WorldWideScience

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. 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,...

  4. 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

  5. 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...

  6. 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.

  7. 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.

  8. 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)

  9. 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...

  10. 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

  11. 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

  12. 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.

  13. 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.

  14. 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

  15. 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...

  16. 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

  17. 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

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. 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

  4. 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.)

  5. 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.)

  6. 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.

  7. 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

  8. 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...

  9. 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.

  10. 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...

  11. 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.

  12. 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.

  13. 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)

  14. 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...

  15. 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)

  16. 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)

  17. 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

  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. 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

  20. 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

  1. 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.

  2. 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

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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

  9. 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

  10. 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.

  11. 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)

  12. Validating the BISON fuel performance code to integral LWR experiments

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, R.L., E-mail: Richard.Williamson@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gamble, K.A., E-mail: Kyle.Gamble@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Perez, D.M., E-mail: Danielle.Perez@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Novascone, S.R., E-mail: Stephen.Novascone@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Pastore, G., E-mail: Giovanni.Pastore@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gardner, R.J., E-mail: Russell.Gardner@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Hales, J.D., E-mail: Jason.Hales@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Liu, W., E-mail: Wenfeng.Liu@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States); Mai, A., E-mail: Anh.Mai@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States)

    2016-05-15

    Highlights: • The BISON multidimensional fuel performance code is being validated to integral LWR experiments. • Code and solution verification are necessary prerequisites to validation. • Fuel centerline temperature comparisons through all phases of fuel life are very reasonable. • Accuracy in predicting fission gas release is consistent with state-of-the-art modeling and the involved uncertainties. • Rod diameter comparisons are not satisfactory and further investigation is underway. - Abstract: BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON's computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Results demonstrate that (1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, (2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and (3) comparison

  13. 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

  14. 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

  15. 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.

  16. 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

  17. 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

  18. 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.

  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. In-pile irradiation test program and safety analysis report of the KAERI fuel for HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Wan; Ryu, Woo Suck; Byun, Taek Sang; Park, Jong Man; Lee, Byung Chul; Kim, Hack No; Park, Hee Tae; Kim, Chang Kyu [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-05-01

    Localization of HANARO fuel has been carried out successfully, and design and fabrication technologies of the fuel are recently arrived the final stage of development. The performance of the fuel which has been fabricated in KAERI is confirmed through out-of-pile characterization, and the quality assurance procedure and assessment criteria are described. In order to verify the KAERI fuel, thus, in-pile irradiation test program of the KAERI fuel is scheduled in HANARO. This report summarizes the in-pile testing schedule, design documents of test rods and assemblies, fabrication history and out-of-pile characteristics of test rods, irradiation test condition and power history, post-irradiation examination scheme, linear power generation distribution, and safety analysis results. The design code for HANARO fuel is used to analyze the centerline temperature and swelling of the KAERI fuels. The results show that at 120 kW/m of linear power the maximum centerline temperature is 267 deg C which is much lower than the limitation temperature of 350 deg C, and that the swelling is 9.3 % at 95 at% lower than criterion of 20 %. Therefore, the KAERI fuels of this in-pile irradiation test is assessed to show good performance of integrity and safety in HANARO. 10 tabs., 7 figs., 3 refs. (Author).

  1. Temperature Profile in Fuel and Tie-Tubes for Nuclear Thermal Propulsion Systems

    Energy Technology Data Exchange (ETDEWEB)

    Vishal Patel

    2015-02-01

    A finite element method to calculate temperature profiles in heterogeneous geometries of tie-tube moderated LEU nuclear thermal propulsion systems and HEU designs with tie-tubes is developed and implemented in MATLAB. This new method is compared to previous methods to demonstrate shortcomings in those methods. Typical methods to analyze peak fuel centerline temperature in hexagonal geometries rely on spatial homogenization to derive an analytical expression. These methods are not applicable to cores with tie-tube elements because conduction to tie-tubes cannot be accurately modeled with the homogenized models. The fuel centerline temperature directly impacts safety and performance so it must be predicted carefully. The temperature profile in tie-tubes is also important when high temperatures are expected in the fuel because conduction to the tie-tubes may cause melting in tie-tubes, which may set maximum allowable performance. Estimations of maximum tie-tube temperature can be found from equivalent tube methods, however this method tends to be approximate and overly conservative. A finite element model of heat conduction on a unit cell can model spatial dependence and non-linear conductivity for fuel and tie-tube systems allowing for higher design fidelity of Nuclear Thermal Propulsion.

  2. 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

  3. Development of a program for evaluating the temperature of SMART-P fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Cheon, Jin Sik; Lee, Byung Ho; Koo, Yang Hyun; Oh, Je Yong; Yim, Jeong Sik; Sohn, Dong Seong

    2003-11-01

    A code for evaluating the temperature of SMART-P fuel rod has been developed. Finite Element (FE) method is adopted for the developed code sharing the user subroutines which has been prepared for the ABAQUS commercial FE code. The developed program for SMART-P fuel rod corresponds to a nonlinear transient heat transfer problem, and uses a sparse matrix solver for FE equations during iterations at every time step. The verifications of the developed program were conducted using the ABAQUS code. Steady state and transient problems were analyzed for 1/8 rod model due to the symmetry of the fuel rod and full model. From the evaluation of temperature for the 1/8 rod model at steady state, maximal error of 0.18 % was present relative to the ABAQUS result. Analysis for the transient problem using the fuel rod model resulted in the same as the variation of centerline temperature from the ABAQUS code during a hypothetical power transient. Also, given a power depression in fuel meat as a function of burnup, its effect on the centerline temperature was more precisely evaluated by the developed program compared to the ABAQUS code. The distribution of heat flux for the entire cross section and surface was almost identical for the two codes.

  4. 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

  5. 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.)

  6. Impact of thermal conductivity models on the coupling of heat transport, oxygen diffusion, and deformation in (U, Pu)O nuclear fuel elements

    Science.gov (United States)

    Mihaila, Bogdan; Stan, Marius; Crapps, Justin; Yun, Di

    2013-02-01

    We study the coupled thermal transport, oxygen diffusion, and thermal expansion in a generic nuclear fuel rod consisting of a (U) fuel pellet separated by a helium gap from zircaloy cladding. Steady-state and time-dependent finite-element simulations with a variety of initial- and boundary-value conditions are used to study the effect of the Pu content, y, and deviation from stoichiometry, x, on the temperature and deformation profiles in this fuel element. We find that the equilibrium radial temperature and deformation profiles are most sensitive to x at small values of y. For larger values of y, the effects of oxygen and Pu content are equally important. Following a change in the heat-generation rate, the centerline temperature, the radial deformation of the fuel pellet, and the centerline deviation from stoichiometry track each other closely in (U,Pu)O, as the characteristic time scales of the heat transport and oxygen diffusion are similar. This result is different from the situation observed in the case of UO fuels.

  7. 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.

  8. Fuel Performance Calculations for FeCrAl Cladding in BWRs

    Energy Technology Data Exchange (ETDEWEB)

    George, Nathan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Sweet, Ryan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Maldonado, G. Ivan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Wirth, Brian D. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor (BWR) cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL (formerly Peregrine) expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behavior of the fuel rod (e.g., strains, centerline fuel temperature, and time to gap closure) were investigated and contrasted.

  9. 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.

  10. 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...

  11. 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

  12. 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.

  13. 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

  14. 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.

  15. 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

  16. 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)

  17. Gas-cooled Fast Reactor (GFR) fuel and In-Core Fuel Management

    International Nuclear Information System (INIS)

    Weaver, K.D.; Sterbentz, J.; Meyer, M.; Lowden, R.; Hoffman, E.; Wei, T.Y.C.

    2004-01-01

    The Gas-Cooled Fast Reactor (GCFR) has been chosen as one of six candidates for development as a Generation IV nuclear reactor based on: its ability to fully utilize fuel resources; minimize or reduce its own (and other systems) actinide inventory; produce high efficiency electricity; and the possibility to utilize high temperature process heat. Current design approaches include a high temperature (2 850 C) helium cooled reactor using a direct Brayton cycle, and a moderate temperature (550 C - 650 C) helium or supercritical carbon dioxide (S-CO 2 ) cooled reactor using direct or indirect Brayton cycles. These design choices have thermal efficiencies that approach 45% to 50%, and have turbomachinery sizes that are much more compact compared to steam plants. However, there are challenges associated with the GCFR, which are the focus of current research. This includes safety system design for decay heat removal, development of high temperature/high fluence fuels and materials, and development of fuel cycle strategies. The work presented here focuses on the fuel and preliminary in-core fuel management, where advanced ceramic-ceramic (cercer) dispersion fuels are the main focus, and average burnups to 266 M Wd/kg appear achievable for the reference Si C/(U,TRU)C block/plate fuel. Solid solution (pellet) fuel in composite ceramic clad (Si C/Si C) is also being considered, but remains as a backup due to cladding fabrication challenges, and high centerline temperatures in the fuel. (Author)

  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. 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

  20. 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.

  1. 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

  2. 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.)

  3. Behavior of mixed-oxide fuel elements during an overpower transient

    International Nuclear Information System (INIS)

    Tsai, H.; Shikakura, S.

    1993-01-01

    A slow-ramp (0.1%/s), extended overpower (∼90%) transient test was conducted in EBR-II on 19 mixed-oxide fuel elements with conservative, moderate, and aggressive designs. Claddings for the elements were Type 316, D9, or PNC-316 stainless steel. Before the transient, the elements were preirradiated under steady-state or steady-state plus duty-cycle (periodic 15% overpower transient) conditions to burnups of 2.5-9.7 at%. Cladding integrity during the transient test was maintained by all fuel elements except one, which had experienced substantial overtemperature in the earlier stedy-state irradiation. Extensive centerline fuel melting occurred in all test elements. Significantly, this melting did not cause any elements to breach, although it did have a strong effect on the other aspects of fuel element behavior. (orig.)

  4. 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.

  5. 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)

  6. Thermomechanical analysis of nuclear fuel elements

    International Nuclear Information System (INIS)

    Hernandez L, H.

    1997-01-01

    This work presents development of a code to obtain the thermomechanical analysis of fuel rods in the fuel assemblies inserted in the core of BWR reactors. The code uses experimental correlations developed in several laboratories. The development of the code is divided in two parts: a) the thermal part and b) the mechanical part, extending both the fuel and the cladding materials. The thermal part consists of finding the radial distribution of temperatures in the pellet, from the fuel centerline up to the coolant, along the total active length, considering one and two phase flow in the coolant, as a result of the pressure drop in the system. The mechanical part analyzes the effects of temperature gradients, pressure and irradiation, to which the fuel rod is subjected. The strains produced by swelling, creep and thermal stress in the fuel material are analyzed. In the same way the strains in the cladding are analyzed, considering the effects produced by the pressure exerted on the cladding by pellet swelling, by the pressure caused by fission gas release toward the cavities, and by the strain produced on the cladding by the pressure changes of the system. (Author)

  7. Out-of-reactor experimental study of fuel-pin failure phenomena

    International Nuclear Information System (INIS)

    Wrona, B.J.; Galvin, T.M.; Stahl, D.

    1976-01-01

    Fundamental experiments have been performed with a direct-electrical-heating apparatus, on both unclad and quartz-clad UO 2 pellet stacks, to study the effect of a radial constraint on solid and molten-fuel motion during power transients. Results of simulated transient over-power experiments show that molten UO 2 can be quite mobile when the fuel centerline temperature exceeds the boiling point, i.e., fuel vapor pressures become a significant driving force for relocating molten fuel. For radially constrained pellet stacks, when an escape path was provided around the top pellet, significant upward axial fuel motion occurred prior to cladding rupture. Thus, the time sequence of events shows that potential exists for providing a negative reactivity-feedback effect, which would promote nuclear reactor safety. The data tend to support the existence of a ''pressurized-bottle'' effect, which was observed in high-speed movies

  8. 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

  9. Power ramping test in the JMTR for PCI study of water reactor fuel

    International Nuclear Information System (INIS)

    Nakata, H.; Kanbara, M.; Ichikawa, M.

    1984-01-01

    Power ramping test is essential for PCI study of water reactor fuel. Boiling water capsules have been used for the tests in the JMTR. Heat generation of fuel rod in the capsule can be changed by the He-3 power control facility during reactor operation. Four specially designed fuel rods have been ramped to about 41-43 kW/m; two of them have small gaps filled with iodine, the other two are equipped with centerline temperature thermocouple. Fuel rod elongation detector is equipped to each capsule. For the fuel rods with small gap, unique contraction followed by ordinary fuel relaxation behaviour was observed right after the fast ramping. None of them failed. Future programme includes a series of tests of fuel rods irradiated in the high-pressure water loop at the JMTR and a verification test of remedy fuel which allows daily-load-following operation of BWRs. (author)

  10. 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)

  11. 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)

  12. 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.)

  13. 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.)

  14. 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

  15. 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

  16. 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)

  17. Novel Processing of Unique Ceramic-Based Nuclear Materials and Fuels

    International Nuclear Information System (INIS)

    Zhang, Hui; Singh, Raman P.

    2008-01-01

    Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These include refractory alloys base on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as those based on silicon carbide (SiCf-SiC); carbon-carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor components is necessary for improved efficiency. Improving thermal conductivity of the materials used in nuclear fuels and other temperature critical components can lower the center-line fuel temperature and thereby enhance durability and reduce the risk of premature failure.

  18. Novel Processing of Unique Ceramic-Based Nuclear Materials and Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hui Zhang; Raman P. Singh

    2008-11-30

    Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These include refractory alloys base on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as those based on silicon carbide (SiCf-SiC); carbon-carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor componets is necessary for improved efficiency. Improving thermal conductivity of the materials used in nuclear fuels and other temperature critical components can lower the center-line fuel temperature and thereby enhance durability and reduce the risk of premature failure.

  19. 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.)

  20. 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)

  1. 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

  2. 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/).

  3. 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)

  4. 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)

  5. Evaluation of thermal margin for HANARO core

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Chae, Hee Taek; Kim Heon Il; Lim, I. C.; Lee, C. S.; Kim, H

    1999-08-01

    During the commissioning and the start-up of the HANARO, various design parameters were confirmed and measured. For safer operation of HANARO and resolution of the CHF penalty issue which is one of unresolved licensing problems, thermal margins for normal and transient conditions were re-evaluated reflecting the commissioning and the start-up test results and the design modifications during operation. The re-evaluation shows that the HANARO meets the design criteria for ONB margin and fuel centerline temperature under normal condition. For upset condition, it also satisfies the safety limits for CHFR and fuel centerline temperature. (Author). 11 refs., 13 tabs., 4 figs.

  6. 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

  7. An innovative fuel design concept for improved Light Water Reactor performance and safety

    International Nuclear Information System (INIS)

    Tulenko, J.S.; Connell, R.G.

    1993-01-01

    The primary goal of this research is to develop a new fuel design which will have improved thermal/mechanical performance characteristics greatly superior to current thermal and mechanical design performance. The mechanical/thermal constraints define the lifetime of the fuel, the maximum power at which the fuel can be operated, the probability of fuel failure over core lifetime, and the integrity of a core during a transient excursion. The thermal/mechanical limits act to degrade fuel integrity when they are violated. The purpose of this project is to investigate a novel design for light water reactor fuel which will extend fuel performance limits and improve reactor safety even further than is currently achieved. This project is investigating liquid metal bonding of LWR fuel in order to radically decrease fuel centerline temperatures which has major performance and safety benefits. The project will verify the compatibility of the liquid metal bond with both the fuel pellets and cladding material, verify the performance enhancement features of the new design over the fuel lifetime, and verify the economic fabricability of the concept and will show how this concept will benefit the LWR nuclear industry

  8. Advances in simulating non-congruent phase transitions of hyperstoichiometric uranium dioxide fuel

    International Nuclear Information System (INIS)

    Welland, M.J.; Thompson, W.T.; Lewis, B.J.

    2007-01-01

    A model is being developed to simulate UO 2 at very high temperatures incorporating the effects of non-congruent phase transitions. In particular, the melting transformation and the possible 'Λ-transition' is being investigated to help support the design and analysis of experimental work being conducted as part of nuclear safety research. This work includes the interpretation of the behaviour of operating CANDU fuel under upset conditions, where centerline melting may potentially occur (particularly if the fuel is oxidized). The model presented here numerically solves a system of coupled nonlinear differential equations as derived from fundamental principles. The results of the model present here compare well against laser flash experiments in recently published literature. (author)

  9. 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

  10. 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

  11. 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.

  12. 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.)

  13. 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.

  14. Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Renfro, David [ORNL; Chandler, David [ORNL; Cook, David [ORNL; Ilas, Germina [ORNL; Jain, Prashant [ORNL; Valentine, Jennifer [ORNL

    2014-10-30

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy’s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the “complex” aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The

  15. Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Renfro, David G [ORNL; Chandler, David [ORNL; Cook, David Howard [ORNL; Ilas, Germina [ORNL; Jain, Prashant K [ORNL; Valentine, Jennifer R [ORNL

    2014-11-01

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the complex aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present

  16. Use of filler materials to aid spent nuclear fuel dry storage

    International Nuclear Information System (INIS)

    Anderson, K.J.

