermal and mechanical submodels brought by the gas models. The code is validated against irradiation data of fuel centerline temperature, FGR, pellet porosity and cladding deformation. (author). 9 refs, 8 figs

  2. Dimensional measurements and eddy currents control of the sheath integrity for a set of irradiated candu fuel elements

    International Nuclear Information System (INIS)

    Gheorghe, G.; Man, I.

    2015-01-01

    During irradiation in the nuclear reactor, fuel elements undergo dimensional and structural changes, and changes of sheath surface condition as well, which can lead to damages and even loss of integrity. This paper presents the results of dimensional measurements and of examination technique with eddy currents for three fuel elements of an irradiated CANDU fuel bundle. One of the fuel elements (FE), which is studied in detail, presented a crack about 40 mm long. The purpose of these nondestructive examination techniques is to determine those parameters that characterize the behavior and performance of nuclear fuel operation. This paper contains images of defects and interpretations of the causes of their occurrence. (authors)

  3. The high temperature out-of-pile test of LVDT for elongation measurement of fuel pellet

    Energy Technology Data Exchange (ETDEWEB)

    Son, J. M.; Kim, B. K.; Jo, M. S.; Joo, K. N.; Park, S. J.; Gang, Y. H.; Kim, Y. J. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), the elongation measurement technique of the fuel pellet is being developed using LVDT(Linear Variable Differential Transformer). The well qualified out-of-pile test were needed to understand the LVDT's detail characteristics at high temperature for the detail design of the fuel irradiation instrumented capsule, because LVDT is very sensitive to variation of temperature. Therefore, the high temperature out-of-pile test system for fuel pellet elongation was developed, and this test was performed under the temperature condition between room temperature and 300 .deg. C with increasing the elongation from 0 to 5 mm. The LVDT's high temperature characteristics and temperature sensitivity of LVDT were analyzed through this experiment. Based on the result of this test, the method for the application of LVDT and elongation detector at high temperature was introduced. It is known that the results will be used to predict accurately the elongation of fuel pellet during irradiation test.

  4. Spent Fuel Test - Climax: technical measurements. Interim report, fiscal year 1982

    International Nuclear Information System (INIS)

    Patrick, W.C.; Ballou, L.B.; Butkovich, T.R.

    1983-02-01

    The Spent Fuel Test - Climax (SFT-C) is located 420 m below surface in the Climax stock granite on the Nevada Test Site. The test is being conducted for the US Department of Energy (DOE) under the technical direction of the Lawrence Livermore National Laboratory (LLNL). Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized April to May 1980, thus initiating a test with a planned 3- to 5-year fuel storage phase. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. Three exchanges of spent fuel between the SFT-C and a surface storage facility furthered this demonstration. Technical objectives of the test led to development of a technical measurements program, which is the subject of this and two previous interim reports. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the first 2-1/2 years of the test on more than 900 channels. Data continue to be acquired from the test. Some data are now available for analysis and are presented here. Highlights of activities this year include analysis of fracture data obtained during site characterization, laboratory studies of radiation effects and drilling damage in Climax granite, improved calculations of near-field heat transfer and thermomechanical response, a ventilation effects study, and further development of the data acquisition and management systems

  5. Method and apparatus for real-time measurement of fuel gas compositions and heating values

    Science.gov (United States)

    Zelepouga, Serguei; Pratapas, John M.; Saveliev, Alexei V.; Jangale, Vilas V.

    2016-03-22

    An exemplary embodiment can be an apparatus for real-time, in situ measurement of gas compositions and heating values. The apparatus includes a near infrared sensor for measuring concentrations of hydrocarbons and carbon dioxide, a mid infrared sensor for measuring concentrations of carbon monoxide and a semiconductor based sensor for measuring concentrations of hydrogen gas. A data processor having a computer program for reducing the effects of cross-sensitivities of the sensors to components other than target components of the sensors is also included. Also provided are corresponding or associated methods for real-time, in situ determination of a composition and heating value of a fuel gas.

  6. A Study on Vehicle Emission Factor Correction Based on Fuel Consumption Measurement

    Science.gov (United States)

    Wang, Xiaoning; Li, Meng; Peng, Bo

    2018-01-01

    The objective of this study is to address the problem of obvious differences between the calculated and measured emissions of pollutants from motor vehicle by using the existing "Environmental Impact Assessment Specification of Highway Construction Projects". First, a field study collects the vehicle composition ratio, speed, slope, fuel consumption and other essential data. Considering practical applications, the emission factors corresponding to 40km/h and 110km/h and 120km/h velocity are introduced by data fitting. Then, the emission factors of motor vehicle are revised based on the measured fuel consumption, and the pollutant emission modified formula was calculated and compared with the standard recommendation formula. The results show the error between calculated and measured values are within 5%, which can better reflect the actual discharge of the motor vehicle.

  7. Burn-up measurements of LEU fuel for short cooling times

    International Nuclear Information System (INIS)

    Pereda B, C.; Henriquez A, C.; Klein D, J.; Medel R, J.

    2005-01-01

    The measurements presented in this work were made essentially at in-pool gamma-spectrometric facility, installed inside of the secondary pool of the RECH-1 research reactor, where the measured fuel elements are under 2 meters of water. The main reason for using the in-pool facility was because of its capability to measure the burning of fuel elements without having to wait so long, that is with only 5 cooling days, which are the usual times between reactor operations. Regarding these short cooling times, this work confirms again the possibility of using the 95 Zr as a promising burnup monitor, in spite of the rough approximations used to do it. These results are statistically reasonable within the range calculated using codes. The work corroborates previous results, presented in Santiago de Chile, and it suggests future improvements in that way. (author)

  8. Measuring head for determining the pressure of fission gases released inside bars of nuclear fuel

    International Nuclear Information System (INIS)

    Granata, S.

    1984-01-01

    A measuring head suitable for determining the pressure of fission gases released inside non-instrumented bars of nuclear fuel (which have reached high irradiation levels), and for connection to said bars by a method which allows no escape of said active gases and does not cause appreciable disturbance either to the fuel or to the released fission gases, is disclosed. The head consists of a tubular casing adapted to be welded at one end to the bar, and having a metal bellows at its other end. A pointed metal bar is used to penetrate the bar by a blow to a pin, whereupon pressure variations within the casing are measured by a pressure measuring device having an iron core, the movement of the core, due to such pressure variations, being recorded by a differential transformer. (author)

  9. In-reactor fuel cladding external corrosion measurement process and results

    International Nuclear Information System (INIS)

    Thomazet, J.; Musante, Y.; Pigelet, J.

    1999-01-01

    Analysis of the zirconium alloy cladding behaviour calls for an on-site corrosion measurement device. In the 80's, a FISCHER probe was used and allowed oxide layer measurements to be taken along the outer generating lines of the peripheral fuel rods. In order to allow measurements on inner rods, a thin Eddy current probe called SABRE was developed by FRAMATOME. The SABRE is a blade equipped with two E.C coils is moved through the assembly rows. A spring allows the measurement coil to be clamped on each of the generating lines of the scanned rods. By inserting this blade on all four assembly faces, measurements can also be performed along several generating lines of the same rod. Standard rings are fitted on the device and allow on-line calibration for each measured row. Signal acquisition and processing are performed by LAGOS, a dedicated software program developed by FRAMATOME. The measurements are generally taken at the cycle outage, in the spent fuel pool. On average, data acquisition calls for one shift per assembly (eight hours): this corresponds to more than 2500 measurement points. These measurements are processed statistically by the utility program SAN REMO. All the results are collected in a database for subsequent behaviour analysis: examples of investigated parameters are the thermal/hydraulic conditions of the reactors, the irradiation history, the cladding material, the water chemistry This analysis can be made easier by comparing the behaviour measurement and prediction by means of the COROS-2 corrosion code. (author)

  10. Quantitative liquid and vapor distribution measurements in evaporating fuel sprays using laser-induced exciplex fluorescence

    International Nuclear Information System (INIS)

    Fansler, Todd D; Drake, Michael C; Gajdeczko, Boguslaw; Düwel, Isabell; Koban, Wieland; Zimmermann, Frank P; Schulz, Christof

    2009-01-01

    Fully quantitative two-dimensional measurements of liquid- and vapor-phase fuel distributions (mass per unit volume) from high-pressure direct-injection gasoline injectors are reported for conditions of both slow and rapid vaporization in a heated, high-pressure spray chamber. The measurements employ the coevaporative gasoline-like fluorobenzene (FB)/diethylmethylamine (DEMA)/hexane exciplex tracer/fuel system. In contrast to most previous laser-induced exciplex-fluorescence (LIEF) experiments, the quantitative results here include regions in which liquid and vapor fuel coexist (e.g. near the injector exit). A unique aspect is evaluation of both vapor- and liquid-phase distributions at varying temperature and pressure using only in situ vapor-phase fluorescence calibration measurements at room temperature and atmospheric pressure. This approach draws on recent extensive measurements of the temperature-dependent spectroscopic properties of the FB–DEMA exciplex system, in particular on knowledge of the quantum efficiencies of the vapor-phase and liquid-phase (exciplex) fluorescence. In addition to procedures necessary for quantitative measurements, we discuss corrections for liquid–vapor crosstalk (liquid fluorescence that overlaps the vapor-fluorescence bandpass), the unknown local temperature due to vaporization-induced cooling, and laser-sheet attenuation by scattering and absorption

  11. Simulations and measurements of adiabatic annular flows in triangular, tight lattice nuclear fuel bundle model

    Energy Technology Data Exchange (ETDEWEB)

    Saxena, Abhishek, E-mail: asaxena@lke.mavt.ethz.ch [ETH Zurich, Laboratory for Nuclear Energy Systems, Department of Mechanical and Process Engineering, Sonneggstrasse 3, 8092 Zürich (Switzerland); Zboray, Robert [Laboratory for Thermal-hydraulics, Nuclear Energy and Safety Department, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Prasser, Horst-Michael [ETH Zurich, Laboratory for Nuclear Energy Systems, Department of Mechanical and Process Engineering, Sonneggstrasse 3, 8092 Zürich (Switzerland); Laboratory for Thermal-hydraulics, Nuclear Energy and Safety Department, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)

    2016-04-01

    High conversion light water reactors (HCLWR) having triangular, tight-lattice fuels bundles could enable improved fuel utilization compared to present day LWRs. However, the efficient cooling of a tight lattice bundle has to be still proven. Major concern is the avoidance of high-quality boiling crisis (film dry-out) by the use of efficient functional spacers. For this reason, we have carried out experiments on adiabatic, air-water annular two-phase flows in a tight-lattice, triangular fuel bundle model using generic spacers. A high-spatial-resolution, non-intrusive measurement technology, cold neutron tomography, has been utilized to resolve the distribution of the liquid film thickness on the virtual fuel pin surfaces. Unsteady CFD simulations have also been performed to replicate and compare with the experiments using the commercial code STAR-CCM+. Large eddies have been resolved on the grid level to capture the dominant unsteady flow features expected to drive the liquid film thickness distribution downstream of a spacer while the subgrid scales have been modeled using the Wall Adapting Local Eddy (WALE) subgrid model. A Volume of Fluid (VOF) method, which directly tracks the interface and does away with closure relationship models for interfacial exchange terms, has also been employed. The present paper shows first comparison of the measurement with the simulation results.

  12. Estimation of irradiation-induced material damage measure of FCM fuel in LWR core

    International Nuclear Information System (INIS)

    Lee, Kyung-Hoon; Lee, Chungchan; Park, Sang-Yoon; Cho, Jin-Young; Chang, Jonghwa; Lee, Won Jae

    2014-01-01

    An irradiation-induced material damage measure on tri-isotropic (TRISO) multi-coating layers of fully ceramic micro-encapsulated (FCM) fuel to replace conventional uranium dioxide (UO 2 ) fuel for existing light water reactors (LWRs) has been estimated using a displacement per atom (DPA) cross section for a FCM fuel performance analysis. The DPA cross sections in 47 and 190 energy groups for both silicon carbide (SiC) and graphite are generated based on the molecular dynamics simulation by SRIM/TRIM. For the selected FCM fuel assembly design with FeCrAl cladding, a core depletion analysis was carried out using the DeCART2D/MASTER code system with the prepared DPA cross sections to evaluate the irradiation effect in the Korean OPR-1000. The DPA of the SiC and IPyC coating layers is estimated by comparing the discharge burnup obtained from the MASTER calculation with the burnup-dependent DPA for each coating layer calculated using DeCART2D. The results show that low uranium loading and hardened neutron spectrum compared to that of high temperature gas-cooled reactor (HTGR) result in high discharge burnup and high fast neutron fluence. In conclusion, it can be seen that the irradiation-induced material damage measure is noticeably increased under LWR operating conditions compared to HTGRs. (author)

  13. Material accountancy measurement techniques in dry-powdered processing of nuclear spent fuels

    International Nuclear Information System (INIS)

    Wolf, S. F.

    1999-01-01

    The paper addresses the development of inductively coupled plasma-mass spectrometry (ICPMS), thermal ionization-mass spectrometry (TIMS), alpha-spectrometry, and gamma spectrometry techniques for in-line analysis of highly irradiated (18 to 64 GWD/T) PWR spent fuels in a dry-powdered processing cycle. The dry-powdered technique for direct elemental and isotopic accountancy assay measurements was implemented without the need for separation of the plutonium, uranium and fission product elements in the bulk powdered process. The analyses allow the determination of fuel burn-up based on the isotopic composition of neodymium and/or cesium. An objective of the program is to develop the ICPMS method for direct fissile nuclear materials accountancy in the dry-powdered processing of spent fuel. The ICPMS measurement system may be applied to the KAERI DUPIC (direct use of spent PWR fuel in CANDU reactors) experiment, and in a near-real-time mode for international safeguards verification and non-proliferation policy concerns

  14. Sources of variance in BC mass measurements from a small marine engine: Influence of the instruments, fuels and loads

    Science.gov (United States)

    Jiang, Yu; Yang, Jiacheng; Gagné, Stéphanie; Chan, Tak W.; Thomson, Kevin; Fofie, Emmanuel; Cary, Robert A.; Rutherford, Dan; Comer, Bryan; Swanson, Jacob; Lin, Yue; Van Rooy, Paul; Asa-Awuku, Akua; Jung, Heejung; Barsanti, Kelley; Karavalakis, Georgios; Cocker, David; Durbin, Thomas D.; Miller, J. Wayne; Johnson, Kent C.

    2018-06-01

    Knowledge of black carbon (BC) emission factors from ships is important from human health and environmental perspectives. A study of instruments measuring BC and fuels typically used in marine operation was carried out on a small marine engine. Six analytical methods measured the BC emissions in the exhaust of the marine engine operated at two load points (25% and 75%) while burning one of three fuels: a distillate marine (DMA), a low sulfur, residual marine (RMB-30) and a high-sulfur residual marine (RMG-380). The average emission factors with all instruments increased from 0.08 to 1.88 gBC/kg fuel in going from 25 to 75% load. An analysis of variance (ANOVA) tested BC emissions against instrument, load, and combined fuel properties and showed that both engine load and fuels had a statistically significant impact on BC emission factors. While BC emissions were impacted by the fuels used, none of the fuel properties investigated (sulfur content, viscosity, carbon residue and CCAI) was a primary driver for BC emissions. Of the two residual fuels, RMB-30 with the lower sulfur content, lower viscosity and lower residual carbon, had the highest BC emission factors. BC emission factors determined with the different instruments showed a good correlation with the PAS values with correlation coefficients R2 >0.95. A key finding of this research is the relative BC measured values were mostly independent of load and fuel, except for some instruments in certain fuel and load combinations.

  15. Time-averaged probability density functions of soot nanoparticles along the centerline of a piloted turbulent diffusion flame using a scanning mobility particle sizer

    KAUST Repository

    Chowdhury, Snehaunshu

    2017-01-23

    In this study, we demonstrate the use of a scanning mobility particle sizer (SMPS) as an effective tool to measure the probability density functions (PDFs) of soot nanoparticles in turbulent flames. Time-averaged soot PDFs necessary for validating existing soot models are reported at intervals of ∆x/D∆x/D = 5 along the centerline of turbulent, non-premixed, C2H4/N2 flames. The jet exit Reynolds numbers of the flames investigated were 10,000 and 20,000. A simplified burner geometry based on a published design was chosen to aid modelers. Soot was sampled directly from the flame using a sampling probe with a 0.5-mm diameter orifice and diluted with N2 by a two-stage dilution process. The overall dilution ratio was not evaluated. An SMPS system was used to analyze soot particle concentrations in the diluted samples. Sampling conditions were optimized over a wide range of dilution ratios to eliminate the effect of agglomeration in the sampling probe. Two differential mobility analyzers (DMAs) with different size ranges were used separately in the SMPS measurements to characterize the entire size range of particles. In both flames, the PDFs were found to be mono-modal in nature near the jet exit. Further downstream, the profiles were flatter with a fall-off at larger particle diameters. The geometric mean of the soot size distributions was less than 10 nm for all cases and increased monotonically with axial distance in both flames.

  16. Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

    Science.gov (United States)

    Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

    2017-08-01

    Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (RERTR-9B experiment. This paper discusses the TEM characterization results for this U-10Mo/Zr/Al6061 monolithic fuel plate (∼59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 °C, respectively. TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (>1 μm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ∼30 at% and ∼7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

  17. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumntation, and measurement techniques in fuel fabrication facilities, P.O.1236909. Final report

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-12-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. Some of the material included has appeared elswhere and it has been summarized. An extensive bibliography is included. A spcific example of application of the accountability methods to a model fuel fabrication facility which is based on the Westinghouse Anderson design

  18. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumntation, and measurement techniques in fuel fabrication facilities, P. O. 1236909. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-12-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. Some of the material included has appeared elswhere and it has been summarized. An extensive bibliography is included. A spcific example of application of the accountability methods to a model fuel fabrication facility which is based on the Westinghouse Anderson design.

  19. Kinetics parameter measurements on RSG-GAS, a low-enriched fuel reactor

    International Nuclear Information System (INIS)

    Jujuratisbela, U; Arbie, B; Pinem, S.; Tukiran; Suparlina, L.; Singh, O.P.

    1995-01-01

    Kinetics parameter measurements, such as reactivity worths of control rods and fuel elements, beam tube void reactivity, power reactivity coefficient and xenon poisoning reactivity have been performed on different cores of Reaktor Serba Guna G.A. Siwabessy (RSG-GAS). In parallel, a programme was also initiated to measure the other kinetics parameters like effective delayed neutron life time, prompt neutron decay constant, validation of period reactivity relationship and zero power frequency response function. The paper provides the results of these measurements. (author)

  20. Measurements of the fuel temperature coefficient of reactivity at Hinkley Point 'B': 1981

    International Nuclear Information System (INIS)

    George, T.A.

    1982-03-01

    Measurements of the fuel temperature coefficient of reactivity made at Hinkley Point 'B' AGR in 1981 are described. These measurements follow earlier tests reported in e.g. RD/B/N4846 and are part of a series of measurements designed to support theoretical estimates of the change of fuel temperature coefficient as a function of core irradiation. Low and high power measurements were made at a mean core irradiation of 1170GWD. As previously, the measurements at both power levels show agreement with theoretical predictions to within the estimated experimental errors. Recent measurements (mean core irradiation >500GWD) show evidence of a small systematic difference between measured and theoretical values with the experimental values being approximately equal to 0.1mN/ 0 C more positive than the theoretical ones. The measured value of αsub(U) at high power was -0.64+-0.10mN/ 0 C and the low power value, corrected theoretically to normal operating conditions, was also -0.64+-0.10mN/ 0 C. (author)

  1. A Centerline Based Model Morphing Algorithm for Patient-Specific Finite Element Modelling of the Left Ventricle.

    Science.gov (United States)

    Behdadfar, S; Navarro, L; Sundnes, J; Maleckar, M; Ross, S; Odland, H H; Avril, S

    2017-09-20

    Hexahedral automatic model generation is a recurrent problem in computer vision and computational biomechanics. It may even become a challenging problem when one wants to develop a patient-specific finite-element (FE) model of the left ventricle (LV), particularly when only low resolution images are available. In the present study, a fast and efficient algorithm is presented and tested to address such a situation. A template FE hexahedral model was created for a LV geometry using a General Electric (GE) ultrasound (US) system. A system of centerline was considered for this LV mesh. Then, the nodes located over the endocardial and epicardial surfaces are respectively projected from this centerline onto the actual endocardial and epicardial surfaces reconstructed from a patient's US data. Finally, the position of the internal nodes is derived by finding the deformations with minimal elastic energy. This approach was applied to eight patients suffering from congestive heart disease. A FE analysis was performed to derive the stress induced in the LV tissue by diastolic blood pressure on each of them. Our model morphing algorithm was applied successfully and the obtained meshes showed only marginal mismatches when compared to the corresponding US geometries. The diastolic FE analyses were successfully performed in seven patients to derive the distribution of principal stresses. The original model morphing algorithm is fast and robust with low computational cost. This low cost model morphing algorithm may be highly beneficial for future patient-specific reduced-order modelling of the LV with potential application to other crucial organs.

  2. Spent fuel test - Climax: technical measurements. Interim report, Fiscal Year 1983

    International Nuclear Information System (INIS)

    Patrick, W.C.; Butkovich, T.R.; Carlson, R.C.

    1984-02-01

    The Spent Fuel Test - Climax (SFT-C) is located 420 m below surface in the Climax stock granite on the Nevada Test Site. The test is being conducted as part of the Nevada Nuclear Waste Storage Investigations. Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized April-May 1980. The spent-fuel canisters were retrieved and the thermal sources were de-energized in March-April 1983 when test data indicated that test objectives were met during the 3-year storage phase. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. In addition to emplacement and retrieval operations, three exchanges of spent-fuel between the SFT-C and a surface storage facility, conducted during the storage phase, furthered this demonstration. Technical objectives of the test led to development of a technical measurements program, which is the subject of this and three previous interim reports. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the 3-1/2 year duration of the test on more than 900 channels. Data acquisition from the test is now limited to instrumentation calibration and evaluation activities. Data now available for analysis are presented here. Highlights of activities this year include a campaign of in situ stress measurements, mineralogical and petrological studies of pretest core samples, microfracture analyses of laboratory irradiated cores, improved calculations of near-field heat transfer and thermomechanical response during the final months of heating as well as during a six-month cool-down period, metallurgical analyses of selected test components, and further development of the data acquisition and data management systems. 27 references, 68 figures, 10 tables

  3. Spent fuel test - Climax: technical measurements. Interim report, Fiscal Year 1983

    Energy Technology Data Exchange (ETDEWEB)

    Patrick, W.C.; Butkovich, T.R.; Carlson, R.C.; Durham, W.B.; Ganow, H.C.; Hage, G.L.; Majer, E.L.; Montan, D.N.; Nyholm, R.A.; Rector, N.L.

    1984-02-01

    The Spent Fuel Test - Climax (SFT-C) is located 420 m below surface in the Climax stock granite on the Nevada Test Site. The test is being conducted as part of the Nevada Nuclear Waste Storage Investigations. Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized April-May 1980. The spent-fuel canisters were retrieved and the thermal sources were de-energized in March-April 1983 when test data indicated that test objectives were met during the 3-year storage phase. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. In addition to emplacement and retrieval operations, three exchanges of spent-fuel between the SFT-C and a surface storage facility, conducted during the storage phase, furthered this demonstration. Technical objectives of the test led to development of a technical measurements program, which is the subject of this and three previous interim reports. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the 3-1/2 year duration of the test on more than 900 channels. Data acquisition from the test is now limited to instrumentation calibration and evaluation activities. Data now available for analysis are presented here. Highlights of activities this year include a campaign of in situ stress measurements, mineralogical and petrological studies of pretest core samples, microfracture analyses of laboratory irradiated cores, improved calculations of near-field heat transfer and thermomechanical response during the final months of heating as well as during a six-month cool-down period, metallurgical analyses of selected test components, and further development of the data acquisition and data management systems. 27 references, 68 figures, 10 tables.

  4. Description of heat flux measurement methods used in hydrocarbon and propellant fuel fires at Sandia.

    Energy Technology Data Exchange (ETDEWEB)

    Nakos, James Thomas

    2010-12-01

    The purpose of this report is to describe the methods commonly used to measure heat flux in fire applications at Sandia National Laboratories in both hydrocarbon (JP-8 jet fuel, diesel fuel, etc.) and propellant fires. Because these environments are very severe, many commercially available heat flux gauges do not survive the test, so alternative methods had to be developed. Specially built sensors include 'calorimeters' that use a temperature measurement to infer heat flux by use of a model (heat balance on the sensing surface) or by using an inverse heat conduction method. These specialty-built sensors are made rugged so they will survive the environment, so are not optimally designed for ease of use or accuracy. Other methods include radiometers, co-axial thermocouples, directional flame thermometers (DFTs), Sandia 'heat flux gauges', transpiration radiometers, and transverse Seebeck coefficient heat flux gauges. Typical applications are described and pros and cons of each method are listed.

  5. A delayed neutron technique for measuring induced fission rates in fresh and burnt LWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, K.A., E-mail: kajordan@gmail.co [Paul Scherrer Institut, Laboratory for Reactor Physics and System Behaviour, 5232 Villigen (Switzerland); Perret, G. [Paul Scherrer Institut, Laboratory for Reactor Physics and System Behaviour, 5232 Villigen (Switzerland)

    2011-04-01

    The LIFE-PROTEUS program at the Paul Scherrer Institut is being undertaken to characterize the interfaces between burnt and fresh fuel assemblies in modern LWRs. Techniques are being developed to measure fission rates in burnt fuel following re-irradiation in the zero-power PROTEUS research reactor. One such technique utilizes the measurement of delayed neutrons. To demonstrate the feasibility of the delayed neutron technique, fresh and burnt UO{sub 2} fuel samples were irradiated in different positions in the PROTEUS reactor, and their neutron outputs were recorded shortly after irradiation. Fission rate ratios of the same sample irradiated in two different positions (inter-positional) and of two different samples irradiated in the same position (inter-sample) were derived from the measurements and compared with Monte Carlo predictions. Derivation of fission rate ratios from the delayed neutron measured signal requires correcting the signal for the delayed neutron source properties, the efficiency of the measurement setup, and the time dependency of the signal. In particular, delayed neutron source properties strongly depend on the fissile and fertile isotopes present in the irradiated sample and must be accounted for when deriving inter-sample fission rate ratios. Measured inter-positional fission rate ratios generally agree within 1{sigma} uncertainty (on the order of 1.0%) with the calculation predictions. For a particular irradiation position, however, a bias of about 2% is observed and is currently under investigation. Calculated and measured inter-sample fission rate ratios have C/E values deviating from unity by less than 1% and within 2{sigma} of the statistical uncertainties. Uncertainty arising from delayed neutron data is also assessed, and is found to give an additional 3% uncertainty factor. The measurement data indicate that uncertainty is overestimated.

