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Sample records for test fuel production

  1. Behavior of fission products released from severely damaged fuel during the PBF severe fuel damage tests

    International Nuclear Information System (INIS)

    Osetek, D.J.; Cronenberg, A.W.; Hagrman, D.L.; Broughton, J.M.; Rest, J.

    1984-01-01

    The results of fission product release behavior during the first two Power Burst Facility Severe Fuel Damage tests are presented. Measured fission product release is compared with calculated release using temperature dependent release rate correlations and FASTGRASS analysis. The test results indicate that release from fuel of the high volatility fission products (Xe, Kr, I, Cs, and Te) is strongly influenced by parameters other than fuel temperature; namely fuel/fission product morphology, fuel and cladding oxidation state, extent of fuel liquefaction, and quench induced fuel shattering. Fission product transport from the test fuel through the sample system was strongly influenced by chemical effects. Holdup of I and Cs was affected by fission product chemistry, and transport time while Te release was primarily influenced by the extent of zircaloy oxidation. Analysis demonstrates that such integral test data can be used to confirm physical, chemical, and mechanistic models of fission product behavior for severe accident conditions

  2. Fission product release measured during fuel damage tests at the Power Burst Facility

    International Nuclear Information System (INIS)

    Osetek, D.J.; Hartwell, J.K.; Vinjamuri, K.; Cronenberg, A.W.

    1985-01-01

    Results are presented of fission product release behavior observed during four severe fuel damage tests on bundles of UO 2 fuel rods. Transient temperatures up to fuel melting were obtained in the tests that included both rapid quench and slow cooldown, low and high (36 GWd/t) burnup fuel and the addition of Ag-In-Cd control rods. Release fractions of major fission product species and release rates of noble gas species are reported. Significant differences in release behavior are discussed between heatup and cooldown periods, low and high burnup fuel and long- and short-lived fission products. Explanations are offered for the probable reasons for the observed differences and recommendations for further studies are given

  3. Fission product behavior during the first two PBF severe fuel damage tests

    International Nuclear Information System (INIS)

    Osetek, D.J.; Cronenberg, A.W.; Hobbins, R.R.; Vinjamuri, K.

    1984-01-01

    The results of the first two severe fuel damage tests performed in the Power Burst Facility are assessed in terms of fission product release and chemical behavior. On-line gamma spectroscopy and grab sample data indicate limited release during solid-phase fuel heatup. Analysis indicates that the fuel morphology conditions for the trace-irradiated fuel employed in these two tests limit initial release. Only upon high temperature fuel restructuring and liquefaction is significant release indicated. Chemical equilibrium predictions, based on steam oxidation or reduction conditions, indicate I to be the primary iodine species during trnsport in the steam environment of the first test and CsI to be the primary species during transport in the hydrogen environment of the second test. However, the higher steam flow rate conditions of the first test transported the released iodine through the sample system; whereas, low-hydrogen flow rate of the second test apparently allowed the vast majority of iodine-bearing compounds to plateout during transport

  4. Chemical Characterization and Reactivity Testing of Fuel-Oxidizer Reaction Product (Test Report)

    Science.gov (United States)

    1996-01-01

    The product of incomplete reaction of monomethylhydrazine (MMH) and nitrogen tetroxide (NTO) propellants, or fuel-oxidizer reaction product (FORP), has been hypothesized as a contributory cause of an anomaly which occurred in the chamber pressure (PC) transducer tube on the Reaction Control Subsystem (RCS) aft thruster 467 on flight STS-51. A small hole was found in the titanium-alloy PC tube at the first bend below the pressure transducer. It was surmised that the hole may have been caused by heat and pressure resulting from ignition of FORP. The NASA Johnson Space Center (JSC) White Sands Test Facility (WSTF) was requested to define the chemical characteristics of FORP, characterize its reactivity, and simulate the events in a controlled environment which may have lead to the Pc-tube failure. Samples of FORP were obtained from the gas-phase reaction of MMH with NTO under laboratory conditions, the pulsed firings of RCS thrusters with modified PC tubes using varied oxidizer or fuel lead times, and the nominal RCS thruster firings at WSTF and Kaiser-Marquardt. Fourier transform infrared spectroscopy (FTIR), differential scanning calorimetry (DSC), accelerating rate calorimetry (ARC), ion chromatography (IC), inductively coupled plasma (ICP) spectrometry, thermogravimetric analysis (TGA) coupled to FTIR (TGA/FTIR), and mechanical impact testing were used to qualitatively and quantitatively characterize the chemical, thermal, and ignition properties of FORP. These studies showed that the composition of FORP is variable but falls within a limited range of compositions that depends on the fuel loxidizer ratio at the time of formation, composition of the post-formation atmosphere (reducing or oxidizing), and reaction or postreaction temperature. A typical composition contains methylhydrazinium nitrate (MMHN), ammonium nitrate (AN), methylammonium nitrate (MAN), and trace amounts of hydrazinium nitrate and 1,1-dimethylhydrazinium nitrate. The thermal decomposition

  5. In-pile tests of HTGR fuel particles and fuel elements

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Kolesov, V.S.; Deryugin, A.I.

    1985-01-01

    Main types of in-pile tests for specimen tightness control at the initial step, research of fuel particle radiation stability and also study of fission product release from fuel elements during irradiation are described in this paper. Schemes and main characteristics of devices used for these tests are also given. Principal results of fission gas product release measurements satisfying HTGR demands are illustrated on the example of fuel elements, manufactured by powder metallurgy methods and having TRISO fuel particles on high temperature pyrocarbon and silicon carbide base. (author)

  6. Preparation results for lifetime test of conversion LEU fuel in plutonium production reactors

    International Nuclear Information System (INIS)

    Vatulin, A.; Stetskiy, Yu.; Kukharkin, N.; Kalougin, A.; Gavrilov, P.; Ivanov, A.

    1999-01-01

    The program of converting Russian production reactors for the purpose to stop their plutonium fabrication is currently in progress. The program also provides for operation of these reactors under the conversion mode with using of low-enriched fuel (LEU). LEU fuel elements were developed and activities related to their preparation for reactor tests were carried out. (author)

  7. Verification tests for CANDU advanced fuel

    International Nuclear Information System (INIS)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D.

    1997-07-01

    For the development of a CANDU advanced fuel, the CANFLEX-NU fuel bundles were tested under reactor operating conditions at the CANDU-Hot test loop. This report describes test results and test methods in the performance verification tests for the CANFLEX-NU bundle design. The main items described in the report are as follows. - Fuel bundle cross-flow test - Endurance fretting/vibration test - Freon CHF test - Production of technical document. (author). 25 refs., 45 tabs., 46 figs

  8. LOFT/LP-FP-2, Loss of Fluid Test, Fission Product Release from Fuel

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The eighth OECD LOFT experiment was conducted on 7 March 1985. It was the second of the two experiments to be performed in the LOFT facility with intentional release of fission products. Its principal objectives were to determine the fission product release from the fuel during a severe fuel damage scenario and the subsequent transport of these fission products in a predominantly vapor/aerosol environment. This was the largest severe fuel damage experiment ever conducted, and serves as an important benchmark between smaller scale tests and the TMI-2 accident. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  9. Testing plutonium fuel assembly production for fast-neutron reactors

    International Nuclear Information System (INIS)

    Nougues, B.; Benhamou, A.; Bertothy, G.; Lepetit, H.

    1975-01-01

    The main characteristics of plutonium fuel elements for fast breeder reactors justify specific test procedures and special techniques. The specific tests relating to the Pu content consist of Pu enrichment and distribution tests, determination of the O/M ratio and external contamination tests. The specific tests performed on fuel configuration are: testing of sintered pellet diameter, testing of pin welding and checking of internal assmbly [fr

  10. Fuel morphology effects on fission product release

    International Nuclear Information System (INIS)

    Osetek, D.J.; Hartwell, J.K.; Cronenberg, A.W.

    1986-01-01

    Results are presented of fission product release behavior observed during four severe fuel damage tests on bundles of UO 2 fuel rods. Transient temperatures up to fuel melting were obtained in the tests that included both rapid and slow cooldown, low and high (36 GWd/t) burnup fuel and the addition of Ag-In-Cd control rods. Release fractions of major fission product species and release rates of noble gas species are reported. Significant differences in release behavior are discussed between heatup and cooldown periods, low and high burnup fuel and long- and short-lived fission products. Explanations for the observed differences are offered that relate fuel morphology changes to the releases

  11. Energy deposition in NSRR test fuels

    International Nuclear Information System (INIS)

    Ohnishi, Nobuaki; Tanzawa, Sadamitsu; Tanzawa, Tomio; Kitano, Teruaki; Okazaki, Shuji

    1978-02-01

    Interpretation of fuel performance data collected during inpile testing in the NSRR requires a knowledge of the energy deposition or enthalpy increase in each sample tested. The report describes the results of absolute measurement of fission products and contents of uranium in irradiated test fuels which were performed to determine the energy deposition. (auth.)

  12. Fuel temperature prediction during high burnup HTGR fuel irradiation test. US-JAERI irradiation test for HTGR fuel

    International Nuclear Information System (INIS)

    Sawa, Kazuhiro; Fukuda, Kousaku; Acharya, R.

    1995-01-01

    This report describes the preirradiation thermal analysis of the HRB-22 capsule designed for an irradiation test in a removable beryllium position of the High Flux Isotope Reactor(HFIR) at Oak Ridge National Laboratory. This test is being carried out under Annex 2 of the Arrangement between the U.S. Department of Energy and the Japan Atomic Energy Research Institute on Cooperation in Research and Development regarding High-Temperature Gas-cooled Reactors. The fuel used in the test is an advanced type. The advanced fuel was designed aiming at burnup of about 10%FIMA(% fissions per initial metallic atom) which was higher than that of the first charge fuel for the High Temperature Engineering Test Reactor(HTTR) and was produced in Japan. CACA-2, a heavy isotope and fission product concentration calculational code for experimental irradiation capsules, was used to determine time-dependent fission power for the fuel compacts. The Heat Engineering and Transfer in Nine Geometries(HEATING) code was used to solve the steady-state heat conduction problem. The diameters of the graphite fuel body, which contains the fuel compacts, and of the primary pressure vessel were determined such that the requirements of running the fuel compacts at an average temperature less than 1250degC and of not exceeding a maximum fuel temperature of 1350degC were met throughout the four cycles of irradiation. The detail design of the capsule was carried out based on this analysis. (author)

  13. Pre-irradiation testing of experimental fuel elements

    International Nuclear Information System (INIS)

    Basova, B.G.; Davydov, E.F.; Dvoretskij, V.G.; Ivanov, V.B.; Syuzev, V.N.; Timofeev, G.A.; Tsykanov, V.A.

    1979-01-01

    The problems of testing of experimental fuel elements of nuclear reactors on the basis of complex accountancy of the factors defining operating capacity of the fuel elements are considered. The classification of the parameters under control and the methods of initial technological testing, including testing of the fuel product, cladding and fished fuel element, is given. The requirements to the apparatus used for complex testing are formulated. One of the possible variants of representation of the information obtained in the form of the input certificate of a single fuel element under study is proposed. The processing flowsheet of the gathered information using the computer is given. The approach under consideration is a methodological basis of investigation of fuel element operating life at the testing stage of the experimental fuel elements

  14. Interim report spent nuclear fuel retrieval system fuel handling development testing

    Energy Technology Data Exchange (ETDEWEB)

    Ketner, G.L.; Meeuwsen, P.V.; Potter, J.D.; Smalley, J.T.; Baker, C.P.; Jaquish, W.R.

    1997-06-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project at the Hanford Site. The project will retrieve spent nuclear fuel, clean and remove fuel from canisters, repackage fuel into baskets, and load fuel into a multi-canister overpack (MCO) for vacuum drying and interim dry storage. The FRS is required to retrieve basin fuel canisters, clean fuel elements sufficiently of uranium corrosion products (or sludge), empty fuel from canisters, sort debris and scrap from whole elements, and repackage fuel in baskets in preparation for MCO loading. The purpose of fuel handling development testing was to examine the systems ability to accomplish mission activities, optimization of equipment layouts for initial process definition, identification of special needs/tools, verification of required design changes to support performance specification development, and validation of estimated activity times/throughput. The test program was set up to accomplish this purpose through cold development testing using simulated and prototype equipment; cold demonstration testing using vendor expertise and systems; and graphical computer modeling to confirm feasibility and throughput. To test the fuel handling process, a test mockup that represented the process table was fabricated and installed. The test mockup included a Schilling HV series manipulator that was prototypic of the Schilling Hydra manipulator. The process table mockup included the tipping station, sorting area, disassembly and inspection zones, fuel staging areas, and basket loading stations. The test results clearly indicate that the Schilling Hydra arm cannot effectively perform the fuel handling tasks required unless it is attached to some device that can impart vertical translation, azimuth rotation, and X-Y translation. Other test results indicate the importance of camera locations and capabilities, and of the jaw and end effector tool design. 5 refs., 35 figs., 3 tabs.

  15. Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

    Energy Technology Data Exchange (ETDEWEB)

    Paul A Demkowicz; Paul Demkowicz; David V Laug

    2010-10-01

    Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

  16. Spent nuclear fuel retrieval system fuel handling development testing. Final report

    International Nuclear Information System (INIS)

    Jackson, D.R.; Meeuwsen, P.V.

    1997-09-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin, clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge), remove the contents from the canisters and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. This report describes fuel handling development testing performed from May 1, 1997 through the end of August 1997. Testing during this period was mainly focused on performance of a Schilling Robotic Systems' Conan manipulator used to simulate a custom designed version, labeled Konan, being fabricated for K-Basin deployment. In addition to the manipulator, the camera viewing system, process table layout, and fuel handling processes were evaluated. The Conan test manipulator was installed and fully functional for testing in early 1997. Formal testing began May 1. The purposes of fuel handling development testing were to provide proof of concept and criteria, optimize equipment layout, initialize the process definition, and identify special needs/tools and required design changes to support development of the performance specification. The test program was set up to accomplish these objectives through cold (non-radiological) development testing using simulated and prototype equipment

  17. Direct fuel cell product design improvement

    Energy Technology Data Exchange (ETDEWEB)

    Maru, H.C.; Farooque, M. [Energy Research Corp., Danbury, CT (United States)

    1996-12-31

    Significant milestones have been attained towards the technology development field testing and commercialization of direct fuel cell power plant since the 1994 Fuel Cell Seminar. Under a 5-year cooperative agreement with the Department of Energy signed in December 1994, Energy Research Corporation (ERC) has been developing the design for a MW-scale direct fuel cell power plant with input from previous technology efforts and the Santa Clara Demonstration Project. The effort encompasses product definition in consultation with the Fuel Cell Commercialization Group, potential customers, as well as extensive system design and packaging. Manufacturing process improvements, test facility construction, cell component scale up, performance and endurance improvements, stack engineering, and critical balance-of-plant development are also addressed. Major emphasis of this product design improvement project is on increased efficiency, compactness and cost reduction to establish a competitive place in the market. A 2.85 MW power plant with an efficiency of 58% and a footprint of 420 m{sup 2} has been designed. Component and subsystem testing is being conducted at various levels. Planning and preparation for verification of a full size prototype unit are in progress. This paper presents the results obtained since the last fuel cell seminar.

  18. Fission products and nuclear fuel behaviour under severe accident conditions part 1: Main lessons learnt from the first VERDON test

    Science.gov (United States)

    Pontillon, Y.; Geiger, E.; Le Gall, C.; Bernard, S.; Gallais-During, A.; Malgouyres, P. P.; Hanus, E.; Ducros, G.

    2017-11-01

    This paper describes the first VERDON test performed at the end of September 2011 with special emphasis on the behaviour of fission products (FP) and actinides during the accidental sequence itself. Two other papers discuss in detail the post-test examination results (SEM, EPMA and SIMS) of the VERDON-1 sample. The first VERDON test was devoted to studying UO2 fuel behaviour and fission product releases under reducing conditions at very high temperature (∼2883 K), which was able to confirm the very good performance of the VERDON loop. The fuel sample did not lose its integrity during this test. According to the FP behaviour measured by the online gamma station (fuel sight), the general classification of the FP in relation to their released fraction is very accurate, and the burn-up effect on the release rate is clearly highlighted.

  19. Fission product release from TRIGA-LEU reactor fuels

    International Nuclear Information System (INIS)

    Baldwin, N.L.; Foushee, F.C.; Greenwood, J.S.

    1980-01-01

    Due to present international concerns over nuclear proliferation, TRIGA reactor fuels will utilize only low-enriched uranium (LEU) (enrichment <20%). This requires increased total uranium loading per unit volume of fuel in order to maintain the appropriate fissile loading. Tests were conducted to determine the fractional release of gaseous and metallic fission products from typical uranium-zirconium hydride TRIGA fuels containing up to 45 wt-% uranium. These tests, performed in late 1977 and early 1978, were similar to those conducted earlier on TRIGA fuels with 8.5 wt-% U. Fission gas release measurements were made on prototypic specimens from room temperature to 1100 deg. C in the TRIGA King Furnace Facility. The fuel specimens were irradiated in the TRIGA reactor at a low power level. The fractional releases of the gaseous nuclides of krypton and xenon were measured under steady-state operating conditions. Clean helium was used to sweep the fission gases released during irradiation from the furnace into a standard gas collection trap for gamma counting. The results of these tests on TRIGA-LEU fuel agree well with data from the similar, earlier tests on TRIGA fuel. The correlation used to calculate the release of fission products from 8.5 wt-% U TRIGA fuel applies equally well for U contents up to 45 wt-%. (author)

  20. The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Laug, David V.; Scates, Dawn M.; Reber, Edward L.; Roybal, Lyle G.; Walter, John B.; Harp, Jason M. [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Morris, Robert N. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831 (United States)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer A system has been developed for safety testing of irradiated coated particle fuel. Black-Right-Pointing-Pointer FACS system is designed to facilitate remote operation in a shielded hot cell. Black-Right-Pointing-Pointer System will measure release of fission gases and condensable fission products. Black-Right-Pointing-Pointer Fuel performance can be evaluated at temperatures as high as 2000 Degree-Sign C in flowing helium. - Abstract: The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 Degree-Sign C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated

  1. Nuclear fuel production

    International Nuclear Information System (INIS)

    Randol, A.G.

    1985-01-01

    The production of new fuel for a power plant reactor and its disposition following discharge from the power plant is usually referred to as the ''nuclear fuel cycle.'' The processing of fuel is cyclic in nature since sometime during a power plant's operation old or ''depleted'' fuel must be removed and new fuel inserted. For light water reactors this step typically occurs once every 12-18 months. Since the time required for mining of the raw ore to recovery of reusable fuel materials from discharged materials can span up to 8 years, the management of fuel to assure continuous power plant operation requires simultaneous handling of various aspects of several fuel cycles, for example, material is being mined for fuel to be inserted in a power plant 2 years into the future at the same time fuel is being reprocessed from a discharge 5 years prior. Important aspects of each step in the fuel production process are discussed

  2. Fission product retention in TRISO coated UO2 particle fuels subjected to HTR simulated core heating tests

    International Nuclear Information System (INIS)

    Baldwin, C.A.; Kania, M.J.

    1991-01-01

    Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbounded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600 deg. C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800 deg. C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800 deg. C and above may exist. (author). 6 refs, 6 figs, 4 tabs

  3. Fission product retention in TRISCO coated UO2 particle fuels subjected to HTR simulated core heating tests

    International Nuclear Information System (INIS)

    Baldwin, C.A.; Kania, M.J.

    1990-11-01

    Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbonded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600 degree C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800 degree C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800 degree C and above may exist. 6 refs., 6 figs., 4 tabs

  4. Spent fuel sabotage aerosol test program :FY 2005-06 testing and aerosol data summary

    International Nuclear Information System (INIS)

    Gregson, Michael Warren; Brockmann, John E.; Nolte, O.; Loiseau, O.; Koch, W.; Molecke, Martin Alan; Autrusson, Bruno; Pretzsch, Gunter Guido; Billone, M. C.; Lucero, Daniel A.; Burtseva, T.; Brucher, W; Steyskal, Michele D.

    2006-01-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program has been underway for several years. This program provides source-term data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. This document focuses on an updated description of the test program and test components for all work and plans made, or revised, primarily during FY 2005 and about the first two-thirds of FY 2006. It also serves as a program status report as of the end of May 2006. We provide details on the significant findings on aerosol results and observations from the recently completed Phase 2 surrogate material tests using cerium oxide ceramic pellets in test rodlets plus non-radioactive fission product dopants. Results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; status on determination of the spent fuel ratio, SFR (the ratio of respirable particles from real spent fuel/respirables from surrogate spent fuel, measured under closely matched test conditions, in a contained test chamber); and, measurements of enhanced volatile fission product species sorption onto respirable particles. We discuss progress and results for the first three, recently performed Phase 3 tests using depleted uranium oxide, DUO 2 , test rodlets. We will also review the status of preparations and the final Phase 4 tests in this program, using short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. These data plus testing results and design are tailored to support and guide, follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments

  5. Power Burst Facility severe-fuel-damage test program

    International Nuclear Information System (INIS)

    McCardell, R.K.; MacDonald, P.E.

    1982-01-01

    As a result of the Three Mile Island Unit 2 (TMI-2) accident, the United States Nuclear Regulatory Commission (USNRC) has initiated a severe fuel damage research program to investigate fuel rod and core response, and fission product and hydrogen release and transport during degraded core cooling accidents. This paper presents a discussion of the expected benefits of the PBF severe fuel damage tests to the nuclear industry, a description of the first five planned experiments, the results of pretest analysis performed to predict the fuel bundle heatup for the first two experiments, and a discussion of Phase II severe fuel damage experiments. Modifications to the fission product detection system envisioned for the later experiments are also described

  6. Power Burst Facility Severe Fuel Damage test series

    International Nuclear Information System (INIS)

    Buescher, B.J.; Osetek, D.J.; Ploger, S.A.

    1982-01-01

    The Severe Fuel Damage (SFD) tests planned for the Power Burst Facility (PBF) are described. Bundles containing 32 zircaloy-clad, PWR-type fuel rods will be subjected to severe overheating transients in a high-pressure, superheated-steam environment. Cladding temperatures are expected to reach 2400 0 K, resulting in cladding ballooning and rupture, severe cladding oxidation, cladding melting, fuel dissolution, fuel rod fragmentation, and possibly, rubble bed formation. An experiment effluent collection system is being installed and the PBF fission product monitoring system is being upgraded to meet the special requirements of the SFD tests. Scoping calculations were performed to evaluate performance of the SFD test design and to establish operational requirements for the PBF loop

  7. Pyro-electrochemical reprocessing of irradiated MOX fast reactor fuel, testing of the reprocessing process with direct MOX fuel production

    Energy Technology Data Exchange (ETDEWEB)

    Kormilitzyn, M.V.; Vavilov, S.K.; Bychkov, A.V.; Skiba, O.V.; Chistyakov, V.M.; Tselichshev, I.V

    2000-07-01

    One of the advanced technologies for fast reactor fuel recycle is pyro-electrochemical molten salt technology. In 1998 we began to study the next phase of the irradiated oxide fuel reprocessing new process MOX {yields} MOX. This process involves the following steps: - Dissolution of irradiated fuel in molten alkaline metal chlorides, - Purification of melt from fission products that are co-deposited with uranium and plutonium oxides, - Electrochemical co-deposition of uranium and plutonium oxides under the controlled cathode potential, - Production of granulated MOX (crushing,salt separation and sizing), and - Purification of melt from fission products by phosphate precipitation. In 1998 a series of experiments were prepared and carried out in order to validate this process. It was shown that the proposed reprocessing flowsheet of irradiated MOX fuel verified the feasibility of its decontamination from most of its fission products (rare earths, cesium) and minor-actinides (americium, curium)

  8. Pyro-electrochemical reprocessing of irradiated MOX fast reactor fuel, testing of the reprocessing process with direct MOX fuel production

    International Nuclear Information System (INIS)

    Kormilitzyn, M.V.; Vavilov, S.K.; Bychkov, A.V.; Skiba, O.V.; Chistyakov, V.M.; Tselichshev, I.V.

    2000-01-01

    One of the advanced technologies for fast reactor fuel recycle is pyro-electrochemical molten salt technology. In 1998 we began to study the next phase of the irradiated oxide fuel reprocessing new process MOX → MOX. This process involves the following steps: - Dissolution of irradiated fuel in molten alkaline metal chlorides, - Purification of melt from fission products that are co-deposited with uranium and plutonium oxides, - Electrochemical co-deposition of uranium and plutonium oxides under the controlled cathode potential, - Production of granulated MOX (crushing,salt separation and sizing), and - Purification of melt from fission products by phosphate precipitation. In 1998 a series of experiments were prepared and carried out in order to validate this process. It was shown that the proposed reprocessing flowsheet of irradiated MOX fuel verified the feasibility of its decontamination from most of its fission products (rare earths, cesium) and minor-actinides (americium, curium)

  9. Product Conversion: The Link between Separations and Fuel Fabrication

    International Nuclear Information System (INIS)

    Felker, L.K.; Vedder, R.J.; Walker, E.A.; Collins, E.D.

    2008-01-01

    Several chemical processing flowsheets are under development for the separation and isolation of the actinide, lanthanide, and fission product streams in spent nuclear fuel. The conversion of these product streams to solid forms, typically oxides, is desired for waste disposition and recycle of product fractions back into transmutation fuels or targets. The modified direct denitration (MDD) process developed at Oak Ridge National Laboratory (ORNL) in the 1980's offers significant advantages for the conversion of the spent fuel products to powder form suitable for direct fabrication into recycle fuels. A glove-box-contained MDD system and a fume-hood-contained system have been assembled at ORNL for the purposes of testing the co-conversion of uranium and mixed-actinide products. The current activities are focused on the conversion of the first products from the processing of spent nuclear fuel in the Coupled End-to-End Demonstration currently being conducted at ORNL. (authors)

  10. Comparative Study of Laboratory-Scale and Prototypic Production-Scale Fuel Fabrication Processes and Product Characteristics

    International Nuclear Information System (INIS)

    Marshall, Douglas W.

    2014-01-01

    An objective of the High Temperature Gas Reactor fuel development and qualification program for the United States Department of Energy has been to qualify fuel fabricated in prototypic production-scale equipment. The quality and characteristics of the tristructural isotropic (TRISO) coatings on fuel kernels are influenced by the equipment scale and processing parameters. The standard deviations of some TRISO layer characteristics were diminished while others have become more significant in the larger processing equipment. The impact on statistical variability of the processes and the products, as equipment was scaled, are discussed. The prototypic production-scale processes produce test fuels meeting all fuel quality specifications. (author)

  11. Irradiation testing of coated particle fuel at Hanaro

    International Nuclear Information System (INIS)

    Goo Kim, Bong; Sung Cho, Moo; Kim, Yong Wan

    2014-01-01

    TRISO-coated particle fuel is developing to support development of VHTR in Korea. From August 2013, the first irradiation testing of coated particle fuel was begun to demonstrate and qualify TRISO fuel for use in VHTR in the HANARO at KAERI. This experiment is currently undergoing under the atmosphere of a mixed inert gas without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one contains nine fuel compacts and the other five compacts and eight graphite specimens. Each compact has 263 coated particles. After a peak burn-up of about 4 at% and a peak fast neutron fluence of about 1.7 x 10 21 n/cm 2 , PIE will be carried out at KAERI's Irradiated Material Examination Facility. This paper is described characteristics of coated particle fuel, the design of test rod and irradiation device for coated particle fuel, discusses the technical results for irradiation testing at HANARO. (authors)

  12. Results of tests with open fuel in KNK II

    International Nuclear Information System (INIS)

    Schmitz, G.

    1987-03-01

    For the operation of Liquid Metal Cooled Fast Breeder Reactors with cladding failures the consequences of increased contamination by fission products and fuel and the possibility of failure propagation to adjacent fuel pins due to fuel swelling have to be envisaged. To clarify some of these problems a KNK II test program involving open fuel was defined with the first experiments of this program being performed between October 1981 and May 1984. After the description of the test equipment and of the test program, the results will be presented on delayed neutron measurements, fission gas measurements and post irradiation examinations. The report will conclude with a discussion of the results [de

  13. Fuel and fission product behaviour in early phases of a severe accident. Part I: Experimental results of the PHEBUS FPT2 test

    Energy Technology Data Exchange (ETDEWEB)

    Barrachin, M., E-mail: marc.barrachin@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); Gavillet, D. [Paul Scherrer Institute, Würenlingen and Villigen, CH-5232 Villigen PSI (Switzerland); Dubourg, R. [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); De Bremaecker, A. [Institute for Nuclear Materials Sciences, SCK-CEN, Boeretang 200, B-2400 Mol (Belgium)

    2014-10-15

    One objective of the FPT2 test of the PHEBUS FP Program was to study the degradation of an irradiated UO{sub 2} fuel test section and the fission product behaviour under conditions of low steam flow. The results of the post-irradiation examinations (PIE) at the upper levels (823 and 900 mm) of the 1-m long test section are presented in this paper. Material interactions leading to local corium formation were identified: firstly between fuel and Zircaloy-4 cladding, notably at 823 mm, where the cladding melting temperature was reached, and secondly between fuel and stainless steel oxides. Regarding fission products, molybdenum left so-called metallic precipitates mainly composed of ruthenium. Xenon and caesium behave similarly whereas barium and molybdenum often seems to be associated in precipitates.

  14. Production process and quality control for the HTTR fuel

    International Nuclear Information System (INIS)

    Yoshimuta, S.; Suzuki, N.; Kaneko, M.; Fukuda, K.

    1991-01-01

    Development of the production and inspection technology for High Temperature Engineering Test Reactor (HTTR) fuel has been carried out by cooperative work between Japan Atomic Energy Research Institute (JAERI) and Nuclear Fuel Industries, Ltd (NFI). The performance and the quality level of the developed fuel are well established to meet the design requirements of the HTTR. For the commercial scale production of the fuel, statistical quality control and quality assurance must be carefully considered in order to assure the safety of the HTTR. It is also important to produce the fuel under well controlled process condition. To meet these requirements in the production of the HTTR fuel, a new production process and quality control system is to be introduced in the new facilities. The main feature of the system is a computer integrated control system. Process control data at each production stage of products and semi-products are all gathered by terminal computers and processed by a host computer. The processed information is effectively used for the production, quality and accountancy control. With the aid of this system, all the products will be easily traceable from starting materials to final stages and the statistical evaluation of the quality of products becomes more reliable. (author). 8 figs

  15. Engineering yeast metabolism for production of fuels and chemicals

    DEFF Research Database (Denmark)

    Nielsen, Jens

    2016-01-01

    faster development of metabolically engineered strains that can be used for production of fuels and chemicals. The yeast Saccharomyces cerevisiae is widely used for production of fuels, chemicals, pharmaceuticals and materials. Through metabolic engineering of this yeast a number of novel industrial...... as for metabolic design. In this lecture it will be demonstrated how the Design-Build-Test cycle of metabolic engineering has allowed for development of yeast cell factories for production of a range of different fuels and chemicals. Some examples of different technologies will be presented together with examples......Metabolic engineering relies on the Design-Build-Test cycle. This cycle includes technologies like mathematical modeling of metabolism, genome editing and advanced tools for phenotypic characterization. In recent years there have been advances in several of these technologies, which has enabled...

  16. Full-length high-temperature severe fuel damage test No. 5

    International Nuclear Information System (INIS)

    Lanning, D.D.; Lombardo, N.J.; Hensley, W.K.; Fitzsimmons, D.E.; Panisko, F.E.; Hartwell, J.K.