    1981-09-01

    The use of filler materials (also known as stabilizer or encapsulating materials) was investigated in conjunction with the dry storage of irradiated light water reactor (LWR) fuel. The results of this investigation appear to be equally valid for the wet storage of fuel. The need for encapsulation and suitable techniques for closing was also investigated. Various materials were reviewed (including solids, liquids, and gases) which were assumed to fill the void areas within a storage can containing either intact or disassembled spent fuel. Materials were reviewed and compared on the basis of cost, thermal characteristics, and overall suitability in the proposed environment. A thermal analysis was conducted to yield maximum centerline and surface temperatures of a design basis fuel encapsulated within various filler materials. In general, air was found to be the most likely choice as a filler material for the dry storage of spent fuel. The choice of any other filler material would probably be based on a desire, or need, to maximize specific selection criteria, such as surface temperatures, criticality safety, or confinement

  17. 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

  18. 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)

  19. 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)

  20. An Innovative High Thermal Conductivity Fuel Design

    Energy Technology Data Exchange (ETDEWEB)

    Jamil A. Khan

    2009-11-21

    Thermal conductivity of the fuel in today's Light Water Reactors, Uranium dioxide, can be improved by incorporating a uniformly distributed heat conducting network of a higher conductivity material, Silicon Carbide. The higher thermal conductivity of SiC along with its other prominent reactor-grade properties makes it a potential material to address some of the related issues when used in UO2 [97% TD]. This ongoing research, in collaboration with the University of Florida, aims to investigate the feasibility and develop a formal methodology of producing the resultant composite oxide fuel. Calculations of effective thermal conductivity of the new fuel as a function of %SiC for certain percentages and as a function of temperature are presented as a preliminary approach. The effective thermal conductivities are obtained at different temperatures from 600K to 1600K. The corresponding polynomial equations for the temperature-dependent thermal conductivities are given based on the simulation results. Heat transfer mechanism in this fuel is explained using a finite volume approach and validated against existing empirical models. FLUENT 6.1.22 was used for thermal conductivity calculations and to estimate reduction in centerline temperatures achievable within such a fuel rod. Later, computer codes COMBINE-PC and VENTURE-PC were deployed to estimate the fuel enrichment required, to maintain the same burnup levels, corresponding to a volume percent addition of SiC.

  1. An Innovative High Thermal Conductivity Fuel Design

    International Nuclear Information System (INIS)

    Khan, Jamil A.

    2009-01-01

    Thermal conductivity of the fuel in today's Light Water Reactors, Uranium dioxide, can be improved by incorporating a uniformly distributed heat conducting network of a higher conductivity material, Silicon Carbide. The higher thermal conductivity of SiC along with its other prominent reactor-grade properties makes it a potential material to address some of the related issues when used in UO2 (97% TD). This ongoing research, in collaboration with the University of Florida, aims to investigate the feasibility and develop a formal methodology of producing the resultant composite oxide fuel. Calculations of effective thermal conductivity of the new fuel as a function of %SiC for certain percentages and as a function of temperature are presented as a preliminary approach. The effective thermal conductivities are obtained at different temperatures from 600K to 1600K. The corresponding polynomial equations for the temperature-dependent thermal conductivities are given based on the simulation results. Heat transfer mechanism in this fuel is explained using a finite volume approach and validated against existing empirical models. FLUENT 6.1.22 was used for thermal conductivity calculations and to estimate reduction in centerline temperatures achievable within such a fuel rod. Later, computer codes COMBINE-PC and VENTURE-PC were deployed to estimate the fuel enrichment required, to maintain the same burnup levels, corresponding to a volume percent addition of SiC.

  2. 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

  3. 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.

  4. 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

  5. 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)

  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. 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. The advanced neutron source three-element-core fuel grading

    International Nuclear Information System (INIS)

    Gehin, J.C.

    1995-01-01

    The proposed Advanced Neutron Source (ANS) pre-conceptual design consists of a two-element 330 MW f nuclear reactor fueled with highly-enriched uranium and is cooled, moderated, and reflected with heavy water. Recently, the ANS design has been changed to a three-element configuration in order to permit a reduction of the enrichment, if required, while maintaining or improving the thermal-hydraulic margins. The core consists of three annular fuel elements composed of involute-shaped fuel plates. Each fuel plate has a thickness of 1.27 mm and consists of a fuel meat region Of U 3 Si 2 -Al (50% enriched in one case that was proposed) and an aluminum filler region between aluminum cladding. The individual plates are separated by a 1.27 mm coolant channel. The three element core has a fuel loading of 31 kg of 235 U which is sufficient for a 17-day fuel cycle. The goal in obtaining a new fuel grading is to maximize important temperature margins. The limits imposed axe: (1) Limit the temperature drop over the cladding oxide layer to less than 119 degrees C to avoid oxide spallation. (2) Limit the fuel centerline temperature to less than 400 degrees C to avoid fuel damage. (3) Limit the cladding wall temperature to less than the coolant. incipient-boiling temperature to avoid coolant boiling. Other thermal hydraulic conditions, such as critical heat flux, are also considered

  9. 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%

  10. 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)

  11. A study of UO2 wafer fuel for very high-power research reactors

    International Nuclear Information System (INIS)

    Hsieh, T.C.; Jankus, V.Z.; Rest, J.; Billone, M.C.

    1983-01-01

    The Reduced Enrichment Research and Test Reactor Program is aimed at reducing fuel enrichment to 2 caramel fuel is one of the most promising new types of reduced-enrichment fuel for use in research reactors with very high power density. Parametric studies have been carried out to determine the maximum specific power attainable without significant fission-gas release for UO 2 wafers ranging from 0.75 to 1.50 mm in thickness. The results indicate that (1) all the fuel designs considered in this study are predicted not to fail under full power operation up to a burnup, of 1.9x10 21 fis/cm 3 ; (2) for all fuel designs, failure is predicted at approximately the same fuel centerline temperature for a given burnup; (3) the thinner the wafer, the wider the margin for fuel specific power between normal operation and increased-power operation leading to fuel failure; (4) increasing the coolant pressure in the reactor core could improve fuel performance by maintaining the fuel at a higher power level without failure for a given burnup; and (5) for a given power level, fuel failure will occur earlier at a higher cladding surface temperature and/or under power-cycling conditions. (author)

  12. 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.

  13. Calculation of radiation dose rates from a spent nuclear fuel shipping cask

    International Nuclear Information System (INIS)

    Chen, S.Y.; Yuan, Y.C.

    1988-01-01

    Radiation doses from a spent nuclear fuel cask are usually from various phases of operations during handling, shipping, and storage of the casks. Assessment of such doses requires knowledge of external radiation dose rates at various locations surrounding a cask. Under current practices, dose rates from gamma photons are usually estimated by means of point- or line-source approaches incorporating the conventional buildup factors. Although such simplified approaches may at times be easy to use, their accuracy has not been verified. For example, those simplified methods have not taken into account influencing factors such as the geometry of the cask and the presence of the ground surface, and the effects of these factors on the calculated dose rates are largely unknown. Moreover, similar empirical equations for buildup factors currently do not exist for neutrons. The objective of this study is to use a more accurate approach in calculating radiation dose rates for both neutrons and gamma photons from a spent fuel cask. The calculation utilizes the more sophisticated transport method and takes into account the geometry of the cask and the presence of the ground surface. The results of a detailed study of dose rates in the near field (within 20 meters) are presented and, for easy application, the cask centerline dose rates are fitted into empirical equations at cask centerline distances up to 2000 meters from the surface of the cask

  14. 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

  15. 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

  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. 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.

  18. 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

  19. 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)

  20. 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

  1. 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

  2. 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.

  3. 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''

  4. The Optimum Plutonium Inert Matrix Fuel Form for Reactor-Based Plutonium Disposition

    International Nuclear Information System (INIS)

    Tulenko, J.S.; Wang, J.; Acosta, C.

    2004-01-01

    The University of Florida has underway an ongoing research program to validate the economic, operational and performance benefits of developing an inert matrix fuel (IMF) for the disposition of the U.S. weapons plutonium (Pu) and for the recycle of reprocessed Pu. The current fuel form of choice for Pu disposition for the Department of Energy is as a mixed oxide (MOX) (PuO2/UO2). We will show analyses that demonstrate that a Silicon Carbide (SiC) IMF offers improved performance capabilities as a fuel form for Pu recycle and disposition. The reason that UF is reviewing various materials to serve as an inert matrix fuel is that an IMF fuel form can offer greatly reduced Pu and transuranic isotope (TRU) production and also improved thermal performance characteristics. Our studies showed that the Pu content is reduced by an order of magnitude while centerline fuel temperatures are reduced approximately 380 degrees centigrade compared to MOX. These reduced temperatures result in reduced stored heat and thermal stresses in the pellet. The reduced stored heat reduces the consequences of the loss of coolant accident, while the reduced temperatures and thermal stresses yield greatly improved fuel performance. Silicon Carbide is not new to the nuclear industry, being a basic fuel material in gas cooled reactors

  5. 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.

  6. 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.

  7. 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

  8. 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.

  9. 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

  10. Application of adjoint sensitivity analysis to nuclear reactor fuel rod performance

    International Nuclear Information System (INIS)

    Wilderman, S.J.; Was, G.S.

    1984-01-01

    Adjoint sensitivity analysis in nuclear fuel behavior modeling is extended to operate on the entire power history for both Zircaloy and stainless steel cladding via the computer codes FCODE-ALPHA/SS and SCODE/SS. The sensitivities of key variables to input parameters are found to be highly non-intuitive and strongly dependent on the fuel-clad gap status and the history of the fuel during the cycle. The sensitivities of five key variables, clad circumferential stress and strain, fission gas release, fuel centerline temperature and fuel-clad gap, to eleven input parameters are studied. The most important input parameters (yielding significances between 1 and 100) are fabricated clad inner and outer radii and fuel radius. The least important significances (less than 0.01) are the time since reactor start-up and fuel-burnup densification rate. Intermediate to these are fabricated fuel porosity, linear heat generation rate, the power history scale factor, clad outer temperature, fill gas pressure and coolant pressure. Stainless steel and Zircaloy have similar sensitivities at start-up but these diverges a burnup proceeds due to the effect of the higher creep rate of Zircaloy which causes the system to be more responsive to changes in input parameters. The value of adjoint sensitivity analysis lies in its capability of uncovering dependencies of fuel variables on input parameters that cannot be determined by a sequential thought process. (orig.)

  11. 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.)

  12. 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

  13. 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.

  14. Reaction of yttria-stabilized zirconia with zirconium, silicon and Zircaloy-4 at high temperature: a compatibility study for cermet fuels

    International Nuclear Information System (INIS)

    Arima, T.; Tateyama, T.; Idemitsu, K.; Inagaki, Y.

    2003-01-01

    Compatibility studies for cermet (ceramic and metal) fuels have been completed for a temperature range of 1073-1423 K. A reaction between yttria-stabilized zirconia (YSZ), as a simulated fuel, and Zr, as a candidate for a metallic matrix, has been observed at temperatures ≥1273 K, which means the formation of a metallic reaction layer at the interface between YSZ and Zr and the occurrence of metallic phases inside the YSZ. Similar results were observed for the YSZ-Zry4 (cladding) system. On the other hand, the degree of reaction was relatively large for the YSZ-Si (metallic matrix) system, and Si diffused into the YSZ. However, the maximum fuel center-line temperature can be predicted to be less than ∼1273 K for cermet fuels. Therefore, compatibility between the ceramic fuel and the metallic matrix should be good under normal reactor operational conditions. Furthermore, since the temperature of the fuel-cladding gap is lower, the cermet fuel and the cladding material are compatible

  15. 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)

  16. 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...

  17. 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

  18. 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

  19. 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.

  20. 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

  1. 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.

  2. Coordinate measurement machines as an alignment tool

    International Nuclear Information System (INIS)

    Wand, B.T.

    1991-03-01

    In February of 1990 the Stanford Linear Accelerator Center (SLAC) purchased a LEITZ PM 12-10-6 CMM (Coordinate measurement machine). The machine is shared by the Quality Control Team and the Alignment Team. One of the alignment tasks in positioning beamline components in a particle accelerator is to define the component's magnetic centerline relative to external fiducials. This procedure, called fiducialization, is critical to the overall positioning tolerance of a magnet. It involves the definition of the magnetic center line with respect to the mechanical centerline and the transfer of the mechanical centerline to the external fiducials. To perform the latter a magnet coordinate system has to be established. This means defining an origin and the three rotation angles of the magnet. The datum definition can be done by either optical tooling techniques or with a CMM. As optical tooling measurements are very time consuming, not automated and are prone to errors, it is desirable to use the CMM fiducialization method instead. The establishment of a magnet coordinate system based on the mechanical center and the transfer to external fiducials will be discussed and presented with 2 examples from the Stanford Linear Collider (SLC). 7 figs

  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. 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)

  5. 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

  6. 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

  7. 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

  8. 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.

  9. Engineering product storage under the advanced fuel cycle initiative. Part I: An iterative thermal transport modeling scheme for high-heat-generating radioactive storage forms

    International Nuclear Information System (INIS)

    Kaminski, Michael D.

    2005-01-01

    The US Department of Energy is developing an integrated nuclear fuel cycle technology under its Advanced Fuel Cycle Initiative (AFCI). Under the AFCI, waste minimization is stressed. Engineered product storage materials will be required to store concentrated radioactive cesium, strontium, americium, and curium for periods of tens to hundreds of years. The fabrication of such engineered products has some precedence but the concept is largely novel. We thus present a theoretical model used to calculate the maximum radial dimensions of right cylinder storage forms under several scenarios. Maximum dimensions are small, comparable to nuclear fuel pins in some cases, to avoid centerline melting temperatures; this highlights the need for a careful strategy for engineered product storage fabrication and storage

  10. 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)

  11. 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.

  12. 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

  13. 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)

  14. 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...

  15. 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.)

  16. Development of thermocouple re-instrumentation technique for irradiated fuel rod. Techniques for making center hole into UO2 pellets and thermocouple re-instrumentation to fuel rod

    International Nuclear Information System (INIS)

    Shimizu, Michio; Saito, Junichi; Oshima, Kunio

    1995-07-01

    The information on FP gas pressure and centerline temperature of fuel pellets during power transient is important to study the pellet clad interaction (PCI) mechanism of high burnup LWR fuel rods. At the Department of JMTR, a re-instrumentation technique of FP gas pressure gage for an irradiated fuel rod was developed in 1990. Furthermore, a thermocouple re-instrumentation technique was successfully developed in 1994. Two steps were taken to carry out the development program of the thermocouple re-instrumentation technique. In the first step, a drilling technique was developed for making a center hole of the irradiated fuel pellets. Various drilling tests were carried out using dummy of fuel rods consisted of Ba 2 FeO 3 pellets and Zry-2 cladding. On this work it is important to keep the pellets just the state cracked at a power reactor. In these tests, the technique to fix the pellets by frozen CO 2 was used during the drilling work. Also, diamond drills were used to make the center hole. These tests were completed successfully. A center hole, 54mm depth and 2.5mm diameter, was realized by these methods. The second step of this program is the in-pile demonstration test on an irradiated fuel rod instrumented dually a thermocouple and FP gas pressure gage. The demonstration test was carried out at the JMTR in 1995. (author)

  17. 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)

  18. 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

  19. 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)

  20. 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.'

  1. 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

  2. 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

  3. 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.

  4. 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

  5. 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...

  6. 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

  7. Steady State and Transient Fuel Rod Performance Analyses by Pad and Transuranus Codes

    International Nuclear Information System (INIS)

    Slyeptsov, O.; Slyeptsov, S.; Kulish, G.; Ostapov, A.; Chernov, I.

    2013-01-01

    The report performed under IAEA research contract No.15370/L2 describes the analysis results of WWER and PWR fuel rod performance at steady state operation and transients by means of PAD and TRANSURANUS codes. The code TRANSURANUS v1m1j09 developed by Institute for of Transuranium Elements (ITU) was used based on the Licensing Agreement N31302. The code PAD 4.0 developed by Westinghouse Electric Company was utilized in the frame of the Ukraine Nuclear Fuel Qualification Project for safety substantiation for the use of Westinghouse fuel assemblies in the mixed core of WWER-1000 reactor. The experimental data for the Russian fuel rod behavior obtained during the steady-state operation in the WWER-440 core of reactor Kola-3 and during the power transients in the core of MIR research reactor were taken from the IFPE database of the OECD/NEA and utilized for assessing the codes themselves during simulation of such properties as fuel burnup, fuel centerline temperature (FCT), fuel swelling, cladding strain, fission gas release (FGR) and rod internal pressure (RIP) in the rod burnup range of (41 - 60) GWD/MTU. The experimental data of fuel behavior at steady-state operation during seven reactor cycles presented by AREVA for the standard PWR fuel rod design were used to examine the code FGR model in the fuel burnup range of (37 - 81) GWD/MTU. (author)

  8. 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)

  9. 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

  10. 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)

  11. 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.

  12. 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)

  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. 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.

  15. 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

  16. 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...

  17. 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

  18. 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.

  19. 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

  20. An Air Bearing Rotating Coil Magnetic Measurement System

    CERN Document Server

    Gottschalk, Stephen C; Taylor, David J; Thayer, William

    2005-01-01

    This paper describes a rotating coil magnetic measurement system supported on air bearings. The design is optimized for measurements of 0.1micron magnetic centerline changes on long, small aperture quadrupoles. Graphite impregnated epoxy resin is used for the coil holder and coil winding forms. Coil holder diameter is 11 mm with a length between supports of 750mm. A pair of coils is used to permit quadrupole bucking during centerline measurements. Coil length is 616mm, inner radius 1.82mm, outer radius 4.74mm. The key features of the mechanical system are simplicity; air bearings for accurate, repeatable measurements without needing warm up time and a vibration isolated stand that uses a steel-topped Newport optical table with air suspension. Coil rotation is achieved by a low noise servo motor controlled by a standalone Ethernet servo board running custom servo software. Coil calibration procedures that correct wire placement errors, tests for mechanical resonances, and other system checks will also be discu...

  1. Core design options for high conversion BWRs operating in Th–233U fuel cycle

    International Nuclear Information System (INIS)

    Shaposhnik, Y.; Shwageraus, E.; Elias, E.