  6. The effect of copper, MDA, and accelerated aging on jet fuel thermal stability as measured by the gravimetric JFTOT

    Energy Technology Data Exchange (ETDEWEB)

    Pande, S.G. [Geo-Centers, Inc., Ft. Washington, MD (United States); Hardy, D.R. [Navy Technology Center for Safety and Survivability, Washington, DC (United States)

    1995-05-01

    Thermally unstable jet fuels pose operational problems. In order to adequately identify such fuels, factors that realistically impact on thermal stability were examined. Evaluation was based on a quantitative method of measuring thermal stability, viz., NRL`s recently developed gravimetric JFTOT. This method gives a quantitative measurement of both the strip deposit and filterables formed. The pertinent factors examined, included the individual and interactive effects of: soluble copper, MDA (metal deactivator), and aging. The latter was accelerated to simulate field conditions of approximately six months aging at ambient temperature and pressure. The results indicate that the individual and interactive effects of copper, MDA, and accelerated aging appear to be fuel dependent. Based on the results, the three test fuels examined (one JP-8 and two JP-5s) were categorized as exhibiting very good, typical, and poor thermal stabilities, respectively. For both the very good and poor thermal stability fuels, the effect of copper in conjunction with accelerated aging did not significantly increase the total thermal deposits of the neat fuels. In contrast, for the typical thermal stability fuel, the combined effects of copper and accelerated aging, did. Furthermore, the addition of MDA prior to aging of the copper-doped, typical stability fuel significantly counteracted the adverse effect of copper and aging. A similar beneficial effect of MDA was not observed for the poor stability fuel. These results focus on the compositional differences among fuels and the need to elucidate these differences (physical and chemical) for a better understanding and prediction of their performance.

  7. Measurement of neutron disadvantage factor for fuel and moderator in the square reactor cell

    International Nuclear Information System (INIS)

    Bosevski, T.; Spiric, V.

    1964-01-01

    Full text: Heterogeneous diffusion treatment for flux distribution was used to define the direction of measurements for obtaining mean neutron flux in the moderator of the reactor cell by single integration. Factor Q for the fuel was determined by using experimental flux distribution in the cell moderator and calculated values for the function X (x;y). Experimental and calculated results are shown as a diagram. All the calculations were done on the ZUSE-Z-23 computer

  8. The measurements of critical mass with uranium fuel elements and thorium rods

    International Nuclear Information System (INIS)

    Yao Zhiquan; Chen Zhicheng; Yao Zewu; Ji Huaxiang; Bao Borong; Zhang Jiahua

    1991-01-01

    The critical experiments with uranium elements and Thorium rods have been performed in zero power reactor at Shanghai Institute of Nuclear Research. The critical masses have been measured in various U/Th ratios. The fuels are 3% 235 U-enriched uranium. The Thorium rods are made from power of ThF 4 . Ratios of calculated values to experimental values are nearly constant at 0.995

  9. Feasibility Demonstration of Exciplex Fluorescence Measurements in Evaporating Laminar Sprays of Diesel Fuel

    Science.gov (United States)

    2011-05-15

    code) 1 FEASIBILITY DEMONSTRATION OF EXCIPLEX FLUORESCENCE MEASUREMENTS IN EVAPORATING LAMINAR SPRAYS OF DIESEL FUEL Final Technical Report Grant...fluorescence is found to increase with temperature up to 538 K and then declines. Fluorescence from the liquid phase, i.e. the exciplex (Naphthalene+TMPD...to have as well characterized a description of the spray environment and assess conclusively the potential of the exciplex approach for more

  10. Measurement of Mixing Rate between Fuel Subchannels: Development of a new Experimental Technique

    International Nuclear Information System (INIS)

    Silin, Nicolas; Barbero, Jose; Bubach, Ernesto; Juanico, Luis

    2000-01-01

    A superficial heater of nickel applied over a ceramic substrate was designed and constructed, together with a system of high sensitivity to measure temperature differentials. The use of both techniques was evaluated and it might allow for the wider use of the method of differential thermal analysis to quantify the turbulent mixing between coupled hydraulic subchannels in fuel elements. Even more, the method presents important advantages as compared to the more complicated techniques known (laser Doppler anemometry)

  11. International interest in the BONAPARTE measurement bench. Post-irradiation examination of lower-enriched fuel plates

    International Nuclear Information System (INIS)

    2014-01-01

    The Belgian Nuclear Research Center SCK-CEN has developed a measurement bench (BONAPARTE) for the non-destructive analysis on fuel plate and rod type fuel elements. BONAPARTE is a modular device that can be employed for many purposes. The article discusses the employment of the BONAPARTE device for the accurate full post-irradiation mapping of fuel plate swelling with degree of precision of just a few micrometers.

  12. Measurement and analysis of vibrational behavior of an SNR-fuel element in sodium flow

    International Nuclear Information System (INIS)

    Hess, B.F.H.; Ruppert, E.; Schmidt, H.; Vinzens, K.

    1975-01-01

    Within the framework of SNR-300 fuel element development programme a complete full size fuel element dummy has been tested thoroughly for nearly 3000 hours at 650 deg C system temperature in the AKB sodium loop at Interatom, Bensberg. It is known that the coolant flow through a subassembly can induce flutter or vibrations of structural parts such as single pins, the wrapper and the total pin bundle all of which have been of interest during this test. To detect these vibrations of different structural parts simultaneously with a minimum of instrumentation only 3 weldable high temperature strain gauges were employed. These strain gauges were especially prepared and bent in such a way as to form a bridge between the inner wrapper and a fuel pin top and spot-welded to both the wrapper and the fuel pin. Although this arrangement seems to be a rather unusual one, the simultaneous-measurement of bundle, wrapper and pin vibrations was possible and periodic flow fluctuations were also detected. The presented results are only relative due to calibration difficulties with these deformed strain gauges which were first used during this test. It is, however, believed that this arrangement, in connection with the proposed anlytical approach, leads to a simple and technical representation of the vibrational behavior of core elements during sodium tests. Detailed information needed for check and calibration of computer codes are however displayed by the respective power spectral density functions

  13. In-situ membrane hydration measurement of proton exchange membrane fuel cells

    Science.gov (United States)

    Lai, Yeh-Hung; Fly, Gerald W.; Clapham, Shawn

    2015-01-01

    Achieving proper membrane hydration control is one of the most critical aspects of PEM fuel cell development. This article describes the development and application of a novel 50 cm2 fuel cell device to study the in-situ membrane hydration by measuring the through-thickness membrane swelling via an array of linear variable differential transducers. Using this setup either as an air/air (dummy) cell or as a hydrogen/air (operating) cell, we performed a series of hydration and dehydration experiments by cycling the RH of the inlet gas streams at 80 °C. From the linear relationship between the under-the-land swelling and the over-the-channel water content, the mechanical constraint within the fuel cell assembly can suppress the membrane water uptake by 11%-18%. The results from the air/air humidity cycling test show that the membrane can equilibrate within 120 s for all RH conditions and that membrane can reach full hydration at a RH higher than 140% in spite of the use of a liquid water impermeable Carbel MP30Z microporous layer. This result confirms that the U.S. DOE's humidity cycling mechanical durability protocol induces sufficient humidity swings to maximize hygrothermal mechanical stresses. This study shows that the novel experimental technique can provide a robust and accurate means to study the in-situ hydration of thin membranes subject to a wide range of fuel cell conditions.

  14. The velocity measurement by LDV at the simulated plate fuel assembly

    International Nuclear Information System (INIS)

    Tae Sung Ha

    2001-01-01

    For a more accurate safety analysis for McMaster Nuclear Reactor (MNR), local velocity measurements in a mock-up of the 18-plate fuel assembly are conducted over the range of M=2.0kg/s to 5.0kg/s (u=0.59m/s to 1.48m/s). To enable the measurement of the mass flow distribution through the channels by Laser Doppler Velocimeter(LDV), the curved fuel plate assembly is modified to flat fuel plates. The experimental result shows that the velocity profile is fairly symmetric for the 1st channel to the 17th subchannel at its center. The velocity in the peripheral area is slightly decreased while that directly above the circular pipe is correspondingly increased due to the effect of blockage by the exit endfitting. The mass flow rate fraction is fairly well distributed from the 1st to the 9th channels; at the outmost channels (1st and 3rd subchannels) the flow is approximately 95-97% of the average channel flow and at the central channels (4th and 8th subchannels) the flow is about 102-105% of the average channel mass flow rate. It is shown that the measured mass flow distribution is consistent with the results of the numerical calculation except 1st and 17th channels. (author)

  15. Neutron radiography characterization of an operating proton exchange membrane fuel cell with localized current distribution measurements

    International Nuclear Information System (INIS)

    Gagliardo, J.J.; Owejan, J.P.; Trabold, T.A.; Tighe, T.W.

    2009-01-01

    Neutron radiography has proven to be a powerful tool to study and understand the effects of liquid water in an operating fuel cell. In the present work, this experimental method is coupled with locally resolved current and ohmic resistance measurements, giving additional insight into water management and fuel cell performance under a variety of conditions. The effects of varying the inlet humidification level and the current density of the 50 cm 2 cell are studied by simultaneously monitoring electrochemical performance with a 10x10 matrix of current sensors, and liquid water volumes are measured using the National Institute of Standards and Technology (NIST) neutron imaging facility. A counter flow, straight channel proton exchange membrane (PEM) fuel cell is used to demonstrate localized performance loss corresponds to water-filled channels that impede gas transport to the catalyst layer, thereby creating an area that has low current density. Furthermore, certain operating conditions causing excess water accumulation in the channels can result in localized proton resistance increase, a result that can only be accurately observed with combined radiography and distributed electrochemical measurements.

  16. Method for in situ carbon deposition measurement for solid oxide fuel cells

    Science.gov (United States)

    Kuhn, J.; Kesler, O.

    2014-01-01

    Previous methods to measure carbon deposition in solid oxide fuel cell (SOFC) anodes do not permit simultaneous electrochemical measurements. Electrochemical measurements supplemented with carbon deposition quantities create the opportunity to further understand how carbon affects SOFC performance and electrochemical impedance spectra (EIS). In this work, a method for measuring carbon in situ, named here as the quantification of gasified carbon (QGC), was developed. TGA experiments showed that carbon with a 100 h residence time in the SOFC was >99.8% gasified. Comparison of carbon mass measurements between the TGA and QGC show good agreement. In situ measurements of carbon deposition in SOFCs at varying molar steam/carbon ratios were performed to further validate the QGC method, and suppression of carbon deposition with increasing steam concentration was observed, in agreement with previous studies. The technique can be used to investigate in situ carbon deposition and gasification behavior simultaneously with electrochemical measurements for a variety of fuels and operating conditions, such as determining conditions under which incipient carbon deposition is reversible.

  17. Study on dynamic measurement of fuel pellet length during loading into cladding tube

    International Nuclear Information System (INIS)

    Zhang Kai

    1993-09-01

    Various methods are presented for measuring the pellet length in the cladding tube (zirconium tube) during the loading process of the preparation of single rod of nuclear fuel assembly. These methods are used in former Soviet Union, west European countries and China in the manufacturing of nuclear power plant element. Different methods of dynamic measurement by using mechanics, optics and electricity and their special features are analysed and discussed. The structure and measuring principle of a developed measuring device,and its measuring precision and system deviation are also introduced. Finally, the length of loaded pellets is checked with analog pellets. The results are as expected and show that the method and principle used in the measuring device are feasible. It is an ideal and advanced method for the pellet loading of single cladding tube. The principle mentioned above can also be used in other industries

  18. Thermal signature measurements for ammonium nitrate/fuel mixtures by laser heating

    International Nuclear Information System (INIS)

    Nazarian, Ashot; Presser, Cary

    2016-01-01

    Highlights: • LDTR is a useful diagnostic for characterizing AN/fuel mixture thermochemical behavior. • Each AN/fuel mixture thermal signature was different. • AN/fuel mixture signature features were defined by the individual constituents. • Baseline signatures changed after an experiment. - Abstract: Measurements were carried out to obtain thermal signatures of several ammonium nitrate/fuel (ANF) mixtures, using a laser-heating technique referred to as the laser-driven thermal reactor (LDTR). The mixtures were ammonium nitrate (AN)/kerosene, AN/ethylene glycol, AN/paraffin wax, AN/petroleum jelly, AN/confectioner's sugar, AN/cellulose (tissue paper), nitromethane/cellulose, nitrobenzene/cellulose, AN/cellulose/nitromethane, AN/cellulose/nitrobenzene. These mixtures were also compared with AN/nitromethane and AN/diesel fuel oil, obtained from an earlier investigation. Thermograms for the mixtures, as well as individual constituents, were compared to better understand how sample thermal signature changes with mixture composition. This is the first step in development of a thermal-signature database, to be used along with other signature databases, to improve identification of energetic substances of unknown composition. The results indicated that each individual thermal signature was associated unambiguously with a particular mixture composition. The signature features of a particular mixture were shaped by the individual constituent signatures. It was also uncovered that the baseline signature was modified after an experiment due to coating of unreacted residue on the substrate surface and a change in the reactor sphere oxide layer. Thus, care was required to pre-oxidize the sphere prior to an experiment. A minimum sample mass (which was dependent on composition) was required to detect the signature characteristics. Increased laser power served to magnify signal strength while preserving the signature features. For the mixtures examined, the thermal

  19. Uncertainty Quantification of Fork Detector Measurements from Spent Fuel Loading Campaigns

    International Nuclear Information System (INIS)

    Vaccaro, S.; De Baere, P.; Schwalbach, P.; Gauld, I.; Hu, J.

    2015-01-01

    With increasing activities at the end of the fuel cycle, the requirements for the verification of spent nuclear fuel for safeguards purposes are continuously growing. In the European Union we are experiencing a dramatic increase in the number of cask loadings for interim dry storage. This is caused by the progressive shut-down of reactors, related to facility ageing but also due to politically motivated phase-out of nuclear power. On the other hand there are advanced plans for the construction of encapsulation plants and geological repositories. The cask loading or the encapsulation process will provide the last occasion to verify the spent fuel assemblies. In this context, Euratom and the US DOE have carried out a critical review of the widely used Fork measurements method of irradiated assemblies. The Nuclear Safeguards directorates of the European Commission's Directorate General for Energy and Oak Ridge National Laboratory have collaborated to improve the Fork data evaluation process and simplify its use for inspection applications. Within the Commission's standard data evaluation package CRISP, we included a SCALE/ORIGEN-based irradiation and depletion simulation of the measured assembly and modelled the fork transfer function to calculate expected count rates based on operator's declarations. The complete acquisition and evaluation process has been automated to compare expected (calculated) with measured count rates. This approach allows a physics-based improvement of the data review and evaluation process. At the same time the new method provides the means for better measurement uncertainty quantification. The present paper will address the implications of the combined approach involving measured and simulated data to the quantification of measurement uncertainty and the consequences of these uncertainties in the possible use of the Fork detector as a partial defect detection method. (author)

  20. Cladding temperature measurement by thermocouples at preirradiated LWR fuel rod samples

    International Nuclear Information System (INIS)

    Leiling, W.

    1981-12-01

    This report describes the technique to measure cladding temperatures of test fuel rod samples, applied during the in-pile tests on fuel rod failure in the steam loop of the FR2 reactor. NiCr/Ni thermocouples with stainless steel and Inconel sheaths, respectively,of 1 mm diameter were resistance spot weld to the outside of the fuel rod cladding. For the pre-irradiated test specimens, welding had to be done under hot-cell conditions, i.e. under remote handling. In order to prevent the formation of eutectics between zirconium and the chemical elements of the thermocouple sheath at elevated temperatures, the thermocouples were covered with a platinum jacket of 1.4 mm outside diameter swaged onto the sheath in the area of the measuring junction. This thermocouple design has worked satisfactorily in the in-pile experiments performed in a steam atmosphere. Even in the heatup phase, in which cladding temperatures up to 1050 0 C were reached, only very few failures occured. This good performance is to a great part due to a careful control and a thorough inspection of the thermocouples. (orig.) [de

  1. Measurement control design and performance assessment in the Integral Fast Reactor fuel cycle

    International Nuclear Information System (INIS)

    Orechwa, Y.; Bucher, R.G.

    1994-01-01

    The Integral Fast Reactor (IFR)--consisting of a metal fueled and liquid metal cooled reactor together with an attendant fuel cycle facility (FCF)--is currently undergoing a phased demonstration of the closed fuel cycle at Argonne National Laboratory. The recycle technology is pyrometalurgical based with incomplete fission product separation and all transuranics following plutonium for recycle. The equipment operates in batch mode at 500 to 1,300 C. The materials are highly radioactive and pyrophoric, thus the FCF requires remote operation. Central to the material control and accounting system for the FCF are the balances for mass measurements. The remote operation of the balances limits direct adjustment. The radiation environment requires that removal and replacement of the balances be minimized. The uniqueness of the facility precludes historical data for design and performance assessment. To assure efficient operation of the facility, the design of the measurement control system has called for procedures which assess the performance of the balances in great detail and will support capabilities for the correction of systematic changes in the performance of the balances through software

  2. A study on the measurement and evaluation of neutron flux using SPNDs during nuclear fuel irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Son, J. M.; Kim, B. K.; Oh, J. M.; Park, S. J.; Lee, B. H.; Seo, C. G.; Kang, Y. H. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    As a part of the development of instrumentation technologies for a nuclear fuel irradiation test in HANARO(High-Flux Advanced Nuclear Application Reactor), a study is performed to measure and evaluate the neutron flux at the same position as the nuclear fuel during irradiation test using the SPND(Self Powered Neutron Detector). To perform this study, rhodium type SPNDs and amplifier are selected suitable to irradiation test, and the selected SPNDs are installed in instrumented fuel capsule(02F-11K). The irradiation test using a instrumented fuel capsule are performed in the OR5 vertical hole of HANARO for about 54 days, and SPND output signals are acquired successfully during irradiation test. Acquired SPND signals are analyzed and evaluated as a reliable data by COSMOS Code, and this will be utilized for the fuel related research together with fuel center temperature and reactor operation data.

  3. System for nondestructive assay of spent fuel subassemblies: comparison of calculations and measurements

    International Nuclear Information System (INIS)

    Ragan, G.L; Ricker, C.W.; Chiles, M.M.; Ingersoll, D.T.; Slaughter, G.G.; Williams, L.R.

    1979-01-01

    A nondestructive assay system was developed for determining the total fissile content of spent fuel subassemblies at the head end of a reprocessing plant. The system can perform an assay in 20 min with an uncertainty of <5%. Antimony-beryllium neutrons interrogate the subassemblies, and proton recoil counters detect the resulting fission neutrons. Pulse-height discrimination differentiates between the low-energy interrogation neutrons and the higher-energy fission neutrons. Calculated and measured results were compared for (1) interrogation-neutron penetrability, (2) fission-neutron detectability, (3) radial variation of assay sensitivity, (4) axial variation of assay sensitivity, and (5) the variation of detector count rate as a function of the number of fuel rods in a special 61-rod, LMFBR-type subassembly

  4. Reactivity and reaction rate measurements in U--D2O lattices with coaxial fuel

    International Nuclear Information System (INIS)

    Pellarin, D.J.; Morris, B.M.

    1976-12-01

    Integral reaction rate parameters, intracell thermal neutron flux profiles, and material bucklings were measured for D 2 O-moderated uniform lattices in the exponential facility at the Savannah River Laboratory. Two different slightly enriched coaxial uranium fuel assemblies were examined over a wide range of triangular lattice pitches. Integral parameters are reported for inner and outer fuel separately providing data for a more detailed and rigorous comparison with computation than has been previously available. Results are compared with RAHAB calculations using ENDF/B-IV cross sections. Large discrepancies in agreement between calculation and experiment, outside of experimental errors and uncertainties in the input cross sections, probably result from the resonance capture models used by RAHAB

  5. A method of surface area measurement of fuel materials by fission gas release at low temperature

    International Nuclear Information System (INIS)

    Kaimal, K.N.G.; Naik, M.C.; Paul, A.R.; Venkateswarlu, K.S.

    1989-01-01

    The present report deals with the development of a method for surface area measurement of nuclear fuel as well as fissile doped materials by fission gas release study at low temperature. The method is based on the evaluation of knock-out release rate of fission 133 Xe from irradiated fuel after sufficient cooling to decay the short lived activity. The report also describes the fabrication of an ampoule breaker unit for such study. Knock-out release rate of 133 Xe has been studied from UO 2 powders having varying surface area 'S' ranging from 270 cm 2 /gm to 4100 cm 2 /gm at two fissioning rates 10 12 f/cm 3 . sec. and 3.2x10 10 f/cm.sec. A relation between K and A has been established and discussed in this report. (author). 6 refs

  6. Interpretation of measurements made by oscillations of irradiated fuels in natural uranium, graphite-gas piles

    International Nuclear Information System (INIS)

    Laponche, Bernard; Luffin, Jean; Brunet, Max; Guerange, Jacques; Tonolli, Jacky

    1969-06-01

    When considering a pile operation, it is interesting to know the evolution of fuel quality with respect to irradiation, i.e. the variation of its fission rate and of its absorption rate. In order to experimentally obtain these features, a method is to introduce an irradiated cartridge into a critical reactor and to measure the induced effect on its reactivity and on the neutron density at the vicinity of the cartridge. An oscillation method presented in another document and based on a periodic introduction of fuel sample into a critical reactor allows, from the measurement of reactivity variation (global signal), and of the neutron density (local signal), effective macroscopic fission and absorption cross sections of this sample to be obtained. As previous studies revealed that the interpretation of the local signal was notably delicate, this information has been replaced by computed information, the fission rate, which is determined by means of the COREGRAF1 code. Thus, the remaining quantity to be obtained is the fuel absorption rate. The authors report studies performed on several sets of cartridges from different reactors, and with an irradiation range from about 700 to 4000 MWJ/T. In a first part, they describe the characteristics of the studied cartridges, their irradiation and measurement conditions, and the use of the evolution code. In a second part, they try to define the interpretation of oscillation-based measurements by using two methods, a first and fast one which gives an approximation of results, and a more elaborated second one which complies with measurement conditions. The last part presents and discusses the obtained results [fr

  7. Equation of state for L.M.F.B.R. fuel (measurement of fission gas release during transients)

    International Nuclear Information System (INIS)

    Combette, P.; Barthelemy, P.

    1979-01-01

    A sample of fuel (UO 2 or UPuO 2 ) can be heated by fission in a heating transient up to energy deposition 4000 j/g, in the Silene reactor. The Kistler type capsule, the calorimeter device and the radiochemical analysis of fission products enable the pressure pulse and the fuel energy deposition to be measured. So, the relationship between the fuel vapour pressure and the fuel specific energy can be deduced. Peaks pressure (about 1 MPa) coming from fresh UO 2 vaporization, have been measured on a 7 milliseconds time scale. There is a good agreement with the E.O.S. for fresh UO 2 , which is well known for low pressure (1 MPa). Numerous tests have been done with 93% enriched UO 2 and a first test with highly active fuel containing plutonium (15 at %) has been performed. The capsule allows the released gas coming from the irradiated fuel to be retained for measurements and analysis. To investigate the mode of fuel disruption, in-pile fission-heated fuel pellets has been recorded by high speed cinematography

  8. Sensors Based Measurement Techniques of Fuel Injection and Ignition Characteristics of Diesel Sprays in DI Combustion System

    Directory of Open Access Journals (Sweden)

    S. Rehman

    2016-09-01

    Full Text Available Innovative sensor based measurement techniques like needle lift sensor, photo (optical sensor and piezoresistive pressure transmitter are introduced and used to measure the injection and combustion characteristics in direct injection combustion system. Present experimental study is carried out in the constant volume combustion chamber to study the ignition, combustion and injection characteristics of the solid cone diesel fuel sprays impinging on the hot surface. Hot surface ignition approach has been used to create variety of advanced combustion systems. In the present study, the hot surface temperatures were varied from 623 K to 723 K. The cylinder air pressures were 20, 30 and 40 bar and fuel injection pressures were 100, 200 and 300 bar. It is found that ignition delay of fuel sprays get reduced with the rise in injection pressure. The ignition characteristics of sprays much less affected at high fuel injection pressures and high surface temperatures. The fuel injection duration reduces with the increase in fuel injection pressures. The rate of heat release becomes high at high injection pressures and it decreases with the increase in injection duration. It is found that duration of burn/combustion decrease with the increase in injection pressure. The use of various sensors is quite effective, reliable and accurate in measuring the various fuel injection and combustion characteristics. The study simulates the effect of fuel injection system parameters on combustion performance in large heavy duty engines.

  9. Measurement and CFD calculation of spacer loss coefficient for a tight-lattice fuel bundle

    International Nuclear Information System (INIS)

    In, Wang Kee; Shin, Chang Hwan; Kwack, Young Kyun; Lee, Chi Young

    2015-01-01

    Highlights: • Experiment and CFD analysis evaluated the pressure drop in a spacer grid. • The measurement and CFD errors for the spacer loss coefficient were estimated. • The spacer loss coefficient for the dual-cooled annular fuel bundle was determined. • The CFD prediction agrees with the measured spacer loss coefficient within 8%. - Abstract: An experiment and computational fluid dynamics (CFD) analysis were performed to evaluate the pressure drop in a spacer grid for a dual-cooled annular fuel (DCAF) bundle. The DCAF bundle for the Korean optimum power reactor (OPR1000) is a 12 × 12 tight-lattice rod array with a pitch-to-diameter ratio of 1.08 owing to a larger outer diameter of the annular fuel rod. An experiment was conducted to measure the pressure drop in spacer grid for the DCAF bundle. The test bundle is a full-size 12 × 12 rod bundle with 11 spacer grid. The test condition covers a Reynolds number range of 2 × 10 4 –2 × 10 5 by changing the temperature and flow rate of water. A CFD analysis was also performed to predict the pressure drop through a spacer grid using the full-size and partial bundle models. The pressure drop and loss coefficient of a spacer grid were predicted and compared with the experimental results. The CFD predictions of spacer pressure drop and loss coefficient agree with the measured values within 8%. The spacer loss coefficient for the DCAF bundle is estimated to be approximately 1.50 at a nominal operating condition of OPR1000, i.e., Re = 4 × 10 5

  10. Design of an optical thermal sensor for proton exchange membrane fuel cell temperature measurement using phosphor thermometry

    Science.gov (United States)

    Inman, Kristopher; Wang, Xia; Sangeorzan, Brian

    Internal temperatures in a proton exchange membrane (PEM) fuel cell govern the ionic conductivities of the polymer electrolyte, influence the reaction rate at the electrodes, and control the water vapor pressure inside the cell. It is vital to fully understand thermal behavior in a PEM fuel cell if performance and durability are to be optimized. The objective of this research was to design, construct, and implement thermal sensors based on the principles of the lifetime-decay method of phosphor thermometry to measure temperatures inside a PEM fuel cell. Five sensors were designed and calibrated with a maximum uncertainty of ±0.6 °C. Using these sensors, surface temperatures were measured on the cathode gas diffusion layer of a 25 cm 2 PEM fuel cell. The test results demonstrate the utility of the optical temperature sensor design and provide insight into the thermal behavior found in a PEM fuel cell.

  11. Calculations of thermal-reactor spent-fuel nuclide inventories and comparisons with measurements

    International Nuclear Information System (INIS)

    Wilson, W.B.; LaBauve, R.J.; England, T.R.