    1993-09-01

    This report describes and presents data from a severe fuel damage test that was conducted in the National Research Universal (NRU) reactor at Chalk River Nuclear Laboratories (CRNL), Ontario, Canada. The test, designated FLHT-5, was the fourth in a series of full-length high-temperature (FLHT) tests on light-water reactor fuel. The tests were designed and performed by staff from the US Department of Energy's Pacific Northwest Laboratory (PNL), operated by Battelle Memorial Institute. The test operation and test results are described in this report. The fuel bundle in the FLHT-5 experiment included 10 unirradiated full-length pressurized-water reactor (PWR) rods, 1 irradiated PWR rod and 1 dummy gamma thermometer. The fuel rods were subjected to a very low coolant flow while operating at low fission power. This caused coolant boilaway, rod dryout and overheating to temperatures above 2600 K, severe fuel rod damage, hydrogen generation, and fission product release. The test assembly and its effluent path were extensively instrumented to record temperatures, pressures, flow rates, hydrogen evolution, and fission product release during the boilaway/heatup transient. Post-test gamma scanning of the upper plenum indicated significant iodine and cesium release and deposition. Both stack gas activity and on-line gamma spectrometer data indicated significant (∼50%) release of noble fission gases. Post-test visual examination of one side of the fuel bundle revealed no massive relocation and flow blockage; however, rundown of molten cladding was evident

  17. HRB-22 capsule irradiation test for HTGR fuel. JAERI/USDOE collaborative irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Minato, Kazuo; Sawa, Kazuhiro; Fukuda, Kousaku [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; and others

    1998-03-01

    As a JAERI/USDOE collaborative irradiation test for high-temperature gas-cooled reactor fuel, JAERI fuel compacts were irradiated in the HRB-22 irradiation capsule in the High Flux Isotope Reactor at the Oak Ridge National Laboratory (ORNL). Postirradiation examinations also were performed at ORNL. This report describes 1) the preirradiation characterization of the irradiation samples of annular-shaped fuel compacts containing the Triso-coated fuel particles, 2) the irradiation conditions and fission gas releases during the irradiation to measure the performance of the coated particle fuel, 3) the postirradiation examinations of the disassembled capsule involving visual inspection, metrology, ceramography and gamma-ray spectrometry of the samples, and 4) the accident condition tests on the irradiated fuels at 1600 to 1800degC to obtain information about fuel performance and fission product release behavior under accident conditions. (author)

  18. Basic research on high-uranium density fuels for research and test reactors

    International Nuclear Information System (INIS)

    Ugajin, M.; Itoh, A.; Akabori, M.

    1992-01-01

    High-uranium density fuels, uranium silicides (U 3 Si 2 , U 3 Si) and U 6 Me-type uranium alloys (Me = Fe, Mn, Ni), were prepared and examined metallurgically as low-enriched uranium (LEU) fuels for research and test reactors. Miniature aluminum-dispersion plate-type fuel (miniplate) and aluminum-clad disk-type fuel specimens were fabricated and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Fuel-aluminum compatibility tests were conducted to elucidate the extent of reaction and to identify reaction products. The relative stability of the fuels in an aluminum matrix was established at 350degC or above. Experiments were also performed to predict the chemical form of the solid fission-products in the uranium silicide (U 3 Si 2 ) simulating a high burnup anticipated for reactor service. (author)

  19. Status of fuel irradiation tests in HANARO

    International Nuclear Information System (INIS)

    Kim, Hark Rho; Lee, Choong Sung; Lee, Kye Hong; Jun, Byung Jin; Lee, Ji Bok

    1999-01-01

    Since 1996 after finishing the long-term operational test, HANARO (High-Flux Advanced Neutron Application Reactor) has been extensively used for material irradiation tests, beam application research, radioisotope production and neutron activation analysis. This paper presents the fuel irradiation test activities which are now conducted or have been finished in HANARO. KAERI developed LEU fuel using an atomization method for the research reactors. Using this LEU, we have set up and conducted three irradiation programs: (1) medium power irradiation test using a short-length mini-assembly made of 3.15 gU/cc U 3 Si, (2) high power irradiation tests using full-length test assemblies made of 3.15 gU/cc U 3 Si, and (3) irradiation test using a short-length mini-plate made of 4.8 gU/cc U 3 Si 2 . DUPIC (Direct Use of spent PWR fuels in CANDU Reactors) simulation fuel pellets, of which compositions are very similar to DUPIC pellets to keep the similarity in the thermo-mechanical property, were developed. Three mini-elements including 5 pellets each were installed in a capsule. This capsule has been irradiated for 2 months and unloaded from the HANARO core at the end of September 1999. Another very important test is the HANARO fuel qualification program at high power, which is required to resolve the licensing issue. This test is imposed on the HANARO operation license due to insufficient test data under high power environment. To resolve this licensing issue, we have been carrying out the required irradiation tests and PIE (Post-irradiation Examination) tests. Through this program, it is believed that the resolution of the licensing issue is achieved. In addition to these programs, several fuel test plans are under way. Through these vigorous activities of fuel irradiation test programs, HANARO is sure to significantly contribute to the national nuclear R and D programs. (author)

  20. Fuel Production from Seawater and Fuel Cells Using Seawater.

    Science.gov (United States)

    Fukuzumi, Shunichi; Lee, Yong-Min; Nam, Wonwoo

    2017-11-23

    Seawater is the most abundant resource on our planet and fuel production from seawater has the notable advantage that it would not compete with growing demands for pure water. This Review focuses on the production of fuels from seawater and their direct use in fuel cells. Electrolysis of seawater under appropriate conditions affords hydrogen and dioxygen with 100 % faradaic efficiency without oxidation of chloride. Photoelectrocatalytic production of hydrogen from seawater provides a promising way to produce hydrogen with low cost and high efficiency. Microbial solar cells (MSCs) that use biofilms produced in seawater can generate electricity from sunlight without additional fuel because the products of photosynthesis can be utilized as electrode reactants, whereas the electrode products can be utilized as photosynthetic reactants. Another important source for hydrogen is hydrogen sulfide, which is abundantly found in Black Sea deep water. Hydrogen produced by electrolysis of Black Sea deep water can also be used in hydrogen fuel cells. Production of a fuel and its direct use in a fuel cell has been made possible for the first time by a combination of photocatalytic production of hydrogen peroxide from seawater and dioxygen in the air and its direct use in one-compartment hydrogen peroxide fuel cells to obtain electric power. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  1. Strategies for fuel cell product development. Developing fuel cell products in the technology supply chain

    International Nuclear Information System (INIS)

    Hellman, H.L.

    2004-01-01

    Due to the high cost of research and development and the broad spectrum of knowledge and competences required to develop fuel cell products, many product-developing firms outsource fuel cell technology, either partly or completely. This article addresses the inter-firm process of fuel cell product development from an Industrial Design Engineering perspective. The fuel cell product development can currently be characterised by a high degree of economic and technical uncertainty. Regarding the technology uncertainty: product-developing firms are more often then not unfamiliar with fuel cell technology technology. Yet there is a high interface complexity between the technology supplied and the product in which it is to be incorporated. In this paper the information exchange in three current fuel cell product development projects is analysed to determine the information required by a product designer to develop a fuel cell product. Technology transfer literature suggests that transfer effectiveness is greatest when the type of technology (technology uncertainty) and the type of relationship between the technology supplier and the recipient are carefully matched. In this line of thinking this paper proposes that the information required by a designer, determined by the design strategy and product/system volume, should be met by an appropriate level of communication interactivity with a technology specialist. (author)

  2. Bioenergy Research Programme. Yearbook 1994. Production of wood fuels

    International Nuclear Information System (INIS)

    Alakangas, E.

    1995-01-01

    BIOENERGIA Research Programme is one of energy technology programmes of the Finnish Ministry of Trade and Industry (in 1995 TEKES, Technology Development Center). The aim of Bioenergy Research Programme is to increase the use of economically profitable and environmentally sound bioenergy by improving the competitiveness of present peat and wood fuels. Research and development projects will also develop new economically competitive biofuels and new equipment and methods for production, handling and using of biofuels. The funding for 1994 was nearly 50 million FIM and projects numbered 60. The main goal of the production of wood fuels research area is to develop new production methods in order to decrease the production costs to the level of imported fuels. The total potential of the wood fuel use should be at least 1.0 million toe/a (5.5 million m 3 ). There were 27 projects in 1994 for research on wood fuel production. This part of the yearbook 1994 presents the main results of these projects. The wood reserves do not limit the obtainability of the target. Research and development work has, however, directed to development of equipment and research on wood fuels production chains. Many devices, designed for both separate and integrated production of wood fuels became ready or were becoming ready for prototyping, to be used for production tests. Results of the biomass harvesting and properties research were obtained for utilization in 1994. According to the results it is possible to obtain the desired targets both in integrated and separated production of wood fuels. (author)

  3. The behavior of fission products during nuclear rocket reactor tests

    International Nuclear Information System (INIS)

    Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

    1991-01-01

    Fission product release from nuclear rocket propulsion reactor fuel is an important consideration for nuclear rocket development and application. Fission product data from the last six reactors of the Rover program are collected in this paper to provide as basis for addressing development and testing issues. Fission product loss from the fuel will depend on fuel composition and reactor design and operating parameters. During ground testing, fission products can be contained downstream of the reactor. The last Rover reactor tested, the Nuclear Furnance, was mated to an effluent clean-up system that was effective in preventing the discharge of fission products into the atmosphere

  4. Alternative Fuels in Cement Production

    DEFF Research Database (Denmark)

    Larsen, Morten Boberg

    The substitution of alternative for fossil fuels in cement production has increased significantly in the last decade. Of these new alternative fuels, solid state fuels presently account for the largest part, and in particular, meat and bone meal, plastics and tyre derived fuels (TDF) accounted...... for the most significant alternative fuel energy contributors in the German cement industry. Solid alternative fuels are typically high in volatile content and they may differ significantly in physical and chemical properties compared to traditional solid fossil fuels. From the process point of view......, considering a modern kiln system for cement production, the use of alternative fuels mainly influences 1) kiln process stability (may accelerate build up of blockages preventing gas and/or solids flow), 2) cement clinker quality, 3) emissions, and 4) decreased production capacity. Kiln process stability...

  5. Out-pile Test of Double Cladding Fuel Rod Mockups for a Nuclear Fuel Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Jaemin; Park, Sungjae; Kang, Younghwan; Kim, Harkrho; Kim, Bonggoo; Kim, Youngki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-05-15

    An instrumented capsule for a nuclear fuel irradiation test has been developed to measure fuel characteristics, such as a fuel temperature, internal pressure of a fuel rod, a fuel pellet elongation and a neutron flux during an irradiation test at HANARO. In the future, nuclear fuel irradiation tests under a high temperature condition are expected from users. To prepare for this request, we have continued developing the technology for a high temperature nuclear fuel irradiation test at HANARO. The purpose of this paper is to verify the possibility that the temperature of a nuclear fuel can be controlled at a high temperature during an irradiation test. Therefore we designed and fabricated double cladding fuel rod mockups. And we performed out-pile tests using these mockups. The purposes of a out-pile test is to analyze an effect of a gap size, which is between an outer cladding and an inner cladding, on the temperature and the effect of a mixture ratio of helium gas and neon gas on the temperature. This paper presents the design and fabrication of double cladding fuel rod mockups and the results of the out-pile test.

  6. MOLTEN CARBONATE FUEL CELL PRODUCT DESIGN IMPROVEMENT

    Energy Technology Data Exchange (ETDEWEB)

    H.C. Maru; M. Farooque

    2003-03-01

    The program efforts are focused on technology and system optimization for cost reduction, commercial design development, and prototype system field trials. The program is designed to advance the carbonate fuel cell technology from full-size field test to the commercial design. FuelCell Energy, Inc. (FCE) is in the later stage of the multiyear program for development and verification of carbonate fuel cell based power plants supported by DOE/NETL with additional funding from DOD/DARPA and the FuelCell Energy team. FCE has scaled up the technology to full-size and developed DFC{reg_sign} stack and balance-of-plant (BOP) equipment technology to meet product requirements, and acquired high rate manufacturing capabilities to reduce cost. FCE has designed submegawatt (DFC300A) and megawatt (DFC1500 and DFC3000) class fuel cell products for commercialization of its DFC{reg_sign} technology. A significant progress was made during the reporting period. The reforming unit design was optimized using a three-dimensional stack simulation model. Thermal and flow uniformities of the oxidant-In flow in the stack module were improved using computational fluid dynamics based flow simulation model. The manufacturing capacity was increased. The submegawatt stack module overall cost was reduced by {approx}30% on a per kW basis. An integrated deoxidizer-prereformer design was tested successfully at submegawatt scale using fuels simulating digester gas, coal bed methane gas and peak shave (natural) gas.

  7. Preliminary results of the BTF-104 experiment: an in-reactor test of fuel behaviour and fission-product release and transport under LOCA/LOECC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Dickson, L W; Elder, P H; Devaal, J W; Irish, J D; Yamazaki, A R [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1996-12-31

    The BTF-104 experiment is one of a series of in-reactor tests being performed to measure fuel behaviour and fission-product release from nuclear fuel subjected to accident conditions. The primary objective of the BTF-104 experiment was to measure fission-product releases from a CANDU-sized fuel element under combined Loss-of-Coolant Accident (LOCA) and Loss-of-Emergency-Core-Cooling (LOECC) conditions at an average fuel temperature of about 1550 deg C. The preliminary results of the BTF-104 experiment are presented in this paper. (author). 6 refs., 12 figs.

  8. Development and testing of the EDF-2 reactor fuel element

    International Nuclear Information System (INIS)

    Delpeyroux, P.

    1964-01-01

    This technical report reviews the work which has been necessary for defining the EDF-2 fuel element. After giving briefly the EDF-2 reactor characteristics and the preliminary choice of parameters which made it possible to draw up a draft plan for the fuel element, the authors consider the research proper: - Uranium studies: tests on the passage into the β phase of an internal crown of a tube, bending of the tube under the effect of a localized force, welding of the end-pellets and testing for leaks. The resistance of the tube to crushing and of the pellets to yielding under the external pressure have been studied in detail in another CEA report. - Can studies: conditions of production and leak proof testing of the can, resistance of the fins to creep due to the effect of the gas flow. - Studies of the extremities of the element: creep under compression and welding of the plugs to the can. - Cartridge studies: determination of the characteristics of the can fuel fixing grooves and of the canning conditions, verification of the resistance of the fuel element to thermal cycling, determination of the temperature drop at the can-fuel interface dealt with in more detail in another CEA report. - Studies of the whole assembly: this work which concerns the graphite jacket, the support and the cartridge vibrations has been carried out by the Mechanical and Thermal Study Service (Mechanics Section). In this field the Fuel Element Study Section has investigated the behaviour of the centering devices in a gas current. The outcome of this research is the defining of the plan of the element the production process and the production specifications. The validity of ail these out-of-pile tests will be confirmed by the in-pile tests already under way and by irradiation of the elements in the EDF-2 reactor itself. In conclusion the programme is given for improving the fuel element and for defining the fuel element for the second charge. (authors) [fr

  9. Irradiation Testing of TRISO-Coated Particle Fuel in Korea

    International Nuclear Information System (INIS)

    Kim, Bong Goo; Yeo, Sunghwan; Jeong, Kyung-Chai; Eom, Sung-Ho; Kim, Yeon-Ku; Kim, Woong Ki; Lee, Young Woo; Cho, Moon Sung; Kim, Yong Wan

    2014-01-01

    In Korea, coated particle fuel is being developed to support development of a VHTR. At the end of March 2014, the first irradiation test in HANARO at KAERI to demonstrate and qualify TRISO-coated particle fuel for use in a VHTR was terminated. This experiment was conducted in an inert gas atmosphere without on-line temperature monitoring and control, or on-line fission product monitoring of the sweep gas. The irradiation device contained two test rods, one has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The duration of irradiation testing at HANARO was about 135 full power days from last August 2013. The maximum average power per particle was about 165 mW/particle. The calculated peak burnup of the TRISO-coated fuel was a little less than 4 atom percent. Post-irradiation examination is being carried out at KAERI’s Irradiated Material Examination Facility beginning in September of 2014. This paper describes characteristics of coated particle fuel, the design of the test rod and irradiation device for this coated particle fuel, and discusses the technical results of irradiation testing at HANARO. (author)

  10. Fuel Cell Electric Vehicle Composite Data Products | Hydrogen and Fuel

    Science.gov (United States)

    Cells | NREL Vehicle Composite Data Products Fuel Cell Electric Vehicle Composite Data Products The following composite data products (CDPs) focus on current fuel cell electric vehicle evaluations Cell Operation Hour Groups CDP FCEV 39, 2/19/16 Comparison of Fuel Cell Stack Operation Hours and Miles

  11. Transient fuel and target performance testing for the HWR-NPR

    International Nuclear Information System (INIS)

    Jicha, J.J. Jr.

    1990-01-01

    This paper describes a five year program of fuel target transient performance testing and model development required for the safety assessment of the HWR new production reactor. Technical issues are described, focusing on fuel and target behavior during extremely low probability transients which can lead to fuel melting. Early work on these issues is reviewed. The program to meet remaining needs is described. Three major transient-testing activities are included: in-cell experiments on small samples of irradiated fuel and target, small-scale phenomenological experiments in the ACRR reactor, and limited-integral experiments in the TREAT reactor. A coordinated development of detailed fuel and target behavior models is also described

  12. Ultra-clean Fischer-Tropsch (F-T) Fuels Production and Demonstration Project

    Energy Technology Data Exchange (ETDEWEB)

    Stephen P. Bergin

    2006-06-30

    The objective of the DOE-NETL Fischer-Tropsch (F-T) Production and Demonstration Program was to produce and evaluate F-T fuel derived from domestic natural gas. The project had two primary phases: (1) fuel production of ultra-clean diesel transportation fuels from domestic fossil resources; and (2) demonstration and performance testing of these fuels in engines. The project also included a well-to-wheels economic analysis and a feasibility study of small-footprint F-T plants (SFPs) for remote locations such as rural Alaska. During the fuel production phase, ICRC partnered and cost-shared with Syntroleum Corporation to complete the mechanical design, construction, and operation of a modular SFP that converts natural gas, via F-T and hydro-processing reactions, into hydrogensaturated diesel fuel. Construction of the Tulsa, Oklahoma plant started in August 2002 and culminated in the production of over 100,000 gallons of F-T diesel fuel (S-2) through 2004, specifically for this project. That fuel formed the basis of extensive demonstrations and evaluations that followed. The ultra-clean F-T fuels produced had virtually no sulfur (less than 1 ppm) and were of the highest quality in terms of ignition quality, saturation content, backend volatility, etc. Lubricity concerns were investigated to verify that commercially available lubricity additive treatment would be adequate to protect fuel injection system components. In the fuel demonstration and testing phase, two separate bus fleets were utilized. The Washington DC Metropolitan Area Transit Authority (WMATA) and Denali National Park bus fleets were used because they represented nearly opposite ends of several spectra, including: climate, topography, engine load factor, mean distance between stops, and composition of normally used conventional diesel fuel. Fuel evaluations in addition to bus fleet demonstrations included: bus fleet emission measurements; F-T fuel cold weather performance; controlled engine dynamometer

  13. Fission product release from high gap-inventory LWR fuel under LOCA conditions

    International Nuclear Information System (INIS)

    Lorenz, R.A.; Collins, J.L.; Osborne, M.F.; Malinauskas, A.P.

    1980-01-01

    Fission product release tests were performed with light water reactor (LWR) fuel rod segments containing large amounts of cesium and iodine in the pellet-to-cladding gap space in order to check the validity of the previously published Source Term Model for this type of fuel. The model describes the release of fission product cesium and iodine from LWR fuel rods for controlled loss-of-coolant accident (LOCA) transients in the temperature range 500 to 1200 0 C. The basis for the model was test data obtained with simulated fuel rods and commercial fuel irradiated to high burnup but containing relatively small amounts of cesium and iodine in the pellet-to-cladding gap space

  14. Gas Test Loop Booster Fuel Hydraulic Testing

    International Nuclear Information System (INIS)

    Gas Test Loop Hydraulic Testing Staff

    2006-01-01

    The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3

  15. Gas Test Loop Booster Fuel Hydraulic Testing

    Energy Technology Data Exchange (ETDEWEB)

    Gas Test Loop Hydraulic Testing Staff

    2006-09-01

    The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3.

  16. HIGH-TEMPERATURE SAFETY TESTING OF IRRADIATED AGR-1 TRISO FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Stempien, John D.; Demkowicz, Paul A.; Reber, Edward L.; Chrisensen, Cad L.

    2016-11-01

    High-Temperature Safety Testing of Irradiated AGR-1 TRISO Fuel John D. Stempien, Paul A. Demkowicz, Edward L. Reber, and Cad L. Christensen Idaho National Laboratory, P.O. Box 1625 Idaho Falls, ID 83415, USA Corresponding Author: john.stempien@inl.gov, +1-208-526-8410 Two new safety tests of irradiated tristructural isotropic (TRISO) coated particle fuel have been completed in the Fuel Accident Condition Simulator (FACS) furnace at the Idaho National Laboratory (INL). In the first test, three fuel compacts from the first Advanced Gas Reactor irradiation experiment (AGR-1) were simultaneously heated in the FACS furnace. Prior to safety testing, each compact was irradiated in the Advanced Test Reactor to a burnup of approximately 15 % fissions per initial metal atom (FIMA), a fast fluence of 3×1025 n/m2 (E > 0.18 MeV), and a time-average volume-average (TAVA) irradiation temperature of about 1020 °C. In order to simulate a core-conduction cool-down event, a temperature-versus-time profile having a peak temperature of 1700 °C was programmed into the FACS furnace controllers. Gaseous fission products (i.e., Kr-85) were carried to the Fission Gas Monitoring System (FGMS) by a helium sweep gas and captured in cold traps featuring online gamma counting. By the end of the test, a total of 3.9% of an average particle’s inventory of Kr-85 was detected in the FGMS traps. Such a low Kr-85 activity indicates that no TRISO failures (failure of all three TRISO layers) occurred during the test. If released from the compacts, condensable fission products (e.g., Ag-110m, Cs-134, Cs-137, Eu-154, Eu-155, and Sr-90) were collected on condensation plates fitted to the end of the cold finger in the FACS furnace. These condensation plates were then analyzed for fission products. In the second test, five loose UCO fuel kernels, obtained from deconsolidated particles from an irradiated AGR-1 compact, were heated in the FACS furnace to a peak temperature of 1600 °C. This test had two

  17. Leak testing fuel stored in the ICPP fuel storage basin

    International Nuclear Information System (INIS)

    Lee, J.L.; Rhodes, D.W.

    1977-06-01

    Irradiated fuel to be processed at the Idaho Chemical Processing Plant is stored under water at the CPP-603 Fuel Storage Facility. Leakage of radionuclides through breaks in the cladding of some of the stored fuels contaminates the water with radionuclides resulting in radiation exposure to personnel during fuel handling operations and contamination of the shipping casks. A leak test vessel was fabricated to test individual fuel assemblies which were suspected to be leaking. The test equipment and procedures are described. Test results demonstrated that a leaking fuel element could be identified by this method; of the eleven fuel assemblies tested, six were estimated to be releasing greater than 0.5 Ci total radionuclides/day to the basin water

  18. Fission-product SiC reaction in HTGR fuel

    International Nuclear Information System (INIS)

    Montgomery, F.

    1981-01-01

    The primary barrier to release of fission product from any of the fuel types into the primary circuit of the HTGR are the coatings on the fuel particles. Both pyrolytic carbon and silicon carbide coatings are very effective in retaining fission gases under normal operating conditions. One of the possible performance limitations which has been observed in irradiation tests of TRISO fuel is chemical interaction of the SiC layer with fission products. This reaction reduces the thickness of the SiC layer in TRISO particles and can lead to release of fission products from the particles if the SiC layer is completely penetrated. The experimental section of this report describes the results of work at General Atomic concerning the reaction of fission products with silicon carbide. The discussion section describes data obtained by various laboratories and includes (1) a description of the fission products which have been found to react with SiC; (2) a description of the kinetics of silicon carbide thinning caused by fission product reaction during out-of-pile thermal gradient heating and the application of these kinetics to in-pile irradiation; and (3) a comparison of silicon carbide thinning in LEU and HEU fuels

  19. Measurement of fission product release during LWR fuel failure

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  20. Spent fuel sabotage test program, characterization of aerosol dispersal : interim final report

    International Nuclear Information System (INIS)

    Gregson, Michael Warren; Brockmann, John E.; Loiseau, Olivier; Klennert, Lindsay A.; Nolte, Oliver; Molecke, Martin Alan; Autrusson, Bruno A.; Koch, Wolfgang; Pretzsch, Gunter Guido; Brucher, Wenzel; Steyskal, Michele D.

    2008-01-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program provides source-term data that are relevant to plausible sabotage scenarios in relation to spent fuel transport and storage casks and associated risk assessments. We present details and significant results obtained from this program from 2001 through 2007. Measured aerosol results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; measurements of volatile fission product species enhanced sorption--enrichment factors onto respirable particles; and, status on determination of the spent fuel ratio, SFR, needed for scaling studies. Emphasis is provided on recent Phase 3 tests using depleted uranium oxide pellets plus non-radioactive fission product dopants in surrogate spent fuel test rodlets, plus the latest surrogate cerium oxide results and aerosol laboratory supporting calibration work. The DUO 2 , CeO 2 , plus fission product dopant aerosol particle results are compared with available historical data. We also provide a status review on continuing preparations for the final Phase 4 in this program, tests using individual short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. The source-term data, aerosol results, and program design have been tailored to support and guide follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage, aerosol test program was performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission. This program has significant input from, and is cooperatively supported and

  1. Spent fuel sabotage test program, characterization of aerosol dispersal : interim final report.

    Energy Technology Data Exchange (ETDEWEB)

    Gregson, Michael Warren; Brockmann, John E.; Loiseau, Olivier (Institut de Radioprotection et de Surete Nucleaire, France); Klennert, Lindsay A.; Nolte, Oliver (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Molecke, Martin Alan; Autrusson, Bruno A. (Institut de Radioprotection et de Surete Nucleaire, France); Koch, Wolfgang (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Brucher, Wenzel (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Steyskal, Michele D.

    2008-03-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program provides source-term data that are relevant to plausible sabotage scenarios in relation to spent fuel transport and storage casks and associated risk assessments. We present details and significant results obtained from this program from 2001 through 2007. Measured aerosol results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; measurements of volatile fission product species enhanced sorption--enrichment factors onto respirable particles; and, status on determination of the spent fuel ratio, SFR, needed for scaling studies. Emphasis is provided on recent Phase 3 tests using depleted uranium oxide pellets plus non-radioactive fission product dopants in surrogate spent fuel test rodlets, plus the latest surrogate cerium oxide results and aerosol laboratory supporting calibration work. The DUO{sub 2}, CeO{sub 2}, plus fission product dopant aerosol particle results are compared with available historical data. We also provide a status review on continuing preparations for the final Phase 4 in this program, tests using individual short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. The source-term data, aerosol results, and program design have been tailored to support and guide follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage, aerosol test program was performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission. This program has significant input from, and is cooperatively

  2. Fuel behaviour and fission product release under realistic hydrogen conditions comparisons between HEVA 06 test results and Vulcain computations

    International Nuclear Information System (INIS)

    Dumas, J.M.; Lhiaubet, G.

    1989-07-01

    The HEVA 06 test was designed to simulate the conditions existing at the time when fission products are released from irradiated fuel under hydrogen conditions occurring in a PWR core at low pressure. The test conditions were defined from results provided by the core degradation module of the ESCADRE system (1): VULCAIN. This computer code has been recently used to analyse the early core degradation of a 900 MWe PWR in the AF accident sequence (as defined in WASH - 1400, USNRC - 1975). In this scenario, the core would begin to uncover about one day after scram with the system pressure at about 0.4 MPa. The fission product release starts 70 minutes after core dewatering. The F.P. are transferred to the core outlet in an increasingly hydrogen-rich steam atmosphere. The carrier gas is nearly pure hydrogen in the time period 100 - 130 minutes after core uncovering. A large release of F.P. is predicted in the upper part of the core when the steam starvation occurs. At that time, two thirds of the cladding have been oxidised on an average. Before each HEVA test a fuel sample with a burn-up of 36 GWd/tU is reirradiated in order to observe the release of short-lived fission products. A pre-oxidation was primarely conducted in the HEVA 06 test at a temperature of 1300 0 C and controlled to reach a 2/3 cladding oxidation state. Then the steam was progressively replaced by hydrogen and a heat-up rate of 1.5 0 C/s was induced to reach a temperature of 2100 0 C. The fuel was maintained at this temperature for half an hour in hydrogen. The volatile F.P. release kinetics were observed by on-line gamma spectrometry. Pre test calculations of F.P. release kinetics performed with the EMIS module based on the CORSOR models (3) are compared with the test results. Measured releases of cesium and iodine are really lower than those predicted. Axial and radial F.P. distributions in the fuel pellets are available from gamma tomography measurements performed after the test. Tellurium seems

  3. Refinery Integration of By-Products from Coal-Derived Jet Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Caroline Clifford; Andre Boehman; Chunshan Song; Bruce Miller; Gareth Mitchell

    2008-03-31

    The final report summarizes the accomplishments toward project goals during length of the project. The goal of this project was to integrate coal into a refinery in order to produce coal-based jet fuel, with the major goal to examine the products other than jet fuel. These products are in the gasoline, diesel and fuel oil range and result from coal-based jet fuel production from an Air Force funded program. The main goal of Task 1 was the production of coal-based jet fuel and other products that would need to be utilized in other fuels or for non-fuel sources, using known refining technology. The gasoline, diesel fuel, and fuel oil were tested in other aspects of the project. Light cycle oil (LCO) and refined chemical oil (RCO) were blended, hydrotreated to removed sulfur, and hydrogenated, then fractionated in the original production of jet fuel. Two main approaches, taken during the project period, varied where the fractionation took place, in order to preserve the life of catalysts used, which includes (1) fractionation of the hydrotreated blend to remove sulfur and nitrogen, followed by a hydrogenation step of the lighter fraction, and (2) fractionation of the LCO and RCO before any hydrotreatment. Task 2 involved assessment of the impact of refinery integration of JP-900 production on gasoline and diesel fuel. Fuel properties, ignition characteristics and engine combustion of model fuels and fuel samples from pilot-scale production runs were characterized. The model fuels used to represent the coal-based fuel streams were blended into full-boiling range fuels to simulate the mixing of fuel streams within the refinery to create potential 'finished' fuels. The representative compounds of the coal-based gasoline were cyclohexane and methyl cyclohexane, and for the coal-base diesel fuel they were fluorine and phenanthrene. Both the octane number (ON) of the coal-based gasoline and the cetane number (CN) of the coal-based diesel were low, relative to

  4. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    International Nuclear Information System (INIS)

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

    2014-01-01

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests

  5. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests.

  6. Fuels for research and test reactors, status review: July 1982

    International Nuclear Information System (INIS)

    Stahl, D.

    1982-12-01

    A thorough review is provided on nuclear fuels for steady-state thermal research and test reactors. The review was conducted to provide a documented data base in support of recent advances in research and test reactor fuel development, manufacture, and demonstration in response to current US policy on availability of enriched uranium. The review covers current fabrication practice, fabrication development efforts, irradiation performance, and properties affecting fuel utilization, including thermal conductivity, specific heat, density, thermal expansion, corrosion, phase stability, mechanical properties, and fission-product release. The emphasis is on US activities, but major work in Europe and elsewhere is included. The standard fuel types discussed are the U-Al alloy, UZrH/sub x/, and UO 2 rod fuels. Among new fuels, those given major emphasis include H 3 Si-Al dispersion and UO 2 caramel plate fuels

  7. 40 CFR 94.108 - Test fuels.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 20 2010-07-01 2010-07-01 false Test fuels. 94.108 Section 94.108... EMISSIONS FROM MARINE COMPRESSION-IGNITION ENGINES Test Procedures § 94.108 Test fuels. (a) Distillate diesel test fuel. (1) The diesel fuels for testing Category 1 and Category 2 marine engines designed to...

  8. Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio programme

    International Nuclear Information System (INIS)

    Molecke, M.A.; Gregson, M.W.; Sorenson, K.B.

    2004-01-01

    We provide a detailed overview of an on-going, multinational test programme that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolised materials plus volatilised fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high-energy/density device. The programme participants in the United States plus Germany, France and the United Kingdom, part of the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC) have strongly supported and coordinated this research programme. Sandia National Laboratories has the lead role for conducting this research programme; test programme support is provided by both the US Department of Energy and the US Nuclear Regulatory Commission. We provide a summary of the overall, multiphase test design and a description of all explosive containment and aerosol collection test components used. We focus on the recently initiated tests on 'surrogate' spent fuel, unirradiated depleted uranium oxide and forthcoming actual spent fuel tests. We briefly summarise similar results from completed surrogate tests that used non-radioactive, sintered cerium oxide ceramic pellets in test rods. (author)

  9. Effects of burnup on fission product release and implications for severe fuel damage events

    International Nuclear Information System (INIS)

    Appelhans, A.D.; Cronenberg, A.W.; Carboneau, M.L.