    2013-01-01

    Highlights: • BWR core operating in a closed self-sustainable Th– 233 U fuel cycle. • Seed blanket optimization that includes assembly size array and axial dimensions. • Fully coupled MC with fuel depletion and thermo-hydraulic feedback modules. • Thermal-hydraulic analysis includes MCPR observation. -- Abstract: Several options of fuel assembly design are investigated for a BWR core operating in a closed self-sustainable Th– 233 U fuel cycle. The designs rely on an axially heterogeneous fuel assembly structure consisting of a single axial fissile zone “sandwiched” between two fertile blanket zones, in order to improve fertile to fissile conversion ratio. The main objective of the study was to identify the most promising assembly design parameters, dimensions of fissile and fertile zones, for achieving net breeding of 233 U. The design challenge, in this respect, is that the fuel breeding potential is at odds with axial power peaking and the core minimum critical power ratio (CPR), hence limiting the maximum achievable core power rating. Calculations were performed with the BGCore system, which consists of the MCNP code coupled with fuel depletion and thermo-hydraulic feedback modules. A single 3-dimensional fuel assembly having reflective radial boundaries was modeled applying simplified restrictions on the maximum centerline fuel temperature and the CPR. It was found that axially heterogeneous fuel assembly design with a single fissile zone can potentially achieve net breeding, while matching conventional BWR core power rating under certain restrictions to the core loading pattern design

  2. 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)

  3. TEM characterization of irradiated U-7Mo/Mg dispersion fuel

    Science.gov (United States)

    Gan, J.; Keiser, D. D.; Miller, B. D.; Jue, J. F.; Robinson, A. B.; Madden, J.

    2017-10-01

    This paper presents the results of transmission electron microscopy (TEM) characterization on neutron-irradiated samples taken from the low-flux and high-flux sides of the same fuel plate with U-7Mo fuel particles dispersed in Mg matrix with aluminum alloy Al6061 as cladding material that was irradiated edge-on to the core in the Advanced Test Reactor. The corresponding local fission density and fission rate of the fuel particles and the average fuel-plate centerline temperature for the low-flux and high-flux samples are estimated to be 3.7 × 1021 f/cm3, 7.4 × 1014 f/cm3/s and 123 °C, and 5.5 × 1021 f/cm3, 11.0 × 1014 f/cm3/s and 158 °C, respectively. Complex interaction layers developed at the Al-Mg interface, consisting of Al3Mg2 and Al12Mg17 along with precipitates of MgO, Mg2Si and FeAl5.3. No interaction between Mg matrix and U-Mo fuel particle was identified. For the U-Mo fuel particles, at low fission density, small elongated bubbles wrapped around the clean areas with a fission gas bubble superlattice, which suggests that bubble coalescence is an important mechanism for converting the fission gas bubble superlattice to large bubbles. At high fission density, no bubbles or porosity were observed in the Mg matrix, and pockets of residual fission gas bubble superlattice were observed in the U-Mo fuel particle interior.

  4. Final safety analysis report for the irradiated fuels storage facility

    International Nuclear Information System (INIS)

    Bingham, G.E.; Evans, T.K.

    1976-01-01

    A fuel storage facility has been constructed at the Idaho Chemical Processing Plant to provide safe storage for spent fuel from two commercial HTGR's, Fort St. Vrain and Peach Bottom, and from the Rover nuclear rocket program. The new facility was built as an addition to the existing fuel storage basin building to make maximum use of existing facilities and equipment. The completed facility provides dry storage for one core of Peach Bottom fuel (804 elements), 1 1 / 2 cores of Fort St. Vrain fuel (2200 elements), and the irradiated fuel from the 20 reactors in the Rover program. The facility is designed to permit future expansion at a minimum cost should additional storage space for graphite-type fuels be required. A thorough study of the potential hazards associated with the Irradiated Fuels Storage Facility has been completed, indicating that the facility is capable of withstanding all credible combinations of internal accidents and pertinent natural forces, including design basis natural phenomena of a 10,000 year flood, a 175-mph tornado, or an earthquake having a bedrock acceleration of 0.33 g and an amplification factor of 1.3, without a loss of integrity or a significant release of radioactive materials. The design basis accident (DBA) postulated for the facility is a complete loss of cooling air, even though the occurrence of this situation is extremely remote, considering the availability of backup and spare fans and emergency power. The occurrence of the DBA presents neither a radiation nor an activity release hazard. A loss of coolant has no effect upon the fuel or the facility other than resulting in a gradual and constant temperature increase of the stored fuel. The temperature increase is gradual enough that ample time (28 hours minimum) is available for corrective action before an arbitrarily imposed maximum fuel centerline temperature of 1100 0 F is reached

  5. 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.)

  6. Development of program for evaluating the temperature of Zr-U metallic fuel rod

    International Nuclear Information System (INIS)

    Chun, J. S.; Lee, B. H.; Ku, Y. H.; Oh, J. Y.; Im, J. S.; Sohn, D. S.

    2003-01-01

    A code for evaluating the temperature of Zr-U metallic rod has been developed. Finite element (FE) method is adopted for the developed code sharing the user subroutines which has been prepared for the ABAQUS commercial FE code. The developed program for the Zr-U metallic fuel rod corresponds to a nonlinear transient heat transfer problem, and uses a sparse matrix solver for FE equations during iterations at every time step. The verifications of the developed program were conducted using the ABAQUS code. Steady state and transient problems were analyzed for 1/8 rod model due to the symmetry of the fuel rod and full model. From the evaluation of temperature for the 1/8 rod model at steady state, maximal error of 0.18 % was present relative to the ABAQUS result. Analysis for the transient problem using the fuel rod model resulted in the same as the variation of centerline temperature from the ABAQUS code during a hypothetical power transient. The distribution of heat flux for the entire cross section and surface was almost identical for the two codes

  7. 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.

  8. 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.

  9. 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.

  10. 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)

  11. 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.

  12. Fabrication of Cr-doped UO2 Fuel Pellet using Liquid Phase Sintering

    International Nuclear Information System (INIS)

    Kim, Dong Joo; Yang, Jae Ho; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Oh, Jang Soo; Koo, Yang Hyun

    2013-01-01

    An enhancement of the thermal conductivity of a pellet can be obtained by the addition of a higher thermal conductive material in the pellet. In addition, the resistance to the PCI can be increased through a plasticity increase of the pellet. Thermal conductivity of ceramic materials is generally lower than that of metallic materials. The thermal conductivity of uranium oxide which is a typical ceramic material is low as well. The steep temperature gradient in the fuel pellet results from the low thermal conductivity. Therefore, the thermal conductivity improvement of a nuclear fuel pellet can enhance the fuel performance in various aspects. The lower centerline temperature of a fuel pellet affects the enhancement of fuel safety as well as fuel pellet integrity during nuclear reactor operation. Besides, the nuclear reactor power can be uprated due to the higher safety margin. So, many researches to enhance the thermal conductivity of nuclear fuel pellet have been performed in various ways. To improve the thermal conductivity of UO 2 pellet, an appropriate arrangement of the high thermal conductive material in UO 2 matrix is one of the various methods. We intended to control a placement of chromium as the high thermal conductive material. The metallic chromium and chromium oxide were arranged in a grain boundary of UO 2 using a liquid phase sintering method. The liquid phase sintering of Cr-doped UO 2 pellet could be adjusted using a control of an oxygen potential in sintering atmosphere

  13. 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%.

  14. 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)

  15. 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

  16. 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

  17. 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...

  18. 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

  19. Assessment of uranium dioxide fuel performance with the addition of beryllium oxide

    Energy Technology Data Exchange (ETDEWEB)

    Muniz, Rafael O.R.; Abe, Alfredo; Gomes, Daniel S.; Silva, Antonio T., E-mail: romuniz@usp.br, E-mail: ayabe@ipen.br, E-mail: danieldesouza@gmail.com, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energética s e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (LabRisco/USP), Sao Paulo, SP (Brazil). Lab. de Análise, Avaliação e Gerenciamento de Risco; Aguiar, Amanda A., E-mail: amanda.abati.aguiar@gmail.com [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil)

    2017-07-01

    The Fukushima Daiichi accident in 2011 pointed the problem related to the hydrogen generation under accident scenarios due to the oxidation of zirconium-based alloys widely used as fuel rod cladding in water-cooled reactors. This problem promoted research programs aiming the development of accident tolerant fuels (ATF) which are fuels that under accident conditions could keep longer its integrity enabling the mitigation of the accident effects. In the framework of the ATF program, different materials have been studied to be applied as cladding to replace zirconium-based alloy; also efforts have been made to improve the uranium dioxide thermal conductivity doping the fuel pellet. This paper evaluates the addition of beryllium oxide (BeO) to the uranium dioxide in order to enhance the thermal conductivity of the fuel pellet. Investigations performed in this area considering the addition of 10% in volume of BeO, resulting in the UO{sub 2}-BeO fuel, have shown good results with the improvement of the fuel thermal conductivity and the consequent reduction of the fuel temperatures under irradiation. In this paper, two models obtained from open literature for the thermal conductivity of UO{sub 2}- BeO fuel were implemented in the FRAPCON 3.5 code and the results obtained using the modified code versions were compared. The simulations were carried out using a case available in the code documentation related to a typical pressurized water reactor (PWR) fuel rod irradiated under steady state condition. The results show that the fuel centerline temperatures decrease with the addition of BeO, when compared to the conventional UO{sub 2} pellet, independent of the model applied. (author)

  20. 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.

  1. 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

  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. Preliminary analysis of in-reactor behavior of three MOX fuel rods in the halden reactor

    International Nuclear Information System (INIS)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong; Joo, Hyung Kook

    1999-09-01

    Preliminary analysis of in-reactor thermal performance for three MOX fuel rods that are going to be irradiated in the Halden reactor from the first quarter of the year 2000 have been conducted by using the computer code COSMOS. Using the assumption that microstructure of MOX fuel fabricated by SBR and dry milling method is the same, parametric studies have been carried out considering four kinds of uncertainties, which are thermal conductivity, linear power, manufacturing parameters, and model constant, to investigate the effect of each of uncertainty on in-reactor behavior. It is found that the uncertainty of model constants for FGR has a greatest impact of the all because the amount of gas released to the gap is one of the parameters that dominantly affects the gap conductance. The parametric analysis shows that, tn the case of MOX-1, calculational results vary widely depending on the choice of model constants for FGR. Therefore, the model constants for FGR for the present test need to be established through the measured fuel centerline temperature, rod internal pressure, stack length if any, and finally thermal conductivity derived from measured data during irradiation. On the other hand, the difference in thermal performance of MOX-3 resulting from the choice of FGR model constants is not so large as that for MOX-1. This might arise, since the temperature of the MOX-3 is high, the capacity of grain boundaries to retain gas atoms is not sufficient enough to accommodate the large amount of gas atoms reaching the grain boundaries through diffusion. (Author). 20 refs., 7 tabs., 47 figs

  4. Preliminary analysis of in-reactor behavior of three MOX fuel rods in the halden reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong; Joo, Hyung Kook

    1999-09-01

    Preliminary analysis of in-reactor thermal performance for three MOX fuel rods that are going to be irradiated in the Halden reactor from the first quarter of the year 2000 have been conducted by using the computer code COSMOS. Using the assumption that microstructure of MOX fuel fabricated by SBR and dry milling method is the same, parametric studies have been carried out considering four kinds of uncertainties, which are thermal conductivity, linear power, manufacturing parameters, and model constant, to investigate the effect of each of uncertainty on in-reactor behavior. It is found that the uncertainty of model constants for FGR has a greatest impact of the all because the amount of gas released to the gap is one of the parameters that dominantlyaffects the gap conductance. The parametric analysis shows that, tn the case of MOX-1, calculational results vary widely depending on the choice of model constants for FGR. Therefore, the model constants for FGR for the present test need to be established through the measured fuel centerline temperature, rod internal pressure, stack length if any, and finally thermal conductivity derived from measured data during irradiation. On the other hand, the difference in thermal performance of MOX-3 resulting from the choice of FGR model constants is not so large as that for MOX-1. This might arise, since the temperature of the MOX-3 is high, the capacity of grain boundaries to retain gas atoms is not sufficient enough to accommodate the large amount of gas atoms reaching the grain boundaries through diffusion. (Author). 20 refs., 7 tabs., 47 figs.

  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. 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

  7. 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...

  8. 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.)

  9. 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)

  10. Analysis of WWER-440 fuel performance under normal operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gunduz, Oe; Koese, S; Akbas, T [Atomenerjisi Komisyonu, Ankara (Turkey); Colak, Ue [Ankara Nuclear Research and Training Center (Turkey)

    1994-12-31

    FRAPCON-2 code originally developed for LWR fuel behaviour simulation is used to analyse the WWER-440 fuel rod behaviour at normal operational conditions. The code is capable of utilizing different models for mechanical analysis and gas release calculations. Heat transfer calculations are accomplished through a collocation technique by the method of weighted residuals. Temperature and burnup element properties are evaluated using MATPRO package. As the material properties of Zr-1%Nb used as cladding in WWER-440s are not provided in the code, Zircaloy-4 is used as a substitute for Zr-1%Nb. Mac-Donald-Weisman model is used for gas release calculation. FRACAS-1 and FRACAS-2 models are used in the mechanical calculations. It is assumed that the reactor was operated for 920 days (three consecutive cycles), the burnup being 42000 Mwd/t U. Results of the fuel rod behaviour analysis are given for three axial nodes: bottom node, central node and top node. The variations of the following characteristic fuel rod parameters are studied through the prescribed power history: unmoved gap thickness, gap heat transfer coefficient, fuel axial elongation, cladding axial elongation, fuel centerline temperature and ZrO-thickness at cladding surface. The value of each parameter is calculated as a function of the effective power days for the three nodes by using FRACAS-1 and FRACAS-2 codes for comparison.The results show that calculations with deformable pellet approximation with FRACAS-II model could provide better information for the behaviour of a typical fuel rod. Calculations indicate that fuel rod failure is not observed during the operation. All fuel rod parameters investigated are found to be within the safety limits. It is concluded, however, that for better assessment of reactor safety these calculations should be extended for transient conditions such as LOCA. 1 tab., 10 figs., 4 refs.

  11. 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

  12. 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

  13. 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

  14. 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.

  15. 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

  16. 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

  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. 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

  19. 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.

  20. USNRC-OECD Halden Project fuel behavior test program: experiment data report for test assemblies IFA-226 and IFA 239

    International Nuclear Information System (INIS)

    Laats, E.T.; MacDonald, P.E.; Quapp, W.J.

    1975-12-01

    The experimental data which were obtained from the IFA-226 and IFA-239 test assemblies during operation in the Halden Boiling Water Reactor are reported. Included are cladding elongation, fuel centerline temperature, internal gas pressure, and power history data from IFA-226 which were obtained from November 1971 through April 1974, and cladding elongation, diametral profile, and power history data from IFA-239 covering the period from March 1973 through April 1974. The data, presented in the form of composite graphs, have been analyzed only to the extent necessary to assure that they are reasonable and correct. A description of these mixed oxide fuel test assemblies and their instrumentation is presented. Test pin fabrication history, instrument calibration data, assembly power calibration methods, and the neutron detector data reduction technique are included as appendices

  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. Thermal aspects of mixed oxide fuel in application to supercritical water-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grande, L.; Peiman, W.; Rodriguez-Prado, A.; Villamere, B.; Mikhael, S.; Allison, L.; Pioro, I., E-mail: lisa.grande@mycampus.uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

    2010-07-01

    SuperCritical Water-cooled nuclear Reactors (SCWRs) are a renewed technology being developed as one of the Generation IV reactor concepts. This reactor type uses a light water coolant at temperatures and pressures above its critical point. These elevated operating conditions will improve Nuclear Power Plant (NPP) thermal efficiencies by 10 - 15% compared to those of current NPPs. Also, SCWRs will have the ability to utilize a direct cycle, thus decreasing NPP capital and operational costs. The SCWR core has 2 configurations: 1) Pressure Vessel (PV) -type enclosing a fuel assembly and 2) Pressure Tube (PT) -type consisting of individual pressurized channels containing fuel bundles. Canada and Russia are developing PT-type SCWRs. In particular, the Canadian SCWR reactor has an output of 1200 MW{sub el} and will operate at a pressure of 25 MPa with inlet and outlet fuel-channel temperatures of 350 and 625°C, respectively. These extreme operating conditions require alternative fuels and materials to be investigated. Current CANadian Deuterium Uranium (CANDU) nuclear reactor fuel-channel design is based on the use of uranium dioxide (UO{sub 2}) fuel; zirconium alloy sheath (clad) bundle, pressure and calandria tubes. Alternative fuels should be considered to supplement depleting world uranium reserves. This paper studies general thermal aspects of using Mixed OXide (MOX) fuel in an Inconel-600 sheath in a generic PT-type SCWR. The bulk fluid, sheath and fuel centerline temperatures along with the Heat Transfer Coefficient (HTC) profiles were calculated at uniform and non-uniform Axial Heat Flux Profiles (AHFPs). (author)

  3. 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

  4. 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.)

  5. Thermally-induced bowing of CANDU fuel elements

    International Nuclear Information System (INIS)

    Suk, H.C.; Sim, K.S.; Park, J.H.; Park, G.S.