    1982-01-01

    Comparisons with integral measurements have demonstrated the accuracy of CINDER codes and libraries in calculating aggregate fission-product properties, including neutron absorption, decay power, and decay spectra. CINDER calculations have, alternatively, been used to supplement measured integral data describing fission-product decay power and decay spectra. Because of the incorporation of the extensive actinide library and the use of ENDF/B-V data, it is desirable to compare the inventory of individual nuclides obtained from tandem EPRI-CELL/CINDER-2 calculations with those determined in documented benchmark inventory measurements of spent reactor fuel. The development of the popular 148 Nd burnup measurement procedure is outlined, and areas of uncertainty in it and lack of clarity in its interpretation are indicated. Six inventory samples of varying quality and completeness are examined. The power histories used in the calculations have been listed for other users

  12. Error Analysis of Ceramographic Sample Preparation for Coating Thickness Measurement of Coated Fuel Particles

    International Nuclear Information System (INIS)

    Liu Xiaoxue; Li Ziqiang; Zhao Hongsheng; Zhang Kaihong; Tang Chunhe

    2014-01-01

    The thicknesses of four coatings of HTR coated fuel particle are very important parameters. It is indispensable to control the thickness of four coatings of coated fuel particles for the safety of HTR. A measurement method, ceramographic sample-microanalysis method, to analyze the thickness of coatings was developed. During the process of ceramographic sample-microanalysis, there are two main errors, including ceramographic sample preparation error and thickness measurement error. With the development of microscopic techniques, thickness measurement error can be easily controlled to meet the design requirements. While, due to the coated particles are spherical particles of different diameters ranged from 850 to 1000μm, the sample preparation process will introduce an error. And this error is different from one sample to another. It’s also different from one particle to another in the same sample. In this article, the error of the ceramographic sample preparation was calculated and analyzed. Results show that the error introduced by sample preparation is minor. The minor error of sample preparation guarantees the high accuracy of the mentioned method, which indicates this method is a proper method to measure the thickness of four coatings of coated particles. (author)

  13. Automated methods for real-time analysis of spent-fuel measurement data

    International Nuclear Information System (INIS)

    Bosler, G.E.; Rinard, P.M.; Klosterbuer, S.F.; Painter, J.

    1988-01-01

    Software has been developed for ''real-time'' analysis of neutron and gamma data from GRAND-1/fork measurements on spent-fuel assemblies. Three modules compose the software package. The modules are linked through a database system of files. The first module is part of a general data-base processing code. This module prepares input data files with inventory and correction-factor information for the second module. The second module, called OLAF, operates on a computer attached to the GRAND-1 electronics unit. In this second module, neutron and gamma data from spent-fuel assemblies are analyzed to verify consistency in the facility operator declarations for exposure (burnup) and cooling time. From the analysis, potential discrepancies in the measurement data are questioned while equipment is still installed at the facility and is available for additional measurements. During the measurements, data are written to an output file, called a results file, which can be processed by the third module of the software package. In the third module, printed reports summarizing the data and results are prepared, and neutron and gamma data are written to files that are process by the Deming curve-fitting code

  14. Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

    Energy Technology Data Exchange (ETDEWEB)

    Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

    2017-08-01

    Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (< 20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. This paper discusses the TEM results of the U-10Mo/Zr/Al6061 monolithic fuel plate (Plate ID: L1P09T, ~ 59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory as part of RERTR-9B irradiation campaign with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 C, respectively. A total of 5 TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (> 1 µm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ~ 30 at% and ~ 7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

  15. Establishing the fuel burn-up measuring system for 106 irradiated assemblies of Dalat reactor by using gamma spectrometer method

    International Nuclear Information System (INIS)

    Nguyen Minh Tuan; Pham Quang Huy; Tran Tri Vien; Trang Cao Su; Tran Quoc Duong; Dang Tran Thai Nguyen

    2013-01-01

    The fuel burn-up is an important parameter needed to be monitored and determined during a reactor operation and fuel management. The fuel burn-up can be calculated using computer codes and experimentally measured. This work presents the theory and experimental method applied to determine the burn-up of the irradiated and 36% enriched VVR-M2 fuel type assemblies of Dalat reactor. The method is based on measurement of Cs-137 absolute specific activity using gamma spectrometer. Designed measuring system consists of a collimator tube, high purity Germanium detector (HPGe) and associated electronics modules and online computer data acquisition system. The obtained results of measurement are comparable with theoretically calculated results. (author)

  16. Field test and evaluation of the IAEA coincidence collar for the measurement of unirradiated BWR fuel assemblies

    International Nuclear Information System (INIS)

    Menlove, H.O.; Keddar, A.

    1982-12-01

    The neutron coincidence counter has been field tested and evaluated for the measurement of boiling-water-reactor (BWR) fuel assemblies at the ASEA-ATOM Fuel Fabrication Facility. The system measures the 235 U content per unit length of full fuel assemblies using neutron interrogation and coincidence counting. The 238 U content is measured in the passive mode without the AmLi neutron interrogatioin source. The field tests included both standard production movable fuel rods to investigate enrichment and absorber variations. Results gave a response standard deviation of 0.9% for the active case and 2.1% for the passive case in 1000-s measurement times. 10 figures, 2 tables

  17. The value of retrofitting carbon-saving measures into fuel poor social housing

    International Nuclear Information System (INIS)

    Jenkins, D.P.

    2010-01-01

    With current fuel poverty and carbon-saving policies continuing to miss their targets in the UK, the synergy between the two problems is investigated to highlight an approach that could be mutually beneficial. Focussing on the 550,000 fuel poor socially housed dwellings in the UK, costs of between Pounds 3.9 and Pounds 17.5 bn are estimated as the required capital investment for achieving deep-cut carbon savings (defined as at least 50%) across this section of the housing stock, with a potential total annual carbon saving of 1.7 MtCO 2 . It is assumed that such costs would be largely (or totally) state-funded, though additional private investment could clearly increase the possible carbon savings across this section of the stock. The use of these socially housed fuel poor dwellings as low-carbon exemplars is discussed, and benefits for the private housing sector are postulated. The study also focuses on the problem of installing non-cost effective measures, i.e. technologies that would not currently be encouraged by existing subsidy schemes, but which might be necessary for achieving large carbon-saving targets.

  18. Fuel ion rotation measurement and its implications on H-mode theories

    International Nuclear Information System (INIS)

    Kim, J.; Burrell, K.H.; Gohil, P.; Groebner, R.J.; Hinton, F.L.; Kim, Y.B.; Seraydarian, R.; Mandl, W.

    1993-10-01

    Poloidal and toroidal rotation of the fuel ions (He 2+ ) and the impurity ions (C 6+ and B 5+ ) in H-mode helium plasmas have been investigated in the DIII-D tokamak by means of charge exchange recombination spectroscopy, resulting in the discovery that the fuel ion poloidal rotation is in the ion diamagnetic drift direction while the impurity ion rotation is in the electron diamagnetic drift direction. The radial electric field obtained from radial force balance analysis of the measured pressure gradients and rotation velocities is shown to be the same regardless of which ion species is used and therefore is a more fundamental parameter than the rotation flows in studying H-mode phenomena. It is shown that the three contributions to the radial electric field (diamagnetic, poloidal rotation, and toroidal rotation terms) are comparable and consequently the poloidal flow does not solely represent the E x B flow. In the high-shear edge region, the density scale length is comparable to the ion poloidal gyroradius, and thus neoclassical theory is not valid there. In view of this new discovery that the fuel and impurity ions rotate in opposite sense, L-H transition theories based on the poloidal rotation may require improvement

  19. Spent fuel test - Climax: technical measurements. Interim report, fiscal year 1981

    International Nuclear Information System (INIS)

    Patrick, W.C.; Ballou, L.B.; Butkovich, T.R.

    1982-01-01

    The Spent Fuel Test-Climax (SFT-C) is located 420 m below surface in the Climax granite stock on the Nevada Test Site. Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized from April to May 1980, initiating the 3- to 5-year-duration test. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. Technical objectives of the test led to development of a technical measurements program, which is the subject of this report. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the first 1-1/2 years of the test on more than 900 channels. Much of the acquired data are now available for analysis and are presented here. Highlights of activities this year include completion of site characterization field work, major modifications to the data acquisition and the management systems, and the addition of instrument evaluation as an explicit objective of the test

  20. Spent fuel test - Climax: technical measurements. Interim report, fiscal year 1981

    Energy Technology Data Exchange (ETDEWEB)

    Patrick, W.C.; Ballou, L.B.; Butkovich, T.R.

    1982-04-30

    The Spent Fuel Test-Climax (SFT-C) is located 420 m below surface in the Climax granite stock on the Nevada Test Site. Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized from April to May 1980, initiating the 3- to 5-year-duration test. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. Technical objectives of the test led to development of a technical measurements program, which is the subject of this report. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the first 1-1/2 years of the test on more than 900 channels. Much of the acquired data are now available for analysis and are presented here. Highlights of activities this year include completion of site characterization field work, major modifications to the data acquisition and the management systems, and the addition of instrument evaluation as an explicit objective of the test.

  1. Gross gamma-ray measurements of light water reactor spent-fuel assemblies in underwater storage arrays

    International Nuclear Information System (INIS)

    Moss, C.E.; Lee, D.M.

    1980-12-01

    Two gross gamma-ray detection systems have been developed for rapid measurement of spent-fuel assemblies in underwater storage racks. One system uses a scintillator as the detector and has a 2% crosstalk between a fuel assembly and an adjacent void. The other system uses an ion chamber as the detector. The measurements with both detectors correlate well with operator-declared burnup and cooling-time values

  2. Determination of a Wear Initiation Cycle by using a Contact Resistance Measurement in Nuclear Fuel Fretting

    International Nuclear Information System (INIS)

    Lee, Young Ho; Kim, Hyung Kyu

    2008-01-01

    In nuclear fuel fretting, the improving of the contact condition with a modified spring shape is a useful method for increasing the wear resistance of the nuclear fuel rod. This is because the fretting wear resistance between the fuel rod and grid spring is mainly affected by the grid spring shape rather than the environment, the contact modes, etc. In addition, the wear resistance is affected by the wear debris behavior between contact surfaces. So, it is expected that the wear initiation of each spring shape should be determined in order to evaluate a wear resistance. However, it is almost impossible to measure the wear behavior in contact surfaces on a real time basis because the contact surfaces are always hidden. Besides, the results of the worn surface observation after the fretting wear tests are restricted to archive the information on the wear debris behavior and the formation mechanism of the wear scar. In order to evaluate the wear behavior during the fretting wear tests, it is proposed that the contact resistance measurement is a useful method for examining the wear initiation cycle and modes. Generally, fretting wear damages are rapidly progressed by a localized plastic deformation between the contact surfaces, crack initiation and fracture of the deformed surface with a strain hardening difference between a surface and a subsurface and finally a detachment of wear debris. After this, wear debris is easily oxidized by frictional heat, test environment, etc. At this time, a small amount of electric current applied between the contact surfaces will be influenced by the wear debris, which could be an obstacle to an electric current flow. So, it is possible to archive the information on the wear behavior by measuring the contact resistance. In order to determine the wear initiation cycle during the fretting wear tests, in this study, fretting wear tests have been performed by applying a constant electric current in room temperature air

  3. Production and measurement of minor actinides in the commercial fuel cycle

    International Nuclear Information System (INIS)

    Stanbro, W.D.

    1997-03-01

    The minor actinide elements, particularly neptunium and americium, are produced as a normal byproduct of the operation of thermal power reactors. Because of the existence of long-lived isotopes of these elements, they constitute the major sources of the residual radiation in spent fuel or in wastes resulting from reprocessing. This has led to examinations by some countries of the possibility of separating the minor actinides from waste products. The papers found in this report address the production of minor actinides in common thermal power reactors as well as approaches to measure these materials in various media. The first paper in this volume, open-quotes Production of Minor Actinides in the Commercial Fuel Cycle,close quotes uses calculations with the ORIGEN2 reactor and decay code to estimate the amounts of minor actinides in spent fuel and separated plutonium as a function of reactor irradiation and the time after discharge. The second paper, open-quotes Destructive Assay of Minor Actinides,close quotes describes a number of promising approaches for the chemical analysis of minor actinides in the various forms in which they are found at reprocessing plants. The next paper, open-quotes Hybrid KED/XRF Measurement of Minor Actinides in Reprocessing Plants,close quotes uses the results of a simulation model to examine the possible applications of the hybrid KED/XRF instrument to the determination of minor actinides in some of the solutions found in reprocessing plants. In open-quotes Calorimetric Assay of Minor Actinides,close quotes the authors show some possible extensions of this powerful technique beyond the normal plutonium assays to include the minor actinides. Finally, the last paper in this volume, open-quotes Environment Measurements of Transuranic Nuclides,close quotes discusses what is known about the levels of the minor actinides in the environment and ways to analyze for these materials in environmental matrices

  4. Measurement of PuO2 spot in the mixed-oxide fuels

    International Nuclear Information System (INIS)

    Yokosuka, Yoshifumi; Kaneda, Kenichiro; Ishikawa, Eiji; Nakajima, Katsuaki; Kashima, Sadamitsu

    1974-01-01

    The homogeneity of mixed-oxide fuel is evaluated with the maximum diameter of PuO 2 spots. PuO 2 spot distribution is influenced by fuel production process and raw materials used. Analytical measurement of the PuO 2 spot distribution was carried out. α-autoradiograph was taken with nitrocellulose film as a detector. The film was exposed for 15 and 30 minutes to the sample pellets, and etched in 10% NaOH for 2 minutes at 60 0 C. Measurement of PuO 2 spots was performed manually with a projector. PuO 2 spot distribution was measured. There is no difference of spot distribution owing to the location in a pellet. However some discrepancy exists in spot number and maximum diameter between surface and inner part. This phenomenon is explained by considering that the diffusion on surface is more effective than in inner part in sintering process. Three pellets were taken out from a lot at random, and α-autoradiograph was taken longitudinally with a pellet and transversely with the others. PuO 2 spot number counted from three autoradiographs was about 150/cm 2 , and the standard deviation was about 1.5. No difference was observed among them. Concerning the influence of mixing method on PuO 2 spot formation, three methods were examined; ball mill mixing (with and without lining) and Kneader mixing. Kneader mixing gave the worst results in maximum spot diameter and spot distribution range. (Tai, I.)

  5. System for measuring spacer pin pitch in a nuclear fuel assembly

    International Nuclear Information System (INIS)

    Isono, Kenji; Tateishi, Yoshinori; Mano, Tadashi.

    1975-01-01

    Object: To reduce the period for discriminating whether or not spacer pin pitch is satisfactory by simultaneously inserting a number of reference rods into a nuclear fuel assembly spacer ring element of a reactor and arranging them such that they can be simultaneously withdrawn to simplify the withdrawing operation. Structure: A spacer provided with a ring element which clamps a nuclear fuel element is supported on a spacer support with a rod secured to the support as a guide and is secured to the support by securing means. A vertically movable structure with a reference rod provided upright and thru-holes formed in two support plates provided in the same row as the spacer ring element is operated by a fluid pressure mechanism to simultaneously insert the reference rod into the spacer ring element. The reference rod is mounted in support plates via ball bearings such that it is slightly movable in the horizontal direction, and it is aligned with respect to the core of the ring element. The intercore distance of the reference rod is measured with the reference rod inserted in the ring element, thereby measuring the space pin pitch. From the results of measurement, discrimination as to whether the spacer is satisfactory or not is made. (Kamimura, M.)

  6. Heat transfer and pressure measurements for the SSME fuel-side turbopump

    Science.gov (United States)

    Dunn, Michael G.

    1990-01-01

    A measurement program is currently underway at the Calspan-UB Research Center (CUBRC) which utilizes the Rocketdyne two-state fuel-side turbine with the engine geometric configuration reproduced. This is a full two-state turbine for which the vane rows and the blades are the engine hardware currently used on the Space Shuttle turbopump. A status report is provided for the experimental program and a description of the instrumentation and the measurements to be performed. The specific items that will be illustrated and described are as follows: (1) the gas flow path, (2) the heat-flux instrumentation, (3) the surface-pressure instrumentation, (4) the experimental conditions for which data will be obtained, and (5) the specific measurements that will be performed.

  7. Relative risk measure suitable for comparison of design alternatives of interim spent nuclear fuel storage facility

    International Nuclear Information System (INIS)

    Ferjencik, M.

    1997-01-01

    Accessible reports on risk assessment of interim spent nuclear fuel storage facilities presume that only releases of radioactive substances represent undesired consequences. However, only certain part of the undesired consequences is represented by them. Many other events are connected with safety and are able to cause losses to the operating company. The following two presumptions are pronounced based on this. 1. Any event causing a disturbance of a safety function of the storage facility is an incident event. 2. Any disturbance of a safety function is an undesired consequence. If the facility safety functions are identified and if the severity of their disturbances is quantified, then it is possible to combine consequence severity quantifications and event frequencies into a risk measure. Construction and application of such a risk measure is described in this paper. The measure is shown to be a tool suitable for comparison of interim storage technology design alternatives. (author)

  8. Measurement of particle size distribution and mass concentration of nuclear fuel aerosols

    International Nuclear Information System (INIS)

    Pickering, S.

    1982-01-01

    The particle size distribution and particle mass concentration of a nuclear fuel aerosol is measured by admitting the aerosol into a vertically-extending container, positioning an alpha particle detector within the container so that its window is horizontal and directed vertically, stopping the admission of aerosol into the container, detecting the alpha-activity of the particles of the aerosol sedimenting onto the detector window (for example in a series of equal time intervals until a constant level is reached), and converting the alpha-activity measurements into particle size distribution and/or particle mass concentration measurements. The detector is attached to a pivotted arm and by raising a counterweight can be lowered from the container for cleaning. (author)

  9. Mind the Gap: The Vicious Circle of Measuring Automobile Fuel Use

    DEFF Research Database (Denmark)

    Figueroa, Maria; author), L. Schipper (main; Price, L.

    1993-01-01

    We review the circularity between estimates of automobile use, fuel consumption and fuel intensity. We find that major gaps exist between estimates of road gasoline, the quantity most often used to represent automobile fuel use in economic studies of transport fuel use, and the actual sales data...

  10. Status of the nuclear measurement stations for the process control of spent fuel reprocessing at AREVA NC/La Hague

    International Nuclear Information System (INIS)

    Eleon, Cyrille; Passard, Christian; Hupont, Nicolas; Estre, Nicolas; Battel, Benjamin; Doumerc, Philippe; Dupuy, Thierry; Batifol, Marc; Grassi, Gabriele

    2015-01-01

    Nuclear measurements are used at AREVA NC/La Hague for the monitoring of spent fuel reprocessing. The process control is based on gamma-ray spectroscopy, passive neutron counting and active neutron interrogation, and gamma transmission measurements. The main objectives are criticality and safety, online process monitoring, and the determination of the residual fissile mass and activities in the metallic waste remained after fuel shearing and dissolution (empty hulls, grids, end pieces), which are put in radioactive waste drums before compaction. The whole monitoring system is composed of eight measurement stations which will be described in this paper. The main measurement stations no. 1, 3 and 7 are needed for criticality control. Before fuel element shearing for dissolution, station no. 1 allows determining the burn-up of the irradiated fuel by gamma-ray spectroscopy with HP Ge (high purity germanium) detectors. The burn-up is correlated to the 137 Cs and 134 Cs gamma emission rates. The fuel maximal mass which can be loaded in one bucket of the dissolver is estimated from the lowest burn-up fraction of the fuel element. Station no. 3 is dedicated to the control of the correct fuel dissolution, which is performed with a 137 Cs gamma ray measurement with a HP Ge detector. Station no. 7 allows estimating the residual fissile mass in the drums filled with the metallic residues, especially in the hulls, from passive neutron counting (spontaneous fission and alpha-n reactions) and active interrogation (fission prompt neutrons induced by a pulsed neutron generator) with proportional 3 He detectors. The measurement stations have been validated for the reprocessing of Uranium Oxide (UOX) fuels with a burn-up rate up to 60 GWd/t. This paper presents a brief overview of the current status of the nuclear measurement stations. (authors)

  11. Status of the nuclear measurement stations for the process control of spent fuel reprocessing at AREVA NC/La Hague

    Energy Technology Data Exchange (ETDEWEB)

    Eleon, Cyrille; Passard, Christian; Hupont, Nicolas; Estre, Nicolas [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 St Paul-lez-Durance (France); Battel, Benjamin; Doumerc, Philippe; Dupuy, Thierry; Batifol, Marc [AREVA NC, La Hague plant - Nuclear Measurement Team, F-50444 Beaumont-Hague (France); Grassi, Gabriele [AREVA NC, 1 place Jean-Millier, 92084 Paris-La-Defense cedex (France)

    2015-07-01

    Nuclear measurements are used at AREVA NC/La Hague for the monitoring of spent fuel reprocessing. The process control is based on gamma-ray spectroscopy, passive neutron counting and active neutron interrogation, and gamma transmission measurements. The main objectives are criticality and safety, online process monitoring, and the determination of the residual fissile mass and activities in the metallic waste remained after fuel shearing and dissolution (empty hulls, grids, end pieces), which are put in radioactive waste drums before compaction. The whole monitoring system is composed of eight measurement stations which will be described in this paper. The main measurement stations no. 1, 3 and 7 are needed for criticality control. Before fuel element shearing for dissolution, station no. 1 allows determining the burn-up of the irradiated fuel by gamma-ray spectroscopy with HP Ge (high purity germanium) detectors. The burn-up is correlated to the {sup 137}Cs and {sup 134}Cs gamma emission rates. The fuel maximal mass which can be loaded in one bucket of the dissolver is estimated from the lowest burn-up fraction of the fuel element. Station no. 3 is dedicated to the control of the correct fuel dissolution, which is performed with a {sup 137}Cs gamma ray measurement with a HP Ge detector. Station no. 7 allows estimating the residual fissile mass in the drums filled with the metallic residues, especially in the hulls, from passive neutron counting (spontaneous fission and alpha-n reactions) and active interrogation (fission prompt neutrons induced by a pulsed neutron generator) with proportional {sup 3}He detectors. The measurement stations have been validated for the reprocessing of Uranium Oxide (UOX) fuels with a burn-up rate up to 60 GWd/t. This paper presents a brief overview of the current status of the nuclear measurement stations. (authors)

  12. Measurements and correlations of turbulent burning velocities over wide ranges of fuels and elevated pressures

    KAUST Repository

    Bradley, Derek; Lawes, Malcolm; Liu, Kexin; Mansour, Morkous S.

    2013-01-01

    The implosion technique has been used to extend measurements of turbulent burning velocities over greater ranges of fuels and pressures. Measurements have been made up to 3.5 MPa and at strain rate Markstein numbers as low as 23. The implosion technique, with spark ignition at two opposite wall positions within a fan-stirred spherical bomb is capable of measuring turbulent burning velocities, at higher pressures than is possible with central ignition. Pressure records and schlieren high speed photography define the rate of burning and the smoothed area of the flame front. The first aim of the study was to extend the previous measurements with ethanol and propane-air, with further measurements over wider ranges of fuels and equivalence ratios with mixtures of hydrogen, methane, 10% hydrogen-90% methane, toluene, and i-octane, with air. The second aim was to study further the low turbulence regime in which turbulent burning co-exists with laminar flame instabilities. Correlations are presented of turbulent burning velocity normalised by the effective rms turbulent velocity acting on the flame front, ut=u0k , with the Karlovitz stretch factor, K, for different strain rate Markstein numbers, a decrease in which increases ut=u0k . Experimental correlations are presented for the present measurements, combined with previous ones. Different burning regimes are also identified, extending from that of mixed turbulence/laminar instability at low values of K to that at high values of K, in which ut=u0k is gradually reduced due to increasing localised flame extinctions. © 2012 The Combustion Institute.

  13. Development of a new measurement method for fast breeder reactor fuel burnup using a shielded ion microprobe analyzer

    International Nuclear Information System (INIS)

    Mizuno, M.; Enokido, Y.; Itaki, T.; Kono, K.; Unno, I.; Yamanouchi, S.

    1985-01-01

    A new method of burnup measurement using a shielded ion microprobe analyzer (SIMA) has been developed. The method is based on the isotope analysis of uranium, plutonium, and fission products in irradiated mixed oxide fuel by means of secondary ion mass spectrometry (SIMS). Fourteen samples irradiated in the Japanese experimental fast reactor JOYO were examined. The maximum local burnup of JOYO MK-I core fuels was about5.1 at. %. The axial burnup distribution of the fuel pin was in good agreement with that of the sibling pin in the same subassembly, measured by surface ionization mass spectrometry, which requires the chemical separation of fission products and heavy metals. The new method facilitates the rapid and accurate measurement of fast breeder reactor fuel burnup without human radiation exposure during sample preparation and analysis

  14. Applying hot-wire anemometry to directly measure the water balance in a proton exchange membrane fuel cell

    DEFF Research Database (Denmark)

    Al Shakhshir, Saher; Andreasen, Søren Juhl; Berning, Torsten

    2016-01-01

    In order to better understand and more accurately measure the water balance in a proton exchange membrane fuel cell, our group has recently proposed to apply hot wire anemometry in the fuel cell's anode outlet. It was theoretically shown that the electrical signal obtained from the hot wire sensor...... can be directly converted into the fuel cell water balance. In this work an ex-situ experimental investigation is performed to examine the effect of the wire diameter and the outlet pipe diameter on the voltage signal. For a laboratory fuel cell where the mass flow rate the anode outlet is small...... number Nu range between m = 0.137 and m = 0.246. In general, it is shown that applying hot wire anemometry yields in fact very clear voltage readings with high frequency, and it can be used as a diagnosis tool in various fuel cell applications....

  15. Effect of Grain Refinement on Structure Evolution, “Floating” Grains, and Centerline Macrosegregation in Direct-Chill Cast AA2024 Alloy Billets

    NARCIS (Netherlands)

    Nadella, R.; Eskin, D.G.; Katgerman, L.

    2007-01-01

    Direct-chill (DC) cast billets 192 mm in diameter of an Al-Cu-Mg alloy were examined in detail with the aim to reveal the effects of grain refining (GR) and casting speed on structure, “floating” grains, and centerline macrosegregation. Experimental results show that grain size and dendrite arm

  16. METHODS TO MEASURE, PREDICT AND RELATE FRICTION, WEAR AND FUEL ECONOMY

    Energy Technology Data Exchange (ETDEWEB)

    Gravante, Steve [Ricardo, Inc.; Fenske, George [Argonne National Lab. (ANL), Argonne, IL (United States); Demas, Nicholas [Argonne National Lab. (ANL), Argonne, IL (United States); Erck, Robert [Argonne National Lab. (ANL), Argonne, IL (United States)

    2018-03-19

    High-fidelity measurements of the coefficient of friction and the parasitic friction power of the power cylinder components have been made for the Isuzu 5.2L 4H on-highway engine. In particular, measurements of the asperity friction coefficient were made with test coupons using Argonne National Lab’s (ANL) reciprocating test rig for the ring-on-liner and skirt-on-liner component pairs. These measurements correlated well with independent measurements made by Electro-Mechanical Associates (EMA). In addition, surface roughness measurements of the Isuzu components were made using white light interferometer (WLI). The asperity friction and surface characterization are key inputs to advanced CAE simulation tools such as RINGPAK and PISDYN which are used to predict the friction power and wear rates of power cylinder components. Finally, motored friction tests were successfully performed to quantify the friction mean effective pressure (FMEP) of the power cylinder components for various oils (High viscosity 15W40, low viscosity 5W20 with friction modifier (FM) and specially blended oil containing consisting of PAO/ZDDP/MoDTC) at 25, 50, and 110 °C. Ricardo's objective is to use this data along with advanced CAE methods to develop empirical characterizations of friction and wear mechanisms in internal combustion engines such that the impact of such mechanisms of engine fuel consumption and/or vehicle fuel economy can be estimated. The value of such predictive schemes is that if one knows how a particular friction reduction technology changes oil viscosity and/or the friction coefficient then the fuel consumption or fuel economy impacts can be estimated without the excessive cost of motored or fired engine tests by utilizing cost effective lab scale tests and in combination with advanced analytical methods. One accomplishment made during this work was the development and validation of a novel technique for quantifying wear using data from WLI through the use of

  17. An evaluation of the influence of fuel design parameters and burnup on pellet/cladding interaction for boiling water reactor fuel rod through in-core diameter measurement

    International Nuclear Information System (INIS)

    Yanagisawa, K.