    1984-01-01

    Xe, Kr, and I fission-product release data from (a) Halden tests where release in intact rods was measured during irradiation at burnups to 18,000 MWd/t and fuel temperatures of 800 to 1800 0 K, and (b) Power Burst Facility (PBF) tests where trace-irradiated fuel (approx. = 90 MWd/t) was driven to temperatures of >2400 0 K and fuel liquefaction occurred are discussed and related to fuel morphology. Results from both indicate that the fission-product morphology and fuel restructuring govern release behavior. The Halden tests show low release at beginning of life with a 10-fold increase at burnups in excess of 10,000 MWd/t, due to the development of grain boundary interlinkage at higher burnups. Such dependence of release on morphology characteristics is consistent with findings from the PBF tests, where for trace-irradiated fuel, the absence of interlinkage accounts for the low release rates observed during initial fuel heatup, with subsequent enhanced Xe, Kr, and I release via liquefaction or quench-induced destruction of the grain structure. Morphology is also shown to influence the chemical release form of I and Cs fission products

  10. Laboratory Scale Coal And Biomass To Drop-In Fuels (CBDF) Production And Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lux, Kenneth [Altex Technologies Corporation, Sunnyvale, CA (United States); Imam, Tahmina [Altex Technologies Corporation, Sunnyvale, CA (United States); Chevanan, Nehru [Altex Technologies Corporation, Sunnyvale, CA (United States); Namazian, Mehdi [Altex Technologies Corporation, Sunnyvale, CA (United States); Wang, Xiaoxing [Pennsylvania State Univ., University Park, PA (United States); Song, Chunshan [Pennsylvania State Univ., University Park, PA (United States)

    2016-06-29

    This Final Technical Report describes the work and accomplishments of the project entitled, “Laboratory Scale Coal and Biomass to Drop-In Fuels (CBDF) Production and Assessment.” The main objective of the project was to fabricate and test a lab-scale liquid-fuel production system using coal containing different percentages of biomass such as corn stover and switchgrass at a rate of 2 liters per day. The system utilizes the patented Altex fuel-production technology, which incorporates advanced catalysts developed by Pennsylvania State University. The system was designed, fabricated, tested, and assessed for economic and environmental feasibility relative to competing technologies.

  11. FUELS IN TOBACCO PRODUCTION

    Directory of Open Access Journals (Sweden)

    M. Čavlek

    2008-09-01

    Full Text Available Energy production from biomass can reduce „greenhouse effect” and contribute to solving energy security especially in the agricultural households which rely on energy from fossil fuels. In Croatia fuel-cured tobacco is produced on about 5000 ha. Gross income for the whole production is about 180 000 000 kn/year. Flue-cured tobacco is a high energy consuming crop. There are two parts of energy consumption, for mechanization used for the field production (11% and, energy for bulk-curing (89%. In each case, presently used fuels of fossil origin need to be substituted by an alternative energy source of organic origin. Hereafter attention is paid to finding a more economic and ecologically acceptable fuel for curing tobacco. Curing flue-cured tobacco is done by heated air in curing burns. Various sources of heat have been used; wood, coal, oil and gas. In each case different burning facilities of different efficiency have been used. This has had an impact on curing costs and ecology. Recently, mostly used fuel has been natural gas. However, gas is getting expensive. Consequently, an alternative fuel for curing tobacco is sought for. According to literature, agricultural crops suitable for the latter purpose could be wheat, barley, maize, sorghum, sugar beet and some other annual and perennial plant species. Wooden pellets (by-products are suitable for combustion too. Ligno-cellulose fuels have been used for heating since long time. However, not sufficient research has been done from an applied point of view (Venturi and Venturi, 2003. Fuel combustion is getting more efficient with developing technological innovations. The curing barn manufacturers are offering technology for combusting wooden pellets (by-products for curing tobacco. The pellets are available on domestic market. The same technology can be used for combustion of maize grain. Within “Hrvatski duhani” research on suitability of using wooden pellets and maize grain and whole

  12. Status of RBCB testing of LMR oxide fuel in EBR-II

    International Nuclear Information System (INIS)

    Strain, R.V.; Bottcher, J.H.; Gross, K.C.; Lambert, J.D.B.; Ukai, S.; Nomura, S.; Shikakura, S.; Katsuragawa, M.

    1991-01-01

    The status is given of the the American-Japanese collaborative program in Experimental Breeder Reactor 2 to determine the run-beyond-cladding-breach performance of (UPu)O 2 fuel pins for liquid-metal cooled reactors. Phase 1 of the collaboration involved eighteen irradiation tests over 1981--86 with 5.84-mm pins in 316 or D9 stainless steel. Emphasis in Phase 2 tests from 1989 onwards is with larger diameter (7.5mm) pins in advanced claddings. Results include delayed neutron and fission gas release data from breached pins, the impact of fuel-sodium reaction product formation on pin performance, and fuel and fission product contamination from failures. 13 refs, 1 fig., 4 tabs

  13. Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition

    International Nuclear Information System (INIS)

    Terry, W.K.; Ryskamp, J.M.; Sterbentz, J.W.

    1995-08-01

    Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification; (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania; (3) The effects of WGPu isotopic composition; (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight; (5) The effects of americium and gallium in WGPu; (6) Fission gas release from MOX fuel pellets made from WGPu; (7) Fuel/cladding gap closure; (8) The effects of power cycling and off-normal events on fuel integrity; (9) Development of radial distributions of burnup and fission products; (10) Power spiking near the interfaces of MOX and urania fuel assemblies; and (11) Fuel performance code validation. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program

  14. Fuel bundle examination techniques for the Phebus fission product test

    International Nuclear Information System (INIS)

    Blanc, J.Y.; Clement, B.; Hardt, P. von der

    1996-01-01

    The paper develops the non-destructive examinations, with a special emphasis on transmission tomography, performed in the Phebus facility, using a linear accelerator associated with a line scan camera based on PCD components. This particular technique enabled the high level of penetration to be obtained, necessary for this high density application. Spatial resolution is not far from the theoretical limit and the density resolution is often adequate. This technique permitted: 1) to define beforehand the cuts on a precise basis, avoiding a long step-by-step choice as in previous in-pile tests; 2) to determine, at an early stage, mass balance, material relocations (in association with axial gamma spectrometry), and FP distribution, as an input into re-calculations of the bundle events. However, classical cuttings, periscopic visual examinations, macrographies, micrographies and EPMA analyses remain essential to give oxidation levels (in the less degraded zones), phase aspect and composition, to distinguish between materials of identical density, and, if possible, to estimate temperatures. Oxidation resistance of sensors (thermocouples or ultrasonic thermometers) is also traced. The EPMA gives access to the molten material chemical analyses, especially in the molten fuel blockage area. The first results show that an important part of the fuel bundle melted (which was one of the objectives of this test) and that the degradation level is close to TIMI-2 with a molten plug under a cavity surrounded by an uranium-rich crust. In lower and upper areas fuel rods are less damaged. Complementaries between these examination techniques and between international teams involved will be major advantages in the Phebus FPT0 test comprehension. 3 refs, 9 figs

  15. Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Ott, Larry J.; Ellis, Ronald James; McDuffee, Joel Lee; Spellman, Donald J.; Bevard, Bruce Balkcom

    2009-01-01

    The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions (i.e., for both fast-spectrum conditions and light-water-reactor conditions). This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.

  16. Foucault current testing of ferritic steel fuel cans

    International Nuclear Information System (INIS)

    Stossel, A.

    1984-10-01

    The analysis of impedance involved by a Foucault current test of ferritic steel tubes, is quite different from the classical analysis which refers to non-magnetic tubes; more particularly, volume defects are considered as magnetic anomalies. Contrarily to current instructions which recommend to test the product in a satured magnetic state, it is very interesting to work with a continuous energizing field, comparatively low, corresponding to a sequenced magnetization, of which value is obtained according to the magnetic structure of the product. This analysis is useful when testing fast reactor fuel cans [fr

  17. Fuel motion in overpower tests of metallic integral fast reactor fuel

    International Nuclear Information System (INIS)

    Rhodes, E.A.; Bauer, T.H.; Stanford, G.S.; Regis, J.P.; Dickerman, C.E.

    1992-01-01

    In this paper results from hodoscope data analyses are presented for transient overpower (TOP) tests M5, M6, and M7 at the Transient Reactor Test Facility, with emphasis on transient feedback mechanisms, including prefailure expansion at the tops of the fuel pins, subsequent dispersive axial fuel motion, and losses in relative worth of the fuel pins during the tests. Tests M5 and M6 were the first TOP tests of margin to cladding breach and prefailure elongation of D9-clad ternary (U-Pu-Zr) integral fast reactor-type fuel. Test M7 extended these results to high-burnup fuel and also initiated transient testing of HT-9-clad binary (U-Zr) Fast Flux Test Facility driver fuel. Results show significant prefailure negative reactivity feedback and strongly negative feedback from fuel driven to failure

  18. Comparative Study of Laboratory-Scale and Prototypic Production-Scale Fuel Fabrication Processes and Product Characteristics

    International Nuclear Information System (INIS)

    2014-01-01

    An objective of the High Temperature Gas Reactor fuel development and qualification program for the United States Department of Energy has been to qualify fuel fabricated in prototypic production-scale equipment. The quality and characteristics of the tristructural isotropic coatings on fuel kernels are influenced by the equipment scale and processing parameters. Some characteristics affecting product quality were suppressed while others have become more significant in the larger equipment. Changes to the composition and method of producing resinated graphite matrix material has eliminated the use of hazardous, flammable liquids and enabled it to be procured as a vendor-supplied feed stock. A new method of overcoating TRISO particles with the resinated graphite matrix eliminates the use of hazardous, flammable liquids, produces highly spherical particles with a narrow size distribution, and attains product yields in excess of 99%. Compact fabrication processes have been scaled-up and automated with relatively minor changes to compact quality to manual laboratory-scale processes. The impact on statistical variability of the processes and the products as equipment was scaled are discussed. The prototypic production-scale processes produce test fuels that meet fuel quality specifications.

  19. Fuel behavior and fission product release under HTGR accident conditions

    International Nuclear Information System (INIS)

    Fukuda, K.; Hayashi, K.; Shiba, K.

    1990-01-01

    In early 1989 a final decision was made over construction of a 30 MWth HTGR called the High Temperature Engineering Test Reactor, HTTR, in Japan in order to utilize it for high temperature gas engineering tests and various nuclear material tests. The HTTR fuel is a pin-in-block type fuel element which is composed of a hexagonal graphite block with dimension of 580 mm in length and 360 mm in face-to-face distance and about 30 of the fuel rods inserted into the coolant channels drilled in the block. The TRISO coated fuel particles for HTTR are incorporated with graphite powder and phenol resin into the fuel compacts, 19 of which are encased into a graphite sleeve as a fuel rod. It is necessary for the HTTR licensing to prove the fuel stability under predicted accidents related to the high temperature events. Therefore, the release of the fission products and the fuel failure have been investigated in the irradiation---and the heating experiments simulating these conditions at JAERI. This report describes the HTTR fuel behavior at extreme temperature, made clear in these experiments

  20. Costs of electronuclear fuel production

    International Nuclear Information System (INIS)

    Flaim, T.; Loose, V.

    1978-07-01

    The Los Alamos Scientific Laboratory (LASL) proposes to study the electronuclear fuel producer (EFP) as a means of producing fissile fuel to generate electricity. The main advantage of the EFP is that it may reduce the risks of nuclear proliferation by breeding 233 U from thorium, thereby avoiding plutonium separation. A report on the costs of electronuclear fuel production based upon two designs considered by LASL is presented. The findings indicate that the EFP design variations considered are not likely to result in electricity generation costs as low as the uranium fuel cycle used in the US today. At current estimates of annual fuel output (500 kg 233 U per EFP), the costs of electricity generation using fuel produced by the EFP are more than three times higher than generating costs using the traditional fuel cycle. Sensitivity analysis indicates that electronuclear fuel production would become cost competitive with the traditional uranium fuel cycle when U 3 O 8 (yellowcake) prices approach $1000 per pound

  1. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    International Nuclear Information System (INIS)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D.

    1999-04-01

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs

  2. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

    1999-04-01

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs.

  3. HTGR Fuels and Core Development Program. Quarterly progress report for the period ending August 31, 1977. [Graphite and fuel irradiation; fission product release

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    The work reported includes studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and data are presented.

  4. Fabrication of Fast Reactor Fuel Pins for Test Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Karsten, G. [Institute for Applied Reactor Physics, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Dippel, T. [Institute for Radiochemistry, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Laue, H. J. [Institute for Applied Reactor Physics, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany)

    1967-09-15

    An extended irradiation programme is being carried out for the fuel element development of the Karlsruhe fast breeder project. A very important task within the programme is the testing of plutonium-containing fuel pins in a fast-reactor environment. This paper deals with fabrication of such pins by our laboratories at Karlsruhe. For the fast reactor test positions at present envisaged a fuel with 15% plutonium and the uranium fully enriched is appropriate. Hie mixed oxide is both pelletized and vibro-compacted with smeared densities between 80 and 88% theoretical. The pin design is, for example, such that there are two gas plena at the top and bottom, and one blanket above the fuel with the fuel zone fitting to the test reactor core length. The specifications both for fuel and cladding have been adapted to the special purpose of a fast-breeder reactor - the outer dimensions, the choice of cladding and fuel types, the data used and the kind of tests outline the targets of the development. The fuel fabrication is described in detail, and also the powder line used for vibro-compaction. The source materials for the fuel are oxalate PuO{sub 2} and UO{sub 2} from the UF{sub 6} process. The special problems of mechanical mixing and of plutonium homogeneity have been studied. The development of the sintering technique and grain characteristics for vibratory compactive fuel had to overcome serious problems in order to reach 82-83% theoretical. The performance of the pin fabrication needed a major effort in welding, manufacturing of fits and decontamination of the pin surfaces. This was a stimulation for the development of some very subtle control techniques, for example taking clear X-ray photographs and the tube testing. In general the selection of tests was a special task of the production routine. In conclusion the fabrication of the pins resulted in valuable experiences for the further development of fast reactor fuel elements. (author)

  5. Development of Out-pile Test Technology for Fuel Assembly Performance Verification

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Tae Hyun; In, W. K.; Oh, D. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)] (and others)

    2007-03-15

    Out-pile tests with full scale fuel assembly are to verify the design and to evaluate the performance of the final products. HTL for the hydraulic tests and FAMeCT for mechanical/structural tests were constructed in this project. The maximum operating conditions of HTL are 30 bar, 320 .deg. C, and 500 m3/hr. This facility can perform the pressure drop test, fuel assembly uplift test, and flow induced vibration test. FAMeCT can perform the bending and vibration tests. The verification of the developed facilities were carried out by comparing the reference data of the fuel assembly which was obtained at the Westinghouse Co. The compared data showed a good coincidence within uncertainties. FRETONUS was developed for high temperature and high pressure fretting wear simulator and performance test. A performance test was conducted for 500 hours to check the integrity, endurance, data acquisition capability of the simulator. The technology of turbulent flow analysis and finite element analysis by computation was developed. From the establishments of out-pile test facilities for full scale fuel assembly, the domestic infrastructure for PWR fuel development has been greatly upgraded.

  6. Transuranics and fission products release from PWR fuels in severe accident conditions. Lessons learnt from VERCORS RT3 and RT4 tests

    International Nuclear Information System (INIS)

    Pontillon, Y.; Ducros, G.; Van Winckel, S.; Christiansen, B.; Kissane, M.P.; Dubourg, R.; Dutheillet, Y.; Andreo, F.

    2006-01-01

    Over the last decades, several experimental programs devoted to the source term of fission products (FP) and actinides released from PWR fuel samples in severe accident (SA) conditions have been initiated throughout the world. In France, in this context, the Institute for Radiological Protection and Safety (IRSN) and Electricite de France (EDF) have supported the analytical VERCORS program which was performed by the Commissariat a l'Energie Atomique (CEA). The VERCORS facility at the LAMA-laboratory (CEA-Grenoble, France) was designed to heat up an irradiated fuel sample - taken from EDF's nuclear power reactors - to fuel relocation, and to capture the fission products released from the fuel and deposited downstream on a series of specific filters (impactors, bead-bed filter). On-line gamma detectors aimed at the fuel position, filters and gas capacity monitored the progress of FP release from the fuel, FP deposition on the filters and the fission gases emitted by the fuel (xenon and krypton). Before and after the test, a longitudinal gamma-scan of the fuel was conducted to measure the initial and final FP inventory in order to evaluate the quantitative fractions of FP emitted by the fuel during the test. All the components of the loop were then gamma-scanned to measure and locate the FPs released during the test and to draw up a mass balance of these FP. 25 annealing tests were performed between 1983 and 2002 on irradiated PWR fuels under various conditions of temperature and atmospheres (oxidising or reducing conditions). The influence of the nature of the fuel (UO 2 versus MOX, burn up) and the fuel morphology (initially intact or fragmented fuel) have also been investigated. This led to an extended data base allowing on the one hand to study mechanisms which promote FP release in SA conditions, and on the other hand to enhance models implemented in SA codes. Because gamma spectrometry is well suited to FP measurement and not to actinides (except neptunium

  7. Post irradiation test report of irradiated DUPIC simulated fuel

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Jung, I. H.; Moon, J. S. and others

    2001-12-01

    The post-irradiation examination of irradiated DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) simulated fuel in HANARO was performed at IMEF (Irradiated Material Examination Facility) in KAERI during 6 months from October 1999 to March 2000. The objectives of this post-irradiation test are i) the integrity of the capsule to be used for DUPIC fuel, ii) ensuring the irradiation requirements of DUPIC fuel at HANARO, iii) performance verification in-core behavior at HANARO of DUPIC simulated fuel, iv) establishing and improvement the data base for DUPIC fuel performance verification codes, and v) establishing the irradiation procedure in HANARO for DUPIC fuel. The post-irradiation examination performed are γ-scanning, profilometry, density, hardness, observation the microstructure and fission product distribution by optical microscope and electron probe microanalyser (EPMA)

  8. Aviation fuel and future oil production scenarios

    International Nuclear Information System (INIS)

    Nygren, Emma; Aleklett, Kjell; Hoeoek, Mikael

    2009-01-01

    Most aviation fuels are jet fuels originating from crude oil. Crude oil must be refined to be useful and jet fuel is only one of many products that can be derived from crude oil. Jet fuel is extracted from the middle distillates fraction and competes, for example, with the production of diesel. Crude oil is a limited natural resource subject to depletion and several reports indicate that the world's crude oil production is close to the maximum level and that it will start to decrease after reaching this maximum. A post-Kyoto political agenda to reduce oil consumption will have the same effect on aviation fuel production as a natural decline in the crude oil production. On the other hand, it is predicted by the aviation industry that aviation traffic will keep on increasing. The industry has put ambitious goals on increases in fuel efficiency for the aviation fleet. Traffic is predicted to grow by 5% per year to 2026, fuel demand by about 3% per year. At the same time, aviation fuel production is predicted to decrease by several percent each year after the crude oil production peak is reached resulting in a substantial shortage of jet fuel by 2026. The aviation industry will have a hard time replacing this with fuel from other sources, even if air traffic remains at current levels.

  9. Refinery Integration of By-Products from Coal-Derived Jet Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Caroline E. Burgess Clifford; Andre Boehman; Chunshan Song; Bruce Miller; Gareth Mitchell

    2007-03-17

    This report summarizes the accomplishments toward project goals during the no cost extension period of the third year of the project to assess the properties and performance of coal based products. These products are in the gasoline, diesel and fuel oil range and result from coal based jet fuel production from an Air Force funded program. Specific areas of progress include generation of coal based material that has been fractionated into the desired refinery cuts for a third round of testing, the use of a research gasoline engine to test coal-based gasoline, and modification of diesel engines for use in evaluating diesel produced in the project. At the pilot scale, the hydrotreating process was modified to separate the heavy components from the LCO and RCO fractions before hydrotreating in order to improve the performance of the catalysts in further processing. Hydrotreating and hydrogenation of the product has been completed, and due to removal of material before processing, yield of the jet fuel fraction has decreased relative to an increase in the gasoline fraction. Characterization of the gasoline fuel indicates a dominance of single ring alkylcycloalkanes that have a low octane rating; however, blends containing these compounds do not have a negative effect upon gasoline when blended in refinery gasoline streams. Characterization of the diesel fuel indicates a dominance of 3-ring aromatics that have a low cetane value; however, these compounds do not have a negative effect upon diesel when blended in refinery diesel streams. Both gasoline and diesel continue to be tested for combustion performance. The desulfurization of sulfur containing components of coal and petroleum is being studied so that effective conversion of blended coal and petroleum streams can be efficiently converted to useful refinery products. Activated carbons have proven useful to remove the heavy sulfur components, and unsupported Ni/Mo and Ni/Co catalysts have been very effective for

  10. Bio-fuel production potential in Romania

    International Nuclear Information System (INIS)

    Laurentiu, F.; Silvian, F.; Dumitru, F.

    2006-01-01

    The paper is based on the ESTO Study: Techno- Economic Feasibility of Large-Scale Production of Bio-Fuels in EU-Candidate Countries. Bio-fuel production has not been taken into account significantly until now in Romania, being limited to small- scale productions of ethanol, used mostly for various industrial purposes. However the climatic conditions and the quality of the soil are very suitable in the country for development of the main crops (wheat, sugar-beet, sunflower and rape-seed) used in bio-ethanol and bio-diesel production. The paper intended to consider a pertinent discussion of the present situation in Romania's agriculture stressing on the following essential items in the estimation of bio-fuels production potential: availability of feed-stock for bio-fuel production; actual productions of bio-fuels; fuel consumption; cost assessment; SWOT approach; expected trends. Our analysis was based on specific agricultural data for the period 1996-2000. An important ethanol potential (due to wheat, sugar-beet and maize cultures), as well as bio-diesel one (due to sun-flower and rape-seed) were predicted for the period 2005-2010 which could be exploited with the support of an important financial and technological effort, mainly from EU countries

  11. Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio program

    Energy Technology Data Exchange (ETDEWEB)

    Molecke, M.A.; Gregson, M.W.; Sorenson, K.B. [Sandia National Labs. (United States); Billone, M.C.; Tsai, H. [Argonne National Lab. (United States); Koch, W.; Nolte, O. [Fraunhofer Inst. fuer Toxikologie und Experimentelle Medizin (Germany); Pretzsch, G.; Lange, F. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (Germany); Autrusson, B.; Loiseau, O. [Inst. de Radioprotection et de Surete Nucleaire (France); Thompson, N.S.; Hibbs, R.S. [U.S. Dept. of Energy (United States); Young, F.I.; Mo, T. [U.S. Nuclear Regulatory Commission (United States)

    2004-07-01

    We provide a detailed overview of an ongoing, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high energy density device, HEDD. The program participants in the U.S. plus Germany, France, and the U.K., part of the international Working Group for Sabotage Concerns of Transport and Storage Casks, WGSTSC have strongly supported and coordinated this research program. Sandia National Laboratories, SNL, has the lead role for conducting this research program; test program support is provided by both the U.S. Department of Energy and Nuclear Regulatory Commission. WGSTSC partners need this research to better understand potential radiological impacts from sabotage of nuclear material shipments and storage casks, and to support subsequent risk assessments, modeling, and preventative measures. We provide a summary of the overall, multi-phase test design and a description of all explosive containment and aerosol collection test components used. We focus on the recently initiated tests on ''surrogate'' spent fuel, unirradiated depleted uranium oxide, and forthcoming actual spent fuel tests. The depleted uranium oxide test rodlets were prepared by the Institut de Radioprotection et de Surete Nucleaire, in France. These surrogate test rodlets closely match the diameter of the test rodlets of actual spent fuel from the H.B. Robinson reactor (high burnup PWR fuel) and the Surry reactor (lower, medium burnup PWR fuel), generated from U.S. reactors. The characterization of the spent fuels and fabrication into short, pressurized rodlets has been performed by Argonne National Laboratory, for testing at SNL. The ratio of the aerosol and respirable particles released from HEDD-impacted spent

  12. Wood fuel production technologies in EU countries

    Energy Technology Data Exchange (ETDEWEB)

    Hakkila, P [Finnish Forest Research Institute, Vantaa (Finland)

    1998-12-31

    The presentation reviews the major technologies used for the production of fuel chips for heating plants in Europe. Three primary options are considered: production of whole-tree chips from young trees for fuel; integrated harvesting of fiber and energy from thinning based on tree-section system; and production of fuel chips from logging residue in clear-cut areas after fully mechanized logging. The characteristics of the available biomass reserve and proven technology for its recovery are discussed. The employment effects of fuel chip production and the costs of wood fuels are also briefly discussed. (author) 3 refs., 3 figs.

  13. Wood fuel production technologies in EU countries

    Energy Technology Data Exchange (ETDEWEB)

    Hakkila, P. [Finnish Forest Research Institute, Vantaa (Finland)

    1997-12-31

    The presentation reviews the major technologies used for the production of fuel chips for heating plants in Europe. Three primary options are considered: production of whole-tree chips from young trees for fuel; integrated harvesting of fiber and energy from thinning based on tree-section system; and production of fuel chips from logging residue in clear-cut areas after fully mechanized logging. The characteristics of the available biomass reserve and proven technology for its recovery are discussed. The employment effects of fuel chip production and the costs of wood fuels are also briefly discussed. (author) 3 refs., 3 figs.

  14. Test Results of PBMR Fuel Spheres

    International Nuclear Information System (INIS)

    Koshcheev, Konstantin; Diakov, Alexander; Beltyukov, Igor; Barybin, Andrey; Chernetsov, Mikhail

    2014-01-01

    Results of pre-irradiation testing of fuel spheres (FS) and coated particles (CP) manufactured by PBMR SOC (Republic of South Africa) are described. The stable high quality level of major characteristics (dimensions, CP coating structure, uranium-235 contamination of the FS matrix graphite and the outer PyC layer of the CP coating) are shown. Results of a methodical irradiation test of two FS in helium and neon medium at temperatures of 800 to 1300 °C with simultaneous determination of release-to-birth ratios for major gaseous fission products (GFP) are described. (author)

  15. Loss-of-flow test L5 on FFTF-type irradiated fuel

    International Nuclear Information System (INIS)

    Simms, R.; Gehl, S.M.; Lo, R.K.; Rothman, A.B.

    1978-03-01

    Test L5 simulated a hypothetical loss-of-flow accident in an LMFBR using three (Pu, U)O 2 fuel elements of the FTR type. The test elements were irradiated before TREAT Test L5 in the General Electric Test Reactor to 8 at. % burnup at about 40 kW/m. The preirradiation in GETR caused a fuel-restructuring range characteristic of moderate-power structure relative to the FTR. The test transient was devised so that a power burst would be initiated at incipient cladding melting after the loss of flow. The test simulation corresponds to a scenario for FTR in which fuel in high-power-structure subassemblies slump, resulting in a power excursion. The remaining subassemblies are subjected to this power burst. Test L5 addressed the fuel-motion behavior of the subassemblies in this latter category. Data from test-vehicle sensors, hodoscope, and post-mortem examinations were used to construct the sequence of events within the test zone. From these observations, the fuel underwent a predominantly dispersive event just after reaching a peak power six times nominal at, or after, scram. The fuel motion was apparently driven by the release of entrained fission-product gases, since fuel vapor pressure was deliberately kept below significant levels for the transient. The test remains show a wide range of microstructural evolution, depending on the extent of heat deposition along the active fuel column. Extensive fuel swelling was also observed as a result of the lack of the cladding restraint. The results of the thermal-hydraulic calculations with the SAS3A code agreed qualitatively with the postmortem results with respect to the extent of the melting and the dispersal of cladding and fuel. However, the calculated times of certain events did not agree with the observed times

  16. Research and Test Reactor Fuel Elements (RTRFE)

    International Nuclear Information System (INIS)

    Pace, Brett W.; Marinak, Edward A.

    1999-01-01

    BWX Technologies Inc. (BWXT) has experienced several production improvements over the past year. The homogeneity yields in 4.8 gU/cc U 3 Si 2 plates have increased over last year's already high yields. Through teamwork and innovative manufacturing techniques, maintaining high quality surface finishes on plates and elements is becoming easier and less expensive. Currently, BWXT is designing a fabrication development plan to reach a fuel loading of 9 gU/cc within 2 - 4 years. This development will involve a step approach requested by ANL to produce plates using U-8Mo at a loading of 6 gU/cc first and qualify the fuel at those levels. In achieving the goal of a very high-density fuel loading of 9 gU/cc, BWXT is considering employing several new, state of the art, ultrasonic testing techniques for fuel core evaluation. (author)

  17. Updated FY12 Ceramic Fuels Irradiation Test Plan

    International Nuclear Information System (INIS)

    Nelson, Andrew T.

    2012-01-01

    The Fuel Cycle Research and Development program is currently devoting resources to study of numerous fuel types with the aim of furthering understanding applicable to a range of reactors and fuel cycles. In FY11, effort within the ceramic fuels campaign focused on planning and preparation for a series of rabbit irradiations to be conducted at the High Flux Isotope Reactor located at Oak Ridge National Laboratory. The emphasis of these planned tests was to study the evolution of thermal conductivity in uranium dioxide and derivative compositions as a function of damage induced by neutron damage. Current fiscal realities have resulted in a scenario where completion of the planned rabbit irradiations is unlikely. Possibilities for execution of irradiation testing within the ceramic fuels campaign in the next several years will thus likely be restricted to avenues where strong synergies exist both within and outside the Fuel Cycle Research and Development program. Opportunities to augment the interests and needs of modeling, advanced characterization, and other campaigns present the most likely avenues for further work. These possibilities will be pursued with the hope of securing future funding. Utilization of synthetic microstructures prepared to better understand the most relevant actors encountered during irradiation of ceramic fuels thus represents the ceramic fuel campaign's most efficient means to enhance understanding of fuel response to burnup. This approach offers many of the favorable attributes embraced by the Separate Effects Testing paradigm, namely production of samples suitable to study specific, isolated phenomena. The recent success of xenon-imbedded thick films is representative of this approach. In the coming years, this strategy will be expanded to address a wider range of problems in conjunction with use of national user facilities novel characterization techniques to best utilize programmatic resources to support a science-based research program.