    1995-01-01

    Considering only the thermally-induced bending moments which are generated both within the sheath and between the fuel and sheath by an asymmetric temperature distribution with respect to the axis of an element, a generalized and explicit analytical formula for the thermally-induced bending is developed in this paper, based on the cases of 1) the bending of an empty tube treated by neglecting of the fuel/sheath mechanical interaction and 2) the fuel/sheath interaction due to the pellet and sheath temperature variations. In each of the cases, the temperature asymmetries in sheath are modelled to be caused by the combined effects of (i) non-uniform coolant temperature due to imperfect coolant mixing, (ii) variable sheath/coolant heat transfer coefficient, (iii) asymmetric heat generation due to neutron flux gradients across an element and so as to inclusively cover the uniform temperature distributions within the fuel and sheath with respect to the axial centerline. Investigating the relative importance of the various parameters affecting fuel element bowing, the element bowing is found to be greatly affected with the variations of element length, sheath diameter, pellet/sheath mechanical interaction and neutron flux depression factors, pellet thermal expansion coefficient, pellet/sheath heat transfer coefficient in comparison with those of other parameters such as sheath thickness, film heat transfer coefficient, sheath thermal expansion coefficient, and sheath and pellet thermal conductivities. Also, the element bowing of the standard 37-element bundle and CANFLEX 43-element bundle for the use in CANDU-6 reactors was analyzed with the formula, which could help to demonstrate the integrity of the fuel. All the required input data for the analyses were generated in terms of the reactor operation conditions on the reactor physics, thermal hydraulics and fuel performance by using various CANDU computer codes. The analysis results indicate that the CANFLEX 43-element

  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. 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

  8. 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.

  9. 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

  10. 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

  11. Optimization of Additive-Powder Characteristics for Metallic Micro-Cell UO{sub 2} Fuel Pellet Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong-Joo; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Oh, Jang Soo; Yang, Jae Ho; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The improvement in the thermal conductivity of the UO{sub 2} fuel pellet can enhance the fuel performance in various aspects. The mobility of the fission gases is reduced by the lower temperature gradient in the UO{sub 2} fuel pellet. That is to say, the capability of the fission gas retention of the fuel pellet can increase. In addition, the lower centerline temperature of the fuel pellet affects the accident tolerance for nuclear fuel as well as the enhancement of fuel safety and fuel pellet integrity under normal operation conditions. The nuclear reactor power can be uprated owing to the higher safety margin. Thus, many researches on enhancing the thermal conductivity of a nuclear fuel pellet for LWRs have been performed. Typically, an enhancement of the thermal conductivity of the UO{sub 2} fuel pellet can be obtained by the addition of a higher thermal conductive material in the fuel pellet. To maximize the effect of the thermal conductivity enhancement, a continuous and uniform channel of the thermal conductive material in the UO{sub 2} matrix must be formed. To enhance the thermal conductivity of a UO{sub 2} fuel pellet, the development of fabrication process of a Cr metallic micro-cell UO{sub 2} pellet with a continuous and uniform channel of the Cr metallic phase was carried out. The formation of the Cr-oxide phases was prevented and the uniformity of the Cr-metal phase distribution was enhanced simultaneously, through the optimization of the additive-powder characteristics. In the results, the Cr metallic micro-cell pellet with continuous and uniform Cr metallic channel could be obtained.

  12. 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

  13. 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...

  14. 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.)

  15. 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...

  16. 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.

  17. 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)

  18. 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

  19. 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)

  20. 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

  1. 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)

  2. 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.

  3. 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

  4. 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)

  5. 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

  6. 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.

  7. The deformation analysis of the KALIMER breakeven core driver fuel pin based on the axial power profile during irradiation

    International Nuclear Information System (INIS)

    Lee, Dong Uk; Lee, Byoung Oon; Kim, Young Kyun; Hong, Ser Gi; Chang, Jin Wook; Lee, Ki Bok; Kim, Young Il

    2003-03-01

    In this study, material properties such as coolant specific heat, film heat transfer coefficient, cladding thermal conductivity, surface diffusion coefficient of the multi-bubble are improved in MACSIS-Mod1. The axial power and flux profile module was also incorporated with irradiation history. The performance and feasibility of the driver fuel pin have been analyzed for nominal parameters based on the conceptual design for the KALIMER breakeven core by MACSIS-MOD1 code. The fuel slug centerline temperature takes the maximum at 700mm from the bottom of the slug in spite of the nearly symmetric axial power distribution. The cladding mid-wall and coolant temperatures take the maximum at the top of the pin. Temperature of the fuel slug surface over the entire irradiation life is much lower than the fuel-clad eutectic reaction temperature. The fission gas release of the driver fuel pin at the End Of Life(EOL) is predicted to be 68.61% and plenum pressure is too low to cause cladding yielding. The probability that the fuel pin would fail is estimated to be much less than that allowed in the design criteria. The maximum radial deformation of the fuel pin is 1.928%, satisfying the preliminary design criterion (3%) for fuel pin deformation. Therefore the conceptual design parameters of the driver fuel pin for the KALIMER breakeven core are expected to satisfy the preliminary criteria on temperature, fluence limit, deformation limit etc

  8. The deformation analysis of the KALIMER breakeven core driver fuel pin based on the axial power profile during irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Uk; Lee, Byoung Oon; Kim, Young Kyun; Hong, Ser Gi; Chang, Jin Wook; Lee, Ki Bok; Kim, Young Il

    2003-03-01

    In this study, material properties such as coolant specific heat, film heat transfer coefficient, cladding thermal conductivity, surface diffusion coefficient of the multi-bubble are improved in MACSIS-Mod1. The axial power and flux profile module was also incorporated with irradiation history. The performance and feasibility of the driver fuel pin have been analyzed for nominal parameters based on the conceptual design for the KALIMER breakeven core by MACSIS-MOD1 code. The fuel slug centerline temperature takes the maximum at 700mm from the bottom of the slug in spite of the nearly symmetric axial power distribution. The cladding mid-wall and coolant temperatures take the maximum at the top of the pin. Temperature of the fuel slug surface over the entire irradiation life is much lower than the fuel-clad eutectic reaction temperature. The fission gas release of the driver fuel pin at the End Of Life(EOL) is predicted to be 68.61% and plenum pressure is too low to cause cladding yielding. The probability that the fuel pin would fail is estimated to be much less than that allowed in the design criteria. The maximum radial deformation of the fuel pin is 1.928%, satisfying the preliminary design criterion (3%) for fuel pin deformation. Therefore the conceptual design parameters of the driver fuel pin for the KALIMER breakeven core are expected to satisfy the preliminary criteria on temperature, fluence limit, deformation limit etc.

  9. 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

  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. 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.

  12. 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)

  13. 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)

  14. 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.)

  15. 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

  16. 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

  17. Ultrasonic measurement of gap between calandria tube and liquid injection shutdown system tube in PHWR

    International Nuclear Information System (INIS)

    Kim, Tae Ryong; Sohn, Seok Man; Lee, Jun Shin; Lee, Sun Ki; Lee, Jong Po

    2001-01-01

    Sag of CT or liquid injection shutdown system tubes in pressurized heavy water reactor is known to occur due to irradiation creep and growth during plant operation. When the sag of CT is big enough, the CT tube possibly comes in contact with liquid injection shutdown system tube (LIN) crossing beneath the CT, which subsequently may prevent the safe operation. It is therefore necessary to check the gap between the two tubes in order to confirm no contacts when using a proper measure periodically during the plant life. An ultrasonic gap measuring probe assembly which can be fed through viewing port installed on the calandria was developed and utilized to measure the sags of both tubes in a pressurized heavy water reactor in Korea. It was found that the centerlines of CT and LIN can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. But the measured gap data observed at the viewing port were actually not the data at the crossing point of CT and LIN. To get the actual gap between two tubes, mathematical modeling for the deflection curves of two tubes was used. The sags of CT and LIN tubes were also obtained by comparison of the present centerlines with the initial elevations at the beginning of plant operation. The gaps between two tubes in the unmeasurable regions were calculated based on the measurement data and the channel power distribution

  18. Behavior of four PWR rods subjected to a simulated loss-of-coolant accient in the power burst facility

    International Nuclear Information System (INIS)

    Cook, T.F.; Hagrman, D.L.; Sepold, L.K.

    1978-01-01

    Cladding deformation characteristics resulting from the first nuclear blowdown tests (LOC-11) conducted in the Power Burst Facility (PBF) are emphasized in this paper. The objective of the LOC-11 tests was to obtain data on the thermal, mechanical, and materials behavior of pressurized and unpressurized fuel rods when exposed to a blowdown similiar to that expected in a pressurized water reactor (PWR) during a hypothesized double-ended cold-leg break. The test hardware consisted of four separately shrouded fresh fuel rods of PWR 15 x 15 design. Initial plenum pressures ranged from atmospheric to 4.8 MPa (representative of end-of-life). During LOC-11C, the four fuel rods were subjected to 6.5 hours of nuclear operation at approximately 67 kW/m average rod power to cause decay heat build-up. Just before the start of blowdown, cladding surface temperatures were about 620 K and fuel centerline temperatures were in the 2500 to 2600 K range. During the 30-second blowdown transient, CHF occurred 2 seconds after initiation. Fuel centerline temperature dropped continuously, while cladding surface temperatures increased. Maximum cladding temperatures of 1030 to 1050 K occurred 15 seconds into the transient. Posttest destructive examination revealed cladding microstructures and oxide thicknesses consistent with the measured cladding temperatures. The cladding surface thermocouples did not appreciably affect cladding temperature distributuion (fin cooling effect) in the vicinity of the thermocouples

  19. 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

  20. 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

  1. 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

  2. Research on Elemental Technology of Advanced Nuclear Fuel Performance Verification

    International Nuclear Information System (INIS)

    Kim, Yong Soo; Lee, Dong Uk; Jean, Sang Hwan; Koo, Min

    2003-04-01

    Most of current properties models and fuel performance models used in the performance evaluation codes are based on the in-pile data up to 33,000 MWd/MtU. Therefore, international experts are investigating the properties changes and developing advanced prediction models for high burn-up application. Current research is to develop high burn-up fission gas release model for the code and to support the code development activities by collecting data and models, reviewing/assessing the data and models together, and benchmarking the selected models against the appropriate in-pile data. For high burn-up applications, two stage two step fission gas release model is developed based on the real two diffusion process in the grain lattice and grain boundaries of the fission gases and the observation of accelerated release rate in the high burn-up. It is found that the prediction of this model is in excellent agreement with the in-pile measurement results, not only in the low burn-up but also in the high burn-up. This research is found that the importance of thermal conductivity of oxide fuel, especially in the high burn-up, is focused again. It is found that even the temperature dependent models differ from one to another and most of them overestimate the conductivity in the high burn-up. An in-pile data benchmarking of high LHGR fuel rod shows that the difference can reach 30%∼40%, which predicts 400 .deg. C lower than the real fuel centerline temperature. Recent models on the thermal expansion and heat capacity of oxide fuel are found to be well-defined. Irradiation swelling of the oxide fuel are now well-understood that in most cases in LWRs solid fission product swelling is dominant. Thus, the accumulation of in-pile data can enhance the accuracy of the model prediction, rather than theoretical modeling works. Thermo-physical properties of Zircaloy cladding are also well-defined and well-understood except the thermal expansion. However, it turns out that even the

  3. 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)

  4. 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)

  5. 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...

  6. 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

  7. 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.

  8. Fabrication of fully ceramic microencapsulated fuel by hot pressing

    International Nuclear Information System (INIS)

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

    2014-01-01

    Fully ceramic microencapsulated(FCM) nuclear fuel is one of the recently suggested concept to enhance stability nuclear fuel itself. The requirements to increase the accident tolerance of nuclear fuel are mainly two parts: First, the performance has to be maintained compared to the existing UO 2 nuclear fuel and zircaloy cladding system under the normal operation condition. Second, under the severe accident condition, the high temperature structural integrity has to be kept and the generation rate of hydrogen has to be decrease largely. FCM nuclear fuel consists of tristructural isotropic(TRISO) fuel particle and SiC matrix. The relative thermal conductivity of the SiC matrix as compared to UO 2 is quite good, yielding as-irradiated fuel centerline temperature compared to high temperature for the existing fuel leading to reduced stored energy in the core and reduced operational release of fission products from the fuel. Generally SiC ceramics are fabricated via liquid phase sintering due to strong covalent bonding property and low self-diffusivity coefficient. Hot pressing is very effective method to conduct sintering of SiC powder including different second phase. In this study, SiC-matrix composite including TRISO particles were sintered by hot pressing with Al 2 O 3 -Y 2 O 3 additive system. Various sintering condition were investigated to obtain high relative density above 95%. The internal distribution of TRISO particles within SiC-matrix composite was observed by x-ray radiograph. From the analysis of the cross-section of SiC-matrix composite, the fracture of TRISO particles was investigated. In order to uniform distribution of TRISO particle embedded in the SiC matrix, SiC powder overcoating is considered. SiC matrix composite including TRISO was fabricated by hot pressing. FCM pallets with full density were obtained with Al 2 O 3 -Y 2 O 3 additive system. From the microstructure image, the effect of the sintering additive contents and sintering mechanism

  9. Maximum Diameter Measurements of Aortic Aneurysms on Axial CT Images After Endovascular Aneurysm Repair: Sufficient for Follow-up?

    International Nuclear Information System (INIS)

    Baumueller, Stephan; Nguyen, Thi Dan Linh; Goetti, Robert Paul; Lachat, Mario; Seifert, Burkhardt; Pfammatter, Thomas; Frauenfelder, Thomas

    2011-01-01

    Purpose: To assess the accuracy of maximum diameter measurements of aortic aneurysms after endovascular aneurysm repair (EVAR) on axial computed tomographic (CT) images in comparison to maximum diameter measurements perpendicular to the intravascular centerline for follow-up by using three-dimensional (3D) volume measurements as the reference standard. Materials and Methods: Forty-nine consecutive patients (73 ± 7.5 years, range 51–88 years), who underwent EVAR of an infrarenal aortic aneurysm were retrospectively included. Two blinded readers twice independently measured the maximum aneurysm diameter on axial CT images performed at discharge, and at 1 and 2 years after intervention. The maximum diameter perpendicular to the centerline was automatically measured. Volumes of the aortic aneurysms were calculated by dedicated semiautomated 3D segmentation software (3surgery, 3mensio, the Netherlands). Changes in diameter of 0.5 cm and in volume of 10% were considered clinically significant. Intra- and interobserver agreements were calculated by intraclass correlations (ICC) in a random effects analysis of variance. The two unidimensional measurement methods were correlated to the reference standard. Results: Intra- and interobserver agreements for maximum aneurysm diameter measurements were excellent (ICC = 0.98 and ICC = 0.96, respectively). There was an excellent correlation between maximum aneurysm diameters measured on axial CT images and 3D volume measurements (r = 0.93, P < 0.001) as well as between maximum diameter measurements perpendicular to the centerline and 3D volume measurements (r = 0.93, P < 0.001). Conclusion: Measurements of maximum aneurysm diameters on axial CT images are an accurate, reliable, and robust method for follow-up after EVAR and can be used in daily routine.

  10. Review of behavior of mixed-oxide fuel elements in extended overpower transient tests in EBR-II

    International Nuclear Information System (INIS)

    Tsai, H.; Neimark, L.A.

    1994-10-01

    From a series of five tests conducted in EBR-II, a substantial data base has been established on the performance of mixed-oxide fuel elements in a liquid-metal-cooled reactor under slow-ramp transient overpower conditions. Each test contained 19 preirradiated fuel elements with varying design and prior operating histories. Elements with aggressive design features, such as high fuel smear density and/or thin cladding, were included to accentuate transient effects. The ramp rates were either 0.1 or 10% ΔP/P/s and the overpowers ranged between ∼60 and 100% of the elements' prior power ratings. Six elements breached during the tests, all with aggressive design parameters. The other elements, including all those with moderate design features for the reference or advanced long-life drivers for PNC's prototype fast reactor Monju, maintained their cladding integrity during the tests. Posttest examination results indicated that fuel/cladding mechanical interaction (FCMI) was the most significant mechanism causing the cladding strain and breach. In contrast, pressure loading from the fission gas in the element plenum was less important, even in high-burnup elements. During an overpower transient, FCMI arises from fuel/cladding differential thermal expansion, transient fuel swelling, and, significantly, the gas pressure in the sealed central cavity of elements with substantial centerline fuel melting. Fuel performance data from these tests, including cladding breaching margin and transient cladding strain, are correlatable with fuel-element design and operating parameters. These correlations are being incorporated into fuel-element behavior codes. At the two tested ramp rates, fuel element behavior appears to be insensitive to transient ramp rate and there appears to be no particular vulnerability to slow ramp transients as previously perceived

  11. 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...

  12. Characterization of fuel swelling in helium-bonded carbide fuel pins

    International Nuclear Information System (INIS)

    Louie, D.L.Y.

    1987-08-01

    This work is not only the first attempt at characterizing the swelling of (U,Pu)C fuel pellets, but it also represents the only detailed examinations on carbide fuel swelling at high fuel burnups (4 to 16 at. %). This characterization includes the contributions of fission gases, cracks and solid fission products to fuel swelling. Significantly, the contributions of fission gases and cracks were determined by using the image analysis technique (IAT) which allows researchers to take areal measurements of the irradiated fuel porosity and cracks from the photographs of metallographic fuel samples. However, because areal measurements for varying depths in the fuel pellet could not be obtained, the crack areal measurements could not be converted into volumetric quantities. Consequently, in this situation, an areal fuel swelling analysis was used. The macroscopic fission-gas induced fuel swelling (MAS) caused by fission-gas bubbles and pores > 1 μm was determined using the measured irradiated fuel porosity because the measuring range of IAT is limited to bubbles and pores >1 μm. Conversely, for fuel swelling induced by fission-gas bubbles < 1 μm, the microscopic fission-gas induced fuel swelling (MIS) was estimated using an areal fuel swelling model

  13. Computation and measurement of calandria tube sag in PHWR

    International Nuclear Information System (INIS)

    Kim, Tae Ryong; Sohn, Seok Man

    2003-01-01

    Calandria tubes and liquid injection shutdown system (LISS) tubes in a pressurized heavy water reactor (PHWR) is known to sag due to irradiation creep and growth during plant operation. When the sag of calandria tube becomes bigger, the calandria tube possibly comes in contact with LISS tube crossing beneath and calandria tube. The contact subsequently may cause the damage on the calandria tube resulting in unpredicted outage of the plant. It is therefore necessary to check the gap between the two tubes in order to periodically confirm no contact by using a proper measure during the plant life. An ultrasonic gap measuring probe assembly which can be inserted into two viewing ports of the calandria was developed in Korea and utilized to measure the sags of both tubes in the PHWR. It was found that the centerlines of calandria tubes and liquid injection shutdown system tubes can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. Based on the irradiation creep equation and the measurement data, a computer program to calculate the sags was also developed. With the computer program, the sag at the end of plant life was predicted. (author)

  14. 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)

  15. 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.