    1986-01-01

    The influence of design parameters and burning on pellet/cladding interaction (PCI) of current boiling water reactor fuel rods was studied through in-core diameter measurement. Thinner cladding and a smaller diametral gap enhanced the PCI during startup. At constant power, fuel with SiO 2 added greatly reduced PCI due to relaxation. The fuel with a small grain size greatly reduced PCI due to densification. Preirradiation of rods up to 23 MWd/kgU caused a large PCI not only in a small gap but also in a large gap rod. Relaxation and permanent deformation was small. In the power increase experiment, one rod experienced PCI failure. The spurt times of coolant radioactivity coincided well with the sudden drop of cladding axial strain and marked crack opening at the rod surface. The estimated hoop stress predicted by FEMAXI-III was 350 MPa at the failure

  18. O/M ratio measurement in pure and mixed oxide fuels - where are we now?

    Energy Technology Data Exchange (ETDEWEB)

    Rubin, J.; Chidester, K.; Thompson, M. [Los Alamos National Lab., NM (United States)

    2001-07-01

    The scale-down in the US and Russian nuclear weapons stockpiles has produced a surplus of weapons grade plutonium and highly enriched uranium. The incorporation into mixed-oxide fuel (MOX) is one of the currently favored routes for surplus weapons-grade plutonium. The use of MOX as a nuclear reactor fuel is well established, particularly in Europe and Japan but not in the US. The primary purpose of this investigation was to evaluate existing analytical techniques for their applicability to O/M (oxygen-to-metal ratio) measurements of MOX derived from excess weapons plutonium. The second objective of this investigation was to bring up-to-date the literature on O/M measurement methods, which has not been undertaken in over 20 years. There are several classification schemes that can be used to organize O/M measurement methods. The most popular schemes are based on (a) whether the analysis is performed in solution (wet chemical) or on solid material (dry), and (b) whether the concentration of major constituents are analyzed directly (direct) or are inferred (indirect). Solid state coulometric titration is currently used extensively in studies of phase equilibria, defect chemistry, thermochemical measurement of oxides, including ferrites. Regardless of which indirect method is used (solid state coulometric titration or thermogravimetry), a primary, direct method will also be required for the establishment of the MO{sub 2} reference state, determination of method bias, and periodic calibration. It was recommended that the following direct method be adapted for this purpose: oxygen measurement by inert gas fusion/carbon reduction, and total U, Pu by controlled potential coulometry. In a table are listed the experimental values of accuracy for about 30 O/M methods. (A.C.)

  19. The lumped parameter model for fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Liu, W S [Ontario Hydro, Toronto, ON (Canada)

    1996-12-31

    The use of a lumped fuel-pin model in a thermal-hydraulic code is advantageous because of computational simplicity and efficiency. The model uses an averaging approach over the fuel cross section and makes some simplifying assumptions to describe the transient equations for the averaged fuel, fuel centerline and sheath temperatures. It is shown that by introducing a factor in the effective fuel conductivity, the analytical solution of the mean fuel temperature can be modified to simulate the effects of the flux depression in the heat generation rate and the variation in fuel thermal conductivity. The simplified analytical method used in the transient equation is presented. The accuracy of the lumped parameter model has been compared with the results from the finite difference method. (author). 4 refs., 2 tabs., 4 figs.

  20. Three-dimensional imaging, change detection, and stability assessment during the centerline trench levee seepage experiment using terrestrial light detection and ranging technology, Twitchell Island, California, 2012

    Science.gov (United States)

    Bawden, Gerald W.; Howle, James; Bond, Sandra; Shriro, Michelle; Buck, Peter

    2014-01-01

    A full scale field seepage test was conducted on a north-south trending levee segment of a now bypassed old meander belt on Twitchell Island, California, to understand the effects of live and decaying root systems on levee seepage and slope stability. The field test in May 2012 was centered on a north-south trench with two segments: a shorter control segment and a longer seepage test segment. The complete length of the trench area measured 40.4 meters (m) near the levee centerline with mature trees located on the waterside and landside of the levee flanks. The levee was instrumented with piezometers and tensiometers to measure positive and negative porewater pressures across the levee after the trench was flooded with water and held at a constant hydraulic head during the seepage test—the results from this component of the experiment are not discussed in this report. We collected more than one billion three-dimensional light detection and ranging (lidar) data points before, during, and after the centerline seepage test to assess centimeter-scale stability of the two trees and the levee crown. During the seepage test, the waterside tree toppled (rotated 20.7 degrees) into the water. The landside tree rotated away from the levee by 5 centimeters (cm) at a height of 2 m on the tree. The paved surface of the levee crown had three regions that showed subsidence on the waterside of the trench—discussed as the northern, central, and southern features. The northern feature is an elongate region that subsided 2.1 cm over an area with an average width of 1.35 m that extends 15.8 m parallel to the trench from the northern end of the trench to just north of the trench midpoint, and is associated with a crack 1 cm in height that formed during the seepage test on the trench wall. The central subsidence feature is a semicircular region on the waterside of the trench that subsided by as much as 6.2 cm over an area 3.4 m wide and 11.2 m long. The southern feature is an elongate

  1. Measurement of fuel corrosion products using planar laser-induced fluorescence

    International Nuclear Information System (INIS)

    Wantuck, P.J.; Sappey, A.D.; Butt, D.P.

    1993-01-01

    Characterizing the corrosion behavior of nuclear fuel material in a high-temperature hydrogen environment is critical for ascertaining the operational performance of proposed nuclear thermal propulsion (NTP) concepts. In this paper, we describe an experimental study undertaken to develop and test non-intrusive, laser-based diagnostics for ultimately measuring the distribution of key gas-phase corrosion products expected to evolve during the exposure of NTP fuel to hydrogen. A laser ablation technique is used to produce high temperature, vapor plumes from uranium-free zirconium carbide (ZrC) and niobium carbide (NbC) forms for probing by various optical diagnostics including planar laser-induced fluorescence (PLIF). We discuss the laser ablation technique, results of plume emission measurements, and we describe both the actual and proposed planar LIF schemes for imaging constituents of the ablated ZrC and NbC plumes. Envisioned testing of the laser technique in rf-heated, high temperature gas streams is also discussed

  2. Expected total counts for the Self-Interrogation Neutron Resonance Densitometry measurements of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Rossa, Riccardo [Belgian nuclear research centre SCK.CEN (Belgium); Universite Libre de Bruxelles (Belgium); Borella, Alessandro; Van der Meer, Klaas [Belgian nuclear research centre SCK.CEN. Boeretang 200, 2400 Mol (Belgium); Labeau, Pierre-Etienne; Pauly, Nicolas [Universite Libre de Bruxelles. Av. F. D. Roosevelt 50, B1050 Brussels (Belgium)

    2015-07-01

    The Self-Interrogation Neutron Resonance Densitometry (SINRD) is a passive neutron technique that aims at a direct quantification of {sup 239}Pu in spent fuel assemblies by measuring the attenuation of the neutron flux in the energy region close to the 0.3 eV resonance of {sup 239}Pu. The {sup 239}Pu mass is estimated by calculating the SINRD signature, that is the ratio between the neutron counts in the fast energy region and around the 0.3 eV resonance region. The SINRD measurement approach in this study consisted in introducing a small neutron detector in the central guide tube of a PWR 17x17 fuel assembly. In order to measure the neutron flux in the energy regions defined in the SINRD signature, different detector types were used. The response of a bare {sup 238}U fission chamber is considered for the determination of the fast neutron flux, while other thermal-epithermal detectors wrapped in neutron absorbers are envisaged to measure the neutron flux around the resonance region. This paper provides an estimation of the total neutron counts that can be achieved with the detector types proposed for the SINRD measurement. In the first section a set of detectors are evaluated in terms of total neutron counts and sensitivity to the {sup 239}Pu content, in order to identify the optimal measurement configuration for each detector type. Then a study is performed to increase the total neutron counts by increasing the detector size. The study shows that the highest total neutron counts are achieved by using either {sup 3}He or {sup 10}B proportional counters because of the high neutron efficiency of these detectors. However, the calculations indicate that the biggest contribution to the measurement uncertainty is due to the measurement of the fast neutron flux. Finally, similar sensitivity to the {sup 239}Pu content is obtained by using the different detector types for the measurement of the neutron flux close to the resonance region. Therefore, the total neutron counts

  3. Expected count rate for the Self- Interrogation Neutron Resonance Densitometry measurements of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Rossa, Riccardo [SCK.CEN, Belgian Nuclear Research Centre, Boeretang, 200 - B2400 Mol (Belgium); Universite libre de Bruxelles, Ecole polytechnique de Bruxelles - Service de Metrologie Nucleaire, CP 165/84, Avenue F.D. Roosevelt, 50 - B1050 Brussels (Belgium); Borella, Alessandro; Van der Meer, Klaas [SCK.CEN, Belgian Nuclear Research Centre, Boeretang, 200 - B2400 Mol (Belgium); Labeau, Pierre-Etienne; Pauly, Nicolas [Universite libre de Bruxelles, Ecole polytechnique de Bruxelles - Service de Metrologie Nucleaire, CP 165/84, Avenue F.D. Roosevelt, 50 - B1050 Brussels (Belgium)

    2015-07-01

    The Self-Interrogation Neutron Resonance Densitometry (SINRD) is a passive neutron technique that aims at a direct quantification of {sup 239}Pu in the fuel assemblies by measuring the attenuation of the neutron flux in the energy region close to the 0.3 eV resonance of {sup 239}Pu. The {sup 239}Pu mass is estimated by calculating the SINRD signature, that is the ratio between the neutron flux integrated over the fast energy region and around the 0.3 eV resonance region. The SINRD measurement approach considered in this study consists in introducing a small neutron detector in the central guide tube of a PWR 17x17 fuel assembly. In order to measure the neutron flux in the energy regions defined in the SINRD signature, different detector types are used. The response of a bare {sup 238}U fission chamber is considered for the determination of the fast neutron flux, while other thermal-epithermal detectors wrapped in neutron absorbers are envisaged to measure the neutron flux around the resonance region. This paper provides an estimation of the count rate that can be achieved with the detector types proposed for the SINRD measurement. In the first section a set of detectors are evaluated in terms of count rate and sensitivity to the {sup 239}Pu content, in order to identify the optimal measurement configuration for each detector type. Then a study is performed to increase the count rate by increasing the detector size. The study shows that the highest count rate is achieved by using either {sup 3}He or {sup 10}B proportional counters because of the high neutron efficiency of these detectors. However, the calculations indicate that the biggest contribution to the measurement uncertainty is due to the measurement of the fast neutron flux. Finally, similar sensitivity to the {sup 239}Pu content is obtained by using the different detector types for the measurement of the neutron flux close to the resonance region. Therefore, the count rate associated to each detector type

  4. Thermal analysis of thermo-gravimetric measurements of spent nuclear fuel oxidation rates

    International Nuclear Information System (INIS)

    Cramer, E.R.

    1997-01-01

    A detailed thermal analysis was completed of the sample temperatures in the Thermo-Gravimetric Analysis (TGA) system used to measure irradiated N Reactor fuel oxidation rates. Sample temperatures during the oxidation process did not show the increase which was postulated as a result of the exothermic reactions. The analysis shows the axial conduction of heat in the sample holder effectively removes the added heat and only a very small, i.e., <10 C, increase in temperature is calculated. A room temperature evaporation test with water showed the sample thermocouple sensitivity to be more than adequate to account for a temperature change of approximately 5 C. Therefore, measured temperatures in the TGA are within approximately 10 C of the actual sample temperatures and no adjustments to reported data to account for the heat input from the oxidation process are necessary

  5. Flow field measurements using LDA and numerical computation for rod bundle of reactor fuel assembly

    International Nuclear Information System (INIS)

    Hu Jun; Zou Zunyu

    1995-02-01

    Local mean velocity and turbulence intensity measurements were obtained with DANTEC 55 X two-dimensional Laser Dopper Anemometry (LDA) for rod bundle of reactor fuel assembly test model which was a 4 x 4 rod bundle. The data were obtained from different experimental cross-sections both upstream and downstream of the model support plate. Measurements performed at test Reynolds numbers of 1.8 x 10 4 ∼3.6 x 10 4 . The results described the local and gross effects of the support plate on upstream and downstream flow distributions. A numerical computation was also given, the experimental results are in good agreement with the numerical one and the others in references. Finally, a few suggestions were proposed for how to use the LDA system well. (11 figs.)

  6. Effects of thermocouple installation and location on fuel rod temperature measurements

    International Nuclear Information System (INIS)

    McCormick, R.D.

    1983-01-01

    This paper describes the results of analyses of nuclear fuel rod cladding temperature data obtained during in-reactor experiments under steady state and transient (simulated loss-of-coolant accident) operating conditions. The objective of the analyses was to determine the effect of thermocouple attachment method and location on measured thermal response. The use of external thermocouples increased the time to critical heat flux (CHF), reduced the blowdown peak temperature, and enhanced rod quench. A comparison of laser welded and resistance welded external thermocouple responses showed that the laser welding technique reduced the indicated cladding steady state temperatures and provided shorter time-to-CHF. A comparison of internal welded and embedded thermocouples indicated that the welded technique gave generally unsatisfactory cladding temperature measurements. The embedded thermocouple gave good, consistent results, but was possibly more fragile than the welded thermocouples. Detailed descriptions of the thermocouple designs, attachment methods and locations, and test conditions are provided

  7. Signs management and acquisition system for measurement of thermo physics properties of nuclear fuel

    International Nuclear Information System (INIS)

    Migliorini, Fabricio Lima; Silva, Egonn Hendrigo Carvalho; Neto Ferreira, Ricardo Alberto; Miranda, Odair; Grossi, Pablo Andrade; Camarano, Denise das Merces

    2009-01-01

    This work presents the management and acquisition system (SGA-LMPT) developed in a Lab View platform implemented at the Laboratory for Measurement of Thermophysics Properties at the Center for Development of Nuclear Technology - Brazilian Nuclear Energy Commission (CDTN/CNEN-MG). The SGA-LMPT is a tool which guarantee the complete fulfilment of the regulatory regulations, allowing a much more quality, productivity, costs reduction and optimization of the laboratory process. Besides, the SGA-LMPT presents the necessary requirements for demonstration a management system capable to produce results technically valid. The determination of the thermal diffusivity of the UO 2 pellet fabricated with microspheres was used for demonstration of the new system adequacy for the measurement of nuclear fuel thermo physics properties. The results were discussed and compared with literature

  8. Sensor for measurement of fuel rod gas pressure during loss-of-fluid-tests

    International Nuclear Information System (INIS)

    Billeter, T.R.

    1979-05-01

    Qualification tests have been conducted of a measurement system for determining the pressure of certain fuel rods in the loss-of-fluid-test (LOFT) reactor. Because of physical size (0.35-in. OD by 5.5-in length) and operational characteristics, an eddy current device was selected as the most promising measurement transducer for the application. The sensor must operate at pressure up to 17.2 MPa (2500 psig) and at temperatures up to 800 0 F. During the reactor transient caused by loss of coolant flow, sensor temperature and applied pressure will vary rapidly and significantly. Consequently, qualification tests included subjection of the sensor to rapid depressurization, temperature transients, and blowdowns in an autoclave, as well as to calibrations and various slow temperature cycles

  9. Deriving fuel-based emission factor thresholds to interpret heavy-duty vehicle roadside plume measurements.

    Science.gov (United States)

    Quiros, David C; Smith, Jeremy D; Ham, Walter A; Robertson, William H; Huai, Tao; Ayala, Alberto; Hu, Shaohua

    2018-04-13

    Remote sensing devices have been used for decades to measure gaseous emissions from individual vehicles at the roadside. Systems have also been developed that entrain diluted exhaust and can also measure particulate matter (PM) emissions. In 2015, the California Air Resources Board (CARB) reported that 8% of in-field diesel particulate filters (DPF) on heavy-duty (HD) vehicles were malfunctioning and emitted about 70% of total diesel PM emissions from the DPF-equipped fleet. A new high-emitter problem in the heavy-duty vehicle fleet had emerged. Roadside exhaust plume measurements reflect a snapshot of real-world operation, typically lasting several seconds. In order to relate roadside plume measurements to laboratory emission tests, we analyzed carbon dioxide (CO 2 ), oxides of nitrogen (NO X ), and PM emissions collected from four HD vehicles during several driving cycles on a chassis dynamometer. We examined the fuel-based emission factors corresponding to possible exceedances of emission standards as a function of vehicle power. Our analysis suggests that a typical HD vehicle will exceed the model year (MY) 2010 emission standards (of 0.2 g NO X /bhp-hr and 0.01 g PM/bhp-hr) by three times when fuel-based emission factors are 9.3 g NO X /kg fuel and 0.11 g PM/kg using the roadside plume measurement approach. Reported limits correspond to 99% confidence levels, which were calculated using the detection uncertainty of emissions analyzers, accuracy of vehicle power calculations, and actual emissions variability of fixed operational parameters. The PM threshold was determined for acceleration events between 0.47 and 1.4 mph/sec only, and the NO X threshold was derived from measurements where aftertreatment temperature was above 200°C. Anticipating a growing interest in real-world driving emissions, widespread implementation of roadside exhaust plume measurements as a compliment to in-use vehicle programs may benefit from expanding this analysis to a larger

  10. Measurement of nuclear reaction rates and spectral indices along the radius of fuel pellets from IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Mura, Luis Felipe Liambos

    2010-01-01

    This work presents the measurements of the nuclear reaction rates along the radial direction of the fuel pellet by irradiation and posterior gamma spectrometry of a thin slice of fuel pellet of UO 2 with 4,3% enrichment. From its irradiation the rate of radioactive capture and fission have been measured as a function of the radius of the pellet disk using a HPGe detector. Lead collimators has been used for this purpose. Simulating the fuel pellet in the pin fuel of the IPEN/MB-01 reactor, a thin UO 2 disk is used. This disk is inserted in the interior of a dismountable fuel rod. This fuel rod is then placed in the central position of the IPEN/MB-01 reactor core and irradiated during 1 hour under a neutron flux of around 9 x 10 8 n/cm 2 s. For gamma spectrometry 10 collimators with different diameters have been used, consequently, the nuclear reactions of radioactive capture that occurs in atoms of 238 U and fissions that occur on both 235 U and 238 U are measured in function of 10 different region (diameter of collimator) of the fuel pellet disk. Corrections in the geometric efficiency due to introduction of collimators on HPGe detection system were estimated using photon transport of MCNP-4C code. Some calculated values of nuclear reaction rate of radioactive capture and fission along of the radial direction of the fuel pellet obtained by Monte Carlo methodology, using the MCNP-4C code, are presented and compared to the experimental data showing very good agreement. Besides nuclear reaction rates, the spectral indices 28 ρ and 25 δ have been obtained at each different radius of the fuel pellet disk. (author)

  11. New measurement capabilities of mass spectrometry in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Perrin, R.E.

    1979-01-01

    Three recent developments, when combined, have the potential for greatly improving accountability measurements in the nuclear fuel cycle. The techniques are particularly valuable when measuring the contents of vessels which are difficult to calibrate by weight or volume. Input dissolver accountability measurements, inparticular, benefit from the application of these techniques. Los Alamos Scientific Laboratory has developed the capability for isotopic analysis of U and Pu samples at the nanogram level with an accuracy of 0.1 relative %. The Central Bureau for Nuclear Materials Measurement in Geel, Belgium has developed the capability of preparing mixed, solid metal U and Pu spikes with an accuracy of better than 0.1 relative %. Idaho Nuclear Energy Laboratory and C.K. Mathews at Bhabha Atomic Research have demonstrated a technique for determining the ratio of sample size to total solution measured which is independent of both the weight and the volume of the solution being measured. The advantages and limitations of these techniques are discussed. An analytical scheme which takes advantage of the special features of these techniques is proposed. 4 refs

  12. Burnup verification measurements on spent fuel assemblies at Arkansas Nuclear One

    International Nuclear Information System (INIS)

    Ewing, R.I.

    1995-01-01

    Burnup verification measurements have been performed using the Fork system at Arkansas Nuclear One, Units 1 and 2, operated by Energy Operations, Inc. Passive neutron and gamma-ray measurements on individual spent fuel assemblies were correlated with the reactor records for burnup, cooling time, and initial enrichment. The correlation generates an internal calibration for the system in the form of a power law determined by a least squares fit to the neutron data. The values of the exponent in the power laws were 3.83 and 4.35 for Units 1 and 2, respectively. The average deviation of the reactor burnup records from the calibration determined from the measurements is a measure of the random error in the burnup records. The observed average deviations were 2.7% and 3.5% for assemblies at Units 1 and 2, respectively, indicating a high degree of consistency in the reactor records. Two non-standard assemblies containing neutron sources were studied at Unit 2. No anomalous measurements were observed among the standard assemblies at either Unit. The effectiveness of the Fork system for verification of reactor records is due to the sensitivity of the neutron yield to burnup, the self-calibration generated by a series of measurements, the redundancy provided by three independent detection systems, and the operational simplicity and flexibility of the design

  13. Post-Irradiation Examination and In-Pile Measurement Techniques for Water Reactor Fuels

    International Nuclear Information System (INIS)

    2009-12-01

    in the 1960s when the construction of NPPs was being started. Evidently it can be assumed that infrastructure with basic unique equipments is old enough, both morally and physically, and needs to be up-graded or replaced. Thus, a sharp increase of the hydrocarbon fuel cost, green-house effect, necessity to construct more safe and efficient NPPs, justification of the lifetime prolongation of the existing NPPs, moral and physical ageing of the hot labs and research reactors equipment lead to the strong necessity to develop more perfect and more precise methods and equipment to examine irradiated components of nuclear reactors, first of all the most expensive one - nuclear fuel. Now the national hot laboratories and material testing reactors usually act as individual independent research establishments without any common and coordinated technical and business strategy towards the future needs and challenges. Even if there are not many joint programs for the development of nuclear power engineering in different countries, the method base and accumulated experience of the in- and post-reactor experiments should be widely shared so as to decrease the cost of this base in each country and to enforce its development. Thus, both problems and results of the application of new techniques to examine nuclear reactor components, as well as the conditions of separate labs should be discussed at the international level. The IAEA technical meetings are one of the most convenient means of arranging such discussion on the problems of the hot labs and research reactors development and application of new original techniques for examination of reactor materials properties. This publication represents a summary and proceedings of the two technical meetings (TMs) organized by IAEA on the subjects of Hot Cell Post-Irradiation Examination (PIE) Techniques and Pool Side Inspection of Water Reactor Fuel Assemblies and Fuel Rod Instrumentation and In-Pile Measurement Techniques. The first TM was

  14. LITGS: a new technique for single shot temperature and fuel concentration measurements in turbulent combusting environments

    Energy Technology Data Exchange (ETDEWEB)

    Fantoni, Roberta; Giorgi, M. [ENEA, Centro Ricerche Frascati, Frascati, RM (Italy). Dipt. Innovazione; De Risi, A.; Laforgia, D. [Lecce Univ., Lecce (Italy). Dipt. di Ingegneria dell' Innovazione

    1999-07-01

    In the present study the possibility to apply time resolved Laser Induced Thermal Grating Spectroscopy (LITGS) to detect fuel concentration and temperature in mixtures and flames at atmospheric pressure or higher is investigated. The resonant IR single photon absorption of two short pulse pump beams is used to initially generate a population grating, decaying into a thermal grating due to relaxation processes in the gas mixture. The thermal grating evolution is followed by monitoring the scattered signal of a cw visible probe beam after the end of the pump pulse. The use of the IR optical transition of diesel fuel assured a high species selectivity and a negligible influence of the visible emission background due to the presence of electronically excited species in flames. Fuel concentration and temperature measurements in a pressurized cell, with pressure ranging between 0.1 an 1.5 MPa, and in a diffusion turbulent flame generated by a burner feed with diesel fuel operating at atmospheric pressure are presented. The experimental investigation shows that LITGS signal increase linearly with gas density. This characteristic makes LITGS a very interesting technique for fuel distribution and temperature measurements in hostile (high-pressure and turbulent flow) environments. Detection limit for diesel fuel at atmospheric pressure is found to be about 40 ppm and it decreases with the increase of the pressure. The low detection limit which can be reached makes this technique suitable also for monitoring minor species and radicals. [Italian] Nel presente studio si investiga la possibilita' di applicare la tecnica LITGS (Laser Induced Thermal Grating Spectroscopy) per misurare la concentrazione e la temperatura di carburante in miscele e fiamme a pressiona atmosferica o superiore. L'assorbimento risonante di un singolo fotone IR proveniente da uno dei due laser impulsati di pompa e' utilizzato per generare inizialmente un reticolo di popolazione, che decade

  15. Development of a Computer Program for the Integrated Control of the Fuel Homogeneity Measurement System

    International Nuclear Information System (INIS)

    Shin, H. S.; Jang, J. W.; Lee, Y. H.; Oh, S. J.; Park, H. D.; Kim, C. K.

    2005-11-01

    The computer program is developed based on Visual C++, which is equipped with a user-friendly interface of the input/output(I/O) and a display function for the measuring conditions. This program consists of three parts which are the port communication, PLC(Programmable Logic Controller) and the MCA(Multi Channel Analyzer) control parts. The communication type between the CPU of the PLC module box and the computer is selected as be the Rs-232 asynchronous type and the thread method is adapted in the development of the first part of the program. The PLC-related program has been developed so that the data communication between the PLC CPU and the computer could be harmonized with the unique commands which have already been defined in the PLC. The measuring space and time intervals, the start and end ROI(region of interest) values, and the allowable error limitation are input at each measurement in this program. Finally the controlling MCA program has been developed by using Canberra's programming library which contains several files including the head files in which the variable and the function of C++ are declared according to the MCA function. The performance test has been carried out through an application of the developed computer program to the homogeneity measurement system. The gamma counts at 28 measuring points of a fuel rod of 700 mm in length are measured for 50 sec at each point. It was revealed that the measurement results are better than the previous ones in respects of the measurement accuracy and a measurement time saving could be achieved. It was concluded that the gamma measurement system can be improved through equipping it with the developed control program

  16. Comparison of neutron dose measured by Albedo TLD and etched tracks detector at PNC plutonium fuel facilities

    International Nuclear Information System (INIS)

    Tsujimura, N.; Momose, T.; Shinohara, K.; Ishiguro, H.