  18. Spent fuel sabotage aerosol ratio program : FY 2004 test and data summary.

    Energy Technology Data Exchange (ETDEWEB)

    Brucher, Wenzel (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Koch, Wolfgang (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Loiseau, Olivier (Institut de Radioprotection et de Surete Nucleaire, France); Mo, Tin (U.S. Nuclear Regulatory Commission, Washington, DC); Billone, Michael C. (Argonne National Laboratory, Argonne, IL); Autrusson, Bruno A. (Institut de Radioprotection et de Surete Nucleaire, France); Young, F. I. (U.S. Nuclear Regulatory Commission, Washington, DC); Coats, Richard Lee; Burtseva, Tatiana (Argonne National Laboratory, Argonne, IL); Luna, Robert Earl; Dickey, Roy R.; Sorenson, Ken Bryce; Nolte, Oliver (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Thompson, Nancy Slater (U.S. Department of Energy, Washington, DC); Hibbs, Russell S. (U.S. Department of Energy, Washington, DC); Gregson, Michael Warren; Lange, Florentin (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Molecke, Martin Alan; Tsai, Han-Chung (Argonne National Laboratory, Argonne, IL)

    2005-07-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program has been underway for several years. This program provides data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. The program also provides significant technical and political benefits in international cooperation. We are quantifying the Spent Fuel Ratio (SFR), the ratio of the aerosol particles released from HEDD-impacted actual spent fuel to the aerosol particles produced from surrogate materials, measured under closely matched test conditions, in a contained test chamber. In addition, we are measuring the amounts, nuclide content, size distribution of the released aerosol materials, and enhanced sorption of volatile fission product nuclides onto specific aerosol particle size fractions. These data are the input for follow-on modeling studies to quantify respirable hazards, associated radiological risk assessments, vulnerability assessments, and potential cask physical protection design modifications. This document includes an updated description of the test program and test components for all work and plans made, or revised, during FY 2004. It also serves as a program status report as of the end of FY 2004. All available test results, observations, and aerosol analyses plus interpretations--primarily for surrogate material Phase 2 tests, series 2/5A through 2/9B, using cerium oxide sintered ceramic pellets are included. Advanced plans and progress are described for upcoming tests with unirradiated, depleted uranium oxide and actual spent fuel test rodlets. This spent fuel sabotage--aerosol test program is coordinated with the international Working Group for Sabotage Concerns of

  19. Spent fuel sabotage aerosol ratio program : FY 2004 test and data summary

    International Nuclear Information System (INIS)

    Brucher, Wenzel; Koch, Wolfgang; Pretzsch, Gunter Guido; Loiseau, Olivier; Mo, Tin; Billone, Michael C.; Autrusson, Bruno A.; Young, F. I.; Coats, Richard Lee; Burtseva, Tatiana; Luna, Robert Earl; Dickey, Roy R.; Sorenson, Ken Bryce; Nolte, Oliver; Thompson, Nancy Slater; Hibbs, Russell S.; Gregson, Michael Warren; Lange, Florentin; Molecke, Martin Alan; Tsai, Han-Chung

    2005-01-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program has been underway for several years. This program provides data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. The program also provides significant technical and political benefits in international cooperation. We are quantifying the Spent Fuel Ratio (SFR), the ratio of the aerosol particles released from HEDD-impacted actual spent fuel to the aerosol particles produced from surrogate materials, measured under closely matched test conditions, in a contained test chamber. In addition, we are measuring the amounts, nuclide content, size distribution of the released aerosol materials, and enhanced sorption of volatile fission product nuclides onto specific aerosol particle size fractions. These data are the input for follow-on modeling studies to quantify respirable hazards, associated radiological risk assessments, vulnerability assessments, and potential cask physical protection design modifications. This document includes an updated description of the test program and test components for all work and plans made, or revised, during FY 2004. It also serves as a program status report as of the end of FY 2004. All available test results, observations, and aerosol analyses plus interpretations--primarily for surrogate material Phase 2 tests, series 2/5A through 2/9B, using cerium oxide sintered ceramic pellets are included. Advanced plans and progress are described for upcoming tests with unirradiated, depleted uranium oxide and actual spent fuel test rodlets. This spent fuel sabotage--aerosol test program is coordinated with the international Working Group for Sabotage Concerns of

  20. Future Transient Testing of Advanced Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack

    2009-09-01

    The transient in-reactor fuels testing workshop was held on May 4–5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat à l'Énergie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric – Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by

  1. Future Transient Testing of Advanced Fuels

    International Nuclear Information System (INIS)

    Carmack, Jon

    2009-01-01

    The transient in-reactor fuels testing workshop was held on May 4-5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat energie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric - Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by the

  2. Fuel-sodium reaction product formation in breached mixed-oxide fuel

    International Nuclear Information System (INIS)

    Bottcher, J.H.; Lambert, J.D.B.; Strain, R.V.; Ukai, S.; Shibahara, S.

    1988-01-01

    The run-beyond-cladding-breach (RBCB) operation of mixed-oxide LMR fuel pins has been studied for six years in the Experimental Breeder Reactor-II (EBR-II) as part of a joint program between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan. The formation of fuel-sodium reaction product (FSRP), Na 3 MO 4 , where M = U/sub 1-y/Pu/sub y/, in the outer fuel regions is the major phenomenon governing RBCB behavior. It increases fuel volume, decreases fuel stoichiometry, modifies fission-product distributions, and alters thermal performance of a pin. This paper describes the morphology of Na 3 MO 4 observed in 5.84-mm diameter pins covering a variety of conditions and RBCB times up to 150 EFPD's. 8 refs., 1 fig

  3. Studies on the fission products behavior during dissolution process of BWR spent fuel

    International Nuclear Information System (INIS)

    Sato, K.; Nakai, E.; Kobayashi, Y.

    1987-01-01

    In order to obtain basic data on fission products behavior in connection with the head end process of fuel reprocessing, especially to obtain better understanding on undissolved residues, small scale dissolution studies were performed by using BWR spent fuel rods which were irradiated as monitoring fuel rods under the monitoring program for LWR fuel assembly performance entitled PROVING TEST ON RELIABILITY OF FUEL ASSEMBLY . The Zircaloy-2 claddings and the fuel pellets were subjected individually to the following studies on 1) release of fission products during dissolution process, 2) characterization of undissolved residues, and 3) analysis of the claddings. This paper presents comprehensive descriptions of the fission products behavior during dissolution process, based on detailed and through PIE conducted by JNFS under the sponsorship of MITI (Ministry of International Trade and Industry)

  4. HANARO fuel irradiation test(II)

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, D. S.; Kim, H. R.; Chae, H. T.; Lee, B. C.; Lee, C. S.; Kim, B. G.; Lee, C. B.; Hwang, W

    2001-04-01

    In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiatied at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%.

  5. Alternative Fuels Data Center: Conventional Natural Gas Production

    Science.gov (United States)

    Conventional Natural Gas Production to someone by E-mail Share Alternative Fuels Data Center : Conventional Natural Gas Production on Facebook Tweet about Alternative Fuels Data Center: Conventional Natural Gas Production on Twitter Bookmark Alternative Fuels Data Center: Conventional Natural Gas Production

  6. Preliminary nuclear design for test MOX Fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyung Kook; Kim, Taek Kyum; Jeong, Hyung Guk; Noh, Jae Man; Cho, Jin Young; Kim, Young Il; Kim, Young Jin; Sohn, Dong Seong

    1997-10-01

    As a part of activity for future fuel development project, test MOX fuel rods are going to be loaded and irradiated in Halden reactor core as a KAERI`s joint international program with Paul Scherrer Institute (PSI). PSI will fabricate test MOX rods with attrition mill device which was developed by KAERI. The test fuel assembly rig contains three MOX rods and three inert matrix rods. One of three MOX rods will be fabricated by BNFL, the other two MOX fuel rods will be manufacturing jointly by KAERI and PSI. Three inert matrix fuel rods will be fabricated with Zr-Y-Er-Pu oxide. Neutronic evaluation was preliminarily performed for test fuel assembly suggested by PSI. The power distribution of test fuel rod in test fuel assembly was analyzed for various fuel rods position in assembly and the depletion characteristic curve for test fuel was also determined. The fuel rods position in test fuel assembly does not effect the rod power distribution, and the proposal for test fuel rods suggested by PSI is proved to be feasible. (author). 2 refs., 13 tabs., 16 figs.

  7. Testing of HTR UO{sub 2} TRISO fuels in AVR and in material test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kania, Michael J., E-mail: MichaelJKania@googlemail.com [Retired from Lockheed Martin Corp, 20 Beach Road, Averill Park, NY 12018 (United States); Nabielek, Heinz, E-mail: heinznabielek@me.com [Retired from Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Verfondern, Karl [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Allelein, Hans-Josef [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, 52072 Aachen (Germany)

    2013-10-15

    The German High Temperature Reactor Fuel Development Program successfully developed, licensed and manufactured many thousands of spherical fuel elements that were used to power the experimental AVR reactor and the commercial THTR reactor. In the 1970s, this program extended the performance envelope of HTR fuels by developing and qualifying the TRISO-coated particle system. Irradiation testing in real-time AVR tests and accelerated MTR tests demonstrated the superior manufacturing process of this fuel and its irradiation performance. In the 1980s, another program direction change was made to a low enriched UO{sub 2} TRISO-coated particle system coupled with high-quality manufacturing specifications designed to meet new HTR plant design needs. These needs included requirements for inherent safety under normal operation and accident conditions. Again, the German fuel development program met and exceeded these challenges by manufacturing and qualifying the low-enriched UO{sub 2} TRISO-fuel system for HTR systems with steam generation, gas-turbine systems and very high temperature process heat applications. Fuel elements were manufactured in production scale facilities that contained near defect free UO{sub 2} TRISO coated particles, homogeneously distributed within a graphite matrix with very low levels of uranium contamination. Good irradiation performance for these elements was demonstrated under normal operating conditions to 12% FIMA and under accident conditions not exceeding 1600 °C.

  8. A nondestructive testing device for determining 235U enrichment in power reactor fuel elements

    International Nuclear Information System (INIS)

    Liu Lanhua; Liu Nangai

    1990-07-01

    The development and application of a nondestructive testing device are presented, which is used for determining the 235 U enrichment in the mixed fuel of fuel elements with UO 2 pellets. The testing efficiency is improved because the passive gamma ray method and a hole-bored NaI crystal and four channel multichannel analyzer are used. The false discrimination rate is reduced as the average comparing method is taken. This device is simple in structure and easy in operation. It has provided a new testing tool for the fuel elements production in China. This device has successfully been used in Qinshan Nuclear Power Plant in testing its fuel elements

  9. Transport of volatile fission products in the fuel-to-sheath gap of defective fuel elements during normal and reactor accident conditions

    International Nuclear Information System (INIS)

    Lewis, B.J.; Bonin, H.W.

    1995-01-01

    An analytical treatment has been used to model the vapour transport of radioactive fission products released into the fuel-to-sheath gap of defective nuclear fuel elements. The model accounts for both diffusive and bulk-convective transport. Convective transport becomes important as a result of a significant release of gaseous fission products into the gap during a high-temperature reactor accident. However, during normal reactor operation, diffusion is shown to be the dominant process of transport. The model is based on an analysis of several in-reactor tests with operating defective fuel elements, and high-temperature annealing experiments with irradiated fuel specimens. ((orig.))

  10. Traversing the mountaintop: world fossil fuel production to 2050.

    Science.gov (United States)

    Nehring, Richard

    2009-10-27

    During the past century, fossil fuels--petroleum liquids, natural gas and coal--were the dominant source of world energy production. From 1950 to 2005, fossil fuels provided 85-93% of all energy production. All fossil fuels grew substantially during this period, their combined growth exceeding the increase in world population. This growth, however, was irregular, providing for rapidly growing per capita production from 1950 to 1980, stable per capita production from 1980 to 2000 and rising per capita production again after 2000. During the past half century, growth in fossil fuel production was essentially limited by energy demand. During the next half century, fossil fuel production will be limited primarily by the amount and characteristics of remaining fossil fuel resources. Three possible scenarios--low, medium and high--are developed for the production of each of the fossil fuels to 2050. These scenarios differ primarily by the amount of ultimate resources estimated for each fossil fuel. Total fossil fuel production will continue to grow, but only slowly for the next 15-30 years. The subsequent peak plateau will last for 10-15 years. These production peaks are robust; none of the fossil fuels, even with highly optimistic resource estimates, is projected to keep growing beyond 2050. World fossil fuel production per capita will thus begin an irreversible decline between 2020 and 2030.

  11. Accident analysis of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.; Chi, D. Y

    1998-03-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. The HANARO fuel test loop was designed to match the CANDU and PWR fuel operating conditions. The accident analysis was performed by RELAP5/MOD3 code based on FTL system designs and determined the detail engineering specification of in-pile test section and out-pile systems. The accident analysis results of FTL system could be used for the fuel and materials designer to plan the irradiation testing programs. (author). 23 refs., 20 tabs., 178 figs.

  12. Locomotive fuel tank structural safety testing program : passenger locomotive fuel tank jackknife derailment load test.

    Science.gov (United States)

    2010-08-01

    This report presents the results of a passenger locomotive fuel tank load test simulating jackknife derailment (JD) load. The test is based on FRA requirements for locomotive fuel tanks in the Title 49, Code of Federal Regulations (CFR), Part 238, Ap...

  13. Bio-diesel fuels production: Feasibility studies

    International Nuclear Information System (INIS)

    Tabasso, L.

    1993-01-01

    This paper reviews the efforts being made by Italy's national government and private industry to develop diesel engine fuels derived from vegetable oils, in particular, sunflower seed oil. These fuels are being promoted in Italy from the environmental protection stand-point in that they don't contain any sulfur, the main cause of acid rain, and from the agricultural stand-point in that they provide Italian farmers, whose food crop production capacity is limited due to European Communities agreements, with the opportunity to use their set-aside land for the production of energy crops. This paper provides brief notes on the key performance characteristics of bio-diesel fuels, whose application doesn't require any modifications to diesel engines, apart from minor adjustments to the air/fuel mix regulating system, and assesses commercialization prospects. Brief mention is made of the problems being encountered by the Government in the establishing fair bio-fuel production tax rebates which are compatible with the marketing practices of the petroleum industry. One of the strategies being considered is to use the bio-fuels as additives to be mixed with conventional fuel oils so as to derive a fuel which meets the new European air pollution standards

  14. HANARO fuel irradiation test (II): revision

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, D. S.; Kim, H.; Chae, H. T.; Lee, C. S.; Kim, B. G.; Lee, C. B

    2001-04-01

    In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiated at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%. This report is the revision of KAERI/TR-1816/2001 on the irradiation test for HANARO fuel.

  15. Fuel Retrieval Sub-Project (FRS) Stuck Fuel Station Performance Test Data Report

    International Nuclear Information System (INIS)

    THIELGES, J.R.

    2000-01-01

    This document provides the test data report for Stuck Fuel Station Performance Testing in support of the Fuel Retrieval Sub-Project. The stuck fuel station was designed to provide a means of cutting open a canister barrel to release fuel elements, etc

  16. Refinery Integration of By-Products from Coal-Derived Jet Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Caroline E. Burgess Clifford; Andre Boehman; Chunshan Song; Bruce Miller; Gareth Mitchell

    2006-05-17

    This report summarizes the accomplishments toward project goals during the first six months of the third year of the project to assess the properties and performance of coal based products. These products are in the gasoline, diesel and fuel oil range and result from coal based jet fuel production from an Air Force funded program. Specific areas of progress include generation of coal based material that has been fractionated into the desired refinery cuts, acquisition and installation of a research gasoline engine, and modification of diesel engines for use in evaluating diesel produced in the project. Characterization of the gasoline fuel indicates a dominance of single ring alkylcycloalkanes that have a low octane rating; however, blends containing these compounds do not have a negative effect upon gasoline when blended in refinery gasoline streams. Characterization of the diesel fuel indicates a dominance of 3-ring aromatics that have a low cetane value; however, these compounds do not have a negative effect upon diesel when blended in refinery diesel streams. The desulfurization of sulfur containing components of coal and petroleum is being studied so that effective conversion of blended coal and petroleum streams can be efficiently converted to useful refinery products. Equipment is now in place to begin fuel oil evaluations to assess the quality of coal based fuel oil. Combustion and characterization of fuel oil indicates that the fuel is somewhere in between a No. 4 and a No. 6 fuel oil. Emission testing indicates the fuel burns similarly to these two fuels, but trace metals for the coal-based material are different than petroleum-based fuel oils. Co-coking studies using cleaned coal are highly reproducible in the pilot-scale delayed coker. Evaluation of the coke by Alcoa, Inc. indicated that while the coke produced is of very good quality, the metals content of the carbon is still high in iron and silica. Coke is being evaluated for other possible uses

  17. Development of Micro-welding Technology of Cladding Tube with Temperature Sensor for Nuclear Fuel Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Soo Sung; Lee, C. Y.; Kim, W. K.; Lee, J. W.; Lee, D. Y

    2006-01-15

    Laser welding technology is widely used to fabricate some products of nuclear fuel in the nuclear industry. Especially, micro-laser welding is one of the key technology to be developed to fabricate precise products of fuel irradiation test. We have to secure laser welding technology to perform various instrumentations for fuel irradiation test. The instrumented fuel irradiation test at a research reactor is needed to evaluate the performance of the developed nuclear fuel. The fuel elements can be designed to measure the center line temperature of fuel pellets during the irradiation test by using temperature sensor. The thermal sensor was composed of thermocouple and sensor sheath. Micro-laser welding technology was adopted to seal between seal tube and sensor sheath with thickness of 0.15mm. The soundness of weld area has to be confirmed to prevent fission gas of the fuel from leaking out of the element during the fuel irradiation test. In this study, fundamental data for micro-laser welding technology was proposed to seal temperature sensor sheath of the instrumented fuel element. And, micro-laser welding for dissimilar metals between sensor sheath and seal tube was characterized by investigating welding conditions. Moreover, the micro-laser welding technology is closely related to advanced industry. It is expected that the laser material processing technology will be adopted to various applications in the industry.

  18. Recent metal fuel safety tests in TREAT

    International Nuclear Information System (INIS)

    Wright, A.E.; Bauer, T.H.; Lo, R.K.; Robinson, W.R.; Palm, R.G.

    1986-01-01

    In-reactor safety tests have been performed on metal-alloy reactor fuel to study its response to transient-overpower conditions, in particular, the margin to cladding breach and the axial self-extrusion of fuel within intact cladding. Uranium-fissium EBR-II driver fuel elements of several burnups were tested, some to cladding breach and others to incipient breach. Transient fuel motions were monitored, and time and location of breach were measured. The test results and computations of fuel extrusion and cladding failure in metal-alloy fuel are described

  19. Fuels and materials testing capabilities in Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Baker, R.B.; Chastain, S.A.; Culley, G.E.; Ethridge, J.L.; Lovell, A.J.; Newland, D.J.; Pember, L.A.; Puigh, R.J.; Waltar, A.E.

    1989-01-01

    The Fast Flux Test Facility (FFTF) reactor, which started operating in 1982, is a 400 MWt sodium-cooled fast neutron reactor located in Hanford, Washington State, and operated by Westinghouse Hanford Co. under contract with U.S. Department of Energy. The reactor has a wide variety of functions for irradiation tests and special tests, and its major purpose is the irradiation of fuel and material for liquid metal reactor, nuclear reactor and space reactor projects. The review first describes major technical specifications and current conditions of the FFTF reactor. Then the plan for irradiation testing is outlined focusing on general features, fuel pin/assembly irradiation tests, and absorber irradiation tests. Assemblies for special tests include the material open test assembly (MOTA), fuel open test assembly (FOTA), closed loop in-reactor assembly (CLIRA), and other special fuel assemblies. An interim examination and maintenance cell (FFTF/IEM cell) and other hot cells are used for nondestructive/destructive tests and physical/mechanical properties test of material after irradiation. (N.K.)

  20. Fuel fragmentation data review and separate effects testing

    International Nuclear Information System (INIS)

    Yueh, Ken. H.; Snis, N.; Mitchell, D.; Munoz-Reja, C.

    2014-01-01

    A simple alternative test has been developed to study the fuel fragmentation process at loss of coolant accident (LOCA) temperatures. The new test heats a short section of fuel, approximately two pellets worth of material, in a tube furnace open to air. An axial slit is cut in the test sample cladding to reduce radial restraint and to simulate ballooned condition. The tube furnace allows the fuel fragmentation process be observed during the experiment. The test was developed as a simple alternative so large number of tests could be conducted quickly and efficiently to identify key variables that influence fuel fragmentation and to zeroing on the fuel fragmentation burn-up threshold. Several tests were conducted, using fuel materials from fuel rods that were used in earlier integral tests to benchmark and validate the test technique. High burn-up fuel materials known to be above the fragmentation threshold was used to evaluate the fragmentation process as a function of temperature. Even with an axial slit and both ends open, no significant fuel detachment/release was detected until above 750°C. Additional tests were conducted with fuel materials at burn-ups closer to the fuel fragmentation burn-up threshold. Results from these tests indicate a minor power history effect on the fuel fragmentation burn-up threshold. An evaluation of available literature and data generated from this work suggest a fuel fragmentation burn-up threshold between 70 and 75 GWd/MTU. (author)

  1. Results from Cycles 1 and 2 of NNWSI Series 2 spent fuel dissolution tests

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1987-05-01

    PWR spent fuel rod segments from the H.B. Robinson Unit 2 and Turkey Point Unit 3 reactors were leach tested in Nevada Nuclear Waste Storage Investigations (NNWSI) reference J-13 water under ambient hot cell conditions. The test matrix included bare fuel plus the cladding, rod segments with artificially induced cladding defects, and undefected rod segments. Radionuclide release results are presented and discussed. The actinides Pu, Am, Cm and Np appear to have been released congruently as the UO 2 oxide fuel matrix dissolved. Preferential U release measured in certain tests may be related to dissolution of oxidized UO/sub 2+x/ from the fuel surface, and/or greater solubility (and mobility) of U relative to the other actinides within defected cladding specimens. Uranium solubility measured in the J-13 water was much greater then that measured in deionized water in previous tests. All of the principal fission products analyzed ( 137 Cs, 129 I, 99 Tc and 90 Sr) were released preferentially relative to the actinides. Preferential release of activation product 14 C was also measured, with a portion of the 14 C release appearing to originate from the cladding exterior surface. Much greater fractional fuel dissolution appeared to have occurred with bare fuel particles than from fuel contained in defected cladding. Actinide release from test specimens containing small (∼200 μm) laser-drilled holes through the cladding was not significantly greater than from undefected specimens

  2. Fast Flux Test Facility fuel and test management: The first 10 years

    International Nuclear Information System (INIS)

    Bennett, R.A.; Bennett, C.L.; Campbell, L.R.; Dobbin, K.D.; Tang, E.L.

    1991-07-01

    Core design and fuel and test management have been performed efficiently at the Fast Flux Test Facility. No outages have been extended to adjust core loadings. Development of mixed oxide fuels for advanced liquid metal breeder reactors has been carried out successfully. In fact, the fuel performance is extraordinary. Failures have been so infrequent that further development and refinement of fuel requirements seem appropriate and could lead to a significant reduction in projected electrical busbar costs. The Fast Flux Test Facility is also involved in early metal fuel development tests and appears to be an ideal test bed for any further fuel development or refinement testing. 3 refs., 4 figs., 2 tabs

  3. Radiochemical analyses of several spent fuel Approved Testing Materials

    International Nuclear Information System (INIS)

    Guenther, R.J.; Blahnik, D.E.; Wildung, N.J.

    1994-09-01

    Radiochemical characterization data are described for UO 2 and UO 2 plus 3 wt% Gd 2 O 3 commercial spent nuclear fuel taken from a series of Approved Testing Materials (ATMs). These full-length nuclear fuel rods include MLA091 of ATM-103, MKP070 of ATM-104, NBD095 and NBD131 of ATM-106, and ADN0206 of ATM-108. ATMs 103, 104, and 106 were all irradiated in the Calvert Cliffs Nuclear Power Plant (Reactor No.1), a pressurized-water reactor that used fuel fabricated by Combustion Engineering. ATM-108 was part of the same fuel bundle designed as ATM-105 and came from boiling-water reactor fuel fabricated by General Electric and irradiated in the Cooper Nuclear Power Plant. Rod average burnups and expected fission gas releases ranged from 2,400 to 3,700 GJ/kgM. (25 to 40 Mwd/kgM) and from less than 1% to greater than 10%, respectively, depending on the specific ATM. The radiochemical analyses included uranium and plutonium isotopes in the fuel, selected fission products in the fuel, fuel burnup, cesium and iodine on the inner surfaces of the cladding, 14 C in the fuel and cladding, and analyses of the gases released to the rod plenum. Supporting examinations such as fuel rod design and material descriptions, power histories, and gamma scans used for sectioning diagrams are also included. These ATMs were examined as part of the Materials Characterization Center Program conducted at Pacific Northwest Laboratory provide a source of well-characterized spent fuel for testing in support of the US Department of Energy Office of Civilian Radioactive Waste Management Program

  4. Grain boundary sweeping and liquefaction-induced fission product behavior in nuclear fuel under severe-core damage accident conditions

    International Nuclear Information System (INIS)

    Rest, J.

    1984-05-01

    The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, tellurium, and cesium release from: (1) irradiated high-burnup LWR fuel in a flowing steam atmosphere during high-temperature, in-cell heating tests performed at Oak Ridge National Laboratory; and (2) trace-irradiated and high-burnup LWR fuel during severe-fuel-damage (SFD) tests performed in the PBF reactor in Idaho. A theory of grain boundary sweeping of gas bubbles, gas bubble behavior during fuel liquefaction (destruction of grain boundaries due to formation of a U-rich melt phase), and U-Zr eutectic melting has been included within the FASTGRASS-VFP formalism. Results of the analyses demonstrate that intragranular fission product behavior during both types of tests can be interpreted in terms of a grain-growth/grain-boundary-sweeping mechanism that enhances the flow of fission products from within the grains to the grain boundaries. Whereas fuel liquefaction leads to an enhanced release of fission products in trace-irradiated fuel, the occurrence of fuel liquefaction in high-burnup fuel can degrade fission product release. This phenomenon is due in part to reduced gas-bubble mobilities in a viscous medium as compared to vapor transport, and in part to a degradation of grain growth rates and the subsequent decrease in grain-boundary sweeping of intragranular fission products into the liquefied lamina. The analysis shows that total UO 2 dissolution due to eutectic melting leads to increased release for both trace-irradiated and high-burnup fuel. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in high-burnup fuel are highlighted

  5. 14 CFR 29.965 - Fuel tank tests.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank tests. 29.965 Section 29.965 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Fuel System § 29.965 Fuel tank tests. (a) Each fuel tank...

  6. 14 CFR 27.965 - Fuel tank tests.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank tests. 27.965 Section 27.965 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: NORMAL CATEGORY ROTORCRAFT Powerplant Fuel System § 27.965 Fuel tank tests. (a) Each fuel tank...

  7. MCO Engineering Test Report Fuel Basket Handling Grapple Acceptance Test

    International Nuclear Information System (INIS)

    CHENAULT, D.M.

    2000-01-01

    Acceptance testing of the production SNF Fuel Basket lift grapples to the required 150 percent maximum lift load is documented herein. The report shows the results affirming the proof test passage. The primary objective of this test was to confirm the load rating of the grapple per applicable requirements of ANSI 14 6 American National Standard For Radioactive Materials Special Lifting Devices for Shipping Containers Weighing 10,000 pounds (4500kg) or More. The above Standard requires a load test of 150% of the design load which must be held for a minimum of 10 minutes followed by a Liquid Penetrant or Magnetic Particle examination of critical areas and welds in accordance with the ANSI/ASME Boiler and Pressure Vessel Code 1989 Section 111 Division 1 section NF 5350

  8. Optimum nuclear design of target fuel rod for Mo-99 production in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun [Kyung Hee University, Seoul (Korea)

    1998-04-01

    Nuclear target design for Mo-99 production in HANARO was performed, KAERI proposed target design was analyzed and its feasibility was shown. Three commercial target designs of Cintichem, ANL and KAERI were tested for the HANARO irradiation an d they all satisfied with design specification. A parametric study was done for target design options and Mo-99 yields ratio and surface heat flux were compared. Tested parameters were target fuel thickness, irradiation location, target axial length, packing density of powder fuel, size of target radius, target geometry, fuel enrichment, fuel composition, and cladding material. Optimized target fuel was designed for both LEU and HEU options. (author). 17 refs., 33 figs., 42 tabs.

  9. Legacy Vehicle Fuel System Testing with Intermediate Ethanol Blends

    Energy Technology Data Exchange (ETDEWEB)

    Davis, G. W.; Hoff, C. J.; Borton, Z.; Ratcliff, M. A.

    2012-03-01

    The effects of E10 and E17 on legacy fuel system components from three common mid-1990s vintage vehicle models (Ford, GM, and Toyota) were studied. The fuel systems comprised a fuel sending unit with pump, a fuel rail and integrated pressure regulator, and the fuel injectors. The fuel system components were characterized and then installed and tested in sample aging test rigs to simulate the exposure and operation of the fuel system components in an operating vehicle. The fuel injectors were cycled with varying pulse widths during pump operation. Operational performance, such as fuel flow and pressure, was monitored during the aging tests. Both of the Toyota fuel pumps demonstrated some degradation in performance during testing. Six injectors were tested in each aging rig. The Ford and GM injectors showed little change over the aging tests. Overall, based on the results of both the fuel pump testing and the fuel injector testing, no major failures were observed that could be attributed to E17 exposure. The unknown fuel component histories add a large uncertainty to the aging tests. Acquiring fuel system components from operational legacy vehicles would reduce the uncertainty.

  10. Design criteria and fabrication in-pile test section of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1997-10-01

    Safety state fuel test loop will be equipped in HANARO to obtain the development and betterments of advanced fuel and materials through the irradiation tests. The objective of this study is to determine the design criteria and technical specification of in-pile test section and to specify the manufacturing requirements of in-pile test section. HANARO fuel test loop was designed to meet the CANDU and PWR fuel testing and in-pile section will be manufactured and installed in HANARO. The design criteria and technical specification of in-pile test section could be used the fuel and materials design with for irradiation testing IPS of HANARO fuel test loop. This results will become guidances for the planning and programming of irradiation testing. (author). 12 refs., tabs., figs.

  11. MOLTEN CARBONATE FUEL CELL PRODUCT DESIGN IMPROVEMENT

    Energy Technology Data Exchange (ETDEWEB)

    H.C. Maru; M. Farooque

    2005-03-01

    The program was designed to advance the carbonate fuel cell technology from full-size proof-of-concept field test to the commercial design. DOE has been funding Direct FuelCell{reg_sign} (DFC{reg_sign}) development at FuelCell Energy, Inc. (FCE, formerly Energy Research Corporation) from an early state of development for stationary power plant applications. The current program efforts were focused on technology and system development, and cost reduction, leading to commercial design development and prototype system field trials. FCE, in Danbury, CT, is a world-recognized leader for the development and commercialization of high efficiency fuel cells that can generate clean electricity at power stations, or at distributed locations near the customers such as hospitals, schools, universities, hotels and other commercial and industrial applications. FCE has designed three different fuel cell power plant models (DFC300A, DFC1500 and DFC3000). FCE's power plants are based on its patented DFC{reg_sign} technology, where a hydrocarbon fuel is directly fed to the fuel cell and hydrogen is generated internally. These power plants offer significant advantages compared to the existing power generation technologies--higher fuel efficiency, significantly lower emissions, quieter operation, flexible siting and permitting requirements, scalability and potentially lower operating costs. Also, the exhaust heat by-product can be used for cogeneration applications such as high-pressure steam, district heating and air conditioning. Several sub-MW power plants based on the DFC design are currently operating in Europe, Japan and the US. Several one-megawatt power plant design was verified by operation on natural gas at FCE. This plant is currently installed at a customer site in King County, WA under another US government program and is currently in operation. Because hydrogen is generated directly within the fuel cell module from readily available fuels such as natural gas and

  12. Refinery Integration of By-Products from Coal-Derived Jet Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Caroline E. Burgess Clifford; Andre' Boehman; Chunshan Song; Bruce Miller; Gareth Mitchell

    2006-09-17

    This report summarizes the accomplishments toward project goals during the second six months of the third year of the project to assess the properties and performance of coal based products. These products are in the gasoline, diesel and fuel oil range and result from coal based jet fuel production from an Air Force funded program. Specific areas of progress include generation of coal based material that has been fractionated into the desired refinery cuts and examination of carbon material, the use of a research gasoline engine to test coal-based gasoline, and modification of diesel engines for use in evaluating diesel produced in the project. At the pilot scale, the hydrotreating process was modified to separate the heavy components from the LCO and RCO fractions before hydrotreating in order to improve the performance of the catalysts in further processing. Characterization of the gasoline fuel indicates a dominance of single ring alkylcycloalkanes that have a low octane rating; however, blends containing these compounds do not have a negative effect upon gasoline when blended in refinery gasoline streams. Characterization of the diesel fuel indicates a dominance of 3-ring aromatics that have a low cetane value; however, these compounds do not have a negative effect upon diesel when blended in refinery diesel streams. Both gasoline and diesel continue to be tested for combustion performance. The desulfurization of sulfur containing components of coal and petroleum is being studied so that effective conversion of blended coal and petroleum streams can be efficiently converted to useful refinery products. Activated carbons have proven useful to remove the heavy sulfur components, and unsupported Ni/Mo and Ni/Co catalysts have been very effective for hydrodesulfurization. Equipment is now in place to begin fuel oil evaluations to assess the quality of coal based fuel oil. Combustion and characterization of the latest fuel oil (the high temperature fraction of RCO

  13. Study of ignition, combustion, and production of harmful substances upon burning solid organic fuel at a test bench with a vortex chamber

    Science.gov (United States)

    Burdukov, A. P.; Chernetskiy, M. Yu.; Dekterev, A. A.; Anufriev, I. S.; Strizhak, P. A.; Greben'kov, P. Yu.