  16. 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.

  17. 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.

  18. 40 CFR 86.884-9 - Smoke measurement system.

    Science.gov (United States)

    2010-07-01

    .... (1) The optical unit of the smokemeter shall be mounted radially to the exhaust pipe so that the... distance from the optical centerline to the exhaust pipe outlet shall be 1 ±0.25 inch. The full flow of the... Regulations for New Diesel Heavy-Duty Engines; Smoke Exhaust Test Procedure § 86.884-9 Smoke measurement...

  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. A Study of Pollutant Formation from the Lean Premixed Combustion of Gaseous Fuel Alternatives to Natural Gas

    Science.gov (United States)

    Fackler, Keith Boyd, Jr.

    The goal of this research is to identify how nitrogen oxide (NO x) emissions and flame stability (blowout) are impacted by the use of fuels that are alternatives to typical pipeline natural gas. The research focuses on lean, premixed combustors that are typically used in state-of-the-art natural gas fueled systems. An idealized laboratory lean premixed combustor, specifically the jet-stirred reactor, is used for experimental data. A series of models, including those featuring detailed fluid dynamics and those focusing on detailed chemistry, are used to interpret the data and understand the underlying chemical kinetic reasons for differences in emissions between the various fuel blends. An ultimate goal is to use these data and interpretive tools to develop a way to predict the emission and stability impacts of changing fuels within practical combustors. All experimental results are obtained from a high intensity, single-jet stirred reactor (JSR). Five fuel categories are studied: (1) pure H 2, (2) process and refinery gas, including combinations of H2, CH4, C2H6, and C3H8, (3) oxygen blown gasified coal/petcoke composed of H2, CO, and CO2, (4) landfill and digester gas composed of CH4, CO2, and N2, and (5) liquified natural gas (LNG)/shale/associated gases composed of CH4, C2H6, and C3 H8. NOx measurements are taken at a nominal combustion temperature of 1800 K, atmospheric pressure, and a reactor residence time of 3 ms. This is done to focus the results on differences caused by fuel chemistry by comparing all fuels at a common temperature, pressure, and residence time. This is one of the few studies in the literature that attempts to remove these effects when studying fuels varying in composition. Additionally, the effects of changing temperature and residence time are investigated for selected fuels. At the nominal temperature and residence time, the experimental and modeling results show the following trends for NOx emissions as a function of fuel type: 1.) NOx

  1. 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)

  2. 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.

  3. 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.

  4. 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

  5. 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)

  6. 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)

  7. Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Ahn, Sung-Ho; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system.

  8. Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

    International Nuclear Information System (INIS)

    Joung, Chang-Young; Ahn, Sung-Ho; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho

    2015-01-01

    Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system

  9. 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.

  10. 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)

  11. 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

  12. 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)

  13. 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.

  14. 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

  15. 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

  16. 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.

  17. 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

  18. Overall models and experimental database for UO2 and MOX fuel increasing performance

    International Nuclear Information System (INIS)

    Bernard, L.C.; Blanpain, P.

    2001-01-01

    COPERNIC is an advanced fuel rod performance code developed by Framatome. It is based on the TRANSURANUS code that contains a clear and flexible architecture, and offers many modeling possibilities. The main objectives of COPERNIC are to accurately predict steady-state and transient fuel operations at high burnups and to incorporate advanced materials such as the Framatome M5-alloy cladding. An extensive development program was undertaken to benchmark the code to very high burnups and to new M5-alloy cladding data. New models were developed for the M5-alloy cladding and the COPERNIC thermal models were upgraded and improved to extend the predictions to burnups over 100 GWd/tM. Since key phenomena, like fission gas release, are strongly temperature dependent, many other models were upgraded also. The COPERNIC qualification range extends to 67, 55, 53 GWd/tM respectively for UO 2 , UO 2 -Gd 2 O 3 , and MOX fuels with Zircaloy-4 claddings. The range extends to 63 GWd/tM with UO 2 fuel and the advanced M5-alloy cladding. The paper focuses on thermal and fission gas release models, and on MOX fuel modeling. The COPERNIC thermal model consists of several submodels: gap conductance, gap closure, fuel thermal conductivity, radial power profile, and fuel rim. The fuel thermal conductivity and the gap closure models, in particular, have been significantly improved. The model was benchmarked with 3400 fuel centerline temperature data from many French and international programs. There are no measured to predicted statistical biases with respect to linear heat generation rate or burnup. The overall quality of the model is state-of-the-art as the model uncertainty is below 10 %. The fission gas release takes into account athermal and thermally activated mechanisms. The model was adapted to MOX and Gadolinia fuels. For the heterogeneous MOX MIMAS fuels, an effective burnup is used for the incubation threshold. For gadolinia fuels, a scaled temperature effect is used. The

  19. 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

  20. Validation of an arterial tortuosity measure with application to hypertension collection of clinical hypertensive patients.

    Science.gov (United States)

    Diedrich, Karl T; Roberts, John A; Schmidt, Richard H; Kang, Chang-Ki; Cho, Zang-Hee; Parker, Dennis L

    2011-10-18

    Hypertension may increase tortuosity or twistedness of arteries. We applied a centerline extraction algorithm and tortuosity metric to magnetic resonance angiography (MRA) brain images to quantitatively measure the tortuosity of arterial vessel centerlines. The most commonly used arterial tortuosity measure is the distance factor metric (DFM). This study tested a DFM based measurement's ability to detect increases in arterial tortuosity of hypertensives using existing images. Existing images presented challenges such as different resolutions which may affect the tortuosity measurement, different depths of the area imaged, and different artifacts of imaging that require filtering. The stability and accuracy of alternative centerline algorithms was validated in numerically generated models and test brain MRA data. Existing images were gathered from previous studies and clinical medical systems by manually reading electronic medical records to identify hypertensives and negatives. Images of different resolutions were interpolated to similar resolutions. Arterial tortuosity in MRA images was measured from a DFM curve and tested on numerically generated models as well as MRA images from two hypertensive and three negative control populations. Comparisons were made between different resolutions, different filters, hypertensives versus negatives, and different negative controls. In tests using numerical models of a simple helix, the measured tortuosity increased as expected with more tightly coiled helices. Interpolation reduced resolution-dependent differences in measured tortuosity. The Korean hypertensive population had significantly higher arterial tortuosity than its corresponding negative control population across multiple arteries. In addition one negative control population of different ethnicity had significantly less arterial tortuosity than the other two. Tortuosity can be compared between images of different resolutions by interpolating from lower to higher

  1. 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)

  2. Spray combustion of biomass-based renewable diesel fuel using multiple injection strategy in a constant volume combustion chamber

    KAUST Repository

    Jing, Wei

    2016-05-26

    Effect of a two-injection strategy associated with a pilot injection on the spray combustion process was investigated under conventional diesel combustion conditions (1000 K and 21% O2 concentration) for a biomass-based renewable diesel fuel, i.e., biomass to liquid (BTL), and a regular No. 2 diesel in a constant volume combustion chamber using multiband flame measurement and two-color pyrometry. The spray combustion flame structure was visualized by using multiband flame measurement to show features of soot formation, high temperature and low temperature reactions, which can be characterized by the narrow-band emissions of radicals or intermediate species such as OH, HCHO, and CH. The objective of this study was to identify the details of multiple injection combustion, including a pilot and a main injection, and to provide further insights on how the two injections interact. For comparison, three injection strategies were considered for both fuels including a two-injection strategy (Case TI), single injection strategy A (Case SA), and single injection strategy B (Case SB). Multiband flame results show a strong interaction, indicated by OH emissions between the pilot injection and the main injection for Case TI while very weak connection is found for the narrow-band emissions acquired through filters with centerlines of 430 nm and 470 nm. A faster flame development is found for the main injection of Case TI compared to Cases SA and SB, which could be due to the high temperature environment and large air entrainment from the pilot injection. A lower soot level is observed for the BTL flame compared to the diesel flame for all three injection types. Case TI has a lower soot level compared to Cases SA and SB for the BTL fuel, while the diesel fuel maintains a similar soot level among all three injection strategies. Soot temperature of Case TI is lower for both fuels, especially for diesel. Based on these results, it is expected that the two-injection strategy could be

  3. 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...

  4. 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.

  5. Considerations in modelling the melting of fuel containing fission products and solute oxides

    International Nuclear Information System (INIS)

    Akbari, F.; Welland, M.J.; Lewis, B.J.; Thompson, W.T.

    2005-01-01

    It is well known that the oxidation of a defected fuel element by steam gives rise to an increase in O/U ratio with a consequent lowering of the incipient melting temperature. Concurrently, the hyperstoichiometry reduces the thermal conductivity thereby raising the centerline fuel pellet temperature for a fixed linear power. The development of fission products soluble in the UO 2 phase or, more important, the deliberate introduction of additive oxides in advanced CANDU fuel bundle designs further affects and generally lowers the incipient melting temperature. For these reasons, the modeling of the molten (hyperstoichiometric) UO 2 phase containing several solute oxides (ZrO 2 , Ln 2 O 3 and AnO 2 ) is advancing in the expectation of developing a moving boundary heat and mass transfer model aimed at better defining the limits of safe operating practice as burnup advances. The paper describes how the molten phase stability model is constructed. The redistribution of components across the solid-liquid interface that attends the onset of melting of a non-stoichiometric UO 2 containing several solutes will be discussed. The issues of how to introduce boundary conditions into heat transfer calculations consistent with the requirements of the Phase Rule will be addressed. The Stefan problem of a moving boundary associated with the solid/liquid interface sets this treatment apart from conventional heat and mass transfer problems. (author)

  6. 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.

  7. 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.

  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. 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

  10. 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

  11. 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)

  12. 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

  13. 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

  14. 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.

  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. 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.

  17. Instrumentation of fuel elements and fuel plates

    International Nuclear Information System (INIS)

    Durand, J.P.; Fanjas, Y.

    1993-01-01

    When controlling the behaviour of a reactor or developing a new fuel concept, it is of utmost interest to have the possibility to confirm the thermohydraulic calculations by actual measurements in the fuel elements or in the fuel plates. For years, CERCA has developed the technology and supplied its customers with fuel elements equipped with pressure or temperature measuring devices according to the requirements. Recent customer projects have led to the development of a new method to introduce thermocouples directly into the fuel plate meat instead of the cladding. The purpose of this paper is to review the various instrumentation possibilities available at CERCA. (author)

  18. Instrumentation of fuel elements and fuel plates

    International Nuclear Information System (INIS)

    Durand, J.P.; Fanjas, Y.

    1994-01-01

    When controlling the behaviour of a reactor or developing a new fuel concept, it is of utmost interest to have the possibility to confirm the thermohydraulic calculations by actual measurements in the fuel elements or in the fuel plates. For years, CERCA has developed the technology and supplied its customers with fuel elements equipped with pressure or temperature measuring devices according to the requirements. Recent customer projects have lead to the development of a new method to introduce thermocouples directly into the fuel plate meat instead of the cladding. The purpose of this paper is to review the various instrumentation possibilities available at CERCA. (author)

  19. 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.)

  20. 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.

  1. 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.

  2. 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)

  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. 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)

  5. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Science.gov (United States)

    Favalli, A.; Vo, D.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S. J.; Trellue, H.; Vaccaro, S.

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF 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 [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity's behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  6. 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)

  7. 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)

  8. 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)

  9. Fuel cell plates with skewed process channels for uniform distribution of stack compression load

    Science.gov (United States)

    Granata, Jr., Samuel J.; Woodle, Boyd M.

    1989-01-01

    An electrochemical fuel cell includes an anode electrode, a cathode electrode, an electrolyte matrix sandwiched between electrodes, and a pair of plates above and below the electrodes. The plate above the electrodes has a lower surface with a first group of process gas flow channels formed thereon and the plate below the electrodes has an upper surface with a second group of process gas flow channels formed thereon. The channels of each group extend generally parallel to one another. The improvement comprises the process gas flow channels on the lower surface of the plate above the anode electrode and the process gas flow channels on the upper surface of the plate below the cathode electrode being skewed in opposite directions such that contact areas of the surfaces of the plates through the electrodes are formed in crisscross arrangements. Also, the plates have at least one groove in areas of the surfaces thereof where the channels are absent for holding process gas and increasing electrochemical activity of the fuel cell. The groove in each plate surface intersects with the process channels therein. Also, the opposite surfaces of a bipolar plate for a fuel cell contain first and second arrangements of process gas flow channels in the respective surfaces which are skewed the same amount in opposite directions relative to the longitudinal centerline of the plate.

  10. Standalone BISON Fuel Performance Results for Watts Bar Unit 1, Cycles 1-3

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pawlowski, Roger [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Stimpson, Shane [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-03-07

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is moving forward with more complex multiphysics simulations and increased focus on incorporating fuel performance analysis methods. The coupled neutronics/thermal-hydraulics capabilities within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) have become relatively stable, and major advances have been made in analysis efforts, including the simulation of twelve cycles of Watts Bar Nuclear Unit 1 (WBN1) operation. While this is a major achievement, the VERA-CS approaches for treating fuel pin heat transfer have well-known limitations that could be eliminated through better integration with the BISON fuel performance code. Several approaches are being implemented to consider fuel performance, including a more direct multiway coupling with Tiamat, as well as a more loosely coupled one-way approach with standalone BISON cases. Fuel performance typically undergoes an independent analysis using a standalone fuel performance code with manually specified input defined from an independent core simulator solution or set of assumptions. This report summarizes the improvements made since the initial milestone to execute BISON from VERA-CS output. Many of these improvements were prompted through tighter collaboration with the BISON development team at Idaho National Laboratory (INL). A brief description of WBN1 and some of the VERA-CS data used to simulate it are presented. Data from a small mesh sensitivity study are shown, which helps justify the mesh parameters used in this work. The multi-cycle results are presented, followed by the results for the first three cycles of WBN1 operation, particularly the parameters of interest to pellet-clad interaction (PCI) screening (fuel-clad gap closure, maximum centerline fuel temperature, maximum/minimum clad hoop stress, and cumulative damage index). Once the mechanics of this capability are functioning, future work will target cycles with

  11. 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

  12. Ducted combustion chamber for direct injection engines and method

    Science.gov (United States)

    Mueller, Charles

    2015-03-03

    An internal combustion engine includes an engine block having a cylinder bore and a cylinder head having a flame deck surface disposed at one end of the cylinder bore. A piston connected to a rotatable crankshaft and configured to reciprocate within the cylinder bore has a piston crown portion facing the flame deck surface such that a combustion chamber is defined within the cylinder bore and between the piston crown and the flame deck surface. A fuel injector having a nozzle tip disposed in fluid communication with the combustion chamber has at least one nozzle opening configured to inject a fuel jet into the combustion chamber along a fuel jet centerline. At least one duct defined in the combustion chamber between the piston crown and the flame deck surface has a generally rectangular cross section and extends in a radial direction relative to the cylinder bore substantially along the fuel jet centerline.

  13. 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

  14. 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

  15. 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.

  16. 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.)

  17. Road and Street Centerlines, StreetLabels-The data set is a text feature consisting of 6329 label points representing street names. It was created to show the names 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. StreetLabels-The data set is a text feature consisting of 6329 label points representing street names. It was...

  18. 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

  19. 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

  20. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Favalli, A., E-mail: afavalli@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.J.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg)

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF 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 [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute {sup 137}Cs count rate and the {sup 154}Eu/{sup 137}Cs, {sup 134}Cs/{sup 137}Cs, {sup 106}Ru/{sup 137}Cs, and {sup 144}Ce/{sup 137}Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  1. 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

  2. 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

  3. 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.

  4. 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.

  5. 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

  6. 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.

  7. 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

  8. 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.

  9. 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

  10. 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.

  11. 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

  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. 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. Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

    Energy Technology Data Exchange (ETDEWEB)

    Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro, E-mail: duvan.castellanos@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: pedro.rossi@ufabc.edu.br, E-mail: pedro.carajilescov10@gmail.com [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil). Centro de Engenharias, Modelagem e Ciências Sociais Aplicadas

    2017-07-01

    To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

  15. Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

    International Nuclear Information System (INIS)

    Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro

    2017-01-01

    To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

  16. 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)

  17. 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.

  18. 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.

  19. 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)

  20. 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

  1. 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),...

  2. Fuel/propellant mixing in an open-cycle gas core nuclear rocket engine

    International Nuclear Information System (INIS)

    Guo, X.; Wehrmeyer, J.A.