    1996-01-01

    Power Reactor and Nuclear Fuel Development Corporation (PNC) has fabricated Plutonium and Uranium Mixed OXide (MOX) fuel for FBR MONJU at Tokai works. In this site, PNC/Panasonic albedo TLDs/1/ are used for personnel neutron monitoring. And a part of workers wore Etched Tracks Detector (ETD) combined with TLD in order to check the accuracy of the neutron dose estimated by albedo TLD. In this paper, the neutron dose measured by TLD and ETD in the routine monitoring is compared at PNC plutonium fuel facilities. (author)

  17. Study of a device for the direct measurement of the fission gas pressure inside an in-pile fuel element

    International Nuclear Information System (INIS)

    Lavaud, B.; Uschanoff, S.

    1964-01-01

    The fission gas pressure inside a fuel element made of a refractory fuel constitutes an important limiting factor for the burn-up. Although it is possible to calculate approximately the volume of gas produced outside the fuel during its life-time; it is nevertheless very difficult to evaluate the pressure since the volume allowed to the fission gases, as well as their temperature are known only very approximately. This physical value, which is essential for the technologist, can only be known by direct in-pile measurement of the pressure. The report describes the equipment which has been developed for this test. (authors) [fr

  18. Fuel pellet relocation behavior in fast reactor uranium-plutonium mixed oxide fuel pin at beginning-of-life

    International Nuclear Information System (INIS)

    Inoue, Masaki; Ukai, Shigeharu; Asaga, Takeo

    1999-08-01

    The effects of fabrication parameters, irradiation conditions and fuel microstructural feature on fuel pellet relocation behavior in fast reactor fuel pins were investigated. This work focused only on beginning-of-life conditions, when fuel centerline temperature depends largely on the behavior. Fuel pellet relocation behavior in Joyo Mk-II driver could not be characterized because of the lack of data. And the behavior in FFTF driver and its larger diameter type fuel pins could not be characterized because of the extensive lot-by-lot scatters. The behavior both in Monju type and in Joyo power-to-melt type fuel pins were similar to each other, and depends largely on the as-fabricated gap width while the effects of linear heat rate and the extent of microstructural evolution were negligible. And fuel pellet centerline melting seems to affect slightly the behavior. The correlation, which describes the extent of relocation both in Monju type and in Joyo power-to-melt type fuel pins, were newly formulated and extrapolated for Joyo Mk-II driver, FFTF driver and its larger diameter type fuel pins. And the behavior in Joyo Mk-II driver seemed to be similar. On the contrary, the similarity with JNC fuel pins was observed case-by-case in FFTF driver and its larger diameter type fuel pins. (author)

  19. Calculation study of the WWER-440 fuel performance for extended burnup

    International Nuclear Information System (INIS)

    Kujal, J.; Pazdera, F.; Barta, O.

    1984-01-01

    The results of preliminary calculational study of extended burnup cycling schemes impact on WWER-440 fuel performance are presented. Two high burnup schemes were proposed with three and four cycles, resp. Comparison was made with three cycle reference case. The thermal mechanical analysis was performed with PIN and RELA codes. The values of rod internal pressure, fuel centerline temperatures and fuel-cladding gap are expressed as function of power history. (author)

  20. Status on Establishing the Feasibility of Lead Slowing Down Spectroscopy for Direct Measurement of Plutonium in Used Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kulisek, Jonathan A.; Anderson, Kevin K.; Casella, Andrew M.; Gesh, Christopher J.; Warren, Glen A.; Gavron, Victor A.; Devlin, M.; Haight, R. C.; O' Donnell, J. M.; Danon, Yaron; Weltz, Adam; Bonebrake, Eric; Imel, G. R.; Harris, Jason; Beller, Dennis; Hatchett, D.; Droessler, J.

    2012-08-30

    Developing a method for the accurate, direct, and independent assay of the fissile isotopes in bulk materials (such as used fuel) from next-generation domestic nuclear fuel cycles is a goal of the Office of Nuclear Energy, Fuel Cycle R&D, Material Protection and Control Technology (MPACT) Campaign. To meet this goal, MPACT supports a multi-institutional collaboration to study the feasibility of Lead Slowing Down Spectroscopy. This technique is an active nondestructive assay method that has the potential to provide independent, direct measurement of Pu and U isotopic masses in used fuel with an uncertainty considerably lower than the approximately 10% typical of today’s confirmatory assay methods. This paper will present efforts on the development of time-spectral analysis algorithms, fast neutron detector advances, and validation and testing measurements.

  1. Reactor physics measurements with 19-element ThOsub(2)-sup(235)UOsub(2) cluster fuel in heavy water moderator

    International Nuclear Information System (INIS)

    French, P.M.

    1985-02-01

    Low power lattice physics measurements have been performed with a single rod of 19-element thorium oxide fuel enriched with 1.45 wt. percent sub(235)UOsub(2) (93 percent enriched) in a simulated heavy water moderated and cooled power reactor core. The experiments were designed to provide data relevant to a power reactor irradiation and to obtain some basic information on the physics of uranium-thorium fuel material. Some theoretical flux calculations are summarized and show reasonable agreement with experiment

  2. Measurements in large JP-4 pool fires

    International Nuclear Information System (INIS)

    Keltner, N.R.; Kent, L.A.; Schneider, M.E.

    1987-01-01

    Over the past four years, Sandia National Laboratories has conducted a number of large pool fire tests to evaluate the design of radioactive material (RAM) shipping containers. Some of these tests have been designed to define the thermal environment and some have been used for certification testing. In each test there have been a number of fire diagnostic measurements. The simplest sets of diagnostics have involved measurements of temperature at several elevations on arrays of towers, measurements of hot wall heat flux with small calorimeters suspended from the towers, the average fuel recession rate, and the wind speed and direction. The most complex sets of diagnostics have included the above and in various tests added radiometers in the lower flame zone, centerline velocity measurements at a number of elevations, radiometers and calorimeters at the fuel surface, large cylindrical and flat plate calorimeters, infrared imaging, time resolved fuel recession rates, and a variety of soot particle concentration and size measurements made in the plume with a tethered balloon and an instrumented airplane. All of the large fires have been conducted in a 9.1 m by 18.3 m pool using JP-4 as the fuel. Typical duration is one-half hour. Covering all of the results is beyond the scope of a single paper. Conditionally sampled temperature and velocity measurements from one fire will be presented; for this fire, a 20 cm layer of fuel was floated on 61 cm of water. Pool surface heat flux, fuel recession rate data, and smoke emission data from a second fire are given. Because the wind has a strong effect on the temperature and velocity measurements, conditional sampling has been used to try to obtain data during periods of low winds. 10 refs., 3 figs

  3. Optimal measurement uncertainties for materials accounting in a fast breeder reactor spent-fuel reprocessing plant

    International Nuclear Information System (INIS)

    Dayem, H.A.; Kern, E.A.; Markin, J.T.

    1982-01-01

    Optimization techniques are used to calculate measurement uncertainties for materials accountability instruments in a fast breeder reactor spent-fuel reprocessing plant. Optimal measurement uncertainties are calculated so that performance goals for detecting materials loss are achieved while minimizing the total instrument development cost. Improved materials accounting in the chemical separations process (111 kg Pu/day) to meet 8-kg plutonium abrupt (1 day) and 40-kg plutonium protracted (6 months) loss-detection goals requires: process tank volume and concentration measurements having precisions less than or equal to 1%; accountability and plutonium sample tank volume measurements having precisions less than or equal to 0.3%, short-term correlated errors less than or equal to 0.04%, and long-term correlated errors less than or equal to 0.04%; and accountability and plutonium sample tank concentration measurements having precisions less than or equal to 0.4%, short-term correlated errors less than or equal to 0.1%, and long-term correlated errors less than or equal to 0.05%

  4. 40 CFR 80.46 - Measurement of reformulated gasoline fuel parameters.

    Science.gov (United States)

    2010-07-01

    ... Method for Total Sulfur in Gaseous Fuels by Hydrogenolysis and Rateometric Colorimetry,” or (ii) ASTM... Total Sulfur in Gaseous Fuels by Hydrogenolysis and Rateometric Colorimetry. (2) [Reserved] [59 FR 7813...

  5. Validation of the TUBRNP model with the radial distribution of plutonium in MOX fuel measured by SIMS and EPMA

    Energy Technology Data Exchange (ETDEWEB)

    O` Carroll, C; Laar, J Van De; Walker, C T [CEC Joint Research Centre, Karlsruhe (Germany)

    1997-08-01

    The new model TUBRNP (TRANSURANUS burnup) predicts the radial power density distribution as a function of burnup (and hence the radial burnup profile as a function of time) together with the radial profile of plutonium. Comparisons between measurements and the prediction of the TUBRNP model have been made for UO{sub 2} LWR fuels: they were found to be in excellent agreement and it is seen that TUBRNP is a marked improved on previous models. A powerful techniques for the characterization of irradiation fuel is Electron Probe Microanalysis (EPMA). Uranium, plutonium and fission product distributions can be analysed quantitatively. A complement, providing isotopic information with a lateral resolution comparable to EPMA, is secondary ion mass spectrometry (SIMS). Recently, the technique has been successfully applied for the measurement of the radial distribution of plutonium isotopes in irradiated nuclear fuel pins. The extension of the TUBRNP model to mixed oxide fuels seems to be the natural step to take. In MOX fuels the picture is greatly complicated by the presence of the (U, Pu)O{sub 2} agglomerates. The rim effect referred to above may be masked by the high concentrations of plutonium in the bulk of the fuel. A detailed investigation of a number of MOX fuel samples has been made using the TUBRNP model. Results are presented for a range of fuels with different enrichment and burnup. Through its participation in the PRIMO and DOMO programmes, PSI in conjunction with the Institute for Transuranium Elements had the opportunity to validate the new theoretical model TUBRNP. The authors with therefore to express their thanks to the organizers and to the numerous European and Japanese organizations which have supported these two international programmes on MOX fuel behavior. 7 refs, 9 figs, 3 tabs.

  6. Measurement of the Diffusion Coefficient of Water in RP-3 and RP-5 Jet Fuels Using Digital Holography Interferometry

    Science.gov (United States)

    Li, Chaoyue; Feng, Shiyu; Shao, Lei; Pan, Jun; Liu, Weihua

    2018-04-01

    The diffusion coefficient of water in jet fuel was measured employing double-exposure digital holographic interferometry to clarify the diffusion process and make the aircraft fuel system safe. The experimental method and apparatus are introduced in detail, and the digital image processing program is coded in MATLAB according to the theory of the Fourier transform. At temperatures ranging from 278.15 K to 333.15 K in intervals of 5 K, the diffusion coefficient of water in RP-3 and RP-5 jet fuels ranges from 2.6967 × 10 -10 m2·s-1 to 8.7332 × 10 -10 m2·s-1 and from 2.3517 × 10 -10 m2·s-1 to 8.0099 × 10-10 m2·s-1, respectively. The relationship between the measured diffusion coefficient and temperature can be well fitted by the Arrhenius law. The diffusion coefficient of water in RP-3 jet fuel is higher than that of water in RP-5 jet fuel at the same temperature. Furthermore, the viscosities of the two jet fuels were measured and found to be expressible in the form of the Arrhenius equation. The relationship among the diffusion coefficient, viscosity and temperature is analyzed according to the classic prediction model, namely the Stokes-Einstein correlation, and this correlation is further revised via experimental data to obtain a more accurate predication result.

  7. Manufacture of a UO2-Based Nuclear Fuel with Improved Thermal Conductivity with the Addition of BeO

    Science.gov (United States)

    Garcia, Chad B.; Brito, Ryan A.; Ortega, Luis H.; Malone, James P.; McDeavitt, Sean M.

    2017-12-01

    The low thermal conductivity of oxide nuclear fuels is a performance-limiting parameter. Enhancing this property may provide a contribution toward establishing accident-tolerant fuel forms. In this study, the thermal conductivity of UO2 was increased through the fabrication of ceramic-ceramic composite forms with UO2 containing a continuous BeO matrix. Fuel with a higher thermal conductivity will have reduced thermal gradients and lower centerline temperatures in the fuel pin. Lower operational temperatures will reduce fission gas release and reduce fuel restructuring. Additions of BeO were made to UO2 fuel pellets in 2.5, 5, 7.5, and 10 vol pct concentrations with the goals of establishing reliable lab-scale processing procedures, minimizing porosity, and maximizing thermal conductivity. The microstructure was characterized with electron probe microanalysis, and the thermal properties were assessed by light flash analysis and differential scanning calorimetry. Reliable, high-density samples were prepared using compaction pressure between 200 and 225 MPa and sintering times between 4 and 6 hours. It was found that the thermal conductivity of UO2 improved approximately 10 pct for each 1 vol pct BeO added over the measured temperature range 298.15 K to 523.15 K (25 °C to 250 °C) with the maximum observed improvement being ˜ 100 pct, or doubled, at 10 vol pct BeO.

  8. Experimental measurements and numerical modeling of marginal burning in live chaparral fuel beds

    Science.gov (United States)

    X. Zhou; D.R. Weise; S Mahalingam

    2005-01-01

    An extensive experimental and numerical study was completed to analyze the marginal burning behavior of live chaparral shrub fuels that grow in the mountains of southern California. Laboratory fire spread experiments were carried out to determine the effects of wind, slope, moisture content, and fuel characteristics on marginal burning in fuel beds of common...

  9. 77 FR 47043 - Work Group on Measuring Systems for Electric Vehicle Fueling

    Science.gov (United States)

    2012-08-07

    ... Systems for Electric Vehicle Fueling AGENCY: National Institute of Standards and Technology, Commerce... electric vehicle fuel. There is no cost for participating in the Work Group. No proprietary information... and sell electricity dispensed as a vehicle fuel) and to ensure that the prescribed methodologies and...

  10. Development of a water boil-off spent-fuel calorimeter system. [To measure decay heat generation rate

    Energy Technology Data Exchange (ETDEWEB)

    Creer, J.M.; Shupe, J.W. Jr.

    1981-05-01

    A calorimeter system was developed to measure decay heat generation rates of unmodified spent fuel assemblies from commercial nuclear reactors. The system was designed, fabricated, and successfully tested using the following specifications: capacity of one BWR or PWR spent fuel assembly; decay heat generation range 0.1 to 2.5 kW; measurement time of < 12 h; and an accuracy of +-10% or better. The system was acceptance tested using a dc reference heater to simulate spent fuel assembly heat generation rates. Results of these tests indicated that the system could be used to measure heat generation rates between 0.5 and 2.5 kW within +- 5%. Measurements of heat generation rates of approx. 0.1 kW were obtained within +- 15%. The calorimeter system has the potential to permit measurements of heat generation rates of spent fuel assemblies and other devices in the 12- to 14-kW range. Results of calorimetry of a Turkey Point spent fuel assembly indicated that the assembly was generating approx. 1.55 kW.

  11. Test of Flow Characteristics in Tubular Fuel Assembly I - Establishment of test loop and measurement validation test

    International Nuclear Information System (INIS)

    Park, Jong Hark; Chae, H. T.; Park, C.; Kim, H.

    2005-12-01

    Tubular type fuel has been developed as one of candidates for Advanced HANARO Reactor(AHR). It is necessary to test the flow characteristics such as velocity in each flow channels and pressure drop of tubular type fuel. A hydraulic test-loop to examine the hydraulic characteristics for a tubular type fuel has been designed and constructed. It consists of three parts; a) piping-loop including pump and motor, magnetic flow meter and valves etc, b) test-section part where a simulated tubular type fuel is located, and 3) data acquisition system to get reading signals from sensors or instruments. In this report, considerations during the design and installation of the facility and the selection of data acquisition sensors and instruments are described in detail. Before doing the experiment to measure the flow velocities in flow channels, a preliminary tests have been done for measuring the coolant velocities using pitot-tube and for validating the measurement accuracy as well. Local velocities of the radial direction in circular tubes are measured at regular intervals of 60 degrees by three pitot-tubes. Flow rate inside the circular flow channel can be obtained by integrating the velocity distribution in radial direction. The measured flow rate was compared to that of magnetic flow meter. According to the results, two values had a good agreement, which means that the measurement of coolant velocity by using pitot-tube and the flow rate measured by the magnetic flow meter are reliable. Uncertainty analysis showed that the error of velocity measurement by pitot-tube is less than ±2.21%. The hydraulic test-loop also can be adapted to others such as HANARO 18 and 36 fuel, in-pile system of FTL(Fuel Test Loop), etc

  12. Establishment of an authenticated physical standard for gamma spectrometric determination of the U-235 content of MTR fuel and evaluation of measurement procedures

    International Nuclear Information System (INIS)

    Fleck, C.M.

    1979-12-01

    Measurements of U-235 content in a standard MTR fuel element were carried out, using scintillation and semi-conductor spectrometers. Three different types of measurement were carried out: a) Comparison of different primary standards among one another and with single fuel plates. b) Calibration of the MTR fuel element as an authenticated physical standard. c) Evaluation of over all errors in assay measurements on MTR fuel elements. The error of the whole assay measurement will be approximately 0.9%. The Uranium distribution in the single fuel plates is the original source of error. In the case of equal Uranium contents in all fuel plates of one fuel assembly, the error of assay measurements would be about 0.3% relative to the primary standards

  13. Microbial Fuel Cell Inoculated with Ochrobactrum Tritici KCC210 for Chromium (VI) Measurement in Electroplating Wastewater

    Science.gov (United States)

    Kuo, Jongtar; Kuo, Juiling; Cheng, Chiuyu; Chung, Yingchien

    2018-01-01

    Many methods/techniques have been developed for Cr(VI) measurement, but they are often conducted offsite or/and cannot provide real-time for Cr(VI) monitoring. A microbial fuel cell (MFC) is a self-sustaining device and has great potential as a biosensor for in situ Cr(VI) measurement, especially for wastewater generated from different electroplating units. In this study, Ochrobactrum tritici KCC210, a facultatively anaerobic, Cr(VI)-reducing, and exoelectrogenic bacterium, was isolated and inoculated into the MFC to evaluate its feasibility as a Cr(VI) biosensor. The results indicated that O. tritici KCC210 exhibited high adaptability to pH, and temperature under anaerobic conditions. The maximum power density of the MFC biosensor was 17.5±0.9 mW/m2 at 2,000 Ω. A good linear relationship was observed between the Cr(VI) concentration (10-80 mg/L) and voltage output. The stable performance of the MFC biosensor indicated its potential as a reliable biosensor system. Moreover, the developed MFC biosensor is a simple device that can accurately measure Cr(VI) concentrations in the actual electroplating wastewater generated from different electroplating units within 15 min with low deviations (-1.8% to 7.8%) in comparison with the values determined using standard method. Thus, the MFC biosensor can measure Cr(VI) concentrations in situ in the effluents and has potential as an early warning detection device.

  14. Experimental Measurement of Relative Permeability Functions for Fuel Cell GDL Materials

    KAUST Repository

    Hussaini, Irfan; Wang, Chao-Yang

    2009-01-01

    Gas diffusion layer in PEM fuel cells plays a pivotal role in water management. Modeling of liquid water transport through the GDL relies on knowledge of relative permeability functions in the in-plane and through-plane directions. In the present work, air and water relative permeabilities are experimentally determined as functions of saturation for typical GDL materials such as Toray-060, -090, -120 carbon paper and E-Tek carbon cloth materials in their plain, untreated forms. Saturation is measured using an ex-situ gravimetric method. Absolute and relative permeability functions in the two directions of interest are presented. Significant departure from the generally assumed cubic function of saturation is observed. ©The Electrochemical Society.

  15. Systems and methods for optically measuring properties of hydrocarbon fuel gases

    Science.gov (United States)

    Adler-Golden, Steven; Bernstein, Lawrence S.; Bien, Fritz; Gersh, Michael E.; Goldstein, Neil

    1998-10-13

    A system and method for optical interrogation and measurement of a hydrocarbon fuel gas includes a light source generating light at near-visible wavelengths. A cell containing the gas is optically coupled to the light source which is in turn partially transmitted by the sample. A spectrometer disperses the transmitted light and captures an image thereof. The image is captured by a low-cost silicon-based two-dimensional CCD array. The captured spectral image is processed by electronics for determining energy or BTU content and composition of the gas. The innovative optical approach provides a relatively inexpensive, durable, maintenance-free sensor and method which is reliable in the field and relatively simple to calibrate. In view of the above, accurate monitoring is possible at a plurality of locations along the distribution chain leading to more efficient distribution.

  16. Combined local current distribution measurements and high resolution neutron radiography of operating direct methanol fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, Alexander; Wippermann, Klaus [Forschungszentrum Juelich GmbH (Germany). Inst. of Energy Research, IEF-3: Fuel Cells; Sanders, Tilman [RWTH Aachen (DE). Inst. for Power Electronics and Electrical Drives (ISEA); Arlt, Tobias [Helmholtz Centre Berlin (Germany). Inst. for Applied Materials

    2010-07-01

    Neutron radiography allows the investigation of the local fluid distribution in direct methanol fuel cells (DMFCs) under operating conditions. Spatial resolutions in the order of some tens of micrometers at the full test cell area are achieved. This offers the possibility to study practice-oriented, large stack cells with an active area of several hundred cm{sup 2} as well as specially designed, small test cells with an area of some cm{sup 2}. Combined studies of high resolution neutron radiography and segmented cell measurements are especially valuable, because they enable a correlation of local fluid distribution and local performance [1, 2]. The knowledge of this interdependency is essential to optimise the water management and performance respecting a homogeneous fluid, current and temperature distribution and to achieve high performance and durability of DMFCs. (orig.)

  17. On site PWR fuel inspection measurements for operational and design verification

    International Nuclear Information System (INIS)

    1996-01-01

    The on-site inspection of irradiated Pressurized Water Reactor (PWR) fuel and Non-Fuel Bearing Components (NFBC) is typically limited to visual inspections during refuelings using underwater TV cameras and is intended primarily to confirm whether the components will continue in operation. These inspections do not normally provide data for design verification nor information to benefit future fuel designs. Japanese PWR utilities and Nuclear Fuel Industries Ltd. designed, built, and performed demonstration tests of on-site inspection equipment that confirms operational readiness of PWR fuel and NFBC and also gathers data for design verification of these components. 4 figs, 3 tabs

  18. Technological problems and counter-measures on equipment materials for reprocessing of high burnup fuels

    International Nuclear Information System (INIS)

    Kiuchi, K.; Kato, T.; Motooka, H.; Hamada, S.

    2002-01-01

    The reliability of structural materials is considered as one of the most important technological issues on the commercial reprocessing of high burnup fuels. The durability prediction study of equipment materials used in commercial purex process has been conducted in the JAERI. From the experimental results obtained by scaled mock-up tests and laboratory tests, the stress corrosion cracking (SCC) for a dissolvor made of zirconium and the trans-passive corrosion of heat transfer tubes for evaporators made of austenitic stainless steels have been clarified as critical issues on the reliability. The susceptibility to these failures increases with the amount of TRU and FP elements included in spent fuels, because Np, Pu, Ru, Pd act as strong oxidizers. As counter-measures against these problems, the development of the modified alloys is going on in the JAERI. It has been found that the intergranular corrosion resistance of stainless steels is possible to be completely improved by purifying the electron beam melting process and by modifying the metallographic structure. The other counter measure is to inhibit the trans-passive corrosion by addition of oxide film former elements such as W and Si. It has also been found that the susceptibility to SCC of Zr can be improved by addition of titanium. However, the addition of titanium decreases the corrosion resistance of Zr. We selected niobium alloys as alternative materials to zirconium. By addition of tungsten to the niobium, the corrosion resistance and the mechanical strength have been improved. This niobium alloy can be used in heavily corrosive nitric acid contaminated with fluorine. It is considered that the difference between corrosion resistance of Zr and Nb-alloys is attributed to the chemical stability of the oxide films (MO 2 on Zr and M 2 O 5 on Nb). (author)

  19. Analysis of experimental measurements of PWR fresh and spent fuel assemblies using Self-Interrogation Neutron Resonance Densitometry

    Energy Technology Data Exchange (ETDEWEB)

    LaFleur, Adrienne M., E-mail: alafleur@lanl.gov; Menlove, Howard O., E-mail: hmenlove@lanl.gov

    2015-05-01

    Self-Interrogation Neutron Resonance Densitometry (SINRD) is a new NDA technique that was developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for LWR fuel assemblies. The SINRD detector consists of four fission chambers (FCs) wrapped with different absorber filters to isolate different parts of the neutron energy spectrum and one ion chamber (IC) to measure the gross gamma rate. As a result, two different techniques can be utilized using the same SINRD detector unit and hardware. These techniques are the Passive Neutron Multiplication Counter (PNMC) method and the SINRD method. The focus of the work described in this paper is the analysis of experimental measurements of fresh and spent PWR fuel assemblies that were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using the SINRD detector. The purpose of these experiments was to assess the following capabilities of the SINRD detector: 1) reproducibility of measurements to quantify systematic errors, 2) sensitivity to water gap between detector and fuel assembly, 3) sensitivity and penetrability to the removal of fuel rods from the assembly, and 4) use of PNMC/SINRD ratios to quantify neutron multiplication and/or fissile content. The results from these simulations and measurements provide valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies. - Highlights: • Experimental measurements of PWR fresh and spent FAs were performed with SINRD. • Good agreement of MCNPX and measured results confirmed accuracy of SINRD model. • For fresh fuel, SINRD and PNMC ratios were not sensitive to water gaps of ≤5-mm. • Practical use of SINRD would be in Fork detector to reduce systematic uncertainties.

  20. Measurement of Quasi-periodic Oscillating Flow Motion in Simulated Dual-cooled Annular Fuel Bundle

    International Nuclear Information System (INIS)

    Lee, Chi Young; Shin, Chang Hwan; Park, Ju Yong; Oh, Dong Seok; Chun, Tae Hyun; In, Wang Kee

    2012-01-01

    In order to increase a significant amount of reactor power in OPR1000, KAERI (Korea Atomic Energy Research Institute) has been developing a dual-cooled annular fuel. The dual-cooled annular fuel is simultaneously cooled by the water flow through the inner and the outer channels. KAERI proposed the 12x12 dual-cooled annular fuel array which was designed to be structurally compatible with the 16x16 cylindrical solid fuel array by maintaining the same array size and the guide tubes in the same locations, as shown in Fig. 1. In such a case, due to larger outer diameter of dual-cooled annular fuel than conventional solid fuel, a P/D (Pitch-to-Diameter ratio) of dual cooled annular fuel assembly becomes smaller than that of cylindrical solid fuel. A change in P/D of fuel bundle can cause a difference in the flow mixing phenomena between the dual-cooled annular and conventional cylindrical solid fuel assemblies. In this study, the rod bundle flow motion appearing in a small P/D case is investigated preliminarily using PIV (Particle Image Velocimetry) for dual-cooled annular fuel application

  1. Measurements of subchannel velocity and pressure drop for HANARO fuel assembly

    International Nuclear Information System (INIS)

    Yang, Sun Kyu; Jeong, Heung Jun; Cho, Suk; Min, Kyung Ho; Jeong, Moon Ki

    1996-07-01

    This report presents the hydraulic test results for HANARO fuel assemblies, which are performed to obtain the axial velocity and pressure drop data to be used to validate the code calculation model. For both 18 and 36-element fuel assemblies axial velocities of the entrance and exit regions are obtained, and developing axial velocity profiles along the flow direction for the fuel region of 18-element fuel assembly are also obtained. Varying the pressure tap locations, pressure drop data for each component of fuel assembly are obtained for various flow conditions. From the pressure drop test results it is noted that the pressure drops across the fuel assembly are 214 kPa and 205 kPa for the 18-element and 36-element fuel assembly respectively. 39 tabs., 12 figs., 5 refs. (Author)

  2. Near-field heat transfer at the spent fuel test-climax: a comparison of measurements and calculations

    International Nuclear Information System (INIS)

    Patrick, W.C.; Montan, D.N.; Ballou, L.B.