    2016-01-01

    Results of investigation of furnace processes upon burning of pulverized fuel at a test bench with a power of 5 MW are presented. The test bench consists of two stages with tangential air and pulverized coal feed, and it is equipped by a vibrocentrifugal mill and a disintegrator. Such milling devices have an intensive mechanical impact on solid organic fuel, which, in a number of cases, increases the reactivity of ground material. The processes of ignition and stable combustion of a mixture of gas coal and sludge (wastes of concentration plant), as well as Ekibastus coal, ground in the disintegrator, were studied at the test bench. The results of experimental burning demonstrated that preliminary fuel grinding in the disintegrator provides autothermal combustion mode even for hardly inflammable organic fuels. Experimental combustion of biomass, wheat straw with different lignin content (18, 30, 60%) after grinding in the disintegrator, was performed at the test bench in order to determine the possibility of supporting stable autothermal burning. Stable biofuel combustion mode without lighting by highly reactive fuel was achieved in the experiments. The influence of the additive GTS-Powder (L.O.M. Leaders Co., Ltd., Republic of Korea) in the solid and liquid state on reducing sulfur oxide production upon burning Mugun coal was studied. The results of experimental combustion testify that, for an additive concentration from 1 to 15% of the total mass of the burned mixture, the maximum SO2 concentration reduction in ejected gases was not more than 18% with respect to the amount for the case of burning pure coal.

  14. Grain boundary sweeping and dissolution effects on fission product behaviour under severe fuel damage accident conditions

    International Nuclear Information System (INIS)

    Rest, J.

    1986-01-01

    The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, tellurium, and cesium release from severe-fuel-damage (SFD) tests performed in the PBF reactor in Idaho. A theory of grain boundary sweeping of gas bubbles, gas bubble behavior during fuel liquefaction (destruction of grain boundaries due to formation of a U-rich melt phase), and during U-Zr eutectic melting has been included within the FASTGRASS-VFP formalism. The grain-boundary-sweeping theory considers the interaction between the moving grain boundary and two distinct size classes of bubbles, those on grain faces and on grain edges. The theory of the effects of fuel liquefaction and U-Zr eutectic melting on fission product behaviour considers the migration and coalescence of fission gas bubbles in either molten uranium, or a Zircaloy-Uranium eutectic melt. Results of the analyses demonstrate that intragranular fission product behavior during the tests can be interpreted in terms of a grain-growth/grain-boundary-sweeping mechanism that enhances the flow of fission products from within the grains to the grain boundaries. Whereas fuel liquefaction leads to an enhanced release of fission products in trace-irradiated fuel, the occurrence of fuel liquefaction in normally-irradiated fuel can degrade fission product release. This phenomenon is due in part to reduced gas-bubble mobilities in a viscous medium as compared to vapor transport, and in part to a degradation of grain growth rates and the subsequent decrease in grain-boundary sweeping of intragranular fission products into the liquified lamina. The analysis shows that total UO 2 dissolution due to eutectic melting leads to increased release for both trace-irradiated and normally-irradiated fuel. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in normally

  15. Evaluation of burnup characteristics and energy deposition during NSRR pulse irradiation tests on irradiated BWR fuels

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio

    2000-11-01

    Pulse irradiation tests of irradiated fuel are performed in the Nuclear Safety Research Reactor (NSRR) to investigate the fuel behavior under Reactivity Initiated Accident Conditions (RIA). The severity of the RIA is represented by energy deposition or peak fuel enthalpy during the power excursion. In case of the irradiated fuel tests, the energy deposition varies depending both on the amounts and distribution of residual fissile and neutron absorbing fission products generated during the base irradiation. Thus, proper fuel burnup characterization, especially for low enriched commercial fuels, is important, because plutonium (Pu) takes a large part of fissile and its generation depends on the neutron spectrum during the base irradiation. Fuel burnup calculations were conducted with ORIGEN2, RODBURN and SWAT codes for the BWR fuels tested in the NSRR. The calculation results were compared with the measured isotope concentrations and used for the NSRR neutron calculations to evaluate energy depositions of the test fuel. The comparison of the code calculations and the measurements revealed that the neutron spectrum change due to difference in void fraction altered Pu generation and energy deposition in the NSRR tests considerably. With the properly evaluated neutron spectrum, the combined burnup and NSRR neutron calculation gave reasonably good evaluation of the energy deposition. The calculations provided radial distributions of the fission product accumulation during the base irradiation and power distribution during the NSRR pulse irradiation, which were important for the evaluation of both burnup characteristics and fission gas release behavior. (author)

  16. Fuel Cell Development and Test Laboratory | Energy Systems Integration

    Science.gov (United States)

    Facility | NREL Fuel Cell Development and Test Laboratory Fuel Cell Development and Test Laboratory The Energy System Integration Facility's Fuel Cell Development and Test Laboratory supports fuel cell research and development projects through in-situ fuel cell testing. Photo of a researcher running

  17. Particle fuel bed tests

    International Nuclear Information System (INIS)

    Horn, F.L.; Powell, J.R.; Savino, J.M.

    1985-01-01

    Gas-cooled reactors, using packed beds of small diameter coated fuel particles have been proposed for compact, high-power systems. The particulate fuel used in the tests was 800 microns in diameter, consisting of a thoria kernel coated with 200 microns of pyrocarbon. Typically, the bed of fuel particles was contained in a ceramic cylinder with porous metallic frits at each end. A dc voltage was applied to the metallic frits and the resulting electric current heated the bed. Heat was removed by passing coolant (helium or hydrogen) through the bed. Candidate frit materials, rhenium, nickel, zirconium carbide, and zirconium oxide were unaffected, while tungsten and tungsten-rhenium lost weight and strength. Zirconium-carbide particles were tested at 2000 K in H 2 for 12 hours with no visible reaction or weight loss

  18. Technical specification of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y

    1998-03-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. HANARO fuel test loop was designed in accordance with the same code and standards of nuclear power plant because HANARO FTL will be operated the high pressure and temperature same as nuclear power plant operation conditions. The objective of this study is to confirm the operation limit, safety limit, operation condition and checking points of HANARO fuel test loop. This results will become guidances for the planning of irradiation testing and operation of HANARO fuel test loop. (author). 13 refs., 13 tabs., 8 figs.

  19. Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

    Energy Technology Data Exchange (ETDEWEB)

    Pasqualini, E.E. [Laboratorio de Nanotecnología Nuclear, Centro Atómico Constituyentes, Comisión Nacional de Energía Atómica, Av. General Paz 1499, B1650KNA, San Martín, Prov. Buenos Aires (Argentina); Robinson, A.B. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Porter, D.L., E-mail: Douglas.Porter@inl.gov [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Wachs, D.M. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Finlay, M.R. [Australian Nuclear Science and Technology Organisation, PMB 1, Menai, NSW, 2234 (Australia)

    2016-10-15

    Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U–(7–10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry–4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction—either from fabrication or in-reactor testing—and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm{sup 3}, 3.8E+21 (peak).

  20. The role of spent fuel test facilities in the fuel cycle strategy

    International Nuclear Information System (INIS)

    Huang, S. T.; Gross, D. L.; Snyder, N. W.; Woods, W. D.

    1988-01-01

    Disposal of commercial spent nuclear fuels in the major industrialized countries may be categorized into two broad approaches: a once-through policy which will dispose of spent fuels and recycle fissile materials. Within reprocess spent fuels and recycle fissile materials. Within each policy, various technical, licensing, institutional and public issues exist. These issues tend to complicate the formulation of an effective and acceptable fuel cycle strategy which will meet various cost, schedule, and legislative constraints. This paper examines overall fuel cycle strategies from the viewpoint of these underlying technical issues and assesses the roles of spent fuel test facilities in the overall fuel cycles steps. Basic functions of such test facilities are also discussed. The main emphasis is placed on the once-through policy although the reprocessing / recycle policy is also discussed. Benefits of utilizing test facilities in the fuel cycle strategies are explored. The results indicate that substantial benefits may be obtained in terms of minimizing programmatic risks, increasing public confidence, and more effective utilization of overall budgetary resources by structuring and highlighting the test facilities as an important element in the overall strategy

  1. Advances in Automated QA/QC for TRISO Fuel Particle Production

    International Nuclear Information System (INIS)

    Hockey, Ronald L.; Bond, Leonard J.; Batishko, Charles R.; Gray, Joseph N.; Saurwein, John J.; Lowden, Richard A.

    2004-01-01

    Fuel in most Generation IV reactor designs typically encompasses billions of the TRISO particles. Present day QA/QC methods, done manually and in many cases destructively, cannot economically test a statistically significant fraction of the large number of the individual fuel particles required. Fully automated inspection technologies are essential to economical TRISO fuel particle production. A combination of in-line nondestructive (NDE) measurements employing electromagnetic induction and digital optical imaging analysis is currently under investigation and preliminary data indicate the potential for meeting the demands of this application. To calibrate high-speed NDE methods, surrogate fuel particle samples are being coated with layers containing a wide array of defect types found to degrade fuel performance and these are being characterized via high-resolution CT and digital radiographic images

  2. Development And Testing Of Biogas-Petrol Blend As An Alternative Fuel For Spark Ignition Engine

    Directory of Open Access Journals (Sweden)

    Awogbemi

    2015-08-01

    Full Text Available Abstract This research is on the development and testing of a biogas-petrol blend to run a spark ignition engine. A2080 ratio biogaspetrol blend was developed as an alternative fuel for spark ignition engine test bed. Petrol and biogas-petrol blend were comparatively tested on the test bed to determine the effectiveness of the fuels. The results of the tests showed that biogas petrol blend generated higher torque brake power indicated power brake thermal efficiency and brake mean effective pressure but lower fuel consumption and exhaust temperature than petrol. The research concluded that a spark ignition engine powered by biogas-petrol blend was found to be economical consumed less fuel and contributes to sanitation and production of fertilizer.

  3. NSRR experiment with un-irradiated uranium-zirconium hydride fuel. Design, fabrication process and inspection data of test fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; Kuroha, Hiroshi; Ikeda, Yoshikazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Aizawa, Keiichi

    1998-08-01

    An experiment plan is progressing in the Nuclear Safety Research Reactor (NSRR) to perform pulse-irradiation with uranium-zirconium hydride (U-ZrH{sub x}) fuel. This fuel is widely used in the training research and isotope production reactor of GA (TRIGA). The objectives of the experiment are to determine the fuel rod failure threshold and to investigate fuel behavior under simulated reactivity initiated accident (RIA) conditions. This report summarizes design, fabrication process and inspection data of the test fuel rods before pulse-irradiation. The experiment with U-ZrH{sub x} fuel will realize precise safety evaluation, and improve the TRIGA reactor performance. The data to be obtained in this program will also contribute development of next-generation TRIGA reactor and its safety evaluation. (author)

  4. Characterization of Cassini GPHS fueled clad production girth welds

    International Nuclear Information System (INIS)

    Franco-Ferreira, E.A.; Moyer, M.W.; Reimus, M.A.H.; Placr, A.; Howard, B.D.

    2000-01-01

    Fueled clads for radioisotope power systems are produced by encapsulating 238 PuO 2 in iridium alloy cups, which are joined at their equators by gas tungsten arc welding. Cracking problems at the girth weld tie-in area during production of the Galileo/Ulysses GPHS capsules led to the development of a first-generation ultrasonic test for girth weld inspection at the Savannah River Plant. A second-generation test and equipment with significantly improved sensitivity and accuracy were jointly developed by the Oak Ridge Y-12 Plant and Westinghouse Savannah River Company for use during the production of Cassini GPHS capsules by the Los Alamos National Laboratory. The test consisted of Lamb wave ultrasonic scanning of the entire girth weld from each end of the capsule combined with a time-of-flight evaluation to aid in characterizing nonrelevant indications. Tangential radiography was also used as a supplementary test for further evaluation of reflector geometry. Each of the 317 fueled GP HS capsules, which were girth welded for the Cassini Program, was subjected to a series of nondestructive tests that included visual, dimensional, helium leak rate, and ultrasonic testing. Thirty-three capsules were rejected prior to ultrasonic testing. Of the 44 capsules rejected by the standard ultrasonic test, 22 were upgraded to flight quality through supplementary testing for an overall process acceptance rate of 82.6%. No confirmed instances of weld cracking were found

  5. 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

  6. Production costs of liquid fuels from biomass

    International Nuclear Information System (INIS)

    Bridgwater, A.V.; Double, J.M.

    1994-01-01

    This project was undertaken to provide a consistent and thorough review of the full range of processes for producing liquid fuels from biomass to compare both alternative technologies and processes within those technologies in order to identify the most promising opportunities that deserve closer attention. Thermochemical conversion includes both indirect liquefaction through gasification, and direct liquefaction through pyrolysis and liquefaction in pressurized solvents. Biochemical conversion is based on a different set of feedstocks. Both acid and enzyme hydrolysis are included followed by fermentation. The liquid products considered include gasoline and diesel hydrocarbons and conventional alcohol fuels of methanol and ethanol. Results are given both as absolute fuel costs and as a comparison of estimated cost to market price. In terms of absolute fuel costs, thermochemical conversion offers the lowest cost products, with the least complex processes generally having an advantage. Biochemical routes are the least attractive. The most attractive processes from comparing production costs to product values are generally the alcohol fuels which enjoy a higher market value. (author)

  7. F2 phenomenological test on fuel motion (Interim report)

    International Nuclear Information System (INIS)

    Palm, R.G.; Fink, C.L.; Stewart, R.R.; Gehl, S.M.; Rothman, A.B.

    1976-09-01

    TREAT F-series tests are being conducted to provide data on fuel motion at accident power levels from one to about ten times design for use in development of fuel motion models. Test F2 was conducted to evaluate motion of high power fuel in a hypothetical LMFBR unprotected TUC (transient undercooling) accident. Fuel and fuel-boundary conditions following coolant boiling and dryout under TUC conditions are achieved in each F-series test with a single fuel element surrounded by a nuclear heated wall in a dry test capsule. Test F2 was conducted with a low burnup but restructured fuel element to investigate the effect of fuel vapor pressure on fuel motion. Results are presented and discussed

  8. Biological production of liquid fuels from biomass

    Energy Technology Data Exchange (ETDEWEB)

    None

    1982-01-01

    A scheme for the production of liquid fuels from renewable resources such as poplar wood and lignocellulosic wastes from a refuse hydropulper was investigated. The particular scheme being studied involves the conversion of a cellulosic residue, resulting from a solvent delignified lignocellulosic feed, into either high concentration sugar syrups or into ethyl and/or butyl alcohol. The construction of a pilot apparatus for solvent delignifying 150 g samples of lignocellulosic feeds was completed. Also, an analysis method for characterizing the delignified product has been selected and tested. This is a method recommended in the Forage Fiber Handbook. Delignified samples are now being prepared and tested for their extent of delignification and susceptibility to enzyme hydrolysis. Work is continuing on characterizing the cellulase and cellobiase enzyme systems derived from the YX strain of Thermomonospora.

  9. Impact of fuel chemistry on fission product behaviour

    International Nuclear Information System (INIS)

    Poortmans, C.; Van Uffelen, P.; Van den Berghe, S.

    1999-01-01

    The report contains a series of papers presented at SCK-CEN's workshop on the impact of fuel chemistry on fission product behaviour. Contributing authors discuss different processes affecting the behaviour of fission products in different types of spent nuclear fuel. In addition, a number of papers discusses the behaviour of actinides and fission products released from spent fuel and vitrified high-level waste in geological disposal conditions

  10. Materials and processes for solar fuel production

    CERN Document Server

    Viswanathan, Balasubramanian; Lee, Jae Sung

    2014-01-01

    This book features different approaches to non-biochemical pathways for solar fuel production. This one-of-a-kind book addresses photovoltaics, photocatalytic water splitting for clean hydrogen production and CO2 conversion to hydrocarbon fuel through in-depth comprehensive contributions from a select blend of established and experienced authors from across the world. The commercial application of solar based systems, with particular emphasis on non-PV based devices have been discussed. This book intends to serve as a primary resource for a multidisciplinary audience including chemists, engineers and scientists providing a one-stop location for all aspects related to solar fuel production. The material is divided into three sections: Solar assisted water splitting to produce hydrogen; Solar assisted CO2 utilization to produce green fuels and Solar assisted electricity generation. The content strikes a balance between theory, material synthesis and application with the central theme being solar fuels.

  11. 30 CFR 36.50 - Tests of fuel tank.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Tests of fuel tank. 36.50 Section 36.50 Mineral... Requirements § 36.50 Tests of fuel tank. The fuel tank shall be inspected and tested to determine whether: (a) It is fuel-tight, (b) the vent maintains atmospheric pressure within the tank, and (c) the vent and...

  12. Fuel Cell Stations Automate Processes, Catalyst Testing

    Science.gov (United States)

    2010-01-01

    Glenn Research Center looks for ways to improve fuel cells, which are an important source of power for space missions, as well as the equipment used to test fuel cells. With Small Business Innovation Research (SBIR) awards from Glenn, Lynntech Inc., of College Station, Texas, addressed a major limitation of fuel cell testing equipment. Five years later, the company obtained a patent and provided the equipment to the commercial world. Now offered through TesSol Inc., of Battle Ground, Washington, the technology is used for fuel cell work, catalyst testing, sensor testing, gas blending, and other applications. It can be found at universities, national laboratories, and businesses around the world.

  13. Production of bio-jet fuel from microalgae

    Science.gov (United States)

    Elmoraghy, Marian

    The increase in petroleum-based aviation fuel consumption, the decrease in petroleum resources, the fluctuation of the crude oil price, the increase in greenhouse gas emission and the need for energy security are motivating the development of an alternate jet fuel. Bio-jet fuel has to be a drop in fuel, technically and economically feasible, environmentally friendly, greener than jet fuel, produced locally and low gallon per Btu. Bic jet fuel has been produced by blending petro-based jet fuel with microalgae biodiesel (Fatty Acid Methyl Ester, or simply FAME). Indoor microalgae growth, lipids extraction and transetrification to biodiesel are energy and fresh water intensive and time consuming. In addition, the quality of the biodiesel product and the physical properties of the bio-jet fuel blends are unknown. This work addressed these challenges. Minimizing the energy requirements and making microalgae growth process greener were accomplished by replacing fluorescent lights with light emitting diodes (LEDs). Reducing fresh water footprint in algae growth was accomplished by waste water use. Microalgae biodiesel production time was reduced using the one-step (in-situ transestrification) process. Yields up to 56.82 mg FAME/g dry algae were obtained. Predicted physical properties of in-situ FAME satisfied European and American standards confirming its quality. Lipid triggering by nitrogen deprivation was accomplished in order to increase the FAME production. Bio-jet fuel freezing points and heating values were measured for different jet fuel to biodiesel blend ratios.

  14. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-02-15

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

  15. 48 CFR 908.7109 - Fuels and packaged petroleum products.

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Fuels and packaged petroleum products. 908.7109 Section 908.7109 Federal Acquisition Regulations System DEPARTMENT OF ENERGY....7109 Fuels and packaged petroleum products. Acquisitions of fuel and packaged petroleum products by DOE...

  16. CANFLEX fuel bundle strength tests (test report)

    International Nuclear Information System (INIS)

    Chang, Seok Kyu; Chung, C. H.; Kim, B. D.

    1997-08-01

    This document outlines the test results for the strength tests of the CANFLEX fuel bundle. Strength tests are performed to determine and verify the amount of the bundle shape distortion which is against the side-stops when the bundles are refuelling. There are two cases of strength test; one is the double side-stop test which simulates the normal bundle refuelling and the other is the single side-stop test which simulates the abnormal refuelling. the strength test specification requires that the fuel bundle against the side-stop(s) simulators for this test were fabricated and the flow rates were controlled to provide the required conservative hydraulic forces. The test rig conditions of 120 deg C, 11.2 MPa were retained for 15 minutes after the flow rate was controlled during the test in two cases, respectively. The bundle loading angles of number 13- number 15 among the 15 bundles were 67.5 deg CCW and others were loaded randomly. After the tests, the bundle shapes against the side-stops were measured and inspected carefully. The important test procedures and measurements were discussed as follows. (author). 5 refs., 22 tabs., 5 figs

  17. CANFLEX fuel bundle impact test

    International Nuclear Information System (INIS)

    Chang, Seok Kyu; Chung, C. H.; Park, J. S.; Hong, S. D.; Kim, B. D.

    1997-08-01

    This document outlines the test results for the impact test of the CANFLEX fuel bundle. Impact test is performed to determine and verify the amount of general bundle shape distortion and defect of the pressure tube that may occur during refuelling. The test specification requires that the fuel bundles and the pressure tube retain their integrities after the impact test under the conservative conditions (10 stationary bundles with 31kg/s flow rate) considering the pressure tube creep. The refuelling simulator operating with pneumatic force and simulated shield plug were fabricated and the velocity/displacement transducer and the high speed camera were also used in this test. The characteristics of the moving bundle (velocity, displacement, impacting force) were measured and analyzed with the impact sensor and the high speed camera system. The important test procedures and measurement results were discussed as follows. 1) Test bundle measurements and the pressure tube inspections 2) Simulated shield plug, outlet flange installation and bundle loading 3) refuelling simulator, inlet flange installation and sensors, high speed camera installation 4) Perform the impact test with operating the refuelling simulator and measure the dynamic characteristics 5) Inspections of the fuel bundles and the pressure tube. (author). 8 refs., 23 tabs., 13 figs

  18. 14 CFR 25.965 - Fuel tank tests.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank tests. 25.965 Section 25.965 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.965 Fuel tank tests. (a) It must be...

  19. Water for energy and fuel production

    CERN Document Server

    Shah, Yatish T

    2014-01-01

    Water, in all its forms, may be the key to an environmentally friendly energy economy. Water is free, there is plenty of it, plus it carries what is generally believed to be the best long-term source of green energy-hydrogen. Water for Energy and Fuel Production explores the many roles of water in the energy and fuel industry. The text not only discusses water's use as a direct source of energy and fuel-such as hydrogen from water dissociation, methane from water-based clathrate molecules, hydroelectric dams, and hydrokinetic energy from tidal waves, off-shore undercurrents, and inland waterways-but also: Describes water's benign application in the production of oil, gas, coal, uranium, biomass, and other raw fuels, and as an energy carrier in the form of hot water and steam Examines water's role as a reactant, reaction medium, and catalyst-as well as steam's role as a reactant-for the conversion of raw fuels to synthetic fuels Explains how supercritical water can be used to convert fossil- and bio-based feed...

  20. Modeling of in-vessel fission product release including fuel morphology effects for severe accident analyses

    International Nuclear Information System (INIS)

    Suh, K.Y.

    1989-10-01

    A new in-vessel fission product release model has been developed and implemented to perform best-estimate calculations of realistic source terms including fuel morphology effects. The proposed bulk mass transfer correlation determines the product of fission product release and equiaxed grain size as a function of the inverse fuel temperature. The model accounts for the fuel-cladding interaction over the temperature range between 770 K and 3000 K in the steam environment. A separate driver has been developed for the in-vessel thermal hydraulic and fission product behavior models that were developed by the Department of Energy for the Modular Accident Analysis Package (MAAP). Calculational results of these models have been compared to the results of the Power Burst Facility Severe Fuel Damage tests. The code predictions utilizing the mass transfer correlation agreed with the experimentally determined fractional release rates during the course of the heatup, power hold, and cooldown phases of the high temperature transients. Compared to such conventional literature correlations as the steam oxidation model and the NUREG-0956 correlation, the mass transfer correlation resulted in lower and less rapid releases in closer agreement with the on-line and grab sample data from the Severe Fuel Damage tests. The proposed mass transfer correlation can be applied for best-estimate calculations of fission products release from the UO 2 fuel in both nominal and severe accident conditions. 15 refs., 10 figs., 2 tabs

  1. RIA tests in CABRI with MOX fuel

    International Nuclear Information System (INIS)

    Schmitz, F.; Papin, J.; Gonnier, C.

    2000-01-01

    Three MOX-fuel tests have been successfully performed within the framework of the CABRI REP-Na test program. From the experimental findings which are presently available, no evidence for thermal effects resulting from the heterogeneous nature of the fuel can be given. There are very clear hints however that fission gas effects are enhanced with regard to the behaviour of UO 2 . The clad rupture observed in REP-Na 7 is of different nature than the failures observed in Cabri tests with UO 2 fuel. Failures of UO 2 fuel rods only occurred when the clad mechanical properties were severely affected by the presence of hydride blisters, while in REP-Na 7 a clear indication is made that the loading potential of the MOX fuel pellets was high enough to break a sound cladding. Concerning the transient fuel behaviour after reaching the critical heat-flux under reactor typical conditions (pressure, temperature and flow), no data base could be provided by the tests in the present sodium test loop (as for the UO 2 fuel behaviour). The IPSN project to implement into the Cabri reactor a pressurised water loop which will allow to simulate the complete RIA accident sequence under PWR reactor typical conditions, aims at providing this missing data base. (author)

  2. Test plan for K-Basin fuel handling tools

    International Nuclear Information System (INIS)

    Bridges, A.E.

    1995-01-01

    The purpose of this document is to provide the test plan and procedures for the acceptance testing of the handling tools enveloped for the removal of an N-Reactor fuel element from its storage canister in the K-Basins storage pool and insertion into the Single fuel Element Can for subsequent shipment to a Hot Cell for examination. Examination of these N-Reactor fuel elements is part of the overall characterization effort. New hand tools were required since previous fuel movement has involved grasping the fuel in a horizontal position. The 305 Building Cold Test Facility will be used to conduct the acceptance testing of the Fuel Handling Tools. Upon completion of this acceptance testing and any subsequent training of operators, the tools will be transferred to the 105 KW Basin for installation and use

  3. Composition and methods for improved fuel production

    Science.gov (United States)

    Steele, Philip H.; Tanneru, Sathishkumar; Gajjela, Sanjeev K.

    2015-12-29

    Certain embodiments of the present invention are configured to produce boiler and transportation fuels. A first phase of the method may include oxidation and/or hyper-acidification of bio-oil to produce an intermediate product. A second phase of the method may include catalytic deoxygenation, esterification, or olefination/esterification of the intermediate product under pressurized syngas. The composition of the resulting product--e.g., a boiler fuel--produced by these methods may be used directly or further upgraded to a transportation fuel. Certain embodiments of the present invention also include catalytic compositions configured for use in the method embodiments.

  4. BNFL pushes fuel production into the 21st century

    International Nuclear Information System (INIS)

    Odell, Mark.

    1992-01-01

    When the New Oxide Fuels Complex (NOFC) begins commercial production in May 1995, its operators will have one of the world's most advanced fuel production facilities, designed to compete on international markets by using the latest technology and exploiting economies of scale. British Nuclear Fuels (BNFL's) multi-million pound investment in its Springfields fuel operation is a bold move, aimed at gaining a greater share of the oxide fuel business worldwide. The complex is designed to achieve fuel cost savings and increase the site's manufacturing capacity. The building itself will bring all current fuel production operations at Springfields under one roof, merging five plants into one. Automation will ensure further economies through the use of automated guided vehicles (AGVs) for product transfer, walking beam furnaces for pellet sintering and on-line inspection and handling of pellets and pins. The use of a binderless route for AGR (advanced gas-cooled reactor) fuel means much of the manufacturing process for both AGR and PWR (pressurised water reactor) fuels will proceed on a single line, thereby reducing the fixed costs of both fuels. Two other features which contribute to cost savings are production flexibility which allows both non-irradiated and oxide reprocessed materials to follow the same route and longer kilns which increase the throughput. It is estimated that overall NOFC will reduce fuel manufacturing costs by some 40%. (author)

  5. Production of metal waste forms from spent fuel treatment

    International Nuclear Information System (INIS)

    Westphal, B.R.; Keiser, D.D.; Rigg, R.H.; Laug, D.V.

    1995-01-01

    Treatment of spent nuclear fuel at Argonne National Laboratory consists of a pyroprocessing scheme in which the development of suitable waste forms is being advanced. Of the two waste forms being proposed, metal and mineral, the production of the metal waste form utilizes induction melting to stabilize the waste product. Alloying of metallic nuclear materials by induction melting has long been an Argonne strength and thus, the transition to metallic waste processing seems compatible. A test program is being initiated to coalesce the production of the metal waste forms with current induction melting capabilities

  6. LEU WWR-M2 fuel assemblies burnable test

    International Nuclear Information System (INIS)

    Kirsanov, G.A.; Konoplev, K.A.; Pikulik, R.G.; Sajkov, Yu. P.; Tchmshkyan, D.V.; Tedoradze, L.V.; Zakharov, A.S.

    2000-01-01

    The results of in-pile irradiation tests of LEU WWR-M2 fuel assemblies with reduced enrichment of fuel are submitted in the report. The tests are made according to the Russian Program on Reduced Enrichment for Research and Test Reactors (RERTR). United States Department of Energy and the Ministry of Atomic Energy of Russian Federation jointly fund this Program. The irradiation tests of 5 WWR-M2 experimental assemblies are carried out at WWR-M reactor of the Petersburg Nuclear Physics Institute (PNPI). The information on assembly design and technique of irradiation tests is presented. In the irradiation tests the integrity of fuel assemblies is periodically measured. The report presents the data for the integrity maintained during the burnup of 5 fuel assemblies up to 45%. These results demonstrate the high reliability of the experimental fuel assemblies within the guaranteed burnup limits specified by the manufacturer. The tests are still in progress; it is planned to test and analyze the change in integrity for burnup of up to 70% - 75% or more. LEU WWR-M2 fuel assemblies are to be offered for export by their Novosibirsk manufacturer. Currently, HEU WWR-M2 fuel assemblies are used in Hungary, Ukraine and Vietnam. LEU WWR-M2 fuel assemblies were designed as a possible replacement for the HEU WWR-M2 fuel assemblies in those countries, but their use can be extended to other research reactors. (author)

  7. Maximum thermal loading test of BWR fuel assembly

    International Nuclear Information System (INIS)

    Nakajima, Yoshitaka; Yoshimura, Kunihiro; Nakamura, Satoshi; Ishizuka, Takao.

    1987-01-01

    Various proving tests on the reliability of nuclear power plants have been conducted at the Nuclear Power Engineering Test Center and at the Japan Power Plant Engineering and Inspection Corporation. The tests were initiated at the request of the Ministry of International Trade and Industry (MITI). Toshiba undertook one of the proving tests on the reliability of nuclear fuel assembly; the maximum thermal loading test of BWR fuel assembly from the Nuclear Power Engineering Test Center. These tests are part of the proving tests mentioned above, and their purpose is to confirm the reliability of the thermal hydraulic engineering techniques. Toshiba has been engaged for the past nine years in the design, fabrication and testing of the equipment. For the project, a test model fuel assembly was used to measure the critical power of the BWR fuel assembly and the void and fluidity of the coolant. From the test results, it has been confirmed that the heat is transferred safely from the fuel assembly to the coolant in the BWR nuclear power plant. In addition, the propriety and reliability of the thermal hydraulic engineering techniques for the fuel assembly have been proved. (author)

  8. FUELS IN SOIL TEST KIT: FIELD USE OF DIESEL DOG SOIL TEST KITS

    Energy Technology Data Exchange (ETDEWEB)

    Susan S. Sorini; John F. Schabron; Joseph F. Rovani, Jr.