    1997-01-01

    A numerical investigation of the mixing of gaseous uranium and hydrogen inside an open-cycle gas core nuclear rocket engine (spherical geometry) is presented. The gaseous uranium fuel is injected near the centerline of the spherical engine cavity at a constant mass flow rate, and the hydrogen propellant is injected around the periphery of the engine at a five degree angle to the wall, at a constant mass flow rate. The main objective is to seek ways to minimize the mixing of uranium and hydrogen by choosing a suitable injector geometry for the mixing of light and heavy gas streams. Three different uranium inlet areas are presented, and also three different turbulent models (k-var-epsilon model, RNG k-var-epsilon model, and RSM model) are investigated. The commercial CFD code, FLUENT, is used to model the flow field. Uranium mole fraction, axial mass flux, and radial mass flux contours are obtained. copyright 1997 American Institute of Physics

  3. 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.)

  4. Fuel mechanical design as a boundary condition for fuel management optimization

    International Nuclear Information System (INIS)

    Wunderlich, F.; Aisch, F.W.; Heins, L.

    1988-01-01

    The incentive to reduce fuel cycle costs as well as the amount of active waste requires, among others, measures to optimize fuel management. Improved fuel management in this sense calls, e.g., for reduction of parasitic neutron absorption, for reduction of neutron leakage, and particularly for burnup extension. Such measures result in increased demands for fuel mechanical design. In the first part of this paper their impact on fuel mechanical behaviour is described. In the second part, some examples of practical importance for the interaction between fuel management optimization and fuel mechanical design are discussed. (orig.) [de

  5. Towing Tank Measurements of Hydrodynamic Performance of a Horizontal Axis Tidal Turbine Under Unsteady Flow Conditions

    Science.gov (United States)

    2013-05-10

    John Zseleczky, Mr. Daniel Rhodes, Mr. Bill Beaver and all staff of US Naval Academy Hydromechanics Laboratory for their contributions in designing...turbine centerline. Tip vortex influence was most prevalent at X/D = 0.19, the closest measured plane to the turbine plane pictured in Figure 31

  6. 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

  7. 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.

  8. 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

  9. 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...

  10. 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

  11. Validation of OPERA3D PCMI Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Jeun, Ji Hoon; Choi, Jae Myung; Yoo, Jong Sung [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of); Cheng, G.; Sim, K. S.; Chassie, Girma [Candu Energy INC.,Ontario (Canada)

    2013-10-15

    This report will describe introduction of validation of OPERA3D code, and validation results that are directly related with PCMI phenomena. OPERA3D was developed for the PCMI analysis and validated using the in-pile measurement data. Fuel centerline temperature and clad strain calculation results shows close expectations with measurement data. Moreover, 3D FEM fuel model of OPERA3D shows slight hour glassing behavior of fuel pellet in contact case. Further optimization will be conducted for future application of OPERA3D code. Nuclear power plant consists of many complicated systems, and one of the important objects of all the systems is maintaining nuclear fuel integrity. However, it is inevitable to experience PCMI (Pellet Cladding Mechanical Interaction) phenomena at current operating reactors and next generation reactors for advanced safety and economics as well. To evaluate PCMI behavior, many studies are on-going to develop 3-dimensional fuel performance evaluation codes. Moreover, these codes are essential to set the safety limits for the best estimated PCMI phenomena aimed for high burnup fuel.

  12. Adjoint-based Mesh Optimization Method: The Development and Application for Nuclear Fuel Analysis

    International Nuclear Information System (INIS)

    Son, Seongmin; Lee, Jeong Ik

    2016-01-01

    In this research, methods for optimizing mesh distribution is proposed. The proposed method uses adjoint base optimization method (adjoint method). The optimized result will be obtained by applying this meshing technique to the existing code input deck and will be compared to the results produced from the uniform meshing method. Numerical solutions are calculated form an in-house 1D Finite Difference Method code while neglecting the axial conduction. The fuel radial node optimization was first performed to match the Fuel Centerline Temperature (FCT) the best. This was followed by optimizing the axial node which the Peak Cladding Temperature (PCT) is matched the best. After obtaining the optimized radial and axial nodes, the nodalization is implemented into the system analysis code and transient analyses were performed to observe the optimum nodalization performance. The developed adjoint-based mesh optimization method in the study is applied to MARS-KS, which is a nuclear system analysis code. Results show that the newly established method yields better results than that of the uniform meshing method from the numerical point of view. It is again stressed that the optimized mesh for the steady state can also give better numerical results even during a transient analysis

  13. Thermochemical aspects of fuel-cladding and fuel-coolant interactions in LMFBR oxide fuel pins

    International Nuclear Information System (INIS)

    Adamson, M.G.; Aitken, E.A.; Caputi, R.W.; Potter, P.E.; Mignanelli, M.A.

    1979-01-01

    This paper examines several thermochemical aspects of the fuel-cladding, fuel-coolant and fuel-fission product interactions that occur in LMFBR austenitic stainless steel-clad mixed (U,Pu)-oxide fuel pins during irradiation under normal operating conditions. Results are reported from a variety of high temperature EMF cell experiments in which continuous oxygen activity measurements on reacting and equilibrium mixtures of metal oxides and (excess) liquid alkali metal (Na, K, Cs) were performed. Oxygen potential and 0:M thresholds for Na-fuel reactions are re-evaluated in the light of new measurements and newly-assessed thermochemical data, and the influence on oxygen potential of possible U-Pu segregation between oxide and urano-plutonate (equilibrium) phases has been analyzed. (orig./RW) [de

  14. 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

  15. 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.

  16. 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.)

  17. Measurement and computation for sag of calandria tube due to irradiation creep in PHWR

    International Nuclear Information System (INIS)

    Son, S. M.; Lee, W. R.; Lee, S. K.; Lee, J. S.; Kim, T. R.; Na, B. K.; Namgung I.

    2003-01-01

    Calandria tubes and Liquid Injection Shutdown System(LISS) tubes in a Pressurized Heavy Water Reactor(PHWR) are to sag due to irradiation creep and growth during plant operation. When the sag of calandria tube becomes bigger, the calandria tube possibly comes in contact with LISS tube crossing beneath the calandria tube. The contact subsequently may cause the damage on the calandria tube resulting in unpredicted outage of the plant. It is therefore necessary to check the gap between the two tubes in order to periodically confirm no contact by using a proper measure during the plant life. An ultrasonic gap measuring probe assembly which can be inserted into two viewing ports of the calandria was developed in Korea and utilized to measure the sags of both tubes in the PHWR. It was found that the centerlines of calandria tubes and liquid injection shutdown system tubes can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. Based on the irradiation creep equation and the measurement data, a computer program to calculate the sags was also developed. With the computer program, the sag at the end of plant life was predicted

  18. 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.)

  19. 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

  20. 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)

  1. Analysis of irradiation temperature in fuel rods of OGL-1 fuel assembly

    International Nuclear Information System (INIS)

    Fukuda, Kousaku; Kobayashi, Fumiaki; Minato, Kazuo; Ikawa, Katsuichi; Iwamoto, Kazumi

    1984-10-01

    Irradiation temperature in the fuel rods of 5th OGL-1 fuel assembly was analysed by the system composed by STPDSP2 and TRUMP codes. As the measured input-data, following parameters were allowed for; circumferential heating distribution around the fuel rod, which was measured in the JMTR critical assembly, axial heating distribution through the fuel rod, ratio of peak heatings of three fuel rods, and pre- and post-irradiation outer radii of the fuel compacts and inner radii of the graphite sleeves, which had been measured in PIE of the 5th OGL-1 fuel assembly. In computation the axial distributions of helium coolant temperature through the fuel rod and the heating value of each fuel rod were, firstly, calculated as input data for TRUMP. The TRUMP calculation yielded the temperatures which were fitted in those measured by all of the thermo-couples installed in the fuel rods, by adjusting only the value of the surface heat transfer coefficient, and consequently, the temperatures in all portions of the fuel rod were obtained. The apparent heat transfer coefficient changed to 60% of the initial values in the middle period of irradiation. For this reduction it was deduced that shoot had covered the surface of the fuel rod during irradiation, which was confirmed in PIE. Beside it, several things were found in this analysis. (author)

  2. 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)

  3. Fission rates measured using high-energy gamma-rays from short half-life fission products in fresh and spent nuclear fuel

    International Nuclear Information System (INIS)

    Kroehnert, H.

    2011-02-01

    In recent years, higher discharge burn-ups and initial fuel enrichments have led to more and more heterogeneous core configurations in light water reactors (LWRs), especially at the beginning of cycle when fresh fuel assemblies are loaded next to highly burnt ones. As this trend is expected to continue in the future, the Paul Scherrer Institute has, in collaboration with the Swiss Association of Nuclear Utilities, swissnuclear, launched the experimental programme LIFE(at)PROTEUS. The LIFE(at)PROTEUS programme aims to better characterise interfaces between burnt and fresh UO 2 fuel assemblies in modern LWRs. Thereby, a novel experimental database is to be made available for enabling the validation of neutronics calculations of strongly heterogeneous LWR core configurations. During the programme, mixed fresh and highly burnt UO 2 fuel lattices will be investigated in the zero-power research reactor PROTEUS. One of the main types of investigations will be to irradiate the fuel in PROTEUS and measure the resulting fission rate distributions across the interface between fresh and burnt fuel zones. The measurement of fission rates in burnt fuel re-irradiated in a zero-power reactor requires, however, the development of new experimental techniques which are able to discriminate against the high intrinsic activity of the fuel. The principal goal of the present research work has been to develop such a new measurement technique. The selected approach is based on the detection of high-energy gamma-ray lines above the intrinsic background (i.e. above 2200 keV), which are emitted by short-lived fission products freshly created in the fuel. The fission products 88 Kr, 142 La, 138 Cs, 84 Br, 89 Rb, 95 Y, 90m Rb and 90 Rb, with half-lives between 2.6 min and 2.8 h, have been identified as potential candidates. During the present research work, the gamma-ray activity of short-lived fission products has, for the first time, been measured and quantitatively evaluated for re

  4. 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

  5. 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.)

  6. 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)

  7. Fuel Assembly Damping Summary

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kanghee; Kang, Heungseok; Oh, Dongseok; Yoon, Kyungho; Kim, Hyungkyu; Kim, Jaeyong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    This paper summary the fuel assembly damping data in air/in still water/under flow, released from foreign fuel vendors, compared our data with the published data. Some technical issues in fuel assembly damping measurement testing are also briefly discussed. Understanding of each fuel assembly damping mechanisms according to the surrounding medium and flow velocity can support the fuel design improvement in fuel assembly dynamics and structural integrity aspect. Because the upgraded requirements of the newly-developed advanced reactor system will demands to minimize fuel design margin in integrity evaluation, reduction in conservatism of fuel assembly damping can contribute to alleviate the fuel design margin for sure. Damping is an energy dissipation mechanism in a vibrating mechanical structure and prevents a resonant structure from having infinite vibration amplitudes. The sources of fuel assembly damping are various from support friction to flow contribution, and it can be increased by the viscosity or drag of surrounding fluid medium or the average velocity of water flowing. Fuel licensing requires fuel design evaluation in transient or accidental condition. Dynamic response analysis of fuel assembly is to show fuel integrity and requires information on assembly-wise damping in dry condition and under wet or water flowing condition. However, damping measurement test for the full-scale fuel assembly prototype is not easy to carry out because of the scale (fuel prototype, test facility), unsteadiness of test data (scattering, random sampling and processing), instrumentation under water flowing (water-proof response measurement), and noise. LWR fuel technology division in KAERI is preparing the infra structure for damping measurement test of full-scale fuel assembly, to support fuel industries and related research activities. Here is a preliminary summary of fuel assembly damping, published in the literature. Some technical issues in fuel assembly damping

  8. Experimental Study of Turbine Fuel Thermal Stability in an Aircraft Fuel System Simulator

    Science.gov (United States)

    Vranos, A.; Marteney, P. J.

    1980-01-01

    The thermal stability of aircraft gas turbines fuels was investigated. The objectives were: (1) to design and build an aircraft fuel system simulator; (2) to establish criteria for quantitative assessment of fuel thermal degradation; and (3) to measure the thermal degradation of Jet A and an alternative fuel. Accordingly, an aircraft fuel system simulator was built and the coking tendencies of Jet A and a model alternative fuel (No. 2 heating oil) were measured over a range of temperatures, pressures, flows, and fuel inlet conditions.

  9. 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.

  10. 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.

  11. 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

  12. Measurement of fission gas release, internal pressure and cladding creep rate in the fuel pins of PHWR bundle of normal discharge burnup

    Energy Technology Data Exchange (ETDEWEB)

    Viswanathan, U.K. [Post Irradiation Examination Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Sah, D.N., E-mail: dnsah@barc.gov.i [Post Irradiation Examination Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Rath, B.N.; Anantharaman, S. [Post Irradiation Examination Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

    2009-08-01

    Fuel pins of a Pressurised Heavy Water Reactor (PHWR) fuel bundle discharged from Narora Atomic Power Station unit no. 1 after attaining a fuel burnup of 7528 MWd/tU have been subjected to two types of studies, namely (i) puncture test to estimate extent of fission gas release and internal pressure in the fuel pin and (ii) localized heating of the irradiated fuel pin to measure the creep rate of the cladding in temperature range 800 deg. C - 900 deg. C. The fission gas release in the fuel pins from the outer ring of the bundle was found to be about 8%. However, only marginal release was found in fuel pins from the middle ring and the central fuel pin. The internal gas pressure in the outer fuel pin was measured to be 0.55 +- 0.05 MPa at room temperature. In-cell isothermal heating of a small portion of the outer fuel pins was carried out at 800 deg. C, 850 deg. C and 900 deg. C for 10 min and the increase in diameter of the fuel pin was measured after heat treatment. Creep rates of the cladding obtained from the measurement of the diameter change of the cladding due to heating at 800 deg. C, 850 deg. C and 900 deg. C were found respectively to be 2.4 x 10{sup -5} s{sup -1}, 24.6 x 10{sup -5} s{sup -1} and 45.6 x 10{sup -5} s{sup -1}.

  13. Core design and fuel rod analyses of a super fast reactor with high power density

    International Nuclear Information System (INIS)

    Ju, Haitao; Cao, Liangzhi; Lu, Haoliang; Oka, Yoshiaki; Ikejiri, Satoshi; Ishiwatari, Yuki

    2009-01-01

    A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) that is presently researched in a Japanese project. One of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. A preliminary core has an average power density of 158.8W/cc. In this paper, the principle of improving the average power density is studied and the core design is improved. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. This power density is competitive with that of typical Liquid Metal Fast Breeder Reactors (LMFBR). In order to ensure the fuel rod integrity of this core design, the fuel rod behaviors on the normal operating condition are analyzed using FEMAXI-6 code. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are taken from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak(MPP), Maximum Discharge Burnup(MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900degC. (2) Maximum cladding stress in circumstance direction should be less than 100MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Cumulative damage faction (CDF) of the cladding should be

  14. Evaluation of the 252Cf-source-driven neutron noise analysis method for measuring the subcriticality of LWR fuel storage casks

    International Nuclear Information System (INIS)

    Mihalczo, J.T.

    1987-01-01

    The 252 Cf-source-driven neutron noise analysis method was evaluated to determine if it could be used to measure the subcriticality of storage casks of burnt LWR fuel submerged in fuel storage pools, fully loaded and as they are being loaded. The motivation for this evaluation was that measurements of k/sub eff/ would provide the parameter most directly related to the criticality safety of storage cask configurations of LWR fuel and could allow proper credit for fuel burnup without reliance on calculations. This in turn could lead to more cost-effective cask designs. Evaluation of the method for this application was based on (1) experiments already completed at a critical experiments facility using arrays of PWR fuel pins typical of the size of storage cask configurations, (2) the existence of neutron detectors that can function in shipping cask environments, and (3) the ability to construct ionization chambers containing 252 Cf of adequate intensity for these measurements. These three considerations are discussed

  15. Assessment of core protection and monitoring systems for an advanced reactor SMART

    International Nuclear Information System (INIS)

    In, Wang Kee; Hwang, Dae Hyun; Yoo, Yeon Jong; Zee, Sung Qunn

    2002-01-01

    Analogue and digital core protection/monitoring systems were assessed for the implementation in an advanced reactor. The core thermal margins to nuclear fuel design limits (departure from nucleate boiling and fuel centerline melting) were estimated using the design data for a commercial pressurized water reactor and an advanced reactor. The digital protection system resulted in a greater power margin to the fuel centerline melting by at least 30% of rated power for both commercial and advanced reactors. The DNB margin with the digital system is also higher than that for the analogue system by 8 and 12.1% of rated power for commercial and advanced reactors, respectively. The margin gain with the digital system is largely due to the on-line calculations of DNB ratio and peak local power density from the live sensor signals. The digital core protection and monitoring systems are, therefore, believed to be more appropriate for the advanced reactor

  16. 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.

  17. 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

  18. 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

  19. Apparatus and method for a gas turbine nozzle

    Science.gov (United States)

    Zuo, Baifang; Ziminsky, Willy Steve; Johnson, Thomas Edward; Intile, John Charles; Lacy, Benjamin Paul

    2013-02-05

    A nozzle includes an inlet, an outlet, and an axial centerline. A shroud surrounding the axial centerline extends from the inlet to the outlet and defines a circumference. The circumference proximate the inlet is greater than the circumference at a first point downstream of the inlet, and the circumference at the first point downstream of the inlet is less than the circumference at a second point downstream of the first point. A method for supplying a fuel through a nozzle directs a first airflow along a first path and a second airflow along a second path separate from the first path. The method further includes injecting the fuel into at least one of the first path or the second path and accelerating at least one of the first airflow or the second airflow.