    1981-01-01

    The Spent Fuel Test in the Climax granitic stock at the DOE Nevada Test Site is a test of the feasibility of storage and retrieval of spent nuclear reactor fuel in a deep geologic environment. Eleven spent fuel elements, together with six thermally identical electrical resistance heaters and 20 peripheral guard heaters, are emplaced 420 m below surface in a three-drift test array. This array was designed to simulate the near-field effects of thousands of canisters of nuclear waste and to evaluate the effects of heat alone, and heat plus ionizing radiation on the rock. Thermal calculations and measurements are conducted to determine thermal transport from the spent fuel and electrical resistance heaters. Calculations associated with the as-built Spent Fuel Test geometry and thermal source histories are presented and compared with thermocouple measurements made throughout the test array. Comparisons in space begin at the spent fuel canister and include the first few metres outside the test array. Comparisons in time begin at emplacement and progress through the first year of thermal loading in this multi-year test

  3. Catalase measurement: A new field procedure for rapidly estimating microbial loads in fuels and water-bottoms

    Energy Technology Data Exchange (ETDEWEB)

    Passman, F.J. [Biodeterioration Control Associates, Inc., Chicago, IL (United States); Daniels, D.A. [Basic Fuel Services, Dover, NJ (United States); Chesneau, H.F.

    1995-05-01

    Low-grade microbial infections of fuel and fuel systems generally go undetected until they cause major operational problems. Three interdependent factors contribute to this: mis-diagnosis, incorrect or inadequate sampling procedures and perceived complexity of microbiological testing procedures. After discussing the first two issues, this paper describes a rapid field test for estimating microbial loads in fuels and associated water. The test, adapted from a procedure initially developed to measure microbial loads in metalworking fluids, takes advantage of the nearly universal presence of the enzyme catalase in the microbes that contaminated fuel systems. Samples are reacted with a peroxide-based reagent; liberating oxygen gas. The gas generates a pressure-head in a reaction tube. At fifteen minutes, a patented, electronic pressure-sensing device is used to measure that head-space pressure. The authors present both laboratory and field data from fuels and water-bottoms, demonstrating the excellent correlation between traditional viable test data (acquired after 48-72 hours incubation) and catalase test data (acquired after 15 min.-4 hours). We conclude by recommending procedures for developing a failure analysis data-base to enhance our industry`s understanding of the relationship between uncontrolled microbial contamination and fuel performance problems.

  4. Hydraulic Hybrid and Conventional Parcel Delivery Vehicles' Measured Laboratory Fuel Economy on Targeted Drive Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Lammert, M. P.; Burton, J.; Sindler, P.; Duran, A.

    2014-10-01

    This research project compares laboratory-measured fuel economy of a medium-duty diesel powered hydraulic hybrid vehicle drivetrain to both a conventional diesel drivetrain and a conventional gasoline drivetrain in a typical commercial parcel delivery application. Vehicles in this study included a model year 2012 Freightliner P100H hybrid compared to a 2012 conventional gasoline P100 and a 2012 conventional diesel parcel delivery van of similar specifications. Drive cycle analysis of 484 days of hybrid parcel delivery van commercial operation from multiple vehicles was used to select three standard laboratory drive cycles as well as to create a custom representative cycle. These four cycles encompass and bracket the range of real world in-use data observed in Baltimore United Parcel Service operations. The NY Composite cycle, the City Suburban Heavy Vehicle Cycle cycle, and the California Air Resources Board Heavy Heavy-Duty Diesel Truck (HHDDT) cycle as well as a custom Baltimore parcel delivery cycle were tested at the National Renewable Energy Laboratory's Renewable Fuels and Lubricants Laboratory. Fuel consumption was measured and analyzed for all three vehicles. Vehicle laboratory results are compared on the basis of fuel economy. The hydraulic hybrid parcel delivery van demonstrated 19%-52% better fuel economy than the conventional diesel parcel delivery van and 30%-56% better fuel economy than the conventional gasoline parcel delivery van on cycles other than the highway-oriented HHDDT cycle.

  5. Employing Hot Wire Anemometry to Directly Measure the Water Balance in a Proton Exchange membrane Fuel Cell

    DEFF Research Database (Denmark)

    Shakhshir, Saher Al; Hussain, Nabeel; Berning, Torsten

    2015-01-01

    Water management in proton exchange membrane fuel cells (PEMFC’s) remains a critical problem for their durability, cost, and performance. Because the anode side of this fuel cell has the tendency to become dehydrated, measuring the water balance can be an important diagnosis tool during fuel cell...... operation. The water balance indicates how much of the product water leaves at the anode side versus the cathode side. Previous methods of determining the fuel cell water balance often relied on condensing the water in the exhaust gas streams and weighing the accumulated mass which is a time consuming...... process that has limited accuracy. Currently, our group is developing a novel method to accurately determine the water balance in a PEMFC in real time by employing hot-wire anemometry. The amount of heat transferred from the wire to the anode exhaust stream can be translated into a voltage signal which...

  6. Studies of Nuclear Fuel Performance Using On-site Gamma-ray Spectroscopy and In-pile Measurements

    International Nuclear Information System (INIS)

    Matsson, Ingvar

    2006-01-01

    Presently there is a clear trend of increasing demands on in-pile performance of nuclear fuel. Higher target burnups, part length rods and various fuel additives are some examples of this trend. Together with an increasing demand from the public for even safer nuclear power utilisation, this implies an increased focus on various experimental, preferably non-destructive, methods to characterise the fuel. This thesis focuses on the development and experimental evaluation of such methods. In its first part, the thesis presents a method based on gamma-ray spectroscopy with germanium detectors that have been used at various power reactors in Europe. The aim with these measurements is to provide information about the thermal power distribution within fuel assemblies in order to validate core physics production codes. The early closure of the Barsebaeck 1 BWR offered a unique opportunity to perform such validations before complete depletion of burnable absorbers in Gd-rods had taken place. To facilitate the measurements, a completely submersible measuring system, LOKET, was developed allowing for convenient in-pool measurements to be performed. In its second part, the thesis describes methods that utilise in-pile measurements. These methods have been used in the Halden test-reactor for determination of fission gas release, pellet-cladding interaction studies and fuel development studies. Apart from the power measurements, the LOKET device has been used for fission gas release (FGR) measurements on single fuel rods. The significant reduction in fission gas release in the modern fuel designs, in comparison with older designs, has been demonstrated in a series of experiments. A FGR database covering a wide range of burnup, power histories and fuel designs has been compiled and used for fuel performance analysis. The fission gas release has been measured on fuel rods with average burnups well above 60 MWd/kgU. The comparison between core physics calculations (PHOENIX-4/POLCA

  7. High concentration tritium gas measurement with small volume ionization chambers for fusion fuel gas monitors

    International Nuclear Information System (INIS)

    Uda, Tatsuhiko; Okuno, Kenji; Matsuda, Yuji; Naruse, Yuji

    1991-01-01

    To apply ionization chambers to fusion fuel gas processing systems, high concentration tritium gas was experimentally measured with small volume 0.16 and 21.6 cm 3 ionization chambers. From plateau curves, the optimum electric field strength was obtained as 100∼200 V/cm. Detection efficiency was confirmed as dependent on the ionization ability of the filled gas, and moreover on its stopping power, because when the range of the β-rays was shortened, the probability of energy loss by collisions with the electrode and chamber wall increased. Loss of ions by recombination was prevented by using a small volume ionization chamber. For example the 0.16 cm 3 ionization chamber gave measurement with linearity to above 40% tritium gas. After the tritium gas measurements, the concentration levels inside the chamber were estimated from their memory currents. Although more than 1/4,000 of the maximum, current was observed as a memory effect, the smaller ionization chamber gave a smaller memory effect. (author)

  8. Measurements of bundle end flux peaking effects in 37-element CANDU PHW fuel

    International Nuclear Information System (INIS)

    French, P.M.

    1977-10-01

    Thermal neutron bundle end flux peaking factors have been measured in fresh 37-element Bruce reactor natural UO 2 clusters in heavy water moderator, both with and without staggered plenums at the fuel stack ends, in representative elements throughout the clusters. The measurements were made at a square lattice pitch of 28.58 cm with heavy water coolant. The results indicate that outer element peaking factors are 1.142 +- 0.009 for bundles containing no plenums, and 1.155 +- 0.006 and 1.177 +- 0.006 at the non-plenum and plenum element ends respectively, for bundles containing staggered plenums, irrespective of the azimuthal orientation between pairs of bundles. Measurements are also reported for bundles containing plenums in every outer element, for bundles separated by a stainless steel flux suppressor, for longer graphite plenums, and for changes in plenum and bundle gap lengths. Some theoretical comparisons with the results, reported by other authors, have been summarized. (author)

  9. The 3rd irradiation test plan of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Song, K. C.; Park, J. H. and others

    2001-05-01

    The objective of the 3rd irradiation test of DUPIC fuel at the HANARO is to estimate the in-core behaviour of a DUPIC pellet that is irradiated up to more than average burnup of CANDU fuel. The irradiation of DUPIC fuel is planned to start at May 21, 2001, and will be continued at least for 8 months. The burnup of DUPIC fuel through this irradiation test is thought to be more than 7,000 MWd/tHE. The DUPIC irradiation rig instrumented with three SPN detectors will be used to accumulate the experience for the instrumented irradiation and to estimate the burnup of irradiated DUPIC fuel more accurately. Under normal operating condition, the maximum linear power of DUPIC fuel was estimated as 55.06 kW/m, and the centerline temperature of a pellet was calculated as 2510 deg C. In order to assess the integrity of DUPIC fuel under the accident condition postulated at the HANARO, safety analyses on the locked rotor and reactivity insertion accidents were carried out. The maximum centerline temperature of DUPIC fuel was estimated 2590 deg C and 2094 deg C for each accident, respectively. From the results of the safety analysis, the integrity of DUPIC fuel during the HANARO irradiation test will be secured. The irradiated DUPIC fuel will be transported to the IMEF. The post-irradiation examinations are planned to be performed at the PIEF and IMEF.

  10. Influence of the temperature in the measurement of the {sup 235}U content in fuel rods in the passive scanner

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Carlos A. da S.; Junqueira, Fabio da S., E-mail: carlossilva@inb.gov.br, E-mail: fabiojunqueira@ing.gov.br [Indústrias Nucleares do Brasil (INB), Resende, RJ (Brazil). Superintendência de Engenharia do Combustível

    2017-07-01

    The fuel rod, composed of a cladding responsible for storing UO{sub 2} pellets, is an indispensable component for the composition of the mentioned fuel assembly, a metallic structure that is installed inside nuclear reactors for the purpose of generating energy. With the change of the technology of fuel elements of the Angra-1 plant, it was necessary to install a new equipment in the Nuclear Fuel Plant of INB in Resende / RJ. This new equipment, called Rod Scanner, is responsible for inspecting, through non-destructive tests, the internal structure of the fuel rod. Such inspection is the measurement of enrichment of the pellets ({sup 235}U content) inside of the rod by detection of gamma radiation by bismuth germanate (BGO) detectors using the scintillation principle. However, throughout the operation of the equipment, it was found that this measurement was not happening on a regular basis. After an intense investigation, it was verified that one of the factors that affects such measurement is the temperature at which the BGO detectors are submitted. In this way, the objective of this paper is to show the experiments and tests performed to reach the conclusion described previously, as well as to show the relation obtained between the signals collected in the detectors versus the internal temperature of the block where they are located. (author)

  11. Optimization of a dual-fuel heating system utilizing an EMS a maintain persistence of measures

    International Nuclear Information System (INIS)

    Wolpert, J.S.; Wolpert, S.B.; Martin, G.

    1993-01-01

    An older small office building was subjected to a program substituting gas for electric heat to reduce energy cost and improve comfort for approximately one year and was permanently instituted, with the installation of an energy management system (EMS) the following year. This paper will present a description of the facility, its usage patterns, and the measures taken to introduce the fuel-switching program. The impacts on energy usage and cost as well as comfort will also be reported. This program was initiated by a preliminary audit of the facility conducted by the service contractor in conjunction with the area gas wholesaler. During the audit it was observed that the heating set points for the gas-fired equipment was kept fairly low. This was the result of the desire to keep the cooling set point low and the use of auto-changeover thermostats. The result of this was that the system utilized the gas heat to come up to 68-70 degrees with the majority of the zones then relying on their electric heat to bring temperatures into the 73-75 degrees range. In addition to impacting energy costs, this approach generated numerous comfort complaints. As a further electric penalty, the low cooling set point resulted in a heavy reliance on electric heat (reheat) all summer. The basis of the proposed strategy was to reduce the heavy usage of electric heat by making the building comfortable through reliance more heavily on gas heat. This was tested by raising the heating set points for the RTUS. The success of this approach, along with the comfort considerations and the desire for further savings, led to the installation of an EMS. This allowed further refinements of the control strategy, which are briefly described. When completed, the fuel-switching led to an increase in annual gas costs of 125% with a corresponding decrease in electric cost of nearly 30% for an annual utility cost savings of over 19%

  12. Thermal Signature Measurements for Ammonium Nitrate/Fuel Mixtures by Laser Heating.

    Science.gov (United States)

    Nazarian, Ashot; Presser, Cary

    2016-01-10

    Measurements were carried out to obtain thermal signatures of several ammonium nitrate/fuel (ANF) mixtures, using a laser-heating technique referred to as the laser-driven thermal reactor (LDTR). The mixtures were ammonium nitrate (AN)/kerosene, AN/ethylene glycol, AN/paraffin wax, AN/petroleum jelly, AN/confectioner's sugar, AN/cellulose (tissue paper), nitromethane/cellulose, nitrobenzene/cellulose, AN/cellulose/nitromethane, AN/cellulose/nitrobenzene. These mixtures were also compared with AN/nitromethane and AN/diesel fuel oil, obtained from an earlier investigation. Thermograms for the mixtures, as well as individual constituents, were compared to better understand how the sample thermal signature changes with mixture composition. This is the first step in development of a thermal-signature database, to be used along with other signature databases, to improve identification of energetic substances of unknown composition. The results indicated that each individual thermal signature was associated unambiguously with a particular mixture composition. The signature features of a particular mixture were shaped by the individual constituent signatures. It was also uncovered that the baseline signature was modified after an experiment due to coating of unreacted residue on the substrate surface and a change in the reactor sphere oxide layer. Thus, care was required to pre-oxidize the sphere prior to an experiment. A minimum sample mass (which was dependent on composition) was required to detect the signature characteristics. Increased laser power served to magnify signal strength while preserving the signature features. For the mixtures examined, the thermal response of each ANF mixture was found to be different, which was based on the mixture composition and the thermal behavior of each mixture constituent.

  13. Determination of the perturbing effect of the measuring device on thermal neutron distribution inside the fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Takac, S M; Krcevinac, S B [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1966-07-15

    The measurement of the thermal neutron distribution in an elementary cell of the reactor core is based on activating some of the existing detectors such as gold, copper, dysprosium, etc., inside the fuel rod and the corresponding part of the slowing-down medium. The techniques of measuring may be classified in two groups: - technique with detector foils, and - technique with a detector wire. The first group includes all experiments based on the so called 'tube technique'. By this technique the detector foils are arranged specifically in the tube by means of spacers and are positioned in a radially bored fuel rod. The 'spiral technique' is also included here. By this technique the fuel rod, which is first radially cut, is axially bored along the spiral and then detector foils inserted in the holes. The second group includes techniques according to which the detector wires may be positioned either in the radially bored hole through the fuel rod or in the spiral groove made in the horizontal cross-section in the fuel rod. To obtain higher resolution the detector wire after activation can be extruded, or before irradiation, spirally wound around a solid core and thus positioned in the radial hole in the fuel rod. In all cases the fuel region is perturbed either by the holes and the detector material, or by the holder of the detector foils. A number of authors have carried out these experiments under different geometrical and nuclear conditions, so obviously this perturbation had different effects on the results. So far the cell perturbation effects have not been discussed in the literature, neither this effect has been corrected in the final results. With respect to this a series of experiments for determining the micro distribution of thermal neutrons inside the fuel rod were made on the heavy-water natural-uranium system for different lattice pitches, with special stress or the investigation of the perturbing effects in the fuel rod which inevitably must be

  14. Determination of the perturbing effect of the measuring device on thermal neutron distribution inside the fuel rod

    International Nuclear Information System (INIS)

    Takac, S.M.; Krcevinac, S.B.

    1966-07-01

    The measurement of the thermal neutron distribution in an elementary cell of the reactor core is based on activating some of the existing detectors such as gold, copper, dysprosium, etc., inside the fuel rod and the corresponding part of the slowing-down medium. The techniques of measuring may be classified in two groups: - technique with detector foils, and - technique with a detector wire. The first group includes all experiments based on the so called 'tube technique'. By this technique the detector foils are arranged specifically in the tube by means of spacers and are positioned in a radially bored fuel rod. The 'spiral technique' is also included here. By this technique the fuel rod, which is first radially cut, is axially bored along the spiral and then detector foils inserted in the holes. The second group includes techniques according to which the detector wires may be positioned either in the radially bored hole through the fuel rod or in the spiral groove made in the horizontal cross-section in the fuel rod. To obtain higher resolution the detector wire after activation can be extruded, or before irradiation, spirally wound around a solid core and thus positioned in the radial hole in the fuel rod. In all cases the fuel region is perturbed either by the holes and the detector material, or by the holder of the detector foils. A number of authors have carried out these experiments under different geometrical and nuclear conditions, so obviously this perturbation had different effects on the results. So far the cell perturbation effects have not been discussed in the literature, neither this effect has been corrected in the final results. With respect to this a series of experiments for determining the micro distribution of thermal neutrons inside the fuel rod were made on the heavy-water natural-uranium system for different lattice pitches, with special stress or the investigation of the perturbing effects in the fuel rod which inevitably must be

  15. NDA measurements on spent fuel assemblies at Tihange 1 by means of the ION 1/FORK

    International Nuclear Information System (INIS)

    Carchon, R.; Smaers, G.; Verrecchia, G.P.D.; Arlt, R.; Stoyanova, I.; Satinet, J.

    1986-06-01

    This report describes field tests performed at Tihange 1 Nuclear Power Station on PWR spent fuel by means of the ION 1-FORK detector. Two detector systems and three electronics systems were used to investigate the same fuel assemblies with various burn-ups and cooling times. The purpose of the exercise was to test the performance of the instrument for as well inspection purposes as for fuel management. The results are presented and discussed. (Author)

  16. Reactivity and neutron emission measurements of burnt PWR fuel rod samples in LWR-PROTEUS phase II

    International Nuclear Information System (INIS)

    Murphy, M. F.; Jatuff, F.; Grimm, P.; Seiler, R.; Brogli, R.; Meier, G.; Berger, H. D.; Chawla, R.

    2004-01-01

    Measurements have been made of the reactivity effects and the neutron emission rates of uranium oxide and mixed oxide burnt fuel samples having a wide range of burnup values and coming from a Pressurised Water Reactor (PWR). The reactivity measurements have been made in a PWR lattice moderated in turn with: water, a water and heavy water mixture, and water containing boron. An interesting relationship has been found between the neutron emission rate and the measured reactivity. (authors)

  17. Fission distribution measurements of Atucha's fuel pellets with solid state track detectors

    International Nuclear Information System (INIS)

    Ricabarra, M.D. Bovisio de; Waisman, Dina.

    1979-08-01

    Distribution of fissions in a UO 2 rod has been measured by means of solid state detectors. Mica muscovite and Makrofol-N detectors were used in the experiment. The merits of mica muscovite relative to the Makrofol-N for the detection of fission fragments have been verified. However both fission track detectors closely agree (0,5%) in the final fission distribution of the UO 2 rod. Sensitivity of the detectors shows to be linear in the range between 50.000and 360.000 fission tracks per square centimeter. Due to the high spatial resolution this method is better than any other technique. Determination were made in UO 2 pellets similar to the fuel element of the Atucha reactor. The average fission rate in the rod has been measured within 0,8% error, and provides an accurate determination for the distribution of fissions in the rod wich is needed for the determination of energy liberated per fission in the natural uranium rod.(author) [es

  18. The Tricky Art of Measuring Fossil Fuel Subsidies: A Critique of Existing Studies

    Directory of Open Access Journals (Sweden)

    Kenneth J. McKenzie

    2011-09-01

    Full Text Available Fossil fuel subsidies are of enormous import to policy-makers and public opinion, making it critical to properly define them. However, traditional methodologies tend to place subsidies in the realm of tax expenditure analysis, presenting a flawed picture. A recent report on government subsidies to the Canadian energy sector prepared for the International Institute for Sustainable Development exemplifies this flawed approach along several dimensions: it is not based on a robust underlying economic framework, it fails to account for complex interactions between tax and royalty systems in existing fiscal policy, and it uses a definition of subsidies that was created for a different purpose. The authors of this paper propose an alternative “economic view”, based on economic rents, which provides a neutral benchmark against which subsidies, royalties and other energy-focused fiscal measures can be measured. Using marginal effective tax rate (METR analysis, the authors show that it is possible to obtain a more accurate picture of energy subsidies and their impact on resource allocation and economic activity. This improved schema will ideally allow governments to better understand subsidies and devise sound policies, leading to less waste and distorted investment choices.

  19. Temperature measurement of the reactor materials samples irradiated in the fuel channels of the RA reactor - Annex 16

    International Nuclear Information System (INIS)

    Nikolic, M.; Djalovic, M.

    1964-01-01

    Reactor materials as graphite, stainless steel, magnox, zirconium alloys, etc. were exposed to fast neutron flux inside the fuel elements specially adapted for this purpose. Samples in the form ampoules were placed in capsules inside the fuel channels and cooled by heavy water which cools the fuel elements. In order to monitor the samples temperature 42 thermocouples were placed in the samples. That was necessary for reactor safety reasons and for further interpretation of measured results. Temperature monitoring was done continuously by multichannel milivoltmeters. This paper describes the technique of introducing the thermocouples, compensation instruments, control of the cold ends and adaptation of the instruments for precision (0.5%) temperature measurement in the range 30 deg - 130 deg C; 30 deg - 280 deg C and 30 deg - 80 deg C [sr

  20. Influence of heat flow on drift closure during Climax granite spent-fuel test: measurements and calculations

    International Nuclear Information System (INIS)

    Butkovich, T.R.; Yow, J.L. Jr.; Montan, D.N.

    1982-01-01

    Measurements of drift closure caused by the thermal load have been made routinely during the Spent Fuel Test in Climax granite since about six weeks after emplacement of the fuel. Horizontal and vertical closure was measured with a manually operated tape extensometer at various locations along the lengths of the drifts. Average closures ranged from 0 to 0.6 mm, horizontal and vertical, out to about 2.2 years since the start of the test. At the same time, displacements from the thermal loads were measured with rod extensometers emplaced to measure relative displacements between hole collars and anchor points in holes drilled from two parallel heater drifts. These data are compared with thermo-elastic finite element calculations which utilized measured properties of the Climax granite. The calculations show that more than half of the closures occur between fuel installation and the first closure measurement. The comparisons show that the results track each other, in that where closure followed by dilation is measured, the calculations also show this effect. The agreement is excellent, considering the averaged measured closures remain within 30% of the total calculated drift closures and the extremely small magnitude of the relative displacements (0.5 mm), measured or calculated

  1. U and Pu Gamma-Ray Measurements of Spent Fuel Using a Gamma-Ray Mirror Band-Pass Filter

    International Nuclear Information System (INIS)

    Ziock, K.-P.; Kisner, R.; Melin, A.; Patton, B.; Alameda, J.; Brejnhold, N.; Decker, T.; Descalle, M.-A.; Fernandez-Perea, M.; Hill, R.; Ruz Armendariz, J.; Soufli, R.

    2015-01-01

    We report the use of grazing incidence gamma-ray mirrors as narrow band-pass filters for advanced non-destructive analysis of spent nuclear fuel. The mirrors limit radiation reaching an HPGe detector to narrow spectral bands around characteristic emission lines from fissile isotopes in the fuel. Ideally, these emissions could be used to determine the fuel's fissile content, but they are normally masked by the overwhelming radiation emitted by short-lived fission by-products. These latter emissions raise the overall background, making direct observation of the fuel with HPGe detectors virtually impossible. Such observations can only be performed using precise collimators that restrict the detector's field of view to very small solid angles. This results in impracticably long dwell times for safeguards measurements targeting the weak isotopic lines of interest. In a proof-ofconcept experiment, a set of simple flat gamma-ray mirrors was used to observe the atomic florescence lines from U and Pu from a spent nuclear fuel pin. For the measurements, the mirrors were placed at the egress of an access port in a hot cell wall. A coarse collimator in the port restricted radiation from a fuel pin placed in front of the port to fully illuminate the front surface of the mirror assembly (0:5 x 3:8 cm2). The mirrors, consisting of highly polished silicon substrates deposited with WC/SiC multilayer coatings, were successfully used to deflect the lines of interest onto an HPGe detector while the intense primary radiation from the spent fuel was blocked by a lead beam stop. The gamma-ray mirror multilayer coatings used here at ∼100 keV, have been experimentally tested at energies as high as 645 keV, indicating that direct observation of nuclear emission lines from 239Pu should be possible with an appropriately designed optic. (author)

  2. Determining plasma-fueling sources with an end-loss ion spectrometer

    International Nuclear Information System (INIS)

    Grubb, D.P.; Foote, J.H.