    2002-09-30

    Western Research Institute (WRI) has developed a new commercial product ready for technology transfer, the Diesel Dog{reg_sign} Portable Soil Test Kit, for performing analysis of fuel-contaminated soils in the field. The technology consists of a method developed by WRI (U.S. Patents 5,561,065 and 5,976,883) and hardware developed by WRI that allows the method to be performed in the field (patent pending). The method is very simple and does not require the use of highly toxic reagents. The aromatic components in a soil extract are measured by absorption at 254 nm with a field-portable photometer. WRI added significant value to the technology by taking the method through the American Society for Testing and Materials (ASTM) approval and validation processes. The method is designated as ASTM Method D 5831-96, Standard Test Method for Screening Fuels in Soils. This ASTM designation allows the method to be used for federal compliance activities. In June 2001, the Diesel Dog technology won an American Chemical Society Regional Industrial Innovations Award. To gain field experience with the new technology, Diesel Dog kits have been used for a variety of site evaluation and cleanup activities. Information gained from these activities has led to improvements in hardware configurations and additional insight into correlating Diesel Dog results with results from laboratory methods. The Wyoming Department of Environmental Quality (DEQ) used Diesel Dog Soil Test Kits to guide cleanups at a variety of sites throughout the state. ENSR, of Acton, Massachusetts, used a Diesel Dog Portable Soil Test Kit to evaluate sites in the Virgin Islands and Georgia. ChemTrack and the U.S. Army Corps of Engineers successfully used a test kit to guide excavation at an abandoned FAA fuel-contaminated site near Fairbanks, Alaska. Barenco, Inc. is using a Diesel Dog Portable Soil Test Kit for site evaluations in Canada. A small spill of diesel fuel was cleaned up in Laramie, Wyoming using a Diesel

  9. Fuel Economy Testing and Data

    Science.gov (United States)

    EPA’s Fuel Economy pages provide information on current standards and how federal agencies work to enforce those laws, testing for national Corporate Average Fuel Economy or CAFE standards, and what you can do to reduce your own vehicle emissions.

  10. Accident situations tests HTR fuel with the device Kufa

    International Nuclear Information System (INIS)

    Kellerbauer, A. I.; Freis, D.

    2010-01-01

    The ceramic and ceramic-like coating materials in modern high-temperature reactor fuel are designed to ensure mechanical stability and retention of fission products under normal and transient conditions, regardless of the radiation damage sustained in-pile. In hypothetical depressurization and loss-of-forced-circulation (D LOFC) accidents, fuel elements of modular high-temperate reactors are exposed to temperatures several hundred degrees higher than during normal operation, causing increased thermo-mechanical stress on the coating layers. At the Institute for Transuranium Elements of the European Commission, a vigorous experimental program is being pursued with the aim of characterizing the performance of irradiated HTR fuel under such accident conditions. A cold finger device (Kufa), operational in ITUs hot cells since 2006, has been used to perform heating experiments on eight irradiated HTR fuel pebbles from the AVR experimental reactor and from dedicated irradiation campaigns at the High-Flux Reactor in Petten, the Netherlands. Gaseous fission products are collected in a cryogenic charcoal trap, while volatiles,are plated out on a water-cooled condensate plate. A quantitative measurement of the release is obtained by gamma spectroscopy. We highlight experimental results from the Kufa testing as well as the on-going development of new experimental facilities. (Author) 9 refs.

  11. Test plan for spent fuel cladding containment credit tests

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1983-11-01

    Lawrence Livermore National Laboratory has chosen Westinghouse Hanford Company as a subcontractor to assist them in determining the requirements for successful disposal of spent fuel rods in the proposed Nevada Test Site repository. An initial scoping test, with the objective of determining whether or not the cladding of a breached fuel rod can be given any credit as an effective barrier to radionuclide release, is described in this test plan. 8 references, 2 figures, 4 tables

  12. The environmentally friendly technology for bio fuel production

    International Nuclear Information System (INIS)

    Bekers, M.; Danilevics, A.; Guriniece, E.; Gulbis, V.

    2003-01-01

    Full text: Bio fuel production and use have been discussed this time in EC and in Latvia as alternative energy sources. The national resources allow producing liquid fuels - bio diesel and bi oethanol from rape seeds and grain correspondingly. Liquid bio fuels can be recommended especially for auto transport in big towns to reduce the pollution of air. A system for environmentally friendly production of bio fuel from agricultural raw materials has been developed, which permit a complex utilization of byproducts an wastes for obtaining of valuable food-stuffs and industrial products, providing the agricultural production requirements and supporting with local mineral fertilizers. Such a bio fuel production includes the agricultural and industrial productions in a united biotechnological system. Production objects of system interact: the products, by-products and wastes from one object are used as raw materials, auxiliary materials or heat carriers in other system's objects. This integrated agro-industrial production system would allow the production of feeds and chemical products, along with bio fuels. In this work, a model of a system for a conventional administrative rural region is presented, exemplified with the case of Latvia. The model is developed for three forms of bio fuel production, i.e. ethanol, bio diesel and biogas as local energy source. Bio diesel is produced using ethanol as transesterifying agent of rape-seed oil fatty acids. This bio diesel is a blend of rape-seed oil fatty acid ethyl esters (REE) and consists solely from renewable raw materials. The capacity of distillery of system is 40 million litters per year and bio diesel 35000 ton. Important for agriculture is protein reach press cakes the byproduct from bio diesel production (66000 t/y). This byproduct can be exported as well. Biogas reactors of system can be used for utilization of wastes from town if necessary. Recommended bio system occupates up to 150.000 ha of agriculture lands

  13. Comparison of fuel production costs for future transportation

    DEFF Research Database (Denmark)

    Ridjan, Iva; Mathiesen, Brian Vad; Connolly, David

    The purpose of this poster is to provide an overview of fuel production costs for two types of synthetic fuels – methanol and methane, along with comparable costs for first and second generation biodiesel, two types of second generation bioethanol, and biogas. The model analysed is a 100% renewable...... scenario of Denmark for 2050, where the data for the transport sector has been changed to estimate the fuel production costs for eight different fuel pathways....

  14. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Grover, S. Blaine

    2009-01-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  15. APEX nuclear fuel cycle for production of LWR fuel and elimination of radioactive waste

    International Nuclear Information System (INIS)

    Steinberg, M.; Powell, J.R.

    1981-08-01

    The development of a nuclear fission fuel cycle is proposed which eliminates all the radioactive fission product waste effluent and the need for geological-age high level waste storage and provides a long term supply of fissile fuel for an LWR power reactor economy. The fuel cycle consists of reprocessing LWR spent fuel (1 to 2 years old) to remove the stable nonradioactive (NRFP, e.g. lanthanides, etc.) and short-lived fission products SLFP e.g. half-lives of (1 to 2 years) and returning, in dilute form, the long-lived fission products, ((LLFPs, e.g. 30 y half-life Cs, Sr, and 10 y Kr, and 16 x 10 6 y I) and the transuranics (TUs, e.g. Pu, Am, Cm, and Np) to be refabricated into fresh fuel elements. Makeup fertile and fissile fuel are to be supplied through the use of a Spallator (linear accelerator spallation-target fuel-producer). The reprocessing of LWR fuel elements is to be performed by means of the Chelox process which consists of Airox treatment (air oxidation and hydrogen reduction) followed by chelation with an organic reagent (β-diketonate) and vapor distillation of the organometallic compounds for separation and partitioning of the fission products

  16. Irradiated test fuel shipment plan for the LWR MOX fuel irradiation test project

    International Nuclear Information System (INIS)

    Shappert, L.B.; Dickerson, L.S.; Ludwig, S.B.

    1998-01-01

    This document outlines the responsibilities of DOE, DOE contractors, the commercial carrier, and other organizations participating in a shipping campaign of irradiated test specimen capsules containing mixed-oxide (MOX) fuel from the Idaho National Engineering and Environmental Laboratory (INEEL) to the Oak Ridge National Laboratory (ORNL). The shipments described here will be conducted according to applicable regulations of the US Department of Transportation (DOT), US Nuclear Regulatory Commission (NRC), and all applicable DOE Orders. This Irradiated Test Fuel Shipment Plan for the LWR MOX Fuel Irradiation Test Project addresses the shipments of a small number of irradiated test specimen capsules and has been reviewed and agreed to by INEEL and ORNL (as participants in the shipment campaign). Minor refinements to data entries in this plan, such as actual shipment dates, exact quantities and characteristics of materials to be shipped, and final approved shipment routing, will be communicated between the shipper, receiver, and carrier, as needed, using faxes, e-mail, official shipping papers, or other backup documents (e.g., shipment safety evaluations). Any major changes in responsibilities or data beyond refinements of dates and quantities of material will be prepared as additional revisions to this document and will undergo a full review and approval cycle

  17. MTR fuel testing in BR2

    International Nuclear Information System (INIS)

    Jacquet, P.; Verwimp, A.; Wirix, S.

    2000-01-01

    New fuel design for MTR 's requires to be qualified under representative conditions, that is geometry, neutron spectrum, heat flux and thermo hydraulic conditions. An irradiation device for fuel plates has been designed to derive the maximum benefit from the BR2 irradiation capacities. The fuel plates can be easily extracted from their support during a shutdown to undergo additional tests. One of these tests is the measurement of the thickness changes along the fuel plate. To that purpose, a facility in the reactor water pool has been designed to measure the fuel swelling with an accuracy of 5 μm using inductive probes. At SCK-CEN, the full range of destructive and non-destructive PIE can be performed, including γ-scanning, wet sipping, surface examination and other methods. (author)

  18. Spent Nuclear Fuel (SNF) Project Product Specification

    International Nuclear Information System (INIS)

    PAJUNEN, A.L.

    2000-01-01

    The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major

  19. Spent Nuclear Fuel (SNF) Project Product Specification

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-12-07

    The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major

  20. Test requirements of locomotive fuel tank blunt impact tests

    Science.gov (United States)

    2013-10-15

    The Federal Railroad Administrations Office of Research : and Development is conducting research into passenger : locomotive fuel tank crashworthiness. A series of impact tests : are planned to measure fuel tank deformation under two types : of dy...

  1. 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)

  2. Product consistency test and toxicity characteristic leaching procedure results of the ceramic waste form from the electrometallurgical treatment process for spent fuel

    International Nuclear Information System (INIS)

    Johnson, S. G.; Adamic, M. L.: DiSanto, T.; Warren, A. R.; Cummings, D. G.; Foulkrod, L.; Goff, K. M.

    1999-01-01

    The ceramic waste form produced from the electrometallurgical treatment of sodium bonded spent fuel from the Experimental Breeder Reactor-II was tested using two immersion tests with separate and distinct purposes. The product consistency test is used to assess the consistency of the waste forms produced and thus is an indicator of a well-controlled process. The toxicity characteristic leaching procedure is used to determine whether a substance is to be considered hazardous by the Environmental Protection Agency. The proposed high level waste repository will not be licensed to receive hazardous waste, thus any waste forms destined to be placed there cannot be of a hazardous nature as defined by the Resource Conservation and Recovery Act. Results are presented from the first four fully radioactive ceramic waste forms produced and from seven ceramic waste forms produced from cold surrogate materials. The fully radioactive waste forms are approximately 2 kg in weight and were produced with salt used to treat 100 driver subassemblies of spent fuel

  3. Vibration test and endurance test for HANARO 36-element fuel assembly

    International Nuclear Information System (INIS)

    Ryu, Jeong Soo; Kim, Heon ll; Chung, Heung June

    1998-06-01

    Vibration test and endurance test for HANARO DU (depleted uranium) 36-element fuel assembly which was fabricated by KAERI were carried out based on the HANARO operation conditions. The endurance test of 22 days was added to the previous 18 days test. The vibration test was performed at various flow rates. Vibration frequency for the 36-element fuel assembly is between 11 to 14.5 Hz. And the maximum vibration displacement is less than 100 μm. From the endurance test result, it can be concluded that the appreciable fretting wear for the 36-element fuel assembly and the hexagonal flow tube was not observed. (author). 4 refs., 5 tabs., 29 figs

  4. Fuel-cycle assessment of selected bioethanol production

    International Nuclear Information System (INIS)

    Wu, M.; Wang, M.; Hong, H.

    2007-01-01

    A large amount of corn stover is available in the U.S. corn belt for the potential production of cellulosic bioethanol when the production technology becomes commercially ready. In fact, because corn stover is already available, it could serve as a starting point for producing cellulosic ethanol as a transportation fuel to help reduce the nation's demand for petroleum oil. Using the data available on the collection and transportation of corn stover and on the production of cellulosic ethanol, we have added the corn stover-to-ethanol pathway in the GREET model, a fuel-cycle model developed at Argonne National Laboratory. We then analyzed the life-cycle energy use and emission impacts of corn stover-derived fuel ethanol for use as E85 in flexible fuel vehicles (FFVs). The analysis included fertilizer manufacturing, corn farming, farming machinery manufacturing, stover collection and transportation, ethanol production, ethanol transportation, and ethanol use in light-duty vehicles (LDVs). Energy consumption of petroleum oil and fossil energy, emissions of greenhouse gases (carbon dioxide [CO 2 ], nitrous oxide [N 2 O], and methane [CH 4 ]), and emissions of criteria pollutants (carbon monoxide [CO], volatile organic compounds [VOCs], nitrogen oxide [NO x ], sulfur oxide [SO x ], and particulate matter with diameters smaller than 10 micrometers [PM 10 ]) during the fuel cycle were estimated. Scenarios of ethanol from corn grain, corn stover, and other cellulosic feedstocks were then compared with petroleum reformulated gasoline (RFG). Results showed that FFVs fueled with corn stover ethanol blends offer substantial energy savings (94-95%) relative to those fueled with RFG. For each Btu of corn stover ethanol produced and used, 0.09 Btu of fossil fuel is required. The cellulosic ethanol pathway avoids 86-89% of greenhouse gas emissions. Unlike the life cycle of corn grain-based ethanol, in which the ethanol plant consumes most of the fossil fuel, farming consumes most

  5. Fuel-cycle assessment of selected bioethanol production.

    Energy Technology Data Exchange (ETDEWEB)

    Wu, M.; Wang, M.; Hong, H.; Energy Systems

    2007-01-31

    A large amount of corn stover is available in the U.S. corn belt for the potential production of cellulosic bioethanol when the production technology becomes commercially ready. In fact, because corn stover is already available, it could serve as a starting point for producing cellulosic ethanol as a transportation fuel to help reduce the nation's demand for petroleum oil. Using the data available on the collection and transportation of corn stover and on the production of cellulosic ethanol, we have added the corn stover-to-ethanol pathway in the GREET model, a fuel-cycle model developed at Argonne National Laboratory. We then analyzed the life-cycle energy use and emission impacts of corn stover-derived fuel ethanol for use as E85 in flexible fuel vehicles (FFVs). The analysis included fertilizer manufacturing, corn farming, farming machinery manufacturing, stover collection and transportation, ethanol production, ethanol transportation, and ethanol use in light-duty vehicles (LDVs). Energy consumption of petroleum oil and fossil energy, emissions of greenhouse gases (carbon dioxide [CO{sub 2}], nitrous oxide [N{sub 2}O], and methane [CH{sub 4}]), and emissions of criteria pollutants (carbon monoxide [CO], volatile organic compounds [VOCs], nitrogen oxide [NO{sub x}], sulfur oxide [SO{sub x}], and particulate matter with diameters smaller than 10 micrometers [PM{sub 10}]) during the fuel cycle were estimated. Scenarios of ethanol from corn grain, corn stover, and other cellulosic feedstocks were then compared with petroleum reformulated gasoline (RFG). Results showed that FFVs fueled with corn stover ethanol blends offer substantial energy savings (94-95%) relative to those fueled with RFG. For each Btu of corn stover ethanol produced and used, 0.09 Btu of fossil fuel is required. The cellulosic ethanol pathway avoids 86-89% of greenhouse gas emissions. Unlike the life cycle of corn grain-based ethanol, in which the ethanol plant consumes most of the fossil

  6. Recent advances in fuel product and manufacturing process development

    International Nuclear Information System (INIS)

    Slember, R.J.; Doshi, P.K.

    1987-01-01

    This paper discusses advancements in commercial nuclear fuel products and manufacturing made by the Westinghouse Electric Corporation in response to the commercial nuclear fuel industry's demand for high reliability, increased plant availability and improved operating flexibility. The features and benefits of Westinghouse's most advanced fuel products--VANTAGE 5 for PWR plants and QUAD+ for BWR plants--are described, as well as 'high performance' fuel concepts now under development for delivery in the late 1980s. The paper also disusses the importance of in-process quality control throughout manufacturing towards reducing product variability and improving fuel reliability. (author)

  7. Steady-state irradiation testing of U-Pu-Zr fuel to >18% burnup

    International Nuclear Information System (INIS)

    Pahl, R.G.; Wisner, R.S.; Billone, M.C.; Hofman, G.L.

    1990-01-01

    Tests of austenitic stainless steel clad U-xP-10Zr fuel (x=o, 8, 19 wt. %) to peak burnups as high as 18.4 at. % have been completed in the EBR-II. Fuel swelling and fractional fission gas release are slowly increasing functions of burnup beyond 2 at. % burnup. Increasing plutonium content in the fuel reduces swelling and decreases the amount of fission gas which diffuses from fuel to plenum. LIFE-METAL code modelling of cladding strains is consistent with creep by fission gas loading and irradiation-induced swelling mechanisms. Fuel/cladding chemical interaction involves the ingress of rare-earth fission products. Constituent redistribution in the fuel had not limited steady-state performance. Cladding breach behavior at closure welds, in the gas plenum, and in the fuel column region have been benign events. 3 refs., 5 figs

  8. Physiological tolerance and stoichiometric potential of cyanobacteria for hydrocarbon fuel production.

    Science.gov (United States)

    Kämäräinen, Jari; Knoop, Henning; Stanford, Natalie J; Guerrero, Fernando; Akhtar, M Kalim; Aro, Eva-Mari; Steuer, Ralf; Jones, Patrik R

    2012-11-30

    Cyanobacteria are capable of directly converting sunlight, carbon dioxide and water into hydrocarbon fuel or precursors thereof. Many biological and non-biological factors will influence the ability of such a production system to become economically sustainable. We evaluated two factors in engineerable cyanobacteria which could potentially limit economic sustainability: (i) tolerance of the host to the intended end-product, and (ii) stoichiometric potential for production. Alcohols, when externally added, inhibited growth the most, followed by aldehydes and acids, whilst alkanes were the least inhibitory. The growth inhibition became progressively greater with increasing chain-length for alcohols, whilst the intermediate C6 alkane caused more inhibition than both C3 and C11 alkane. Synechocystis sp. PCC 6803 was more tolerant to some of the tested chemicals than Synechococcus elongatus PCC 7942, particularly ethanol and undecane. Stoichiometric evaluation of the potential yields suggested that there is no difference in the potential productivity of harvestable energy between any of the studied fuels, with the exception of ethylene, for which maximal stoichiometric yield is considerably lower. In summary, it was concluded that alkanes would constitute the best choice metabolic end-product for fuel production using cyanobacteria if high-yielding strains can be developed. Copyright © 2012 Elsevier B.V. All rights reserved.

  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. A comprehensive in-pile test of PWR fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang Rixin; Zhang Shucheng; Chen Dianshan (Academia Sinica, Beijing (China). Inst. of Atomic Energy)

    1991-02-01

    An in-pile test of PWR fuel bundle has been conducted in HWRR at IAE of China. This paper describes the structure of the test bundle (3x3-2), fabrication process and quality control of the fuel rod, irradiation conditions and the main Post Irradiation Examination (PIE) results. The test fuel bundle was irradiated under the PWR operation and water chemistry conditions with an average linear power of 381 W/cm and reached an average burnup of 25010 MWd/tU of the fuel bundle. After the test, destructive and non-destructive examination of the fuel rods was conducted at hot laboratories. The fission gas release was 10.4-23%. The ridge height of cladding was 3 to 8 {mu}m. The hydrogen content of the cladding was 80 to 140 ppm. The fuel stack height was increased by 2.9 to 3.3 mm. The relative irradiation growth was about 0.11 to 0.17% of the fuel rod length. During the irradiation test, no fuel rod failure or other abnormal phenomena had been found by the on-line fuel failure monitoring system of the test loop and water sampling analysis. The structure of the test fuel assembly was left undamaged without twist and detectable deformation. (orig.).

  11. Irradiation test and performance evaluation of DUPIC fuel

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Song, K. C.; Moon, J. S.

    2002-05-01

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

  12. Full-length U-xPu-10Zr (x = 0, 8, 19 wt.%) fast reactor fuel test in FFTF

    Energy Technology Data Exchange (ETDEWEB)

    Porter, D.L., E-mail: Douglas.Porter@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Tsai Hanchung [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439-4803 (United States)

    2012-08-15

    The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt.%) metallic fast reactor test with commercial-length (91.4-cm active fuel-column length) conducted to date. With few remaining test reactors, there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning-of-life (BOL) peak cladding temperature of the hottest pin was 608 Degree-Sign C, cooling to 522 Degree-Sign C at end-of-life (EOL). Selected fuel pins were examined non-destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta-gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3-cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at {approx}0.7 X/L axial location along the fuel column. This resulted from a higher production of rare-earth fission products at this location and a higher {Delta}T between fuel center and cladding than at core center, together providing more rare earths at the cladding and

  13. An Evaluation of a Fission Product Inventory for CANDU Fuels

    International Nuclear Information System (INIS)

    Jung, Jong Yeob; Park, Joo Hwan

    2007-01-01

    Fission products are released by two processes when a single channel accident occurs. One is a 'prompt release' and the other is a 'delayed release'. Prompt release assumes that the gap inventory of the fuel elements is released by a fuel element failure at the time of an accident. Delayed release assumes that the inventories within the grain or at the grain boundary are released after a accident due to a diffusion through grains, an oxidation of the fuel and an interaction between the fuel and the Zircaloy sheath. Therefore, the calculation of a fission product inventory and its distribution in a fuel during a normal operating is the starting point for the assessment of a fission product release for single channel accidents. In this report, the fission product inventories and their distributions within s fuel under a normal operating condition are evaluated for three types of CANDU fuels such as the 37 element fuel, CANFLEX-NU and CANFLEX-RU fuel bundles in the 'limiting channel'. To accomplish the above mentioned purposes, the basic power histories for each type of CANDU fuel were produced and the fission product inventories were calculated by using the ELESTRES code

  14. Visual observations of a degraded bundle of irradiated fuel: the Phebus FPT1 test

    International Nuclear Information System (INIS)

    Barrachin, M.; Bottomley, P.D.

    1999-01-01

    The international Phebus-FP (Fission Product) project is managed by the Institut de Protection et Surete Nucleaire in collaboration with Electricite de France (EDF), the European Commission (EC), the USNRC (USA), COG (Canada), NUPEC and JAERI (Japan), KAERI (South Korea), PSI and HSK (Switzerland). It is designed to measure the source-term and to study the degradation of irradiated UO 2 fuel in conditions typical of a severe loss of coolant accident in a pressurised water reactor (PWR). In the first test (FPT0), performed in December '93, a bundle of 20 fresh fuel rods and a central Ag-In-Cd control rod underwent a short 15-day irradiation to generate fission products before testing in the Phebus reactor in Cadarache. The second test (FPT1) was performed in July '96, in the same conditions and geometry, but using irradiated fuel (-23 GWd/tU). In the FPT1 test, the bundle was heated to an estimated 3000 K over a period of 30 minutes in order to induce a substantial liquefaction of the bundle. After the test, the bundle was embedded in epoxy and cut at different levels to investigate the mechanisms of the core degradation. This paper reports the visual observations of the degraded FPT1 bundle, very preliminary interpretations about the scenario of degradation and a comparison between the behaviour of the fuel in the FPT0 and FPT1 tests. (author)

  15. Hydrogen and synthetic fuel production using pressurized solid oxide electrolysis cells

    DEFF Research Database (Denmark)

    Jensen, Søren Højgaard; Sun, Xiufu; Ebbesen, Sune

    2010-01-01

    Wind and solar power is troubled by large fluctuations in delivery due to changing weather. The surplus electricity can be used in a Solid Oxide Electrolyzer Cell (SOEC) to split CO2 + H2O into CO + H2 (+O2). The synthesis gas (CO + H2) can subsequently be catalyzed into various types of synthetic...... fuels using a suitable catalyst. As the catalyst operates at elevated pressure the fuel production system can be simplified by operating the SOEC at elevated pressure. Here we present the results of a cell test with pressures ranging from 0.4 bar to 10 bar. The cell was tested both as an SOEC...

  16. On Cherenkov light production by irradiated nuclear fuel rods

    International Nuclear Information System (INIS)

    Branger, E.; Grape, S.; Svärd, S. Jacobsson; Jansson, P.; Sundén, E. Andersson

    2017-01-01

    Safeguards verification of irradiated nuclear fuel assemblies in wet storage is frequently done by measuring the Cherenkov light in the surrounding water produced due to radioactive decays of fission products in the fuel. This paper accounts for the physical processes behind the Cherenkov light production caused by a single fuel rod in wet storage, and simulations are presented that investigate to what extent various properties of the rod affect the Cherenkov light production. The results show that the fuel properties have a noticeable effect on the Cherenkov light production, and thus that the prediction models for Cherenkov light production which are used in the safeguards verifications could potentially be improved by considering these properties. It is concluded that the dominating source of the Cherenkov light is gamma-ray interactions with electrons in the surrounding water. Electrons created from beta decay may also exit the fuel and produce Cherenkov light, and e.g. Y-90 was identified as a possible contributor to significant levels of the measurable Cherenkov light in long-cooled fuel. The results also show that the cylindrical, elongated fuel rod geometry results in a non-isotropic Cherenkov light production, and the light component parallel to the rod's axis exhibits a dependence on gamma-ray energy that differs from the total intensity, which is of importance since the typical safeguards measurement situation observes the vertical light component. It is also concluded that the radial distributions of the radiation sources in a fuel rod will affect the Cherenkov light production.

  17. CANFLEX fuel bundle cross-flow endurance test (test report)

    International Nuclear Information System (INIS)

    Hong, Sung Deok; Chung, C. H.; Chang, S. K.; Kim, B. D.

    1997-04-01

    As part of the normal refuelling sequence of CANDU nuclear reactor, both new and irradiated bundles can be parked in the cross-flow region of the liner tubes. This situation occurs normally for a few minutes. The fuel bundle which is subjected to the cross-flow should be capable of withstanding the consequences of cross flow for normal periods, and maintain its mechanical integrity. The cross-flow endurance test was conducted for CANFLEX bundle, latest developed nuclear fuel, at CANDU-Hot Test Loop. The test was carried out during 4 hours at the inlet cross-flow region. After the test, the bundle successfully met all acceptance criteria after the 4 hours cross-flow test. (author). 2 refs., 3 tabs

  18. CANFLEX fuel bundle cross-flow endurance test (test report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Deok; Chung, C. H.; Chang, S. K.; Kim, B. D.

    1997-04-01

    As part of the normal refuelling sequence of CANDU nuclear reactor, both new and irradiated bundles can be parked in the cross-flow region of the liner tubes. This situation occurs normally for a few minutes. The fuel bundle which is subjected to the cross-flow should be capable of withstanding the consequences of cross flow for normal periods, and maintain its mechanical integrity. The cross-flow endurance test was conducted for CANFLEX bundle, latest developed nuclear fuel, at CANDU-Hot Test Loop. The test was carried out during 4 hours at the inlet cross-flow region. After the test, the bundle successfully met all acceptance criteria after the 4 hours cross-flow test. (author). 2 refs., 3 tabs.

  19. Fuel integrity project: analysis of light water reactor fuel rods test results

    International Nuclear Information System (INIS)

    Dallongeville, M.; Werle, J.; McCreesh, G.

    2004-01-01

    BNFL Nuclear Sciences and Technology Services and COGEMA LOGISTICS started in the year 2000 a joint project known as FIP (Fuel Integrity Project) with the aim of developing realistic methods by which the response of LWR fuel under impact accident conditions could be evaluated. To this end BNFL organised tests on both unirradiated and irradiated fuel pin samples and COGEMA LOGISTICS took responsibility for evaluating the test results. Interpretation of test results included simple mechanical analysis as well as simulation by Finite Element Analysis. The first tests that were available for analysis were an irradiated 3 point bending commissioning trial and a lateral irradiated hull compression test, both simulating the loading during a 9 m lateral regulatory drop. The bending test span corresponded roughly to a fuel pin intergrid distance. The outcome of the test was a failure starting at about 35 mm lateral deflection and a few percent of total deformation. Calculations were carried out using the ANSYS code employing a shell and brick model. The hull lateral compaction test corresponds to a conservative compression by neighbouring pins at the upper end of the fuel pin. In this pin region there are no pellets inside. The cladding broke initially into two and later into four parts, all of which were rather similar. Initial calculations were carried out with LS-DYNA3D models. The models used were optimised in meshing, boundary conditions and material properties. The calculation results compared rather well with the test data, in particular for the detailed ANSYS approach of the 3 point bending test, and allowed good estimations of stresses and deformations under mechanical loading as well as the derivation of material rupture criteria. All this contributed to the development of realistic numerical analysis methods for the evaluation of LWR fuel rod behaviour under both normal and accident transport conditions. This paper describes the results of the 3 point bending

  20. Fuel integrity project: analysis of light water reactor fuel rods test results

    Energy Technology Data Exchange (ETDEWEB)

    Dallongeville, M.; Werle, J. [COGEMA Logistics (AREVA Group) (France); McCreesh, G. [BNFL Nuclear Sciences and Technology Services (United Kingdom)

    2004-07-01

    BNFL Nuclear Sciences and Technology Services and COGEMA LOGISTICS started in the year 2000 a joint project known as FIP (Fuel Integrity Project) with the aim of developing realistic methods by which the response of LWR fuel under impact accident conditions could be evaluated. To this end BNFL organised tests on both unirradiated and irradiated fuel pin samples and COGEMA LOGISTICS took responsibility for evaluating the test results. Interpretation of test results included simple mechanical analysis as well as simulation by Finite Element Analysis. The first tests that were available for analysis were an irradiated 3 point bending commissioning trial and a lateral irradiated hull compression test, both simulating the loading during a 9 m lateral regulatory drop. The bending test span corresponded roughly to a fuel pin intergrid distance. The outcome of the test was a failure starting at about 35 mm lateral deflection and a few percent of total deformation. Calculations were carried out using the ANSYS code employing a shell and brick model. The hull lateral compaction test corresponds to a conservative compression by neighbouring pins at the upper end of the fuel pin. In this pin region there are no pellets inside. The cladding broke initially into two and later into four parts, all of which were rather similar. Initial calculations were carried out with LS-DYNA3D models. The models used were optimised in meshing, boundary conditions and material properties. The calculation results compared rather well with the test data, in particular for the detailed ANSYS approach of the 3 point bending test, and allowed good estimations of stresses and deformations under mechanical loading as well as the derivation of material rupture criteria. All this contributed to the development of realistic numerical analysis methods for the evaluation of LWR fuel rod behaviour under both normal and accident transport conditions. This paper describes the results of the 3 point bending

  1. Drop-in capsule testing of plutonium-based fuels in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Chang, G.S.; Ryskamp, J.M.; Terry, W.K.; Ambrosek, R.G.; Palmer, A.J.; Roesener, R.A.

    1996-09-01

    The most attractive way to dispose of weapons-grade plutonium (WGPu) is to use it as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PuO[sub 2]) mixed with urania (UO[sub 2]). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. The proposed weapons-grade MOX fuel is unusual, even relative to ongoing foreign experience with reactor-grade MOX power reactor fuel. Some demonstration of the in- reactor thermal, mechanical, and fission gas release behavior of the prototype fuel will most likely be required in a limited number of test reactor irradiations. The application to license operation with MOX fuel must be amply supported by experimental data. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) is capable of playing a key role in the irradiation, development, and licensing of these new fuel types. The ATR is a 250- MW (thermal) LWR designed to study the effects of intense radiation on reactor fuels and materials. For 25 years, the primary role of the ATR has been to serve in experimental investigations for the development of advanced nuclear fuels. Both large- and small-volume test positions in the ATR could be used for MOX fuel irradiation. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. Furthermore, these data can be obtained more quickly by using ATR instead of testing in a commercial LWR. Our previous work in this area has demonstrated that it is technically feasible to perform MOX fuel testing in the ATR. This report documents our analyses of sealed drop-in capsules containing plutonium-based test specimens placed in various ATR positions

  2. Compact methanol reformer test for fuel-cell powered light-duty vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Emonts, B; Hoehlein, B; Peters, R [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Energieverfahrenstechnik (IEV); Hansen, J B; Joergensen, S L [Haldor Topsoe A/S, Lyngby (Denmark)

    1998-03-15

    On-board production of hydrogen from methanol based on a steam reformer in connection with the use of low-temperature fuel-cells (PEMFC) is an attractive option as energy conversion unit for light-duty vehicles. A steam reforming process at higher pressures with an external burner offers advantages in comparison to a steam reformer with integrated partial oxidation in terms of total efficiency for electricity production. The main aim of a common project carried out by the Forschungszentrum Juelich (FZJ), Haldor Topsoee A/S (HTAS) and Siemens AG is to design, to construct and to test a steam reformer reactor concept (HTAS) with external catalytic burner (FZJ) as heat source as well as catalysts for heterogeneously catalyzed hydrogen production (HTAS), concepts for gas treatment (HTAS, FZJ) and a low-temperature fuel cell (Siemens). Based on the experimental results obtained so far concerning methanol reformers, catalytic burners and gas conditioning units, our report describes the total system, a test unit and preliminary test results related to a hydrogen production capacity of 50 kW (LHV) and dynamic operating conditions. This hydrogen production system is aimed at reducing the specific weight (<2 kg/kW{sub th} or 4 kg/kW{sub el}) combined with high efficiency for net electricity generation from methanol (about 50%) and low specific emissions. The application of Pd-membranes as gas cleaning unit fulfill the requirements with high hydrogen permeability and low cost of the noble metal. (orig.)