  20. Freshly induced short-lived gamma-ray activity as a measure of fission rates in lightly re-irradiated spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kroehnert, H., E-mail: hanna.kroehnert@psi.c [Paul Scherrer Institut (PSI), OPRA-E07, CH-5232 Villigen (Switzerland); Perret, G., E-mail: gregory.perret@psi.c [Paul Scherrer Institut (PSI), OPRA-E07, CH-5232 Villigen (Switzerland); Murphy, M.F., E-mail: mike.murphy@psi.c [Paul Scherrer Institut (PSI), OPRA-E07, CH-5232 Villigen (Switzerland); Chawla, R., E-mail: rakesh.chawla@epfl.c [Paul Scherrer Institut (PSI), OPRA-E07, CH-5232 Villigen (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland)

    2010-12-01

    A new measurement technique has been developed to determine fission rates in burnt fuel, following re-irradiation in a zero-power research reactor. The development has been made in the frame of the LIFE-PROTEUS program at the Paul Scherrer Institute, which aims at characterizing the interfaces between fresh and highly burnt fuel assemblies in modern LWRs. To discriminate against the high intrinsic gamma-ray activity of the burnt fuel, the proposed measurement technique uses high-energy gamma-rays, above 2000 keV, emitted by short-lived fission products freshly produced in the fuel. To demonstrate the feasibility of this technique, a fresh UO{sub 2} sample and a 36 GWd/t burnt UO{sub 2} sample were irradiated in the PROTEUS reactor and their gamma-ray activities were recorded directly after irradiation. For both fresh and the burnt fuel samples, relative fission rates were derived for different core positions, based on the short-lived {sup 142}La (2542 keV), {sup 89}Rb (2570 keV), {sup 138}Cs (2640 keV) and {sup 95}Y (3576 keV) gamma-ray lines. Uncertainties on the inter-position fission rate ratios were mainly due to the uncertainties on the net-area of the gamma-ray peaks and were about 1-3% for the fresh sample, and 3-6% for the burnt one. Thus, for the first time, it has been shown that the short-lived gamma-ray activity, induced in burnt fuel by irradiation in a zero-power reactor, can be used as a quantitative measure of the fission rate. For both fresh and burnt fuel, the measured results agreed, within the uncertainties, with Monte Carlo (MCNPX) predictions.

  1. Safety analysis of MOX fuels by fuel performance code

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    Performance of plutonium rick mixed oxide fuels specified for the Reduced-Moderation Water Reactor (RMWR) has been analysed by modified fuel performance code. Thermodynamic properties of these fuels up to 120 GWd/t burnup have not been measured and estimated using existing uranium fuel models. Fission product release, pressure rise inside fuel rods and mechanical loads of fuel cans due to internal pressure have been preliminarily assessed based on assumed axial power distribution history, which show the integrity of fuel performance. Detailed evaluation of fuel-cladding interactions due to thermal expansion or swelling of fuel pellets due to high burnup will be required for safety analysis of mixed oxide fuels. Thermal conductivity and swelling of plutonium rich mixed oxide fuels shall be taken into consideration. (T. Tanaka)

  2. Thermal property change of MOX and UO{sub 2} irradiated up to high burnup of 74 GWd/t

    Energy Technology Data Exchange (ETDEWEB)

    Nakae, Nobuo, E-mail: nakae-nobuo@jnes.go.jp [Japan Nuclear Energy Safety Organization (JNES), Toranomon Towers Office, 4-1-28, Toranomon, Minato-ku, Tokyo 105-0001 (Japan); Akiyama, Hidetoshi; Miura, Hiromichi; Baba, Toshikazu; Kamimura, Katsuichiro [Japan Nuclear Energy Safety Organization (JNES), Toranomon Towers Office, 4-1-28, Toranomon, Minato-ku, Tokyo 105-0001 (Japan); Kurematsu, Shigeru; Kosaka, Yuji [Nuclear Development Corporation (NDC), 622-12, Funaishikawa, Tokai-mura, Ibaraki 319-1111 (Japan); Yoshino, Aya; Kitagawa, Takaaki [Mitsubishi Nuclear Fuel Co., LTD. (MNF), 12-1, Yurakucho 1-Chome, Chiyoda-ku, Tokyo 100-0006 (Japan)

    2013-09-15

    Thermal property is important because it controls fuel behavior under irradiation. The thermal property change at high burnup of more than 70 GWd/t is examined. Two kinds of MOX fuel rods, which were fabricated by MIMAS and SBR methods, and one referenced UO{sub 2} fuel rod were used in the experiment. These rods were taken from the pre-irradiated rods (IFA 609/626, of which irradiation test were carried out by Japanese PWR group) and re-fabricated and re-irradiated in HBWR as IFA 702 by JNES. The specification of fuel corresponds to that of 17 × 17 PWR type fuel and the axially averaged linear heat rates (LHR) of MOX rods are 25 kW/m (BOL of IFA 702) and 20 kW/m (EOL of IFA 702). The axial peak burnups achieved are about 74 GWd/t for both of MOX and UO{sub 2}. Centerline temperature and plenum gas pressure were measured in situ during irradiation. The measured centerline temperature is plotted against LHR at the position where thermocouples are fixed. The slopes of MOX are corresponded to each other, but that of UO{sub 2} is higher than those of MOX. This implies that the thermal conductivity of MOX is higher than that of UO{sub 2} at high burnup under the condition that the pellet–cladding gap is closed during irradiation. Gap closure is confirmed by the metallography of the postirradiation examinations. It is understood that thermal conductivity of MOX is lower than that of UO{sub 2} before irradiation since phonon scattering with plutonium in MOX becomes remarkable. A phonon scattering with plutonium decreases in MOX when burnup proceeds. Thus, thermal conductivity of MOX becomes close to that of UO{sub 2}. A reverse phenomenon is observed at high burnup region. The phonon scattering with fission products such as Nd and Zr causes a degradation of thermal conductivity of burnt fuel. It might be speculated that this scattering effect causes the phenomenon and the mechanism is discussed here.

  3. 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.

  4. 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.

  5. Comparison of the effects of streptokinase and tissue plasminogen activator on regional wall motion after first myocardial infarction: analysis by the centerline method with correction for area at risk.

    Science.gov (United States)

    Cross, D B; Ashton, N G; Norris, R M; White, H D

    1991-04-01

    In a trial of streptokinase versus recombinant tissue-type plasminogen activator (rt-PA) for a first myocardial infarction, 270 patients were randomized. Regional left ventricular function was assessed in 214 patients at 3 weeks. The infarct-related artery was the left anterior descending artery in 78 patients, the right coronary artery in 122 and a dominant left circumflex artery in 14. Analysis was by the centerline method with a novel correction for the area of myocardium at risk, whereby the search region was determined by the anatomic distribution of the infarct-related artery. Infarct-artery patency at 3 weeks was 73% in the streptokinase group and 71% in the rt-PA group. Global left ventricular function did not differ between the two groups. Mean chord motion (+/- SD) in the most hypokinetic half of the defined search region was similar in the streptokinase and rt-PA groups (-2.4 +/- 1.5 versus -2.3 +/- 1.3, p = 0.63). There were no differences in hyperkinesia of the noninfarct zone. Compared with conventional centerline analysis, regional wall motion in the defined area at risk was significantly more abnormal. The two methods correlated strongly, however (r = 0.99, p less than 0.0001), and both methods produced similar overall results. Patients with a patent infarct-related artery and those with an occluded artery at the time of catheterization had similar levels of global function (ejection fraction 58 +/- 12% versus 57 +/- 12%, p = 0.58).(ABSTRACT TRUNCATED AT 250 WORDS)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-01-11

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

  7. 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)

  8. Benefits of barrier fuel on fuel cycle economics

    International Nuclear Information System (INIS)

    Crowther, R.L.; Kunz, C.L.

    1988-01-01

    Barrier fuel rod cladding was developed to eliminate fuel rod failures from pellet/cladding stress/corrosion interaction and to eliminate the associated need to restrict the rate at which fuel rod power can be increased. The performance of barrier cladding has been demonstrated through extensive testing and through production application to many boiling water reactors (BWRs). Power reactor data have shown that barrier fuel rod cladding has a significant beneficial effect on plant capacity factor and plant operating costs and significantly increases fuel reliability. Independent of the fuel reliability benefit, it is less obvious that barrier fuel has a beneficial effect of fuel cycle costs, since barrier cladding is more costly to fabricate. Evaluations, measurements, and development activities, however, have shown that the fuel cycle cost benefits of barrier fuel are large. This paper is a summary of development activities that have shown that application of barrier fuel significantly reduces BWR fuel cycle costs

  9. 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)

  10. Fission rates measured using high-energy gamma-rays from short half-life fission products in fresh and spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kroehnert, H.

    2011-02-15

    In recent years, higher discharge burn-ups and initial fuel enrichments have led to more and more heterogeneous core configurations in light water reactors (LWRs), especially at the beginning of cycle when fresh fuel assemblies are loaded next to highly burnt ones. As this trend is expected to continue in the future, the Paul Scherrer Institute has, in collaboration with the Swiss Association of Nuclear Utilities, swissnuclear, launched the experimental programme LIFE(at)PROTEUS. The LIFE(at)PROTEUS programme aims to better characterise interfaces between burnt and fresh UO{sub 2} fuel assemblies in modern LWRs. Thereby, a novel experimental database is to be made available for enabling the validation of neutronics calculations of strongly heterogeneous LWR core configurations. During the programme, mixed fresh and highly burnt UO{sub 2} fuel lattices will be investigated in the zero-power research reactor PROTEUS. One of the main types of investigations will be to irradiate the fuel in PROTEUS and measure the resulting fission rate distributions across the interface between fresh and burnt fuel zones. The measurement of fission rates in burnt fuel re-irradiated in a zero-power reactor requires, however, the development of new experimental techniques which are able to discriminate against the high intrinsic activity of the fuel. The principal goal of the present research work has been to develop such a new measurement technique. The selected approach is based on the detection of high-energy gamma-ray lines above the intrinsic background (i.e. above 2200 keV), which are emitted by short-lived fission products freshly created in the fuel. The fission products {sup 88}Kr, {sup 142}La, {sup 138}Cs, {sup 84}Br, {sup 89}Rb, {sup 95}Y, {sup 90m}Rb and {sup 90}Rb, with half-lives between 2.6 min and 2.8 h, have been identified as potential candidates. During the present research work, the gamma-ray activity of short-lived fission products has, for the first time, been

  11. Sensor system for fuel transport vehicle

    Science.gov (United States)

    Earl, Dennis Duncan; McIntyre, Timothy J.; West, David L.

    2016-03-22

    An exemplary sensor system for a fuel transport vehicle can comprise a fuel marker sensor positioned between a fuel storage chamber of the vehicle and an access valve for the fuel storage chamber of the vehicle. The fuel marker sensor can be configured to measure one or more characteristics of one or more fuel markers present in the fuel adjacent the sensor, such as when the marked fuel is unloaded at a retail station. The one or more characteristics can comprise concentration and/or identity of the one or more fuel markers in the fuel. Based on the measured characteristics of the one or more fuel markers, the sensor system can identify the fuel and/or can determine whether the fuel has been adulterated after the marked fuel was last measured, such as when the marked fuel was loaded into the vehicle.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  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. 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.

  20. Postirradiation examination of HTR fuel

    International Nuclear Information System (INIS)

    Nabielek, H.; Reitsamer, G.; Kania, M.J.

    1986-01-01

    Fuel for the High Temperature Reactor (HTR) consists of 1 mm diameter coated particles uniformly distributed in a graphite matrix within a cold-molded 60 mm diameter spherical fuel element. Fuel performance demonstrations under simulated normal operation conditions are conducted in accelerated neutron environments available in Material Test Reactors and in real-time environments such as the Arbeitsgemeinschaft Versuchsreaktor (AVR) Juelich. Postirradiation examinations are then used to assess fuel element behavior and the detailed performance of the coated particles. The emphasis in postirradiation examination and accident testing is on assessment of the capability for fuel elements and individual coated particles to retain fission products and actinide fuel materials. To accomplish this task, techniques have been developed which measures fission product and fuel material distributions within or exterior to the particle: Hot Gas Chlorination - provides an accurate method to measure total fuel material concentration outside intact particles; Profile Electrolytic Deconsolidation - permits determination of fission product distribution along fuel element diameter and retrieval of fuel particles from positions within element; Gamma Spectrometry - provides nondestructive method to measure defect particle fractions based on retention of volatile metallic fission products; Particle Cracking - permits a measure of the partitioning of fission products between fuel kernel and particle coatings, and the derivation of diffusion parameters in fuel materials; Micro Gas Analysis - provides gaseous fission product and reactive gas inventory within free volume of single particles; and Mass-spectrometric Burnup Determination - utilizes isotope dilution for the measurement of heavy metal isotope abundances

  1. 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)

  2. Analysis of the GCFR pin streaming experiment performed at the TSF

    International Nuclear Information System (INIS)

    Slater, C.O.; Bartine, D.E.

    1976-01-01

    An experiment is described which was performed to provide benchmark data to test GCFR fuel pin streaming calculations. The experiment, performed at the Tower Shielding Facility, consisted of 902 UO 2 fuel pins arranged on a triangular pitch with a void fraction comparable to that of the GCFR. A spectrum modifier consisting of a spectrum modifier was used to provide a spectrum similar to that of a fast reactor. Spectral measurements tended to show a strong streaming effect with the total flux showing a sharp drop over small angular traverses from the centerline. Two-dimensional calculations employing both homogeneous and heterogeneous models were used to calculate neutron spectra. Data are presented and compared

  3. Thermocouples for conditions of aggressive environments

    International Nuclear Information System (INIS)

    Blanc, J.Y.

    1988-01-01

    Two new kinds of thermocouples have been chosen for temperature measurements in the in-pile safety program for light water reactors performed in France. They must give fuel centerline or roc cladding temperatures and withstand steam oxidation between 1000 0 C and 1800 0 C or higher, under severe fuel damage conditions. We describe briefly both types, then we emphasize on improvements under way concerning the tungsten-rhenium legs, the hafnia insulation and the sheaths materials. Oxidation resistance is achieved mainly by silicides layers, but other possibilities are considered, such as iridium coatings. Some details of insulators manufacturing or sensor assembly are given, as well as other high temperature applications for these thermocouples

  4. 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.

  5. 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.

  6. Process and equipment for locating defective fuel rods of a reactor fuel element

    International Nuclear Information System (INIS)

    Jester, A.; Honig, H.

    1977-01-01

    By this equipment, well-known processes for determining defective fuel rods of a reactor fuel element are improved in such a fashion that defective fuel rods can be located individually, so that it is possible to replace them. The equipment consists of a cylindrical test vessel open above, which accommodates the element to be tested, so that an annular space is left between the latter's external circumference and the wall of the vessel, and so that the fuel rods project above the vessel. A bell in the shape of a frustrum of a cone is inverted over the test vessel, which has an infra-red measuring equipment at a certain distance above the tops of the fuel rods. The fuel element to be tested together with the test vessel and hood are immersed in a basin full of water, which displaces water by means of gas from the hood. The post-shutdown heat increases the temperature in the water space of the test vessel, which is stabilised at 100 0 C. In each defective fuel rod the water which has penetrated the defective fuel rod previously, or does so now, starts to boil. The steam rising in the fuel rod raises the temperature of the defective fuel rod compared to all the sound ones. The subsequent measurement easily determines this. Where one can expect interference with the measurement by appreciable amounts of gamma rays, the measuring equipment is removed from the path of radiation by mirror deflection in a suitably shaped measuring hood. (FW) [de

  7. 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.)

  8. 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)

  9. Validation and Improvement of the FEMAXI-JNES Code by Using PIE Data at Extended Burnup. Final Report for FUMEX-III

    International Nuclear Information System (INIS)

    Hirose, Tsutomu; Miura, Hiromichi; Kitamura, Toshiya; Kamimura, Katsuichiro

    2013-01-01

    Japan Nuclear Energy Safety Organization (JNES) has participated in the IAEA FUMEX-III Coordinated Research Project (CRP) on the Improvement of Computer Codes Used for Fuel Behaviour Simulation for the following purpose. 1. Cooperate between member states and exchange information and expertise for understanding of fuel modelling and improvement 2. Develop and improve the FEMAXI-JNES code as an audit code for Japanese safety licensing review of fuel rod design, especially, - High burnup fuel - MOX fuel 3. Set the standard models for the FEMAXI-JNES code to provide best-estimate predictions of the thermal and mechanical performance of LWR fuel rod This is the JNES's final report for the FUMEX-III CRP. During the period of the CRP, JNES has modified pellet swelling and fission gas release models, and demonstrated the predictive capability relative to fuel centerline temperature, fission gas release, fuel rod internal gas pressure, cladding diametral deformation and cladding elongation by comparisons of integral code predictions of these parameters to experimental (measured) data from OECD/NEA IFPE database. (author)

  10. 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)

  11. 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.

  12. Neutron measurement method for the detection of transuranic elements in the nuclear fuel cycle; Neutronenmessverfahren fuer den Nachweis von Transuranen im Kernbrennstoff-Kreislauf

    Energy Technology Data Exchange (ETDEWEB)

    Sokcic-Kostic, Marina; Schultheis, Roland [NUKEM Technologies GmbH, Alzenau (Germany)

    2014-04-15

    By handling and storing burned-down fuel elements it is duty to measure the existing nuclear fuel content. For the criticality analysis of the interim storage, for instance, it is imperative to know the nuclear fuel inventory in order to give a detailed description on the safety of storage to the supervisory authority. Among other things, it is necessary to consider that the possibility of mixing up stored fuel elements in the fuel pool was not able to be excluded. Due to high penetration of matter and its origin from decay or spontaneous fission of transuranic elements neutron verification methods are suited best for the proof of fission material as long as it has been burned-down beforehand. If fission chambers are additionally used as detectors, measurements can be even carried out in environments with high gamma levels. A highly improved measuring quality can be achieved, by comparing measurement results with the results of computer-aided simulations such as e.g. burn-up programs or MCNP- calculations. Hereby the impact on the measurement result by special marginal conditions of the measuring environment (e.g. addition of boric acid into the water of the fuel pool) can be estimated and thus revised. It is shown, that the passive neutron measurement is much easier to manage as an active measurement. As restriction it is to be considered, that measurements refer essentially to transuranic elements. Uranium such as U-235, however, is difficult to detect. For fuel elements applies, that the creation of transuranic elements is directly linked to the burn-up of U-235. Hence direct conclusions to the burn-up of U-235 can be drawn, by measuring transuranic content. (orig.)