    1986-01-01

    To help identify the major sources of fueling gas in Tandem Mirror Experiment-Upgrade (TMX-U), we mounted a mass-sensitive, E parallel to B, end-loss ion spectrometer (ELIS) near the machine's centerline. We set the electric field in the ELIS to simultaneously measure the axial loss currents of both hydrogen and deuterium. We then initiated plasma discharges, where we injected either hydrogen or deuterium gas into the central cell. We also selected and deselected the central-cell neutral beams that were fueled with hydrogen gas. The end-cell neutral beams were always selected and fueled with deuterium. By taking the ratio of the hydrogen end-loss current to the deuterium end-loss current (with a known deuterium-gas feed rate), we were able to infer the effective fueling rates that were due to wall reflux, central-cell beams, and end-cell beams. The results were the following: wall reflux, 6 Torr.l/s; central-cell beams, 15 Torr.l/s; and end-cell beams 1 Torr.l/s. 3 refs., 3 figs., 1 tab

  3. Analysis of reactivity worths of highly-burnt PWR fuel samples measured in LWR-PROTEUS Phase II

    Energy Technology Data Exchange (ETDEWEB)

    Grimm, Peter; Murphy, Michael F.; Jatuff, Fabian; Seiler, Rudolf [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland)

    2008-07-01

    The reactivity loss of PWR fuel with burnup has been determined experimentally by inserting fresh and highly-burnt fuel samples in a PWR test lattice in the framework of the LWR-PROTEUS Phase II programme. Seven UO{sub 2} samples irradiated in a Swiss PWR plant with burnups ranging from approx40 to approx120 MWd/kg and four MOX samples with burnups up to approx70 MWd/kg were oscillated in a test region constituted of actual PWR UO{sub 2} fuel rods in the centre of the PROTEUS zero-power experimental facility. The measurements were analyzed using the CASMO-4E fuel assembly code and a cross section library based on the ENDF/B-VI evaluation. The results show close proximity between calculated and measured reactivity effects and no trend for a deterioration of the quality of the prediction at high burnup. The analysis thus demonstrates the high accuracy of the calculation of the reactivity of highly-burnt fuel. (authors)

  4. Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ludewigt, Bernhard A; Quiter, Brian J.; Ambers, Scott D.

    2011-01-14

    The Next Generation Safeguard Initiative (NGSI) of the U.S Department of Energy is supporting a multi-lab/university collaboration to quantify the plutonium (Pu) mass in spent nuclear fuel (SNF) assemblies and to detect the diversion of pins with non-destructive assay (NDA) methods. The following 14 NDA techniques are being studied: Delayed Neutrons, Differential Die-Away, Differential Die-Away Self-Interrogation, Lead Slowing Down Spectrometer, Neutron Multiplicity, Passive Neutron Albedo Reactivity, Total Neutron (Gross Neutron), X-Ray Fluorescence, {sup 252}Cf Interrogation with Prompt Neutron Detection, Delayed Gamma, Nuclear Resonance Fluorescence, Passive Prompt Gamma, Self-integration Neutron Resonance Densitometry, and Neutron Resonance Transmission Analysis. Understanding and maturity of the techniques vary greatly, ranging from decades old, well-understood methods to new approaches. Nuclear Resonance Fluorescence (NRF) is a technique that had not previously been studied for SNF assay or similar applications. Since NRF generates isotope-specific signals, the promise and appeal of the technique lies in its potential to directly measure the amount of a specific isotope in an SNF assay target. The objectives of this study were to design and model suitable NRF measurement methods, to quantify capabilities and corresponding instrumentation requirements, and to evaluate prospects and the potential of NRF for SNF assay. The main challenge of the technique is to achieve the sensitivity and precision, i.e., to accumulate sufficient counting statistics, required for quantifying the mass of Pu isotopes in SNF assemblies. Systematic errors, considered a lesser problem for a direct measurement and only briefly discussed in this report, need to be evaluated for specific instrument designs in the future. Also, since the technical capability of using NRF to measure Pu in SNF has not been established, this report does not directly address issues such as cost, size

  5. Measuring the effect of fuel treatments on forest carbon using landscape risk analysis

    Science.gov (United States)

    A.A. Ager; M.A. Finney; A. McMahan; J. Carthcart

    2010-01-01

    Wildfire simulation modelling was used to examine whether fuel reduction treatments can potentially reduce future wildfire emissions and provide carbon benefits. In contrast to previous reports, the current study modelled landscape scale effects of fuel treatments on fire spread and intensity, and used a probabilistic framework to quantify wildfire effects on carbon...

  6. Measurement of gamma and neutron radiations inside spent fuel assemblies with passive detectors

    Czech Academy of Sciences Publication Activity Database

    Viererbl, L.; Lahodová, Z.; Voljanskij, A.; Klupák, V.; Koleška, M.; Cabalka, M.; Turek, Karel

    2011-01-01

    Roč. 652, č. 1 (2011), s. 90-93 ISSN 0168-9002 Institutional research plan: CEZ:AV0Z10480505 Keywords : fuel assembly * spent fuel * track detector Subject RIV: JF - Nuclear Energetics Impact factor: 1.207, year: 2011

  7. The feasibility of express in situ measurement of the isotopic composition of uranium in fresh WWER-1000 fuel

    International Nuclear Information System (INIS)

    Christoskov, I.; Tsankov, L.; Ivanov, N.

    2011-01-01

    A study of the feasibility and accuracy of measurement of the enrichment of fresh VVER-1000 fuel assemblies using a miniature CdZnTe probe is performed.The possibilities of improvement of the analytical procedure are briefly discussed. (authors)

  8. Wine ethanol C-14 as a tracer for fossil fuel CO2 emissions in Europe : Measurements and model comparison

    NARCIS (Netherlands)

    Palstra, Sanne W. L.; Karstens, Ute; Streurman, Harm-Jan; Meijer, Harro A. J.

    2008-01-01

    C-14 (radiocarbon) in atmospheric CO2 is the most direct tracer for the presence of fossil-fuel-derived CO2 (CO2-ff). We demonstrate the C-14 measurement of wine ethanol as a way to determine the relative regional atmospheric CO2-ff concentration compared to a background site ("regional CO2-ff

  9. Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly

    Directory of Open Access Journals (Sweden)

    Waseem

    2016-01-01

    Full Text Available Fuel assembly of the PWR nuclear power plant is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the fuel assembly (FA of Chashma Nuclear Power Plant-1 (CHASNUPP-1 at room temperature in air. The non-linear contact and structural tensile analysis have been performed using ANSYS 13.0, in order to determine the fuel assembly (FA elongation behaviour as well as the location and values of the stress intensity and stresses developed in axial direction under applied tensile load of 9800 N or 2 g being the fuel assembly handling or lifting load [Y. Zhang et al., Fuel assembly design report, SNERDI, China, 1994]. The finite element (FE model comprises spacer grids, fuel rods, flexible contacts between the fuel rods and grid's supports system and guide thimbles with dash-pots and flow holes, in addition to the spot welds between spacer grids and guide thimbles, has been developed using Shell181, Conta174 and Targe170 elements. FA is a non-straight structure. The actual behavior of the geometry is non-linear due to its curvature or design tolerance. It has been observed that fuel assembly elongation values obtained through FE analysis and experiment [SNERDI Tech. Doc., Mechanical strength and calculation for fuel assembly, Technical Report, F3.2.1, China, 1994] under applied tensile load are comparable and show approximately linear behaviors. Therefore, it seems that the permanent elongation of fuel assembly may not occur at the specified load. Moreover, the values of stresses obtained at different locations of the fuel assembly are also comparable with the stress values of the experiment determined at the same locations through strain gauges. Since the results of both studies (analytical and experimental are comparable, therefore, validation of the FE methodology is confirmed. The stress intensity of the FE model and maximum stresses developed along the guide thimbles in axial direction are

  10. Measurements and observations on microscopic swelling in MX-type fuels

    International Nuclear Information System (INIS)

    Ronchi, C.; Ray, I.L.F.; Thiele, H.; Laar, J. van de.

    1978-01-01

    Microscopic swelling has been investigated by electron microscopy in several MX-type fuels, irradiated in fast and thermal neutron flux. The results show that fission gas bubbles in these compounds grow to large sizes if the in-pile fuel temperature rises above a critical value (swelling critical temperature Tsub(C)). A comparison has been made of the swelling rates in fuels of different composition, showing that Tsub(C) increases from carbides to nitrides. In fuels subjected to in-pile restructuring (highly rated) He-bonded pins microscopic swelling is affected by pore and grain boundary migration. The influence of these phenomena on the fuel swelling performance has been discussed

  11. Measurements of Fission Cross Sections for the Isotopes relevant to the Thorium Fuel Cycle

    CERN Multimedia

    2002-01-01

    The present concern about a sustainable energy supply is characterised by a considerable uncertainty: the green house effect and foreseeable limits in fossil fuel resources on the one hand, the concern about the environmental impact of nuclear fission energy and the long term fusion research on the other hand, have led to the consideration of a variety of advanced strategies for the nuclear fuel cycle and related nuclear energy systems. The present research directories concern such strategies as the extension of the life span of presently operating reactors, the increase of the fuel burn-up, the plutonium recycling, and in particular the incineration of actinides and long-Lived fission products, the accelerator driven systems (ADS), like the "Energy Amplifier" (EA) concept of C. Rubbia, and the possible use of the Thorium fuel cycle. The detailed feasibility study and safety assessment of these strategies requires the accurate knowledge of neutron nuclear reaction data. Both, higher fuel burn-up and especiall...

  12. Fission product release measured during fuel damage tests at the Power Burst Facility

    International Nuclear Information System (INIS)

    Osetek, D.J.; Hartwell, J.K.; Vinjamuri, K.; Cronenberg, A.W.

    1985-01-01

    Results are presented of fission product release behavior observed during four severe fuel damage tests on bundles of UO 2 fuel rods. Transient temperatures up to fuel melting were obtained in the tests that included both rapid quench and slow cooldown, low and high (36 GWd/t) burnup fuel and the addition of Ag-In-Cd control rods. Release fractions of major fission product species and release rates of noble gas species are reported. Significant differences in release behavior are discussed between heatup and cooldown periods, low and high burnup fuel and long- and short-lived fission products. Explanations are offered for the probable reasons for the observed differences and recommendations for further studies are given

  13. Measurement of the Tracer Gradient and Sampling System Bias of the Hot Fuel Examination Facility Stack Air Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Glissmeyer, John A.; Flaherty, Julia E.

    2011-07-20

    This report describes tracer gas uniformity and bias measurements made in the exhaust air discharge of the Hot Fuel Examination Facility at Idaho National Laboratory. The measurements were a follow-up on earlier measurements which indicated a lack of mixing of the two ventilation streams being discharged via a common stack. The lack of mixing is detrimental to the accuracy of air emission measurements. The lack of mixing was confirmed in these new measurements. The air sampling probe was found to be out of alignment and that was corrected. The suspected sampling bias in the air sample stream was disproved.

  14. SFCOMPO 2.0 - A relational database of spent fuel isotopic measurements, reactor operational histories, and design data

    Science.gov (United States)

    Michel-Sendis, Franco; Martinez-González, Jesus; Gauld, Ian

    2017-09-01

    SFCOMPO-2.0 is a database of experimental isotopic concentrations measured in destructive radiochemical analysis of spent nuclear fuel (SNF) samples. The database includes corresponding design description of the fuel rods and assemblies, relevant operating conditions and characteristics of the host reactors necessary for modelling and simulation. Aimed at establishing a thorough, reliable, and publicly available resource for code and data validation of safety-related applications, SFCOMPO-2.0 is developed and maintained by the OECD Nuclear Energy Agency (NEA). The SFCOMPO-2.0 database is a Java application which is downloadable from the NEA website.

  15. SFCOMPO 2.0 – A relational database of spent fuel isotopic measurements, reactor operational histories, and design data

    Directory of Open Access Journals (Sweden)

    Michel-Sendis Franco

    2017-01-01

    Full Text Available SFCOMPO-2.0 is a database of experimental isotopic concentrations measured in destructive radiochemical analysis of spent nuclear fuel (SNF samples. The database includes corresponding design description of the fuel rods and assemblies, relevant operating conditions and characteristics of the host reactors necessary for modelling and simulation. Aimed at establishing a thorough, reliable, and publicly available resource for code and data validation of safety-related applications, SFCOMPO-2.0 is developed and maintained by the OECD Nuclear Energy Agency (NEA. The SFCOMPO-2.0 database is a Java application which is downloadable from the NEA website.

  16. Road and Street Centerlines, All data was acquired by GPS observation & manually corrected, where needed, by observing the linework on aerial photography, Published in 2012, Not Applicable scale, Chippewa County Government.

    Data.gov (United States)

    NSGIC Local Govt | GIS Inventory — Road and Street Centerlines dataset current as of 2012. All data was acquired by GPS observation & manually corrected, where needed, by observing the linework on...

  17. Road Edge of Pavement, Pavement Condition Ratings and Maintenance Designations per Street Centerline. Updated frequently by Public Works Streets Division, Published in 2006, 1:1200 (1in=100ft) scale, City of Asheville Government.

    Data.gov (United States)

    NSGIC Local Govt | GIS Inventory — Road Edge of Pavement dataset current as of 2006. Pavement Condition Ratings and Maintenance Designations per Street Centerline. Updated frequently by Public Works...

  18. Road and Street Centerlines, Street-The data set is a line feature consisting of 13948 line segments representing streets. It was created to maintain the location of city and county based streets., Published in 1989, Davis County Government.

    Data.gov (United States)

    NSGIC Local Govt | GIS Inventory — Road and Street Centerlines dataset current as of 1989. Street-The data set is a line feature consisting of 13948 line segments representing streets. It was created...

  19. Tracer-based laser-induced fluorescence measurement technique for quantitative fuel/air-ratio measurements in a hydrogen internal combustion engine.

    Science.gov (United States)

    Blotevogel, Thomas; Hartmann, Matthias; Rottengruber, Hermann; Leipertz, Alfred

    2008-12-10

    A measurement technique for the quantitative investigation of mixture formation processes in hydrogen internal combustion engines (ICEs) has been developed using tracer-based laser-induced fluorescence (TLIF). This technique can be employed to fired and motored engine operation. The quantitative TLIF fuel/air-ratio results have been verified by means of linear Raman scattering measurements. Exemplary results of the simultaneous investigation of mixture formation and combustion obtained at an optical accessible hydrogen ICE are shown.

  20. Qualification of power determination and in-pile measurements in the UO{sub 2} Gd{sub 2} 0{sub 2} fuel irradiation test IFA 636

    Energy Technology Data Exchange (ETDEWEB)

    Tverberg, T.; Volkov, B.; Kim, J-C.

    2004-04-15

    IFA-S36 is irradiated with the main objective of extending the database on the performance of UO{sub 2}Gd{sub 2}O{sub 2} fuel (with 8% absorbing gadolinia isotopes) compared with commercial UO{sub 2}. The rig carries 6 rods in the lower cluster (including three Gd-doped fuel rods) and 3 rods in the upper cluster (one rod with Gd-doped fuel). The rods are instrumented with expansion thermometers (ETs), fuel and cladding elongation detectors (EFs and ECs) and pressure transducers (PFs). Repeated calorimetric power measurements, physics calculations by the HELIOS code and gamma scans of selected rods in both clusters enabled the power and burnup determination to be qualified and corrected. The data suggest that as of May 2004 the power ratings in both fuels are much alike and burnups are about 30 and 34 MW/kgUO{sub 2} in the Gd-doped and ordinary UO{sub 2} rods respectively. Analysis of in-pile measurements compared with calculations shows that neutron absorption affects fuel temperature, power and burnup radial distributions in Gd-doped fuel at BOL compared with UO{sub 2} fuel. Sensitivity analyses performed with the HELIOS and FTEMP3 codes show that fuel centreline temperature in Gd-doped fuel is influenced by radial power depression, depletion of fissile materials and absorbing Gd isotopes as well as thermal conductivity of the fuel matrix and its degradation during irradiation. Analysis of the fuel dimension changes revealed densification only in the UO{sub 2} fuel whereas fuel elongation measurements in the Gd-doped fuel rods indicated essentially constant swelling with burnup. At burnups above 5 MWd/kgUO{sub 2} the swelling rate was about 0.5-O.fi % DELTAV/V per 10 MWd/kgUO{sub 2} for both fuel types. Internal pressure measured in the Gd-doped rod at BOL showed slight fuel densification and possibly He gas absorption, whereas derived swelling rate was somewhat Iarger than values obtained from the fuel elongation measurements. Cladding elongation measurements

  1. Measuring and monitoring fuel poverty in the UK: National and regional perspectives

    International Nuclear Information System (INIS)

    Liddell, Christine; Morris, Chris; McKenzie, S.J.P.; Rae, Gordon

    2012-01-01

    Throughout the industrialised world, fuel poverty is the most commonly accepted term with which to describe a household's inability to afford basic standards of heat, power and light. Whilst the term gained widespread acceptance with the publication of the UK's Fuel Poverty Strategy in 2001, little is known about the origins of the term itself. This paper traces the earliest formulations of the concept, focusing particularly on the 10% needs to spend threshold which was adopted in 1991 and remains in place some 20 years later. This paper argues that understanding more about the origins of this threshold yields a more critical understanding of why fuel poverty targets in the UK have not been reached, and enables a more informed approach to setting realistic targets for the future. It also provides an opportunity to explore regional disparities in UK fuel poverty prevalence, highlighting the extent to which rigid adherence to a 10% threshold has created an unstable regional mosaic of over-estimation and under-estimation. - Highlights: ► We highlight the origins of the 10% threshold used to define fuel poverty in the UK. ► This sheds new light on why fuel poverty targets in the UK have not been reached. ► It also helps interpret regional disparities in UK fuel poverty prevalence. ► The threshold has created an unstable mosaic of over- and under-estimation.

  2. Measuring of Traction and Speed Characteristics as Well as of Fuel Economy of a Car in Road Conditions

    OpenAIRE

    Krivtsov, Sergey; Syrbakov, Andrey Pavlovich; Korchuganova, Mariya Anatolievna

    2016-01-01

    This article is devoted to the identification of traction and speed characteristics as well as of fuel economy of motor vehicles in road conditions. Among common variants of measuring of the above stated values, the preference was given to the immediate gaining of factors by means of a computer-aided measuring system. There is a theoretical justification given to the suggested approach as well as methods and results allowing to provide a practically sufficient solution accuracy of the problem.

  3. Measuring of Traction and Speed Characteristics as Well as of Fuel Economy of a Car in Road Conditions

    Science.gov (United States)

    Krivtsov, Sergey N.; Syrbakov, Andrey P.; Korchuganova, Marina A.

    2016-08-01

    This article is devoted to the identification of traction and speed characteristics as well as of fuel economy of motor vehicles in road conditions. Among common variants of measuring of the above stated values, the preference was given to the immediate gaining of factors by means of a computer-aided measuring system. There is a theoretical justification given to the suggested approach as well as methods and results allowing to provide a practically sufficient solution accuracy of the problem.

  4. On the problems relating to the accuracy of the measurement of fuel pin diameters by neutron radiography

    International Nuclear Information System (INIS)

    Matfield, R.

    1983-01-01

    The paper identifies the sources of error in the neutron radiographic system and attempts to estimate some of these errors. The sources of error are in the fuel pin materials, the radiographic set-up, the radiographic equipment, image formation, the microdensitometer, the edge criteria and the dimensional measurement from the microdensitometer trace. However, the critical problem area is that of determining a representative edge criteria and upon this will depend the ability of the method to achieve the required measurement accuracy. (Auth.)

  5. Measuring device for the distribution of burn-up degree in fuel assembly irradiated in nuclear reactor

    International Nuclear Information System (INIS)

    Kumanomido, Hironori

    1989-01-01

    The object of the invention is to measure the distribution of burn-up degree, of fuel assemblies irradiated in a nuclear reactor in a short time and exactly. That is, the device comprises a device main body having substantially the same length as that for the axial length of a fuel assembly and a detector container disposed axially slidably to the main body. A plurality of radiation detectors are arranged at an equi-axial pitch and contained in the container. The container is caused to slide at a pitch equal to the equi-axial distance of the detectors. In the device having thus been constituted, measurement is conducted at least for twice at an axial position on the side of a fuel assembly irradiated in the nuclear reactor and a position caused to slide therefrom by one pitch. Based on the result, the sensitivities between each of the detectors are compared and the relative sensitivity of the radiation detectors is calibrated. Accordingly, the sensitivity between each of the detectors can be calibrated rapidly and easily. As a result, the distribution of the burn-up degree, etc of irradiated fuel assembly can be measured exactly. (K.M.)

  6. Determination of technical details concerning measures for transportation of nuclear fuel materials in the works or the enterprise

    International Nuclear Information System (INIS)

    1979-01-01

    The determination is defined under the regulations concerning the fabricating business of nuclear fuel materials, the regulation concerning installation and operation of test reactor, the regulations concerning the reprocessing business of spent fuel and the regulations concerning the uses of nuclear source materials. The notification determining technical details concerning measures for transportation of nuclear fuel materials is abolished. Measures for prevention of hazard designated by the Director General of Science and Technology Agency include such ones not to let radioactive materials easily fly about or leak in regular transport, not to let rain water easily penetrate or make each exterior side of a cubic load more than 10 centi-meters. The application for permission shall be filed for transportation of things highly difficult to be sealed in a vessel, listing name and address of the applicant, kind, quantity, form and nature of the load contaminated by nuclear fuel materials, date and route of transfer and measures taken for prevention of hazard in transport. Radiation doses of load and transporting apparatus are stipulated by the Director for an hour as 200 mili-rem on the surface of load, 10 mili-rem at the distance of 1 meter from the surface of load, and 200 mili-rem on the surface of the vehicle, etc. Dangerous things, signals and radiation dose of particular loads are specified respectively. (Okada, K.)

  7. On the combination of delayed neutron and delayed gamma techniques for fission rate measurement in nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Perret, G.; Jordan, K. A. [Paul Scherrer Institut, Villigen, 5232 (Switzerland)

    2011-07-01

    Novel techniques to measure newly induced fissions in spent fuel after re-irradiation at low power have been developed and tested at the Proteus zero-power research reactor. The two techniques are based on the detection of high energy gamma-rays emitted by short-lived fission products and delayed neutrons. The two techniques relate the measured signals to the total fission rate, the isotopic composition of the fuel, and nuclear data. They can be combined to derive better estimates on each of these parameters. This has potential for improvement in many areas. Spent fuel characterisation and safeguard applications can benefit from these techniques for non-destructive assay of plutonium content. Another application of choice is the reduction of uncertainties on nuclear data. As a first application of the combination of the delayed neutron and gamma measurement techniques, this paper shows how to reduce the uncertainties on the relative abundances of the longest delayed neutron group for thermal fissions in {sup 235}U, {sup 239}Pu and fast fissions in {sup 238}U. The proposed experiments are easily achievable in zero-power research reactors using fresh UO{sub 2} and MOX fuel and do not require fast extraction systems. The relative uncertainties (1{sigma}) on the relative abundances are expected to be reduced from 13% to 4%, 16% to 5%, and 38% to 12% for {sup 235}U, {sup 238}U and {sup 239}Pu, respectively. (authors)

  8. Measuring the distribution of equity in terms of energy, environmental, and economic costs in the fuel cycles of alternative fuel vehicles with hydrogen pathway scenarios

    Science.gov (United States)

    Meyer, Patrick E.

    Numerous analyses exist which examine the energy, environmental, and economic tradeoffs between conventional gasoline vehicles and hydrogen fuel cell vehicles powered by hydrogen produced from a variety of sources. These analyses are commonly referred to as "E3" analyses because of their inclusion of Energy, Environmental, and Economic indicators. Recent research as sought a means to incorporate social Equity into E3 analyses, thus producing an "E4" analysis. However, E4 analyses in the realm of energy policy are uncommon, and in the realm of alternative transportation fuels, E4 analyses are extremely rare. This dissertation discusses the creation of a novel E4 simulation tool usable to weigh energy, environmental, economic, and equity trade-offs between conventional gasoline vehicles and alternative fuel vehicles, with specific application to hydrogen fuel cell vehicles. The model, dubbed the F uel Life-cycle Analysis of Solar Hydrogen -- Energy, Environment, Economic & Equity model, or FLASH-E4, is a total fuel-cycle model that combines energy, environmental, and economic analysis methodologies with the addition of an equity analysis component. The model is capable of providing results regarding total fuel-cycle energy consumption, emissions production, energy and environmental cost, and level of social equity within a population in which low-income drivers use CGV technology and high-income drivers use a number of advanced hydrogen FCV technologies. Using theories of equity and social indicators conceptually embodied in the Lorenz Curve and Gini Index, the equity of the distribution of societal energy and environmental costs are measured for a population in which some drivers use CGVs and other drivers use FCVs. It is found, based on baseline input data representative of the United States (US), that the distribution of energy and environmental costs in a population in which some drivers use CGVs and other drivers use natural gas-based hydrogen FCVs can be

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

    Science.gov (United States)

    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.

  10. Transcontinental methane measurements: Part 2. Mobile surface investigation of fossil fuel industrial fugitive emissions

    Science.gov (United States)

    Leifer, Ira; Culling, Daniel; Schneising, Oliver; Farrell, Paige; Buchwitz, Michael; Burrows, John P.

    2013-08-01

    The potent greenhouse gas, methane, CH4, has a wide variety of anthropogenic and natural sources. Fall, continental-scale (Florida to California) surface CH4 data were collected to investigate the importance of fossil fuel industrial (FFI) emissions in the South US. A total of 6600 measurements along 7020-km of roadways were made by flame ion detection gas chromatography onboard a nearly continuously moving recreational vehicle in 2010. A second, winter survey in Southern California measured CH4 at 2 Hz with a cavity ring-down spectrometer in 2012. Data revealed strong and persistent FFI CH4 sources associated with refining, oil/gas production, a presumed major pipeline leak, and a coal loading plant. Nocturnal CH4 mixing ratios tended to be higher than daytime values for similar sources, sometimes significantly, which was attributed to day/night meteorological differences, primarily changes in the boundary layer height. The highest CH4 mixing ratio (39 ppm) was observed near the Kern River Oil Field, California, which uses steam reinjection. FFI CH4 plume signatures were distinguished as stronger than other sources on local scales. On large (4°) scales, the CH4 trend was better matched spatially with FFI activity than wetland spatial patterns. Qualitative comparison of surface data with SCIAMACHY and GOSAT satellite retrievals showed agreement of the large-scale CH4 spatial patterns. Comparison with inventory models and seasonal winds suggests for some seasons and some portions of the Gulf of Mexico a non-negligible underestimation of FFI emissions. For other seasons and locations, qualitative interpretation is not feasible. Unambiguous quantitative source attribution is more complex, requiring transport modeling.

  11. Carbon conversion predictor for fluidized bed gasification of biomass fuels - from TGA measurements to char gasification particle model

    Energy Technology Data Exchange (ETDEWEB)

    Konttinen, J.T. [University of Jyvaeskylae, Department of Chemistry, Renewable Energy Programme, POB 35, Jyvaeskylae (Finland); Moilanen, A. [VTT Technical Research Centre of Finland, POB 1000, Espoo (Finland); Martini, N. de; Hupa, M. [Abo Akademi University, Process Chemistry Centre, Combustion and Materials Chemistry, Turku (Finland)

    2012-09-15

    When a solid fuel particle is injected into a hot fluidized bed, the reactivity of fuel char in gasification reactions (between char carbon and steam and CO{sub 2}) plays a significant role for reaching a good carbon conversion. In this paper, the gasification reactivity data of some solid waste recovered fuels (SRF) obtained from thermogravimetric analysis (TGA) experiments is presented. Gas mixtures (H{sub 2}O, H{sub 2}, CO{sub 2}, CO), were used in the experiments to find the inhibitive effects of CO and H{sub 2}. Average char gasification reactivity values are determined from the TGA results. Kinetic parameters for char carbon gasification reactivity correlations are determined from this data. The Uniform Conversion model is used to account for the change of gasification reaction rate as function of carbon conversion. Some discrepancies, due to complicated ash-carbon interactions, are subjects of further research. In the carbon conversion predictor, laboratory measured reactivity numbers are converted into carbon conversion numbers in a real-scale fluidized bed gasifier. The predictor is a relatively simple and transparent tool for the comparison of the gasification reactivity of different fuels in fluidized bed gasification. The residence times for solid fuels in fluidized bed gasifiers are simulated. Simulations against some pilot-scale results show reasonable agreement. (orig.)