  3. Environmental costs of fossil fuel energy production

    International Nuclear Information System (INIS)

    Riva, A.; Trebeschi, C.

    1997-01-01

    The costs of environmental impacts caused by fossil fuel energy production are external to the energy economy and normally they are not reflected in energy prices. To determine the environmental costs associated with an energy source a detailed analysis of all environmental impacts of the complete energy cycle is required. The economic evaluation of environmental damages is presented caused by atmospheric emissions produced by fossil fuel combustion for different uses. Considering the emission factors of sulphur oxides, nitrogen oxides, dust and carbon dioxide and the economic evaluation of their environmental damages reported in literature, a range of environmental costs associated with different fossil fuels and technologies is presented. A comparison of environmental costs resulting from atmospheric emissions produced by fossil-fuel combustion for energy production shows that natural gas has a significantly higher environmental value than other fossil fuels. (R.P.)

  4. Bio-fuels production and the environmental indicators

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Marcos Sebastiao de Paula [Mechanical Engineering Department/Pontifical Catholic University of Rio de Janeiro - PUC-Rio, Rua Marques de Sao Vicente 225, Gavea, CEP 22453-900, Rio de Janeiro, RJ (Brazil); Muylaert de Araujo, Maria Silvia [Energy and Environment Planning Program/Federal University of Rio de Janeiro - COPPE/UFRJ, Cidade Universitaria, Centro de Tecnologia, Bloco C, sala 211, Ilha do Fundao, CEP: 21945-970, Caixa Postal: 68501, Rio de Janeiro, RJ (Brazil)

    2009-10-15

    The paper evaluates the role of the bio-fuels production in the transportation sector in the world, for programs of greenhouse gases emissions reductions and sustainable environmental performance. Depending on the methodology used to account for the local pollutant emissions and the global greenhouse gases emissions during the production and consumption of both the fossil and bio-fuels, the results can show huge differences. If it is taken into account a life cycle inventory approach to compare the different fuel sources, these results can present controversies. A comparison study involving the American oil diesel and soybean diesel developed by the National Renewable Energy Laboratory presents CO{sub 2} emissions for the bio-diesel which are almost 20% of the emissions for the oil diesel: 136 g CO{sub 2}/bhp-h for the bio-diesel from soybean and 633 g CO{sub 2}/bhp-h for the oil diesel [National Renewable Energy Laboratory - NREL/SR-580-24089]. Besides that, important local environmental impacts can also make a big difference. The water consumption in the soybean production is much larger in comparison with the water consumption for the diesel production [National Renewable Energy Laboratory - NREL/SR-580-24089]. Brazil has an important role to play in this scenario because of its large experience in bio-fuels production since the seventies, and the country has conditions to produce bio-fuels for attending great part of the world demand in a sustainable pathway. (author)

  5. Fission product induced swelling of U–Mo alloy fuel

    International Nuclear Information System (INIS)

    Kim, Yeon Soo; Hofman, G.L.

    2011-01-01

    Highlights: ► We measured fuel swelling of U–Mo alloy by fission products at temperatures below 250 °C. ► We quantified the swelling portion of U–Mo by fission gas bubbles. ► We developed an empirical model as a function of fission density. - Abstract: Fuel swelling of U–Mo alloy was modeled using the measured data from samples irradiated up to a fission density of ∼7 × 10 27 fissions/m 3 at temperatures below ∼250 °C. The overall fuel swelling was measured from U–Mo foils with as-fabricated thickness of 250 μm. Volume fractions occupied by fission gas bubbles were measured and fuel swelling caused by the fission gas bubbles was quantified. The portion of fuel swelling by solid fission products including solid and liquid fission products as well as fission gas atoms not enclosed in the fission gas bubbles is estimated by subtracting the portion of fuel swelling by gas bubbles from the overall fuel swelling. Empirical correlations for overall fuel swelling, swelling by gas bubbles, and swelling by solid fission products were obtained in terms of fission density.

  6. Straight-chain halocarbon forming fluids for TRISO fuel kernel productionTests with yttria-stabilized zirconia microspheres

    Energy Technology Data Exchange (ETDEWEB)

    Baker, M.P. [Nuclear Science and Engineering Program, Metallurgical and Materials Engineering Department, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States); King, J.C., E-mail: kingjc@mines.edu [Nuclear Science and Engineering Program, Metallurgical and Materials Engineering Department, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States); Gorman, B.P. [Metallurgical and Materials Engineering Department, Colorado Center for Advanced Ceramics, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States); Braley, J.C. [Nuclear Science and Engineering Program, Chemistry and Geochemistry Department, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States)

    2015-03-15

    Highlights: • YSZ TRISO kernels formed in three alternative, non-hazardous forming fluids. • Kernels characterized for size, shape, pore/grain size, density, and composition. • Bromotetradecane is suitable for further investigation with uranium-based precursor. - Abstract: Current methods of TRISO fuel kernel production in the United States use a sol–gel process with trichloroethylene (TCE) as the forming fluid. After contact with radioactive materials, the spent TCE becomes a mixed hazardous waste, and high costs are associated with its recycling or disposal. Reducing or eliminating this mixed waste stream would not only benefit the environment, but would also enhance the economics of kernel production. Previous research yielded three candidates for testing as alternatives to TCE: 1-bromotetradecane, 1-chlorooctadecane, and 1-iodododecane. This study considers the production of yttria-stabilized zirconia (YSZ) kernels in silicone oil and the three chosen alternative formation fluids, with subsequent characterization of the produced kernels and used forming fluid. Kernels formed in silicone oil and bromotetradecane were comparable to those produced by previous kernel production efforts, while those produced in chlorooctadecane and iodododecane experienced gelation issues leading to poor kernel formation and geometry.

  7. A method to evaluate fission gas release during irradiation testing of spherical fuel - HTR2008-58184

    International Nuclear Information System (INIS)

    Van Der Merwet, H.; Venter, J.

    2008-01-01

    The evaluation of fission gas release from spherical fuel during irradiation testing is critical to understand expected fuel performance under real reactor conditions. Online measurements of Krypton and Xenon fission products explain coated particle performance and contributions from graphitic matrix materials used in fuel manufacture and irradiation rig materials. Methods that are being developed to accurately evaluate fission gas release are described here together with examples of evaluations performed on irradiation tests HFR-K5, -K6 and EU1bis. (authors)

  8. Production of jet fuel from alternative source

    Energy Technology Data Exchange (ETDEWEB)

    Eller, Zoltan; Papp, Anita; Hancsok, Jenoe [Pannonia Univ., Veszprem (Hungary). MOL Dept. of Hydrocarbon and Coal Processing

    2013-06-01

    Recent demands for low aromatic content jet fuels have shown significant increase in the last 20 years. This was generated by the growing of aviation. Furthermore, the quality requirements have become more aggravated for jet fuels. Nowadays reduced aromatic hydrocarbon fractions are necessary for the production of jet fuels with good burning properties, which contribute to less harmful material emission. In the recent past the properties of gasolines and diesel gas oils were continuously severed, and the properties of jet fuels will be more severe, too. Furthermore, it can become obligatory to blend alternative components into jet fuels. With the aromatic content reduction there is a possibility to produce high energy content jet fuels with the desirable properties. One of the possibilities is the blending of biocomponents from catalytic hydrogenation of triglycerides. Our aim was to study the possibilities of producing low sulphur and aromatic content jet fuels in a catalytic way. On a CoMo/Al{sub 2}O{sub 3} catalyst we studied the possibilities of quality improving of a kerosene fraction and coconut oil mixture depending on the change of the process parameters (temperature, pressure, liquid hourly space velocity, volume ratio). Based on the quality parameters of the liquid products we found that we made from the feedstock in the adequate technological conditions products which have a high smoke point (> 35 mm) and which have reduced aromatic content and high paraffin content (90%), so these are excellent jet fuels, and their stack gases damage the environment less. (orig.)

  9. Characteristic test technology for PWR fuel and its components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho; Jeong, Yong Hwan; Park, Sang Yoon; Kim, Kyeng Ho; Nam, Cheol; Baek, Jong Hyuk; Lee, Myung Ho; Choi, Byoung Kwon; Song, Kun Woo; Kang, Ki Won; Kim, Keon Sik; Kim, Jong Hun; Kim, Young Min; Yang, Jae Ho; Song, Kee Nam; Kim, Hyung Kyu; Kang, Heung Seok; Yoon, Kyung Ho; Chun, Tae Hyun; In, Wang Kee; Oh, Dong Seok [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-01-01

    Characteristic tests of fuel assembly and its components being developed in the Advanced LWR Fuel Development Project supported by the mid-long term nuclear R and D program are described in this report. Performance verification of fuel and its components by the characteristic tests are essential to their development. Fuel components being developed in the Advanced LWR Fuel Development Project are zirconium alloy cladding, UO{sub 2} and burnable absorber pellets, spacer grid and top and bottom end pieces. Detailed test plans for those fuel components are described in this report, and test procedures of cladding and pellet are also described in the Appendix. Examples of the described tests are in- and out-of- pile corrosion and mechanical tests such as creep and burst tests for the cladding, in-pile capsule and ramp tests for the pellet, mechanical tests such as strength and vibration, and thermal-hydraulic tests such as pressure drop and critical heat flux for the spacer grid and top and bottom end pieces. It is expected that this report could be used as the standard reference for the performance verification tests in the development of LWR fuel and its components. 11 refs., 9 figs., 2 tabs. (Author)

  10. Parametric Sensitivity Tests- European PEM Fuel Cell Stack Test Procedures

    DEFF Research Database (Denmark)

    Araya, Samuel Simon; Andreasen, Søren Juhl; Kær, Søren Knudsen

    2014-01-01

    performed based on test procedures proposed by a European project, Stack-Test. The sensitivity of a Nafion-based low temperature PEMFC stack’s performance to parametric changes was the main objective of the tests. Four crucial parameters for fuel cell operation were chosen; relative humidity, temperature......As fuel cells are increasingly commercialized for various applications, harmonized and industry-relevant test procedures are necessary to benchmark tests and to ensure comparability of stack performance results from different parties. This paper reports the results of parametric sensitivity tests......, pressure, and stoichiometry at varying current density. Furthermore, procedures for polarization curve recording were also tested both in ascending and descending current directions....

  11. Fuel and fission product behaviour in early phases of a severe accident. Part II: Interpretation of the experimental results of the PHEBUS FPT2 test

    Energy Technology Data Exchange (ETDEWEB)

    Dubourg, R. [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); Barrachin, M., E-mail: marc.barrachin@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); Ducher, R. [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); Gavillet, D. [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); De Bremaecker, A. [Institute for Nuclear Materials Sciences, SCK-CEN, Boeretang 200, B-2400 Mol (Belgium)

    2014-10-15

    One objective of the FPT2 test of the PHEBUS FP Program was to study the degradation of an irradiated UO{sub 2} fuel bundle and the fission product behaviour under conditions of low steam flow. The results of the post-irradiation examinations (PIE) at the upper levels (823 mm and 900 mm) of the test section previously reported are interpreted in the present paper. Solid state interactions between fuel and cladding have been compared with the characteristics of interaction identified in the previous separate-effect tests. Corium resulting from the interaction between fuel and cladding was formed. The uranium concentration in the corium is compared to analytical tests and a scenario for the corium formation is proposed. The analysis showed that, despite the rather low fuel burn up, the conditions of temperature and oxygen potential reached during the starvation phase are able to give an early very significant release fraction of caesium. A significant part (but not all) of the molybdenum was segregated at grain boundaries and trapped in metallic inclusions from which they were totally removed in the final part of the experiment. During the steam starvation phase, the conditions of oxygen potential were favourable for the formation of simple Ba and BaO chemical forms but the temperature was too low to provoke their volatility. This is one important difference with out-of-pile experiments such as VERCORS for which only a combination of high temperature and low oxygen potential induced a significant barium release. Finally another significant difference with analytical out-of-pile experiments comes from the formation of foamy zones due to the fission gas presence in FPT2-type experiments which give an additional possibility for the formation of stable fission product compounds.

  12. Spent nuclear fuel project product specification

    International Nuclear Information System (INIS)

    Pajunen, A.L.

    1998-01-01

    Product specifications are limits and controls established for each significant parameter that potentially affects safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for transport to dry storage. The product specifications in this document cover the spent fuel packaged in MultiCanister Overpacks (MCOs) to be transported throughout the SNF Project. The SNF includes N Reactor fuel and single-pass reactor fuel. The FRS removes the SNF from the storage canisters, cleans it, and places it into baskets. The MCO loading system places the baskets into MCO/Cask assembly packages. These packages are then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the MCO cask packages are transferred to the Canister Storage Building (CSB), where the MCOs are removed from the casks, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The key criteria necessary to achieve these goals are documented in this specification

  13. Radioactive characteristics of spent fuels and reprocessing products in thorium fueled alternative cycles

    International Nuclear Information System (INIS)

    Maeda, Mitsuru

    1978-09-01

    In order to provide one fundamental material for the evaluation of Th cycle, compositions of the spent fuels were calculated with the ORIGEN code on following fuel cycles: (1) PWR fueled with Th- enriched U, (2) PWR fueled with Th-denatured U, (3) CANDU fueled with Th-enriched U and (4) HTGR fueled with Th-enriched U. Using these data, product specifications on radioactivity for their reprocessing were calculated, based on a criterion that radioactivities due to foreign elements do not exceed those inherent in nuclear fuel elements, due to 232 U in bred U or 228 Th in recovered Th, respectively. Conclusions are as the following: (1) Because of very high contents of 232 U and 228 Th in the Th cycle fuels from water moderated reactors, especially from PWR, required decontamination factors for their reprocessing will be smaller by a factor of 10 3 to 10 4 , compared with those from U-Pu fueled LWR cycle. (2) These less stringent product specifications on the radioactivity of bred U and recovered Th will justify introduction of some low decontaminating process, with additional advantage of increased proliferation resistance. (3) Decontamination factors required for HTGR fuel will be 10 to 30 times higher than for the other fuels, because of less 232 U and 228 Th generation, and higher burn-up in the fuel. (author)

  14. Life Cycle Analysis of the Production of FT-Fuels. 4 Different Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Blinge, M. [Chalmers University of Technology (Sweden); Rehnlund, B. [Atrax Energi AB (Sweden); Larsen, U.; Lundorf, P.; Ivarsson, A.; Schramm, J. [Technical University of Denmark (Denmark)

    2006-11-15

    This paper deals with aspects concerning the life cycle aspects regarding Fischer-Tropsch (FT) fuels. Four different scenarios are being analysed based on Life Cycle Assessment (LCA) figures. The results etc presented below emanates from a project undertaken by the International Energy Agency's (IEA) Implementing Agreement on Advanced Motor Fuels (IEA/AMF). The project has been carried out as an IEA/AMF annex, number XXXI, with financial support from the USA, Finland and Denmark. Some important results from the scenario studies based on the evaluated LCA data are: Production and use of GTL fuel has the potential of contributing about the same or slightly less greenhouse gas to the atmosphere than production and use of conventional diesel; Production and use of CTL emits more than twice as much greenhouse gases compared to traditional fuels; Production and use of BTL reduces the emissions of greenhouse gases by 60-90 %; To substitute 15 % of the EU 15 countries fuel consumption would an area of 310 000 km2 be cultivated with Salix. This corresponds to an area of the size of Poland. It would also require 122 FTplants of 1,6 GW; Theoretically, it is possible supply the worlds need for energy with biomass. However, planning the production, the localization of plants, building the infrastructure, this will take time and requires heavy long-term investments; The demand for Natural gas is increasing and there is no way for the US to meet an increased demand from supplying the vehicle fleet with F-T fuels from domestic reserves. With the political situation in the Middle East and in Venezuela, it doesn't seem likely that this solution will ease the US problems with reducing their oil dependences. The IEA/AMF project has also included emission tests on road vehicles fuelled by FT-Gasoline. These tests have been performed by The Technical University of Denmark and will be presented in another presentation at the ISAF XVI, 'Emissions from Road Vehicles Fuelled

  15. 25 years of NDE in fabrication of zirconium alloy mill products and nuclear fuel in the Nuclear Fuel Complex

    International Nuclear Information System (INIS)

    Mistry, R.K.; Laxminarayana, B.; Srivastava, R.K.

    1996-01-01

    Failure of nuclear fuel is highly undesirable from both economic and operational aspects. Hence all the components require rigorous QC and inspection checks. NDT plays a major role in assuring the quality of the products both at final and intermediate stages. This paper gives an overall review of NDT methods employed in achieving the integrity of nuclear products. The NDE procedures followed in NFC are visual inspection, radiography, penetrant testing, eddy current testing, ultrasonic testing and helium leak testing. NFC's quality assurance programme is organised to achieve the desired objectives by carrying out in process and final inspection at all critical steps of fabrication. (author)

  16. Full-length high-temperature severe fuel damage test No. 2

    International Nuclear Information System (INIS)

    Hesson, G.M.; Lombardo, N.J.; Pilger, J.P.; Rausch, W.N.; King, L.L.; Hurley, D.E.; Parchen, L.J.; Panisko, F.E.

    1993-09-01

    Hazardous conditions associated with performing the Full-Length High- Temperature (FLHT). Severe Fuel Damage Test No. 2 experiment have been analyzed. Major hazards that could cause harm or damage are (1) radioactive fission products, (2) radiation fields, (3) reactivity changes, (4) hydrogen generation, (5) materials at high temperature, (6) steam explosion, and (7) steam pressure pulse. As a result of this analysis, it is concluded that with proper precautions the FLHT- 2 test can be safely conducted

  17. Bio energy: Bio fuel - Properties and Production

    International Nuclear Information System (INIS)

    Wilhelmsen, Gunnar; Martinsen, Arnold Kyrre; Sandberg, Eiliv; Fladset, Per Olav; Kjerschow, Einar; Teslo, Einar

    2001-01-01

    This is Chapter 3 of the book ''Bio energy - Environment, technique and market''. Its main sections are: (1) Definitions and properties, (2) Bio fuel from the forest, (3) Processed bio fuel - briquettes, pellets and powder, (4) Bio fuel from agriculture, (5) Bio fuel from agro industry, (6) Bio fuel from lakes and sea, (7) Bio fuel from aquaculture, (8) Bio fuel from wastes and (9) Hydrogen as a fuel. The exposition largely describes the conditions in Norway. The chapter on energy from the forest includes products from the timber and sawmill industry, the pulp and paper industry, furniture factories etc. Among agricultural sources are straw, energy forests, vegetable oil, bio ethanol, manure

  18. Post irradiation examination on test fuel pins for PWR

    International Nuclear Information System (INIS)

    Fogaca Filho, N.; Ambrozio Filho, F.

    1981-01-01

    Certain aspects of irradiation technology on test fuel pins for PWR, are studied. The results of post irradiation tests, performed on test fuel pins in hot cells, are presented. The results of the tests permit an evaluation of the effects of irradiation on the fuel and cladding of the pin. (Author) [pt

  19. Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

    International Nuclear Information System (INIS)

    D.R. Jackson; G.R. Kiebel

    1999-01-01

    Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training

  20. Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

    Energy Technology Data Exchange (ETDEWEB)

    D.R. Jackson; G.R. Kiebel

    1999-08-24

    Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training.

  1. The design of in-pile test section for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, K. N.; Lee, J. M.; Shim, B. S.; Zee, D. Y.; Park, S. H.; Ahn, S. H.; Lee, J. Y.; Kim, Y. J. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    As an equipment for nuclear fuel's general performance irradiation test in HANARO, Fuel Test Loop(FTL) has been developed that can irradiate the pin to the maximum number of 3 at the core irradiation hole(IR1 hole) by considering for it's utility and user's irradiation requirement. 3-Pin FTL consists of In-Pile Test Section (IPS) and Out-of-Pile System (OPS). IPS consists for IPS Vessel assembly, In-Pool Piping, IPS Support, In-Pool Piping Support etc. Design that such IPS considers interference item consisted to do not bear in existing facilities by one. IVA that is connected to the OPS are controlled and regulated by means of system pressure, system temperature and the water quality. IPS Vessel assembly is consisted of outer pressure vessel, inner pressure vessel, IPS head, inner assembly and test fuel carrier. After 3-Pin FTL development which is expected to be finished by the 2006, FTL will be used for the irradiation test of the new PWR-type fuel and can maximize the usage of HANARO.

  2. Initial results from dissolution rate testing of N-Reactor spent fuel over a range of potential geologic repository aqueous conditions

    International Nuclear Information System (INIS)

    Gray, W.J.; Einziger, R.E.

    1998-04-01

    Hanford N-Reactor spent nuclear fuel (HSNF) may ultimately be placed in a geologic repository for permanent disposal. To determine whether the engineered barrier system that will be designed for emplacement of light-water-reactor (LWR) spent fuel will also suffice for HSNF, aqueous dissolution rate measurements were conducted on the HSNF. The purpose of these tests was to determine whether HSNF dissolves faster or slower than LWR spent fuel under some limited repository-relevant water chemistry conditions. The tests were conducted using a flowthrough method that allows the dissolution rate of the uranium matrix to be measured without interference by secondary precipitation reactions that would confuse interpretation of the results. Similar tests had been conducted earlier with LWR spent fuel, thereby allowing direct comparisons. Two distinct corrosion modes were observed during the course of these 12 tests. The first, Stage 1, involved no visible corrosion of the test specimen and produced no undissolved corrosion products. The second, Stage 2, resulted in both visible corrosion of the test specimen and left behind undissolved corrosion products. During Stage 1, the rate of dissolution could be readily determined because the dissolved uranium and associated fission products remained in solution where they could be quantitatively analyzed. The measured rates were much faster than has been observed for LWR spent fuel under all conditions tested to date when normalized to the exposed test specimen surface areas. Application of these results to repository conditions, however, requires some comparison of the physical conditions of the different fuels. The surface area of LWR fuel that could potentially be exposed to repository groundwater is estimated to be approximately 100 times greater than HSNF. Therefore, when compared on the basis of mass, which is more relevant to repository conditions, the HSNF and LWR spent fuel dissolve at similar rates

  3. Interim results from UO2 fuel oxidation tests in air

    International Nuclear Information System (INIS)

    Campbell, T.K.; Gilbert, E.R.; Thornhill, C.K.; White, G.D.; Piepel, G.F.; Griffin, C.W.j.

    1987-08-01

    An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to extend the characterization of spent fuel oxidation in air. To characterize oxidation behavior of irradiated UO 2 , fuel oxidation tests were performed on declad light-water reactor spent fuel and nonirradited UO 2 pellets in the temperature range of 135 to 250 0 C. These tests were designed to determine the important independent variables that might affect spent fuel oxidation behavior. The data from this program, when combined with the test results from other programs, will be used to develop recommended spent fuel dry-storage temperature limits in air. This report describes interim test results. The initial PNL investigations of nonirradiated and spent fuels identified the important testing variables as temperature, fuel burnup, radiolysis of the air, fuel microstructure, and moisture in the air. Based on these initial results, a more extensive statistically designed test matrix was developed to study the effects of temperature, burnup, and moisture on the oxidation behavior of spent fuel. Oxidation tests were initiated using both boiling-water reactor and pressurized-water reactor fuels from several different reactors with burnups from 8 to 34 GWd/MTU. A 10 5 R/h gamma field was applied to the test ovens to simulate dry storage cask conditions. Nonirradiated fuel was included as a control. This report describes experimental results from the initial tests on both the spent and nonirradiated fuels and results to date on the tests in a 10 5 R/h gamma field. 33 refs., 51 figs., 6 tabs

  4. Fuel Testing for Sylvatex: Cooperative Research and Development Final Report, CRADA Number CRD-16-636

    Energy Technology Data Exchange (ETDEWEB)

    Burton, Jonathan L. [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2018-02-07

    Sylvatex is a green nano-chemistry company that has developed a platform technology utilizing renewable, non-toxic inputs to create a stable nanoparticle that can be used in multiple applications. Their mission is to increase the use of renewables globally, to empower a cleaner and healthier future. The main application is a fuel technology product - MicroX - that utilizes proprietary knowledge to scale low-cost, cleaner-burning renewable diesel fuel and additives by using a co-location commercial model. The aspects of this project will include testing of two Sylvatex MicroX fuels on an engine dynamometer platform. Industry standard ultra-low sulfur diesel (ULSD) B3 fuel and a ULSD B20 will both be used for comparison of the Sylvatex fuels (U.S. standard diesel fuel at the pump contains an average of approximately 3% biodiesel; this is why B3 would be used as a baseline comparison). Sylvatex is currently using a prototype formulation (MicroX 1) that applies a high cost surfactant. An experimental formulation (MicroX 2) that uses lower cost materials is under development. The MicroX 1 will be blended at a 10% level into the B3 ULSD fuel and the MicroX 2 will be blended at a 10% level into both the B3 and the B20 ULSD fuels for study on the engine dynamometer test platform. All fuel blends will be tested over the FTP transient engine test cycle and a steady state ramped modal engine test cycle. Each test cycle will be performed a minimum of 3 times for each fuel. Tailpipe and/or engine out gaseous exhaust emissions (CO2, CO, NOx, THC, O2,), engine out PM emissions, and brake-specific fuel consumption rates will be evaluated for all test cycles.

  5. Fission products and nuclear fuel behaviour under severe accident conditions part 3: Speciation of fission products in the VERDON-1 sample

    Science.gov (United States)

    Le Gall, C.; Geiger, E.; Gallais-During, A.; Pontillon, Y.; Lamontagne, J.; Hanus, E.; Ducros, G.

    2017-11-01

    Qualitative and quantitative analyses on the VERDON-1 sample made it possible to obtain valuable information on fission product behaviour in the fuel during the test. A promising methodology based on the quantitative results of post-test characterisations has been implemented to assess the release fraction of non γ-emitter fission products. The order of magnitude of the estimated release fractions for each fission product was consistent with their class of volatility.

  6. Conventional fuel tank blunt impact tests : test and analysis results

    Science.gov (United States)

    2014-04-02

    The Federal Railroad Administrations Office of Research : and Development is conducting research into fuel tank : crashworthiness. A series of impact tests are planned to : measure fuel tank deformation under two types of dynamic : loading conditi...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-06-15

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

  8. Reprocessing free nuclear fuel production via fusion fission hybrids

    Energy Technology Data Exchange (ETDEWEB)

    Kotschenreuther, Mike, E-mail: mtk@mail.utexas.edu [Intitute for Fusion Studies, University of Texas at Austin (United States); Valanju, Prashant; Mahajan, Swadesh [Intitute for Fusion Studies, University of Texas at Austin (United States)

    2012-05-15

    Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively 'new' cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th{sup 232}-U{sup 233} conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO{sub 2} matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U{sup 235} fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.

  9. Reprocessing free nuclear fuel production via fusion fission hybrids

    International Nuclear Information System (INIS)

    Kotschenreuther, Mike; Valanju, Prashant; Mahajan, Swadesh

    2012-01-01

    Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively “new” cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th 232 –U 233 conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO 2 matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U 235 fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.

  10. Test reports for K Basins vertical fuel handling tools

    Energy Technology Data Exchange (ETDEWEB)

    Meling, T.A.

    1995-02-01

    The vertical fuel handling tools, for moving N Reactor fuel elements, were tested in the 305 Building Cold Test Facility (CTF) in the 300 Area. After fabrication was complete, the tools were functionally tested in the CTF using simulated N Reactor fuel rods (inner and outer elements). The tools were successful in picking up the simulated N Reactor fuel rods. These tools were also load tested using a 62 pound dummy to test the structural integrity of each assembly. The tools passed each of these tests, based on the performance objectives. Finally, the tools were subjected to an operations acceptance test where K Basins Operations personnel operated the tool to determine its durability and usefulness. Operations personnel were satisfied with the tools. Identified open items included the absence of a float during testing, and documentation required prior to actual use of the tools in the 100 K fuel storage basin.

  11. Seed production for fuel oils

    International Nuclear Information System (INIS)

    Mosca, G.

    1992-01-01

    With the aim of assessing commercialization prospects for vegetable oils to be used as diesel fuel alternatives, this paper provides maps indicating regional production quantities for soybean, rape and sunflower seeds in Italy. It then tables and discusses the results of energy input-output analyses carried out for rape and soybean oil production

  12. Results of industrial tests of carbonate additive to fuel oil

    Science.gov (United States)

    Zvereva, E. R.; Dmitriev, A. V.; Shageev, M. F.; Akhmetvalieva, G. R.

    2017-08-01

    Fuel oil plays an important role in the energy balance of our country. The quality of fuel oil significantly affects the conditions of its transport, storage, and combustion; release of contaminants to atmosphere; and the operation of main and auxiliary facilities of HPPs. According to the Energy Strategy of Russia for the Period until 2030, the oil-refining ratio gradually increases; as a result, the fraction of straight-run fuel oil in heavy fuel oils consistently decreases, which leads to the worsening of performance characteristics of fuel oil. Consequently, the problem of the increase in the quality of residual fuel oil is quite topical. In this paper, it is suggested to treat fuel oil by additives during its combustion, which would provide the improvement of ecological and economic indicators of oil-fired HPPs. Advantages of this method include simplicity of implementation, low energy and capital expenses, and the possibility to use production waste as additives. In the paper, the results are presented of industrial tests of the combustion of fuel oil with the additive of dewatered carbonate sludge, which is formed during coagulation and lime treatment of environmental waters on HPPs. The design of a volume delivery device is developed for the steady additive input to the boiler air duct. The values are given for the main parameters of the condition of a TGM-84B boiler plant. The mechanism of action of dewatered carbonate sludge on sulfur oxides, which are formed during fuel oil combustion, is considered. Results of industrial tests indicate the decrease in the mass fraction of discharged sulfur oxides by 36.5%. Evaluation of the prevented damage from sulfur oxide discharged into atmospheric air shows that the combustion of the fuel oil of 100 brand using carbonate sludge as an additive (0.1 wt %) saves nearly 6 million rubles a year during environmental actions at the consumption of fuel oil of 138240 t/year.

  13. Production of wood fuels from young forests

    International Nuclear Information System (INIS)

    Korpilahti, A.

    1998-01-01

    National forest invention data shows that more than 200 000 ha of thinnings should be carried out annually. The stemwood accumulation corresponding to this is about 13 million m 3 . The share of industrial wood is about 5.7 million m 3 , so the energy wood potential is about 7.0 million m 3 . Because the growing stock can use the nutrients liberated from logging residues the topwood mass should not be totally harvested, and at the barren areas it should not be harvested at all. Even the difficult terrain restricts in some extent the harvesting of logging residues. After these reductions the economically harvestible energy wood potential has been estimated to be 5.1 million m 3 corresponding to about 0.9 million toe. The amount of first thinnings has during the last few years been only about one third of the need. The accumulation in the first thinning phase could be about 40-80 m 3 /ha. The annual young stand treatment area has usually been about 200 000 ha, but during the last few years it has remained to a little over 100 000 ha. Harvesting of wood fuels from young stands, based on a lot-chipping method and the traditional production chains, was investigated in the national Bioenergy Research Programme. Equipment of suitable size and price are needed for harvesting of small-diameter trees. The profitability of mechanized harvesting can be improved significantly if the single-tree processing is replaced with multi- tree processing. Multi-tree harvesting can be carried out in all production chains, felling-bunching, in partial and pulpwood harvesting, as well as with bare felling machines and harvesters. About 60 % of the stems were processed with a prototype machine, tested in treatment of young forests. About 70 % of fellings in felling-bunching, already in commercial use, was processed as multi- tree processing, and about 80 % in the partial-tree harvesting. The felling of pulpwood as partial trees was about 25-30 % faster as multi-tree processing than with

  14. A Status of Art-Report on the Fission Products Behavior Released from Spent Fuel at High Temperature Conditions

    International Nuclear Information System (INIS)

    Park, Geun Il; Kim, J. H.; Lee, J. W.