  13. 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

  14. 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

  15. 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...

  16. 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.

  17. Application of Integral Ex-Core and Differential In-Core Neutron Measurements for Adjustment of Fuel Burn-Up Distributions in VVER-1000

    Science.gov (United States)

    Borodkin, Pavel G.; Borodkin, Gennady I.; Khrennikov, Nikolay N.

    2010-10-01

    The paper deals with calculational and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Time-integrated neutron source distributions used for DORT calculations were prepared by two different approaches based on a) calculated fuel burn-up (standard routine procedure) and b) in-core measurements by means of SPD & TC (new approach). Taking into account that fuel burn-up distributions in operating VVER may be evaluated now by analytical methods (calculations) only it is needed to develop new approaches for testing and correction of calculational evaluations. Results presented in this paper allow to consider a reverse task of alternative estimation of fuel burn-up distributions. The approach proposed is based on adjustment (fitting) of time-integrated neutron source distributions, and hence fuel burn-up patterns in some part of reactor core, on the base of ex-core neutron leakage measurement, neutron-physical calculation and in-core SPD & TC measurement data.

  18. 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.

  19. 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

  20. 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.

  1. 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

  2. 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)

  3. 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)

  4. 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....

  5. 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...

  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. 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

  8. 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)

  9. 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.

  10. 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

  11. 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)

  12. 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

  13. Planar laser-induced fluorescence fuel imaging during gas-turbine relight

    DEFF Research Database (Denmark)

    Read, Robert; Rogerson, J.W.; Hochgreb, S.

    2013-01-01

    of ignition when the airflow rate is high. In the presence of flame, medium-sized burning droplets are observed close to the injector centerline. Flame interference resulting from fluorescence of polycyclic aromatic hydrocarbons is apparent, but small, suggesting that kerosene planar laser...

  14. Mixed U/Pu oxide fuel fabrication facility co-processed feed, pelletized fuel

    International Nuclear Information System (INIS)

    1978-09-01

    Two conceptual MOX fuel fabrication facilities are discussed in this study. The first facility in the main body of the report is for the fabrication of LWR uranium dioxide - plutonium dioxide (MOX) fuel using co-processed feed. The second facility in the addendum is for the fabrication of co-processed MOX fuel spiked with 60 Co. Both facilities produce pellet fuel. The spiked facility uses the same basic fabrication process as the conventional MOX plant but the fuel feed incorporates a high energy gamma emitter as a safeguard measure against diversion; additional shielding is added to protect personnel from radiation exposure, all operations are automated and remote, and normal maintenance is performed remotely. The report describes the fuel fabrication process and plant layout including scrap and waste processing; and maintenance, ventilation and safety measures

  15. 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)

  16. 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

  17. 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

  18. 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 thickness of the coating layers of 196 coated particles was measured using an Image Plate detector, high resolution scanner and digital image processing techniques. The experimental results are as follows. - An X-ray image was acquired for 196 simulated TRISO-coated fuel particles with ZrO 2 kernel using an Image Plate with high resolution in a reduced amount of time. - We could observe clear boundaries between coating layers for 196 particles. - The geometric distortion error was compensated for the calculation. - The coating thickness of the TRISO-coated fuel particles can be nondestructively measured using X-ray radiography and digital image processing technology. - We can increase the number of TRISO-coated particles to be inspected by increasing the number of Image Plate detectors. A TRISO-coated fuel particle for an HTGR (high temperature gas-cooled reactor) is composed of a nuclear fuel kernel and outer coating layers. The coating layers consist of buffer PyC (pyrolytic carbon), inner PyC (I-PyC), SiC, and outer PyC (O-PyC) layer. The coating thickness is measured to evaluate the soundness of the coating layers. X-ray radiography is one of the nondestructive alternatives for measuring the coating thickness without generating a radioactive waste. Several billion particles are subject to be loaded in a reactor. A lot of sample particles should be tested as much as possible. The acquired X-ray images for the measurement of coating thickness have included a small number of particles because of the restricted resolution and size of the X-ray detector. We tried to test many particles for an X-ray exposure to reduce the measurement time. In this experiment, an X-ray image was acquired for 196 simulated TRISO-coated fuel particles using an image plate and high resolution scanner with a pixel size of 25Χ25 μm 2 . The coating thickness for the particles could be measured on the image

  19. 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

  20. Uncertainty quantification and sensitivity analysis with CASL Core Simulator VERA-CS

    International Nuclear Information System (INIS)

    Brown, C.S.; Zhang, Hongbin

    2016-01-01

    VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis. A 2 × 2 fuel assembly model was developed and simulated by VERA-CS, and uncertainty quantification and sensitivity analysis were performed with fourteen uncertain input parameters. The minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surface temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. Parameters used as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.

  1. 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.

  2. Residual stresses determination in an 8 mm Incoloy 800H weld via neutron diffraction

    International Nuclear Information System (INIS)

    Chen, Xizhang; Zhang, Shu Yan; Wang, Jingjun; Kelleher, Joe F.

    2015-01-01

    Highlights: • Stress through thickness at 5 mm from weld centerline indicates a “U” distribution. • Declining of tensile stress through thickness occurred at weld centerline. • Residual stress between layers is the lowest. - Abstract: To investigate the distribution of residual stresses, the 8 mm 800H alloy was joined by multi-layer butt TIG process. Residual stresses in the longitudinal, transverse and normal directions were measured via neutron diffraction. These residual stress measurements were taken at a series of points 2 mm below the top surface, covering the fusion zone, heat affected zone (HAZ) and base metal. In addition, two lines of longitudinal residual stress values at the weld centerline and 5 mm from weld centerline through thickness were measured. Results show that both the longitudinal and transverse stresses from the weld centerline to base metal are mainly tensile stresses. The longitudinal residual stress is the largest, with a maximum value of 330 MPa. As for the normal residual stress, the weld zone shows tensile stress, while the HAZ shows compressive stress. The middle of the thickness shows compressive residual stress along the thickness direction. The longitudinal stress at weld centerline through thickness reveals the interlayer heat treat effects leads to a declining of tensile stress. While the stress at 5 mm from weld centerline indicates a “U” distribution due to the mixed microstructure close to fusion line. With the increasing distance from weld seam, the residual stress decreases gradually

  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. 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

  5. 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

  6. Measurement of dose rates and Monte Carlo analysis of neutrons in a spent-fuel shipping vessel

    International Nuclear Information System (INIS)

    Ueki, K.; Namito, Y.; Fuse, T.

    1986-01-01

    On-board experiments were carried out in a spent-fuel shipping vessel, the Pacific Swan, in which 13 casks of TN-12A and Excellox 3 were loaded in five holds, and neutron and gamma-ray dose rates were measured on the hatch covers of the holds. Before shipping those casks, dose rates were also measured on the cask surfaces, one by one, to eliminate radiation from other casks. The Monte Carlo coupling technique was employed successfully to analyze the measured neutron dose rate distributions in the spent-fuel shipping vessel. Through this study, the Monte Carlo coupling code system, MORSE-CG/CASK-VESSEL, on which the MORSE-CG code was based, was established. The agreement between the measured and the calculated neutron dose rates on the TN-12A cask surface was quite satisfactory. The calculated neutron dose rates agreed with the measured values within a factor of 1.5 on the hold 3 hatch cover and within a factor of 2 on the hold 5 hatch cover in which the concrete shield was fixed in the Pacific Swan

  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. 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

  9. Effects of Operating Parameters on Measurements of Biochemical Oxygen Demand Using a Mediatorless Microbial Fuel Cell Biosensor

    Directory of Open Access Journals (Sweden)

    Min-Chi Hsieh

    2015-12-01

    Full Text Available The conventional Biochemical Oxygen Demand (BOD method takes five days to analyze samples. A microbial fuel cell (MFC may be an alternate tool for rapid BOD determination in water. However, a MFC biosensor for continuous BOD measurements of water samples is still unavailable. In this study, a MFC biosensor inoculated with known mixed cultures was used to determine the BOD concentration. Effects of important parameters on establishing a calibration curve between the BOD concentration and output signal from the MFC were evaluated. The results indicate monosaccharides were good fuel, and methionine, phenylalanine, and ethanol were poor fuels for electricity generation by the MFC. Ions in the influent did not significantly affect the MFC performance. CN− in the influent could alleviate the effect of antagonistic electron acceptors on the MFC performance. The regression equation for BOD concentration and current density of the biosensor was y = 0.0145x + 0.3317. It was adopted to measure accurately and continuously the BOD concentration in actual water samples at an acceptable error margin. These results clearly show the developed MFC biosensor has great potential as an alternative BOD sensing device for online measurements of wastewater BOD.

  10. HEU and Leu FueL Shielding Comparative Study Applied for Spent Fuel Transport

    International Nuclear Information System (INIS)

    Margeanu, C.A.; Margeanu, S.; Barbos, D.

    2009-01-01

    INR Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies. The dual core concept involves the operation of a 14 MW TRIGA steady-state, high flux research and material testing reactor at one end of a large pool, and the independent operation of an annular-core pulsing reactor (TRIGA-ACPR) at the other end of the pool. The steady-state reactor is mostly used for long term testing of power reactor fuel components (pellets, pins, subassemblies and fuel assemblies) followed by post-irradiation examination. Following the general trend to replace the He fuel type (High Enriched Uranium) by Leu fuel type (Low Enriched Uranium), in the light of international agreements between IAEA and the states using He fuel in their nuclear reactors, Inr Past's have been accomplished the TRIGA research reactor core full conversion on May 2006. The He fuel repatriation in US in the frame of Foreign Research Reactor Spent Nuclear Fuel Return Programme effectively started in 1999, the final stage being achieved in summer of 2008. Taking into account for the possible impact on the human and environment, in all activities associated to nuclear fuel cycle, the spent fuel or radioactive waste characteristics must be well known. Shielding calculations basic tasks consist in radiation doses calculation, in order to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper is a comparative study of Leu and He fuel utilization effects for the shielding analysis during spent fuel transport. A comparison against the measured data for He spent fuel, available from the last stage of the spent fuel repatriation, is presented. All the geometrical and material data related on the spent fuel shipping cask were considered according to the Nac-Lt Cask approved model. The shielding analysis estimates radiation doses to shipping cask wall surface

  11. 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

  12. Additional measurements of the radiation environment at the Los Alamos Spallation Radiation Effects Facility at LAMPF

    International Nuclear Information System (INIS)

    Davidson, D.R.; Reedy, R.C.; Greenwood, L.R.; Sommer, W.F.; Wechsler, M.S.

    1987-01-01

    Foil activation dosimetry experiments were conducted in a ''rabbit'' system at the completed Los Alamos Spallation Effects Facility(LASREF). The ''rabbit'' system contains four test tubes spaced radially outward 0.12, 0.18, 0.27, and 0.38 m off beam centerline. Foils were irradiated for 3 to 62 h to measure the neutron flux and energy spectrum radially from beam centerline, along the beamline, and the effect of isotope production (IP) target loadings on the neutron flux in the neutron irradiation locations. Irradiations showed a decrease in the radial flux by a factor of 6 in 0.15 m of iron outside the IP targets. An enhancement was seen in the 24-keV energy region outside 0.15m. There was little difference in the shape of the spectra outside the IP targets and the beam stop with the exception of the high energy tail (energies above 20 MeV). The decrease in the high energy tail outside the beam stop is due to the degradation of the energy of the proton beam in the IP targets. Irradiations outside the beam stop with zero and eight IP targets gave the same spectral shape with the exception of the high energy tail. The magnitude of the integral flux decreased by a factor of two when eight IP targets were present. Irradiations with five ''rabbits'' stacked on top of each other showed no difference in the integral flux below, on and above beam centerline

  13. Additional measurements of the radiation environment at the Los Alamos Spallation Radiation Effects Facility at LAMPF

    International Nuclear Information System (INIS)

    Davidson, D.R.; Reedy, R.C.; Greenwood, L.R.; Sommer, W.F.; Wechsler, M.S.

    1986-01-01

    Foil activation dosimetry experiments were conducted in a ''rabbit'' system at the completed Los Alamos Spallation Radiation Effects Facility (LASREF). The ''raffit'' system contains four tubes spaced radially outward 0.12, 0.18, 0.27, and 0.38 meters off beam centerline. Foils were irradiated for 3 to 62 hours to measure the neutron flux and energy spectrum radially from beam centerline, along the beamline, and the effect of the Isotope Production (IP) target loadings on the neutron flux in the neutron irradiation locations. Irradiations showed a decrease in the radial flux by a factor of 6 in 0.15 meters of iron outside the IP targets. An enchancement was seen in the 24-keV energy region outside 0.15 meters. There was little difference in the shape of the spectra outside the IP targets and the beam stop with the exception of the high energy tail (energies above 20 MeV). The decrease in the high energy tail outside the beam stop is due to the degradation of the energy of the proton beam in the IP targets. Irradiations outside the beam stop with zero and eight IP targets gave the same spectral shape with the exception of the high energy tail. The magnitude of the integral flux decreased by a factor of 2 when eight IP targets were present. Irradiations with five ''rabbits'' stacked on top of each other showed no difference in the integral flux below, on and above beam centerline

  14. 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.

  15. 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.

  16. Fuel loading and homogeneity analysis of HFIR design fuel plates loaded with uranium silicide fuel

    International Nuclear Information System (INIS)

    Blumenfeld, P.E.

    1995-08-01

    Twelve nuclear reactor fuel plates were analyzed for fuel loading and fuel loading homogeneity by measuring the attenuation of a collimated X-ray beam as it passed through the plates. The plates were identical to those used by the High Flux Isotope Reactor (HFIR) but were loaded with uranium silicide rather than with HFIR's uranium oxide fuel. Systematic deviations from nominal fuel loading were observed as higher loading near the center of the plates and underloading near the radial edges. These deviations were within those allowed by HFIR specifications. The report begins with a brief background on the thermal-hydraulic uncertainty analysis for the Advanced Neutron Source (ANS) Reactor that motivated a statistical description of fuel loading and homogeneity. The body of the report addresses the homogeneity measurement techniques employed, the numerical correction required to account for a difference in fuel types, and the statistical analysis of the resulting data. This statistical analysis pertains to local variation in fuel loading, as well as to ''hot segment'' analysis of narrow axial regions along the plate and ''hot streak'' analysis, the cumulative effect of hot segment loading variation. The data for all twelve plates were compiled and divided into 20 regions for analysis, with each region represented by a mean and a standard deviation to report percent deviation from nominal fuel loading. The central regions of the plates showed mean values of about +3% deviation, while the edge regions showed mean values of about -7% deviation. The data within these regions roughly approximated random samplings from normal distributions, although the chi-square (χ 2 ) test for goodness of fit to normal distributions was not satisfied

  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. 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.)

  19. Fuel Cell Power Plant Initiative. Volume 1; Solid Oxide Fuel Cell/Logistics Fuel Processor 27 kWe Power System Demonstration for ARPA

    Science.gov (United States)

    Veyo, S.E.

    1997-01-01

    This report describes the successful testing of a 27 kWe Solid Oxide Fuel Cell (SOFC) generator fueled by natural gas and/or a fuel gas produced by a brassboard logistics fuel preprocessor (LFP). The test period began on May 24, 1995 and ended on February 26, 1996 with the successful completion of all program requirements and objectives. During this time period, this power system produced 118.2 MWh of electric power. No degradation of the generator's performance was measured after 5582 accumulated hours of operation on these fuels: local natural gas - 3261 hours, jet fuel reformate gas - 766 hours, and diesel fuel reformate gas - 1555 hours. This SOFC generator was thermally cycled from full operating temperature to room temperature and back to operating temperature six times, because of failures of support system components and the occasional loss of test site power, without measurable cell degradation. Numerous outages of the LFP did not interrupt the generator's operation because the fuel control system quickly switched to local natural gas when an alarm indicated that the LFP reformate fuel supply had been interrupted. The report presents the measured electrical performance of the generator on all three fuel types and notes the small differences due to fuel type. Operational difficulties due to component failures are well documented even though they did not affect the overall excellent performance of this SOFC power generator. The final two appendices describe in detail the LFP design and the operating history of the tested brassboard LFP.

  20. Developing Spent Fuel Assembly for Advanced NDA Instrument Calibration - NGSI Spent Fuel Project

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Banfield, James [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Skutnik, Steven [Univ. of Tennessee, Knoxville, TN (United States)

    2014-02-01

    This report summarizes the work by Oak Ridge National Laboratory to investigate the application of modeling and simulation to support the performance assessment and calibration of the advanced nondestructive assay (NDA) instruments developed under the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Advanced NDA instrument calibration will likely require reference spent fuel assemblies with well-characterized nuclide compositions that can serve as working standards. Because no reference spent fuel standard currently exists, and the practical ability to obtain direct measurement of nuclide compositions using destructive assay (DA) measurements of an entire fuel assembly is prohibitive in the near term due to the complexity and cost of spent fuel experiments, modeling and simulation will be required to construct such reference fuel assemblies. These calculations will be used to support instrument field tests at the Swedish Interim Storage Facility (Clab) for Spent Nuclear Fuel.