  12. Synchronized droplet size measurements for Coal-Water-Slurry (CWS) diesel sprays of an electronically-controlled fuel injection system

    Science.gov (United States)

    Kihm, K. D.; Terracina, D. P.; Payne, S. E.; Caton, J. A.

    Experiments were completed to study intermittent coal-water slurry (CWS) fuel sprays injected from an electronically-controlled accumulator injector system. A laser diffraction particle analyzing (LDPA) technique was used to measure the spray diameters (Sauter mean diameter, SMD) assuming the Rosin-Rammler two parameter model. In order to ensure an accurate synchronization of the measurement with the intermittent sprays, a new synchronization technique was developed using the light extinction signal as a triggering source for the data taking initiation. This technique allowed measurement of SMD's near the spray tip where the light extinction was low and the data were free from the multiscattering bias. Coal-water slurry fuel with 50% coal loading in mass containing 5 (mu)m mass median diameter coal particulates was considered. Injection pressures ranging from 28 to 110 MPa, two different nozzle orifice diameters, 0.2 ad 0.4 mm, and four axial measurement locations from 60 to 120 mm from the nozzle orifice were studied. Measurements were made for pressurized (2.0 MPa in gauge) and for ambient chamber conditions. The spray SMD showed an increase with the distance of the axial measurement location and with the ambient gas density, and showed a decrease with increasing injection pressure. A correlation of the Sauter mean diameter with the injection conditions was determined. The results were also compared with previous SMD correlations that were available only for diesel fuel sprays.

  13. Measurements of moisture content in wood fuels with dual energy x-ray; Maetning av fukt i biobraenslen med dubbelenergiroentgen

    Energy Technology Data Exchange (ETDEWEB)

    Nordell, Anders [Industriellt Utvecklingscentrum i Karlskoga, Karlskoga (Sweden); Vikterloef, K J [Oerebro Radiofysik AB, Oerebro (Sweden)

    2000-04-01

    Good methods enabling quick determination of moisture in wood fuels are lacking. A standardized method generally used by the biofuel industry means drying samples in furnaces during at least 16 hours. The idea of this work has been to investigate if x-ray technique can offer a superior alternative for fast moisture determination. Analysis with x-ray has many advantages. The x-rays go completely through the whole test specimen. Large samples or many sequential samples can rapidly be measured. The penetration is extraordinary good; all parts of the tested mass will be analyzed. The measurement relies on the absorption of gamma photons in the tested mass. The x-ray absorption is carried out deep down in the atoms and therefore is independent of the materials equation of state as temperature, grain size or molecule binding. The disadvantages of the method are the general anxiety which can be for gamma rays, the relative high cost for equipment and the fact that an x-ray tube will wear with time and needs replacement. The method used here for moisture measurement is known as dual energy x-ray. The test sample is simultaneously exposed to transmission of two separate x-ray energies. The two energies are absorbed differently by the sample. The ratio of the absorption in the two respective energies is calculated and from that conclusions on the composition of the material can be drawn. The equipment used in this test is a standard medical apparatus adapted for the study of human body composition. When instead used for measurements of moisture in wood material the discrimination signal is about half of that achieved in the normal medical case. Despite that, the signal has been good enough for giving satisfactory accuracy in the moisture estimation. A little more than sixty approved results are presented in the report comprehending ten different fuel types. The results are grouped into three main categories depending on fuel type, wood fuels, peat fuels and salix including

  14. Comparison of Calculated and Measured Neutron Fluence in Fuel/Cladding Irradiation Experiments in HFIR

    International Nuclear Information System (INIS)

    Ellis, Ronald James

    2011-01-01

    A recently-designed thermal neutron irradiation facility has been used for a first series of irradiations of PWR fuel pellets in the high flux isotope reactor (HFIR) at Oak Ridge National Laboratory. Since June 2010, irradiations of PWR fuel pellets made of UN or UO 2 , clad in SiC, have been ongoing in the outer small VXF sites in the beryllium reflector region of the HFIR, as seen in Fig. 1. HFIR is a versatile, 85 MW isotope production and test reactor with the capability and facilities for performing a wide variety of irradiation experiments. HFIR is a beryllium-reflected, light-water-cooled and -moderated, flux-trap type reactor that uses highly enriched (in 235 U) uranium (HEU) as the fuel. The reactor core consists of a series of concentric annular regions, each about 2 ft (0.61 m) high. A 5-in. (12.70-cm)-diam hole, referred to as the flux trap, forms the center of the core. The fuel region is composed of two concentric fuel elements made up of many involute-shaped fuel plates: an inner element that contains 171 fuel plates, and an outer element that contains 369 fuel plates. The fuel plates are curved in the shape of an involute, which provides constant coolant channel width between plates. The fuel (U 3 O 8 -Al cermet) is nonuniformly distributed along the arc of the involute to minimize the radial peak-to-average power density ratio. A burnable poison (B 4 C) is included in the inner fuel element primarily to reduce the negative reactivity requirements of the reactor control plates. A typical HEU core loading in HFIR is 9.4 kg of 235 U and 2.8 g of 10 B. The thermal neutron flux in the flux trap region can exceed 2.5 x 10 15 n/cm 2 · s while the fast flux in this region exceeds 1 x 10 15 n/cm 2 · s. The inner and outer fuel elements are in turn surrounded by a concentric ring of beryllium reflector approximately 1 ft (0.30 m) thick. The beryllium reflector consists of three regions: the removable reflector, the semi-permanent reflector, and the

  15. In-core instrumentation and in-situ measurement in connection with fuel behaviour. Working material

    International Nuclear Information System (INIS)

    1996-01-01

    The subject of this meeting has been touched on briefly in most of the Specialist's and topical meetings related to fuel behaviour. On the basis of the conclusions and recommendations of these meetings the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended the Agency to organize a dedicated Specialist's Meeting on the subject. The twenty one papers covered the instrumentation, sensors, methods and computer codes currently used in Material Test Reactor (MTR) and power reactors as well as improved instrumentation and methods. The meeting acknowledged the fast development of fuel modelling and therefore the growing need of dedicated high burnup fuel experiments carried out in MTR reactors on refabricated rods from power reactors. In order to reduce safety margins in power reactors, thus improving economics, the necessity to develop more sophisticated on-line calculations, based on improved sensors, was recognized, although this development is limited by insufficient knowledge of the mechanisms involved. Refs, figs, tabs

  16. Ignition delay measurements of light naphtha: A fully blended low octane fuel

    KAUST Repository

    Javed, Tamour; Nasir, Ehson Fawad; Ahmed, Ahfaz; Badra, Jihad; Djebbi, Khalil; Beshir, Mohamed; Ji, Weiqi; Sarathy, Mani; Farooq, Aamir

    2016-01-01

    . To the best of our knowledge, this is the first fundamental autoignition study on the reactivity of a low-octane fully blended fuel and the use of a suitably formulated multi-component surrogate to model its behavior.

  17. Thermal performance of fresh mixed-oxide fuel in a fast flux LMR [liquid metal reactor

    International Nuclear Information System (INIS)

    Ethridge, J.L.; Baker, R.B.

    1985-01-01

    A test was designed and irradiated to provide power-to-melt (heat generation rate necessary to initiate centerline fuel melting) data for fresh mixed-oxide UO 2 -PuO 2 fuel irradiated in a fast neutron flux under prototypic liquid metal reactor (LMR) conditions. The fuel pin parameters were selected to envelope allowable fabrication ranges and address mass production of LMR fuel using sintered-to-size techniques. The test included fuel pins with variations in fabrication technique, pellet density, fuel-to-cladding gap, Pu concentration, and fuel oxygen-to-metal ratios. The resulting data base has reestablished the expected power-to-melt in mixed-oxide fuels during initial reactor startup when the fuel temperatures are expected to be the highest. Calibration of heat transfer models of fuel pin performance codes with these data are providing more accurate capability for predicting steady-state thermal behavior of current and future mixed-oxide LMR fuels

  18. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    Energy Technology Data Exchange (ETDEWEB)

    Evans, Louise G., E-mail: evanslg@ornl.gov [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Swinhoe, Martyn T.; Menlove, Howard O. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Schwalbach, Peter; Baere, Paul De [European Commission, Euratom Safeguards Office (Luxembourg); Browne, Michael C. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2013-11-21

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd{sub 2}O{sub 3}) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available {sup 241}AmLi (α,n) interrogation source strength of 5.7×10{sup 4} s{sup −1}. Furthermore, the calibration range of the new collar has been extended to verify {sup 235}U content in variable PWR fuel

  19. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    International Nuclear Information System (INIS)

    Evans, Louise G.; Swinhoe, Martyn T.; Menlove, Howard O.; Schwalbach, Peter; Baere, Paul De; Browne, Michael C.

    2013-01-01

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd 2 O 3 ) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241 AmLi (α,n) interrogation source strength of 5.7×10 4 s −1 . Furthermore, the calibration range of the new collar has been extended to verify 235 U content in variable PWR fuel designs in the presence of up to

  20. Development and Validation of Capabilities to Measure Thermal Properties of Layered Monolithic U-Mo Alloy Plate-Type Fuel

    Science.gov (United States)

    Burkes, Douglas E.; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

    2014-07-01

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium to low enriched uranium. One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the thermal-conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify functionality of equipment installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, refine procedures to operate the equipment, and validate models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures, and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a Zr diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

  1. The high temperature out-of-pile test of LVDT for internal pressure measurement of nuclear fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Son, J. M.; Kim, B. K.; Kim, D. S.; Yoon, K. B.; Sin, Y. T.; Park, S. J.; Kang, Y. H. [KAERI, Taejon (Korea, Republic of)

    2002-10-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), the internal pressure measurement technique of the nuclear fuel rod is being developed using LVDT(Linear Variable Differential Transformer). As the results of out-of-pile test at room temperature, it was concluded that the well qualified out-of-pile tests were needed to understand the LVDT's detail characteristics at high temperature for the detail design of the fuel irradiation capsule, because LVDT is very sensitive to variation of temperature. Therefore, the high temperature out-of-pile test system for pressure measurement was developed, and this test was performed under the temperature condition between room temperature and 300 .deg. C increasing the pressure from 0 bar to 30 bar. The LVDT's high temperature characteristics and temperature sensitivity of LVDT were analyzed through this experiment. Based on the result of this test, the method for the application of LVDT at high temperature was introduced. It is known that the results will be used to predict accurately the internal pressure of fuel rod during irradiation test.

  2. Laboratory measurements of trace gas emissions from biomass burning of fuel types from the southeastern and southwestern United States

    Science.gov (United States)

    Burling, I. R.; Yokelson, R. J.; Griffith, D. W. T.; Johnson, T. J.; Veres, P.; Roberts, J. M.; Warneke, C.; Urbanski, S. P.; Reardon, J.; Weise, D. R.; Hao, W. M.; de Gouw, J.

    2010-11-01

    Vegetation commonly managed by prescribed burning was collected from five southeastern and southwestern US military bases and burned under controlled conditions at the US Forest Service Fire Sciences Laboratory in Missoula, Montana. The smoke emissions were measured with a large suite of state-of-the-art instrumentation including an open-path Fourier transform infrared (OP-FTIR) spectrometer for measurement of gas-phase species. The OP-FTIR detected and quantified 19 gas-phase species in these fires: CO2, CO, CH4, C2H2, C2H4, C3H6, HCHO, HCOOH, CH3OH, CH3COOH, furan, H2O, NO, NO2, HONO, NH3, HCN, HCl, and SO2. Emission factors for these species are presented for each vegetation type burned. Gas-phase nitrous acid (HONO), an important OH precursor, was detected in the smoke from all fires. The HONO emission factors ranged from 0.15 to 0.60 g kg-1 and were higher for the southeastern fuels. The fire-integrated molar emission ratios of HONO (relative to NOx) ranged from approximately 0.03 to 0.20, with higher values also observed for the southeastern fuels. The majority of non-methane organic compound (NMOC) emissions detected by OP-FTIR were oxygenated volatile organic compounds (OVOCs) with the total identified OVOC emissions constituting 61 ± 12% of the total measured NMOC on a molar basis. These OVOC may undergo photolysis or further oxidation contributing to ozone formation. Elevated amounts of gas-phase HCl and SO2 were also detected during flaming combustion, with the amounts varying greatly depending on location and vegetation type. The fuels with the highest HCl emission factors were all located in the coastal regions, although HCl was also observed from fuels farther inland. Emission factors for HCl were generally higher for the southwestern fuels, particularly those found in the chaparral biome in the coastal regions of California.

  3. Velocity measurements in the near field of a diesel fuel injector by ultrafast imagery

    Science.gov (United States)

    Sedarsky, David; Idlahcen, Saïd; Rozé, Claude; Blaisot, Jean-Bernard

    2013-02-01

    This paper examines the velocity profile of fuel issuing from a high-pressure single-orifice diesel injector. Velocities of liquid structures were determined from time-resolved ultrafast shadow images, formed by an amplified two-pulse laser source coupled to a double-frame camera. A statistical analysis of the data over many injection events was undertaken to map velocities related to spray formation near the nozzle outlet as a function of time after start of injection. These results reveal a strong asymmetry in the liquid profile of the test injector, with distinct fast and slow regions on opposite sides of the orifice. Differences of ˜100 m/s can be observed between the `fast' and `slow' sides of the jet, resulting in different atomization conditions across the spray. On average, droplets are dispersed at a greater distance from the nozzle on the `fast' side of the flow, and distinct macrostructure can be observed under the asymmetric velocity conditions. The changes in structural velocity and atomization behavior resemble flow structures which are often observed in the presence of string cavitation produced under controlled conditions in scaled, transparent test nozzles. These observations suggest that widely used common-rail supply configurations and modern injectors can potentially generate asymmetric interior flows which strongly influence diesel spray morphology. The velocimetry measurements presented in this work represent an effective and relatively straightforward approach to identify deviant flow behavior in real diesel sprays, providing new spatially resolved information on fluid structure and flow characteristics within the shear layers on the jet periphery.

  4. Measuring the effect of fuel treatments on forest carbon using landscape risk analysis

    Directory of Open Access Journals (Sweden)

    A. A. Ager

    2010-12-01

    Full Text Available Wildfire simulation modelling was used to examine whether fuel reduction treatments can potentially reduce future wildfire emissions and provide carbon benefits. In contrast to previous reports, the current study modelled landscape scale effects of fuel treatments on fire spread and intensity, and used a probabilistic framework to quantify wildfire effects on carbon pools to account for stochastic wildfire occurrence. The study area was a 68 474 ha watershed located on the Fremont-Winema National Forest in southeastern Oregon, USA. Fuel reduction treatments were simulated on 10% of the watershed (19% of federal forestland. We simulated 30 000 wildfires with random ignition locations under both treated and untreated landscapes to estimate the change in burn probability by flame length class resulting from the treatments. Carbon loss functions were then calculated with the Forest Vegetation Simulator for each stand in the study area to quantify change in carbon as a function of flame length. We then calculated the expected change in carbon from a random ignition and wildfire as the sum of the product of the carbon loss and the burn probabilities by flame length class. The expected carbon difference between the non-treatment and treatment scenarios was then calculated to quantify the effect of fuel treatments. Overall, the results show that the carbon loss from implementing fuel reduction treatments exceeded the expected carbon benefit associated with lowered burn probabilities and reduced fire severity on the treated landscape. Thus, fuel management activities resulted in an expected net loss of carbon immediately after treatment. However, the findings represent a point in time estimate (wildfire immediately after treatments, and a temporal analysis with a probabilistic framework used here is needed to model carbon dynamics over the life cycle of the fuel treatments. Of particular importance is the long-term balance between emissions from the

  5. Measuring the effect of fuel treatments on forest carbon using landscape risk analysis

    Science.gov (United States)

    Ager, A. A.; Finney, M. A.; McMahan, A.; Cathcart, J.

    2010-12-01

    Wildfire simulation modelling was used to examine whether fuel reduction treatments can potentially reduce future wildfire emissions and provide carbon benefits. In contrast to previous reports, the current study modelled landscape scale effects of fuel treatments on fire spread and intensity, and used a probabilistic framework to quantify wildfire effects on carbon pools to account for stochastic wildfire occurrence. The study area was a 68 474 ha watershed located on the Fremont-Winema National Forest in southeastern Oregon, USA. Fuel reduction treatments were simulated on 10% of the watershed (19% of federal forestland). We simulated 30 000 wildfires with random ignition locations under both treated and untreated landscapes to estimate the change in burn probability by flame length class resulting from the treatments. Carbon loss functions were then calculated with the Forest Vegetation Simulator for each stand in the study area to quantify change in carbon as a function of flame length. We then calculated the expected change in carbon from a random ignition and wildfire as the sum of the product of the carbon loss and the burn probabilities by flame length class. The expected carbon difference between the non-treatment and treatment scenarios was then calculated to quantify the effect of fuel treatments. Overall, the results show that the carbon loss from implementing fuel reduction treatments exceeded the expected carbon benefit associated with lowered burn probabilities and reduced fire severity on the treated landscape. Thus, fuel management activities resulted in an expected net loss of carbon immediately after treatment. However, the findings represent a point in time estimate (wildfire immediately after treatments), and a temporal analysis with a probabilistic framework used here is needed to model carbon dynamics over the life cycle of the fuel treatments. Of particular importance is the long-term balance between emissions from the decay of dead trees

  6. Automated nondestructive assay system for the measurement of irradiated Rover fuel

    International Nuclear Information System (INIS)

    Augustson, R.H.; Menlove, H.O.; Smith, D.B.; Bond, A.L.; Durrill, D.C.; Hollowell, W.P.; Bromley, C.P.

    1975-01-01

    With the termination of the Nuclear Rocket Propulsion (Rover) Program, and associated reactor testing at the Nuclear Rocket Development Station (NRDS), Nevada, plans are progressing to recover the 93 percent enriched uranium contained in irradiated fuel from twenty various test reactors. This fuel is being packaged into 7-cm-dia by 137-cm-long cardboard tubes, using the remote handling facilities (E-MAD Bldg) of NRDS. After packaging, the fuel is shipped to Allied Chemical Corporation, Idaho Falls, Idaho, for uranium recovery. About 4000 tubes will be needed to package and ship the inventory of fuel elements presently at NRDS. This represents a total of approximately 2500 kg of enriched uranium. To complete the accounting records each tube is being nondestructively assayed and records kept on a reactor-by-reactor basis where possible. The assayed values for a reactor are then compared with original input inventory values and discrepancies resolved. The tubes are being assayed by an active neutron interrogation system designed and fabricated by Los Alamos Scientific Laboratory (LASL) and operated by Westinghouse Astronuclear Laboratory (WANL)-Nevada Operations personnel. WANL is the operating contractor in charge of loading and shipping this fuel. (U.S.)

  7. Spent fuel test-climax: technical measurements interim report, FY 1980

    International Nuclear Information System (INIS)

    Carlson, R.C.; Patrick, W.C.; Wilder, D.G.; Brough, W.G.; Montan, D.N.; Harben, P.E.; Ballou, L.B.; Heard, H.C.

    1980-01-01

    The Spent Fuel Test--Climax (SFT-C), a test of the retrievable geologic storage of spent fuel assemblies from an operating commercial power reactor, is under way at the Nevada Test Site of the US Department of Energy. Although the main thrust of the project is a demonstration of the feasibility of packaging, handling, storing, and retrieving the highly radioactive fuel assemblies, over 800 data channels have been installed to monitor the response of the rock to the heat and radiation produced by the fuel assemblies and to distinguish in that response the effect due to heat alone. Temperatures in the test array are tracking well with thermal modeling calculations performed before the test was started. The fuel assemblies have been in place since May 1980. The canisters have passed through skin temperature maxima of about 145 0 C and are currently declining in temperature. Evidence is emerging that the thermomechanical response of the rock surrounding the SFT-C is strongly affected by fractures and other discontinuities inthe rock. Most of the effort to date has been in project construction, design, and installation of the instrumentation. Although the data are available in raw form for verification purposes, the data are not as yet in a suitable form for detailed analyses. Work continues on the data management aspects of the project and in continued monitoring of the test

  8. Study of coupled heat and water transfer in proton exchange membrane fuel cells by the way of internal measurements

    International Nuclear Information System (INIS)

    Thomas, A; Maranzana, G; Didierjean, S; Dillet, J; Lottin, O

    2012-01-01

    Measurements of electrode temperatures within a proton exchange membrane fuel cell were performed using platinum wires. A temperature difference of 7°C between the electrodes and the bipolar plates was observed for a cell operating at a current density of 1.5 A.cm −2 . These measurements show a strong non-uniformity of the temperature profile through membrane electrode assembly (MEA) that future phenomenological models must take into account. In addition, the simultaneous measurements of heat and water flux through the MEA leads to the conclusion that produced water crosses the diffusion layer in vapor phase. A very simple heat transfer model is proposed.

  9. Modelling Accident Tolerant Fuel Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Hales, Jason Dean [Idaho National Laboratory; Gamble, Kyle Allan Lawrence [Idaho National Laboratory

    2016-05-01

    The catastrophic events that occurred at the Fukushima-Daiichi nuclear power plant in 2011 have led to widespread interest in research of alternative fuels and claddings that are proposed to be accident tolerant. The United States Department of Energy (DOE) through its Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is a three-year project to perform research on two accident tolerant concepts. The final outcome of the ATF HIP will be an in-depth report to the DOE Advanced Fuels Campaign (AFC) giving a recommendation on whether either of the two concepts should be included in their lead test assembly scheduled for placement into a commercial reactor in 2022. The two ATF concepts under investigation in the HIP are uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (Idaho National Laboratory, Los Alamos National Laboratory, and Argonne National Laboratory), a comprehensive multiscale approach to modeling is being used that includes atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. Model development and fuel performance analysis are critical since a full suite of experimental studies will not be complete before AFC must prioritize concepts for focused development. In this paper, we present simulations of the two proposed accident tolerance fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. Sensitivity analyses are completed using Sandia National Laboratories’ Dakota software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). We also outline the multiscale modelling approach being employed. Considerable additional work is required prior to preparing the recommendation report for the Advanced

  10. Estimation of Membrane Hydration Status for Standby Proton Exchange Membrane Fuel Cell Systems by Impedance Measurement: First Results on Stack Characterization

    DEFF Research Database (Denmark)

    Bidoggia, Benoit; Kær, Søren Knudsen

    Fuel cells have started replacing traditional lead-acid battery banks in backup systems. Although these systems are characterized by long periods of standby, they must be able to start at any instant in the shortest time. In the case of low temperature proton exchange membrane fuel cell systems......, a precise estimation of hydration status of the fuel cell during standby is important for a fast and safe startup. In this article, the measurement of the complex impedance of the fuel cell is suggested as a method to estimate the membrane hydration status. A 56-cell fuel cell stack has been symmetrically...... fed with air whose temperature and relative humidity were controlled, and its complex impedance was measured at different frequencies and for different values of relative humidity. After showing that the experiment was repeatable, the fuel cell stack was characterized, a power regression model...

  11. Measures of the Environmental Footprint of the Front End of the Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    Carlsen, Brett; Tavrides, Emily; Schneider, Erich

    2010-01-01

    Previous estimates of environmental impacts associated with the front end of the nuclear fuel cycle have focused primarily on energy consumption and CO2 emissions. Results have varied widely. Section 2 of this report provides a summary of historical estimates. This study revises existing empirical correlations and their underlying assumptions to fit to a more complete set of existing data. This study also addresses land transformation, water withdrawals, and occupational and public health impacts associated with the processes of the front end of the once-through nuclear fuel cycle. These processes include uranium mining, milling, refining, conversion, enrichment, and fuel fabrication. Metrics are developed to allow environmental impacts to be summed across the full set of front end processes, including transportation and disposition of the resulting depleted uranium.

  12. Measures of the Environmental Footprint of the Front End of the Nuclear Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Brett Carlsen; Emily Tavrides; Erich Schneider

    2010-08-01

    Previous estimates of environmental impacts associated with the front end of the nuclear fuel cycle have focused primarily on energy consumption and CO2 emissions. Results have varied widely. Section 2 of this report provides a summary of historical estimates. This study revises existing empirical correlations and their underlying assumptions to fit to a more complete set of existing data. This study also addresses land transformation, water withdrawals, and occupational and public health impacts associated with the processes of the front end of the once-through nuclear fuel cycle. These processes include uranium mining, milling, refining, conversion, enrichment, and fuel fabrication. Metrics are developed to allow environmental impacts to be summed across the full set of front end processes, including transportation and disposition of the resulting depleted uranium.

  13. Measuring a ballooning gap size of irradiated fuels by the indirect method of neutron radiography

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Cheul Muu; Lee, Seung Wook; Lim, In Cheol; Hong, Kwang Pyo; Kim, Young Jin

    2003-11-01

    An indirect method of a neutron radiography is mobilized for inspecting post irradiated nuclear fuel pins, UO{sub 2}Si clad with Al, which swallowing, crack, ballooning, plug gap, thinning and so on, occurred. The system of an indirect method consists of a cask of carrying fuel pins, Dy converter, linear controller of converter, camera of monitoring fuel cassette. The nuclear sample pins of RISO and KAERI are exposed for 25 min. at the first exposure room, 10{sup 7} cm{sup 2}/sec flux. An activation image formed in the Dy foil is subsequently transferred in a dark room for a more than 8 hours to SR film using the decay radiation. Due to L/D ratio an unsharpness of 9.82{approx}14{mu}m and a magnification of 1.0003 are given. After digitizing an image of SR film, the ballooning gap of plug is discernible by H/V filter of image processing.

  14. Measuring a ballooning gap size of irradiated fuels by the indirect method of neutron radiography

    International Nuclear Information System (INIS)

    Sim, Cheul Muu; Lee, Seung Wook; Lim, In Cheol; Hong, Kwang Pyo; Kim, Young Jin

    2003-11-01

    An indirect method of a neutron radiography is mobilized for inspecting post irradiated nuclear fuel pins, UO 2 Si clad with Al, which swallowing, crack, ballooning, plug gap, thinning and so on, occurred. The system of an indirect method consists of a cask of carrying fuel pins, Dy converter, linear controller of converter, camera of monitoring fuel cassette. The nuclear sample pins of RISO and KAERI are exposed for 25 min. at the first exposure room, 10 7 cm 2 /sec flux. An activation image formed in the Dy foil is subsequently transferred in a dark room for a more than 8 hours to SR film using the decay radiation. Due to L/D ratio an unsharpness of 9.82∼14μm and a magnification of 1.0003 are given. After digitizing an image of SR film, the ballooning gap of plug is discernible by H/V filter of image processing

  15. Coating Thickness Measurement of the Simulated TRISO-Coated Fuel Particles using an Image Plate and a High Resolution Scanner

    International Nuclear Information System (INIS)

    Kim, Woong Ki; Kim, Yeon Ku; Jeong, Kyung Chai; Lee, Young Woo; Kim, Bong Goo; Eom, Sung Ho; Kim, Young Min; Yeo, Sung Hwan; Cho, Moon Sung

    2014-01-01

    In this study, the thic