    2003-04-01

    The experiments on the fission products release behavior from spent fuel at high temperature assuming reactor accident conditions have been carried out at Oak Ridge Nation Laboratory of USA in HI/VI tests, CEA of France in HEVA/VERCOS tests, AEA of England and CRNL of Canada in HOX test. The VEGA program to study the fission product release behavior from LWR irradiated fuel was recently initiated at JAERI. The key parameter affecting the fission product(FP) release behavior is temperature. In addition, other parameters such as fuel oxidation, burnup, pre-transient conditions are found to affect the FP releases considerably in the earlier tests. The atmosphere conditions such as oxidizing atmosphere (steam or air) or reducing atmosphere (hydrogen) can cause significant change of FPs release and transport behavior due to chemical forms of the reactive FPs which is dependent on the oxidation potential. The effect of fuel burnup on the Kr-85 or Cs-137 release showed that the release rates of these radionuclides increased with the increase of burnup, meaning that release rates are dominated by the atomic diffusions in the grains and they are primarily a function of temperature. However, the data on FPs release behavior using higher burnups above 50,000 MWD/MTU are not so many reported up to now. This report summarizes the test results of FPs release behavior in reactor accident conditions produced from other countries mentioned above. This review and analysis on earlier studies would be useful for predicting the release characteristics of FPs from domestic spent fuel. The release rates of fission gas or FPs from spent fuel at high temperature conditions during fabrication process of dry recycling fuel were also analyzed using many data obtained from earlier tests

  15. BWR 9 X 9 Fuel Assembly Thermal-Hydraulic Tests (2): Hydraulic Vibration Test

    International Nuclear Information System (INIS)

    Yoshiaki Tsukuda; Katsuichiro Kamimura; Toshiitsu Hattori; Akira Tanabe; Noboru Saito; Masahiko Warashina; Yuji Nishino

    2002-01-01

    Nuclear Power Engineering Corporation (NUPEC) conducted thermal-hydraulic projects for verification of thermal-hydraulic design reliability for BWR high-burnup 8 x 8 and 9 x 9 fuel assemblies, entrusted by the Ministry of Economy, Trade and Industry (METI). As a part of the NUPEC thermal-hydraulic projects, hydraulic vibration tests using full-scale test assemblies simulating 9 x 9 fuel assemblies were carried out to evaluate BWR fuel integrity. The test data were applied to development of a new correlation for the estimation of fuel rod vibration amplitude. (authors)

  16. Production of nuclear ceramic fuel for nuclear power plants at 'Ulba metallurgical plant' OSC

    International Nuclear Information System (INIS)

    Khadeev, V.G.

    2000-01-01

    The paper describes the flow-sheet of production of uranium dioxide powders and nuclear ceramic fuel pellets of them existing at the facility. 'UMP' OSC applies ADU extraction process of UO2 powders production. An indisputable success of the process is the possibility of use of the wide range of raw materials. Uranium hexafluoride, uranium oxides, uranium metal, uranium tetrafluoride, uranyl salts, uranium ore concentrates, all possible types of uranium-containing materials the processing of which by routine methods is difficult (ashes, scraps, etc.) are used as the raw materials. In addition, a reprocessed nuclear fuel can be used for fuel production. The quality of uranium dioxide powder produced does not depend on the type of uranium raw material used. High selectivity of extraction refining makes possible to obtain material with rather low impurities content that meets practically all specifications for uranium dioxide known to us. Ceramic and process features of uranium dioxide powders, namely, specific surface, bulk density, grain size and sinterability make possible to produce nuclear ceramic fuel with specified features. Quality of uranium dioxide powders produced by 'UMP' OSC was highly rated by General Electric company that is one of the leading companies from fuel manufactures in the USA market . It has certified 'UMP' OSC as its supplier. Currently, our company makes great efforts on establishing production of uranium dioxide powders with natural isotopes content for production of fuel for CANDU reactors. Trial lots of such powders are under tests at some companies manufacturing fuel for this type reactors in Canada, USA and Corea

  17. Review of WWER fuel and material tests in the Halden reactor

    International Nuclear Information System (INIS)

    Volkov, B.; Kolstad, E.

    2006-01-01

    A review of the tests with WWER fuels and materials conducted in HBWR over the years of cooperation with Russia is presented. The first test with old generation WWER-440 fuel and PWR specification fuel was carried out from 1995 to 1998. Some differences between these fuels regarding irradiation induced densification and pellet design as well as similar fuel thermal behaviour, swelling and FGR were revealed during the test. The data from this test are reviewed and compared with PIE recently performed to confirm the in-pile measurements. The second test was started in March 1999 with the main objective to study different modified WWER fuels also in comparison with PWR fuel. The results indicated that all these modified WWER fuels exhibit improved densification properties relative to earlier tested fuel. In-pile data on fuel densification have been analysed with respect to as fabricated fuel microstructure and can be used for verification of fuel behaviour models. Corrosion and creep tests in the Halden reactor encompass WWER cladding alloys and some results are given. Prospective WWER fuel and material tests foreseen within the frame of the joint program of OECD HRP are also presented. (authors)

  18. Selection and development of advanced nuclear fuel products

    International Nuclear Information System (INIS)

    Stucker, David L.; Miller, Richard S.; Arnsberger, Peter L.

    2004-01-01

    The highly competitive international marketplace requires a continuing product development commitment, short development cycle times and timely, on-target product development to assure customer satisfaction and continuing business. Westinghouse has maintained its leadership position within the nuclear fuel industry with continuous developments and improvements to fuel assembly materials and design. This paper presents a discussion of the processes used by Westinghouse in the selection and refinement of advanced concepts for deployment in the highly competitive US and international nuclear fuel fabrication marketplace. (author)

  19. PIE Report on the KOMO-3 Irradiation Test Fuels

    International Nuclear Information System (INIS)

    Park, Jong Man; Ryu, H. J.; Yang, J. H.

    2009-04-01

    In the KOMO-3, in-reactor irradiation test had been performed for 12 kinds of dispersed U-Mo fuel rods, a multi wire fuel rod and a tube fuel rod. In this report we described the PIE results on the KOMO-3 irradiation test fuels. The interaction layer thickness between fuel particle and matrix could be reduced by using a large size U-Mo fuel particle or introducing Al-Si matrix or adding the third element in the U-Mo particle. Monolithic fuel rod of multi-wire or tube fuel was also effective in reducing the interaction layer thickness

  20. Technical and Economic Evaluation of Macroalgae Cultivation for Fuel Production (Draft)

    Energy Technology Data Exchange (ETDEWEB)

    Feinberg, D. A.; Hock, S. M.

    1985-04-01

    The potential of macroalgae as sources of renewable liquid and gaseous fuels is evaluated. A series of options for production of macroalgae feedstock is considered. Because of their high carbohydrate content, the fuel products for which macroalgae are most suitable are methane and ethanol. Fuel product costs were compared with projected fuel costs in the year 1995.

  1. Traditional technologies of fuels production for air-jet engines

    Directory of Open Access Journals (Sweden)

    Бойченко С. В.

    2013-07-01

    Full Text Available Available energy resources for various fuels, mainly for gas-turbine engines are presented in the given article. Traditional technologies for jet fuels production from nonrenewable raw materials, such as crude oil, coal, natural gas, oil-shales and others are analyzed in details. Cause and effect relationship between production and use of such fuels and their impact on natural environment is defined. The timeliness and necessity for development of alternative technologies of aviation biofuels production are determined in the given article.

  2. Engineering cyanobacteria for fuels and chemicals production.

    Science.gov (United States)

    Zhou, Jie; Li, Yin

    2010-03-01

    The world's energy and global warming crises call for sustainable, renewable, carbon-neutral alternatives to replace fossil fuel resources. Currently, most biofuels are produced from agricultural crops and residues, which lead to concerns about food security and land shortage. Compared to the current biofuel production system, cyanobacteria, as autotrophic prokaryotes, do not require arable land and can grow to high densities by efficiently using solar energy, CO(2), water, and inorganic nutrients. Moreover, powerful genetic techniques of cyanobacteria have been developed. For these reasons, cyanobacteria, which carry out oxygenic photosynthesis, are attractive hosts for production of fuels and chemicals. Recently, several chemicals including ethanol, isobutanol and isoprene have been produced by engineered cyanobacteria directly using solar energy, CO(2), and water. Cyanobacterium is therefore a potential novel cell factory for fuels and chemicals production to address global energy security and climate change issues.

  3. Selection and properties of alternative forming fluids for TRISO fuel kernel production

    Energy Technology Data Exchange (ETDEWEB)

    Baker, M.P. [Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States); King, J.C., E-mail: kingjc@mines.edu [Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States); Gorman, B.P. [Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States); Marshall, D.W. [Idaho National Laboratory, 2525 N. Fremont Avenue, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer Forming fluid selection criteria developed for TRISO kernel production. Black-Right-Pointing-Pointer Ten candidates selected for further study. Black-Right-Pointing-Pointer Density, viscosity, and surface tension measured for first time. Black-Right-Pointing-Pointer Settling velocity and heat transfer rates calculated. Black-Right-Pointing-Pointer Three fluids recommended for kernel production testing. - Abstract: Current Very High Temperature Reactor (VHTR) designs incorporate TRi-structural ISOtropic (TRISO) fuel, which consists of a spherical fissile fuel kernel surrounded by layers of pyrolytic carbon and silicon carbide. An internal sol-gel process forms the fuel kernel using wet chemistry to produce uranium oxyhydroxide gel spheres by dropping a cold precursor solution into a hot column of trichloroethylene (TCE). Over time, gelation byproducts inhibit complete gelation, and the TCE must be purified or discarded. The resulting TCE waste stream contains both radioactive and hazardous materials and is thus considered a mixed hazardous waste. Changing the forming fluid to a non-hazardous alternative could greatly improve the economics of TRISO fuel kernel production. Selection criteria for a replacement forming fluid narrowed a list of {approx}10,800 chemicals to yield ten potential replacement forming fluids: 1-bromododecane, 1-bromotetradecane, 1-bromoundecane, 1-chlorooctadecane, 1-chlorotetradecane, 1-iododecane, 1-iodododecane, 1-iodohexadecane, 1-iodooctadecane, and squalane. The density, viscosity, and surface tension for each potential replacement forming fluid were measured as a function of temperature between 25 Degree-Sign C and 80 Degree-Sign C. Calculated settling velocities and heat transfer rates give an overall column height approximation. 1-bromotetradecane, 1-chlorooctadecane, and 1-iodododecane show the greatest promise as replacements, and future tests will verify their ability to form satisfactory

  4. Ultraclean Fuels Production and Utilization for the Twenty-First Century: Advances toward Sustainable Transportation Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Fox, Elise B.; Liu, Zhong-Wen; Liu, Zhao-Tie

    2013-11-21

    Ultraclean fuels production has become increasingly important as a method to help decrease emissions and allow the introduction of alternative feed stocks for transportation fuels. Established methods, such as Fischer-Tropsch, have seen a resurgence of interest as natural gas prices drop and existing petroleum resources require more intensive clean-up and purification to meet stringent environmental standards. This review covers some of the advances in deep desulfurization, synthesis gas conversion into fuels and feed stocks that were presented at the 245th American Chemical Society Spring Annual Meeting in New Orleans, LA in the Division of Energy and Fuels symposium on "Ultraclean Fuels Production and Utilization".

  5. Test plan for Series 3 NNWSI spent fuel leaching/dissolution tests

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1986-04-01

    The Series 3 tests will differ from the Series 2 tests in that the Series 3 tests will be run at 85 0 C (J-13 water) in sealed 304 stainless steel (SS) test vessels. The current NNWSI reference spent fuel container material is 304L SS. The candidate NNWSI repository horizon is above the water table, and 95 0 C (boiling temperature at the repository elevation) is the maximum liquid water temperature expected to contact spent fuel in the repository

  6. 14 CFR 23.965 - Fuel tank tests.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank tests. 23.965 Section 23.965 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Powerplant Fuel System § 23.965 Fuel...

  7. Situation of test and research reactors' spent fuels

    International Nuclear Information System (INIS)

    Shimizu, Kenichi; Uchiyama, Junzo; Sato, Hiroshi

    1996-01-01

    The U.S. DOE decided a renewal Off-Site Fuel Policy for stopping to spread a highly enriched uranium which was originally enriched at the U.S., the policy declared that to receive all HEU spent fuels from Test and Research reactors in all the world. In Japan, under bilateral agreement of cooperation between the government of the United States and the government of Japan concerning peaceful uses of nuclear energy, the highly enriched uranium of Test and Research Reactors' fuels was purchased from the U.S. and the fuels had been manufactured in Japan, America, Germany and France. On the other hand, a former president of the U.S. J. Carter proposed that to convert the fuels from HEU to LEU concerning a nonproliferation of nuclear materials in 1978, and Japan absolutely supported this policy. Under this condition, the U.S. stopped to receive the spent fuels from the other countries concerning legal action to the Off-Site Fuels Policy. As a result, the spent fuels are increasing, and to cross to each reactor's storage capacity, and if this policy start, a faced crisis of Test and Research Reactors will be avoided. (author)

  8. Current and prospective fuel test programmes in the MIR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izhutov, A.L.; Burukin, A.V.; Iljenko, S.A.; Ovchinnikov, V.A.; Shulimov, V.N.; Smirnov, V.P. [State Scientific Centre of Russia Research Institute of Atomic Reactors, Ulyanovsk region (Russian Federation)

    2007-07-01

    MIR reactor is a heterogeneous thermal reactor with a moderator and a reflector made of metal beryllium, it has a channel-type design and is placed in a water pool. MIR reactor is mainly designed for testing fragments of fuel elements and fuel assemblies (FA) of different nuclear power reactor types under normal (stationary and transient) operating conditions as well as emergency situations. At present six test loop facilities are being operated (2 PWR loops, 2 BWR loops and 2 steam coolant loops). The majority of current fuel tests is conducted for improving and upgrading the Russian PWR fuel, these tests involve issues such as: -) long term tests of short-size rods with different modifications of cladding materials and fuel pellets; -) further irradiation of power plant re-fabricated and full-size fuel rods up to achieving 80 MW*d/kg U; -) experiments with leaking fuel rods at different burnups and under transient conditions; -) continuation of the RAMP type experiments at high burnup of fuel; and -) in-pile tests with simulation of LOCA and RIA type accidents. Testing of the LEU (low enrichment uranium) research reactor fuel is conducted within the framework of the RERTR programme. Upgrading of the gas cooled and steam cooled loop facilities is scheduled for testing the HTGR fuel and sub-critical water-cooled reactor, correspondingly. The present paper describes the major programs of the WWER high burn-up fuel behavior study in the MIR reactor, capabilities of the applied techniques and some results of the performed irradiation tests. (authors)

  9. Solid oxide fuel cells and hydrogen production

    International Nuclear Information System (INIS)

    Dogan, F.

    2009-01-01

    'Full text': A single-chamber solid oxide fuel cell (SC-SOFC), operating in a mixture of fuel and oxidant gases, provides several advantages over the conventional SOFC such as simplified cell structure (no sealing required). SC-SOFC allows using a variety of fuels without carbon deposition by selecting appropriate electrode materials and cell operating conditions. The operating conditions of single chamber SOFC was studied using hydrocarbon-air gas mixtures for a cell composed of NiO-YSZ / YSZ / LSCF-Ag. The cell performance and catalytic activity of the anode was measured at various gas flow rates. The results showed that the open-circuit voltage and the power density increased as the gas flow rate increased. Relatively high power densities up to 660 mW/cm 2 were obtained in a SC-SOFC using porous YSZ electrolytes instead of dense electrolytes required for operation of a double chamber SOFC. In addition to propane- or methane-air mixtures as a fuel source, the cells were also tested in a double chamber configuration using hydrogen-air mixtures by controlling the hydrogen/air ratio at the cathode and the anode. Simulation of single chamber conditions in double chamber configurations allows distinguishing and better understanding of the electrode reactions in the presence of mixed gases. Recent research efforts; the effect of hydrogen-air mixtures as a fuel source on the performance of anode and cathode materials in single-chamber and double-chamber SOFC configurations,will be presented. The presentation will address a review on hydrogen production by utilizing of reversible SOFC systems. (author)

  10. Behavior of metallic fuel in treat transient overpower tests

    International Nuclear Information System (INIS)

    Bauer, T.H.; Wright, A.E.; Robinson, W.R.; Klickman, A.E.

    1988-01-01

    Results and analyses are reported for TREAT in-pile transient overpower tests of margin to cladding failure and pre-failure axial expansion of metallic fuel. In all cases the power rise was exponential on an 8 s period until either incipient or actual cladding failure was achieved. Test fuel included EBR-II driver fuel and ternary alloy, the reference fuel of the Intergral Fast Reactor concept. Test pin burnup spanned the widest range available. The nature of the observed cladding failure and resultant fuel dispersals is described. Simple models are presented which describe observed cladding failures and pre-failure axial expansions yet are general enough to apply to all metal fuel types

  11. Remote helium leak test of the DUPIC fuel rod

    International Nuclear Information System (INIS)

    Kim, W. K; Kim, S. S.; Lim, S. P.; Lee, J. W.; Yang, M. S.

    1998-01-01

    DUPIC(Direct Use of spent PWR fuel In CANDU reactor) is one of dry reprocessing fuel cycles to reuse irradiated PWR fuel in CANDU power plant. DUPIC fuel is so radioactive that DUPIC fuel is remotely fabricated at hot cell such as IMEF hot cell in which radiation is shielded and remote operation is possible. In this study, Helium leakage has been tested for the simulated DUPIC fuel rod manufactured by Nd:YAG laser end-cap welding at simulated hot cell. The remote inspection technique has been developed to evaluate the soundness of DUPIC fuel fabricated through new processes. Vacuum chamber has been developed to be remotely operated by manipulators at hot cell. As the result of remote test, Helium leakage of DUPIC fuel rod is around background level, CANDU specification has been satisfied. In the result of the study, remote test has been successfully performed at the simulated hot cell, and the soundness of DUPIC fuel rod welded by Nd:YAG laser has been confirmed

  12. HFR irradiation testing of light water reactor (LWR) fuel

    International Nuclear Information System (INIS)

    Markgraf, J.F.W.

    1985-01-01

    For the materials testing reactor HFR some characteristic information with emphasis on LWR fuel rod testing capabilities and hot cell investigation is presented. Additionally a summary of LWR fuel irradiation programmes performed and forthcoming programmes are described. Project management information and a list of publications pertaining to LWR fuel rod test programmes is given

  13. Fluid flow test for KMRR fuel assemblies

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Yang, Sun Kyu; Chung, Chang Hwan; Chun, See Young; Song, Chul Hha; Jun, Hyung Gil; Chung, Heung Joon; Won, Soon Yeun; Cho, Young Rho; Kim, Bok Deuk

    1991-01-01

    Hydraulic and velocity measurment tests were carried out for the KMRR fuel assembly. Two types of the KMRR fuel assembly are consist of longitudinally finned rods. Experimental data of the pressure drops and friction factors for the KMRR fuel assemlby were produced. The measurement technique for the turbulent flow structure in subchannels using the LDV was obtained. The measurement of the experimental constant of the thermal hydraulic analysis code was investigated. The results in this study are used as the basic data for the development of an analysis code. The measurement technique acquired in this study can be applied to the KMRR thermal hydraulic commissioning test and development of the domestic KMRR fuel fabrication. (Author)

  14. Spent fuel drying system test results (first dry-run)

    International Nuclear Information System (INIS)

    Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

    1998-07-01

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site. Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 7.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first dry-run test, which was conducted without a fuel element. The empty test apparatus was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The data from this dry-run test can serve as a baseline for the first two fuel element tests, 1990 (Run 1) and 3128W (Run 2). The purpose of this dry-run was to establish the background levels of hydrogen in the system, and the hydrogen generation and release characteristics attributable to the test system without a fuel element present. This test also serves to establish the background levels of water in the system and the water release characteristics. The system used for the drying test series was the Whole Element Furnace Testing System, described in Section 2.0, which is located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in section 3.0, and the experimental

  15. The coupled kinetics of grain growth and fission product behavior in nuclear fuel under degraded-core accident conditions

    International Nuclear Information System (INIS)

    Rest, J.

    1985-01-01

    The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, and cesium release from (1) irradiated high-burnup LWR fuel in a flowing steam atmosphere during high-temperature, in-cell heating tests (performed at Oak Ridge National Laboratory) and (2) trace-irratiated LWR fuel during severe-fuel-damage (SFD) tests (performed in the PBF reactor in Idaho). A theory of grain boundary sweeping of gas bubbles has been included within the FASTGRASS-VFP formalism. This theory considers the interaction between the moving grain boundary and two distinct size classes of bubbles, those on grain faces and on grain edges, and provides a means of determining whether gas bubbles are caught up and moved along by a moving grain boundary or whether the grain boundary is only temporarily retarded by the bubbles and then breaks away. In addition, as FASTGRASS-VFP provides for a mechanistic calculation of intra- and intergranular fission product behavior, the coupled calculation between fission gas behavior and grain growth is kinetically comprehensive. Results of the analyses demonstrate that intragranular fission product behavior during both types of tests can be interpreted in terms of a grain-growth/grain-boundary-sweeping mechanism that enhances the flow of fission products from within the grains to the grain boundaries. The effect of fuel oxidation by steam on fission product and grain growth behavior is also considered. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in high-burnup fuel are highlighted. (orig.)

  16. SP-100 Fuel Pin Performance: Results from Irradiation Testing

    Science.gov (United States)

    Makenas, Bruce J.; Paxton, Dean M.; Vaidyanathan, Swaminathan; Marietta, Martin; Hoth, Carl W.

    1994-07-01

    A total of 86 experimental fuel pins with various fuel, liner, and cladding candidate materials have been irradiated in the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF) reactor as part of the SP-100 fuel pin irradiation testing program. Postirradiation examination results from these fuel pins are key in establishing performance correlations and demonstrating the lifetime and safety of the reactor fuel system. This paper provides a brief description of the in-reactor fuel pin tests and presents the most recent irradiation data on the performance of wrought rhenium (Re) liner material and high density UN fuel at goal burnup of 6 atom percent (at. %). It also provides an overview of the significant variety of other fuel/liner/cladding combinations which were irradiated as part of this program and which may be of interest to more advanced efforts.

  17. Preliminary evaluation of fungicidal and termiticidal activities of filtrates from biomass slurry fuel production

    Energy Technology Data Exchange (ETDEWEB)

    Kartal, S.N. [Istanbul University (Turkey). Forestry Faculty; Imamura, Y. [Kyoto University (Japan). Wood Research Institute; Tsuchiya, F.; Ohsato, K. [JGC Corporation, Yokohama (Japan)

    2004-10-01

    Biomass slurry fuel (BSF) production has recently been developed as a natural energy for the conversion of solid biomass into fuel. In addition to using fuel, filtrates from BSF production may also serve a chemical source with several organic compounds. There is an increasing interest in the research and application of biomass-based filtrates. In this study, fungicidal and termiticidal properties of filtrates from BSF production using sugi (Cryptomeria japonica) and acacia (Acacia mangium) wood were evaluated in laboratory decay and termite resistance tests. Wood blocks treated with the filtrates showed increased resistance against brown-rot fungus, Formitopsis palustris. However the filtrates from sugi wood processed at 270{sup o}C which contained less phenolic compounds than the other filtrates were effective against white-rot fungus, Trametes versicolor. Phenolic compounds of filtrates seemed to play a role in the decay resistance tests however the filtrates did not increase the durability of the wood blocks against subterranean termites Coptotermes formosanus. Despite high acetic and lactic acid content of the filtrates, vanillin content of the filtrates may have served as an additional food source and promoted termite attack. It can be concluded that filtrates with phenolic compounds from lignin degradation during BSF production can be considered for targeted inhibition of brown-rot. (author)

  18. Power ramp tests of BWR-MOX fuels

    International Nuclear Information System (INIS)

    Asahi, K.; Oguma, M.; Higuchi, S.; Kamimua, K.; Shirai, Y.; Bodart, S.; Mertens, L.

    1996-01-01

    Power ramp test of BWR-MOX and UO 2 fuel rods base irradiated up to about 60 GWd/t in Dodewaard reactor have been conducted in BR2 reactor in the framework of the international DOMO programme. The MOX pellets were provided by BN (MIMAS process) and PNC (MH method). The MOX fuel rods with Zr-liner and non-liner cladding and the UO 2 fuel rods with Zr-liner cladding remained intact during the stepwise power ramp tests to about 600 W/cm, even at about 60 GWd/t

  19. FCTESTNET - Testing fuel cells for transportation

    NARCIS (Netherlands)

    Winkel, R.G.; Foster, D.L.; Smokers, R.T.M.

    2006-01-01

    FCTESTNET (Fuel Cell Testing and Standardization Network) is an ongoing European network project within Framework Program 5. It is a three-year project that commenced January 2003, with 55 partners from European research centers, universities, and industry, working in the field of fuel cell R and D.

  20. Chemical Characterization and Reactivity of Fuel-Oxidizer Reaction Product

    Science.gov (United States)

    David, Dennis D.; Dee, Louis A.; Beeson, Harold D.

    1997-01-01

    Fuel-oxidizer reaction product (FORP), the product of incomplete reaction of monomethylhydrazine and nitrogen tetroxide propellants prepared under laboratory conditions and from firings of Shuttle Reaction Control System thrusters, has been characterized by chemical and thermal analysis. The composition of FORP is variable but falls within a limited range of compositions that depend on three factors: the fuel-oxidizer ratio at the time of formation; whether the composition of the post-formation atmosphere is reducing or oxidizing; and the reaction or post-reaction temperature. A typical composition contains methylhydrazinium nitrate, ammonium nitrate, methylammonium nitrate, and trace amounts of hydrazinium nitrate and 1,1-dimethylhydrazinium nitrate. Thermal decomposition reactions of the FORP compositions used in this study were unremarkable. Neither the various compositions of FORP, the pure major components of FORP, nor mixtures of FORP with propellant system corrosion products showed any unusual thermal activity when decomposed under laboratory conditions. Off-limit thruster operations were simulated by rapid mixing of liquid monomethylhydrazine and liquid nitrogen tetroxide in a confined space. These tests demonstrated that monomethylhydrazine, methylhydrazinium nitrate, ammonium nitrate, or Inconel corrosion products can induce a mixture of monomethylhydrazine and nitrogen tetroxide to produce component-damaging energies. Damaging events required FORP or metal salts to be present at the initial mixing of monomethylhydrazine and nitrogen tetroxide.

  1. Target-fueled nuclear reactor for medical isotope production

    Science.gov (United States)

    Coats, Richard L.; Parma, Edward J.

    2017-06-27

    A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days.

  2. Interpretation and modelling of fission product Ba and Mo releases from fuel

    Science.gov (United States)

    Brillant, G.

    2010-02-01

    The release mechanisms of two fission products (namely barium and molybdenum) in severe accident conditions are studied using the VERCORS experimental observations. Barium is observed to be mostly released under reducing conditions while molybdenum release is most observed under oxidizing conditions. As well, the volatility of some precipitates in fuel is evaluated by thermodynamic equilibrium calculations. The polymeric species (MoO 3) n are calculated to largely contribute to molybdenum partial pressure and barium volatility is greatly enhanced if the gas atmosphere is reducing. Analytical models of fission product release from fuel are proposed for barium and molybdenum. Finally, these models have been integrated in the ASTEC/ELSA code and validation calculations have been performed on several experimental tests.

  3. Posttest examination results of recent treat tests on metal fuel

    International Nuclear Information System (INIS)

    Holland, J.W.; Wright, A.E.; Bauer, T.H.; Goldman, A.J.; Klickman, A.E.; Sevy, R.H.

    1986-01-01

    A series of in-reactor transient tests is underway to study the characteristics of metal-alloy fuel during transient-overpower-without-scam conditions. The initial tests focused on determining the margin to cladding breach and the axial fuel motions that would mitigate the power excursion. The tests were conducted in flowing-sodium loops with uranium - 5% fissium EBR-II Mark-II driver fuel elements in the TREAT facility. Posttest examination of the tests evaluated fuel elongation in intact pins and postfailure fuel motion. Microscopic examination of the intact pins studied the nature and extent of fuel/cladding interaction, fuel melt fraction and mass distribution, and distribution of porosity. Eutectic penetration and failure of the cladding were also examined in the failed pins

  4. Nuclear fuels for material test reactors

    International Nuclear Information System (INIS)

    Ramanathan, L.V.; Durazzo, M.; Freitas, C.T. de

    1982-01-01

    Experimental results related do the development of nuclear fuels for reactors cooled and moderated by water have been presented cylindrical and plate type fuels have been described in which the core consists of U compouns dispersed in an Al matrix and is clad with aluminium. Fabrication details involving rollmilling, swaging or hot pressing have been described. Corrosion and irradiation test results are also discussed. The performance of the different types of fuels indicates that it is possible to locally fabricate fuel plates with U 3 O 8 +Al cores (20% enriched U) for use in operating Brazilian research reactors. (Author) [pt

  5. Operational reliability testing of FBR fuel in EBR-II

    International Nuclear Information System (INIS)

    Asaga, Takeo; Ukai, Shigeharu; Nomura, Shigeo; Shikakura, Sakae

    1991-01-01

    The operational reliability testing of FBR fuel has been conducting in EBR-II as a DOE/PNC collaboration program. This paper reviews the achieved summary of Phase-I test as well as outline of progressing Phase-II test. In Phase-I test, the reliability of FBR fuel pins including 'MONJU' fuel was demonstrated at the event of operational transient. Continued operation of the failed pins was also shown to be feasible without affecting the plant operation. The objectives of the Phase-II test is to extend the data base relating with the operational reliability for long life fuel, and to supply the highly quantitative evaluation. The valuable insight obtained in Phase-II test are considerably expected to be useful toward the achievement of commercial FBR. (author)

  6. Stand for visual ultrasonic testing of spent fuel

    International Nuclear Information System (INIS)

    Czajkowski, W.; Borek-Kruszewska, E.

    2001-01-01

    A stand for visual and ultrasonic testing of spent fuel, constructed under Strategic Governmental Programme for management of spent fuel and radioactive waste, is presented in the paper. The stand, named 'STEND-1', built up at the Institute of Atomic Energy in Swjerk, is appointed for underwater visual testing of spent fuel elements type MR6 and WWR by means of TV-CCD camera and image processing system and for ultrasonic scanning of external surface of these elements by means of video scan immersion transducer and straight UHT connector. 'STEND-1' is built using flexible in use, high-tensile, anodized aluminum profiles. All the profiles feature longitudinal grooves to accommodate connecting elements and for the attachment of accessories at any position. They are also characterised by straight-through core bores for use with standard fastening elements and to accommodate accessory components. Stand, equipped with automatic control and processing system based on personal computer, may be manually or automatically controlled. Control system of movements of the camera in the vertical axis and rotational movement of spent fuel element permits to fix chosen location of fuel element with accuracy better than 0.1 mm. High resolution of ultrasonic method allows to record damages of outer surface of order 0.1 mm. The results of visual testing of spent fuel are recorded on video tape and then may be stored on the hard disc of the personal computer and presented in shape of photo or picture. Only selected damage surfaces of spent fuel elements are tested by means of ultrasonic scanning. All possibilities of the stand and results of visual testing of spent fuel type WWR are presented in the paper. (author)

  7. Continuous process of powder production for MOX fuel fabrication according to ''granat'' technology

    International Nuclear Information System (INIS)

    Morkovnikov, V.E.; Raginskiy, L.S.; Pavlinov, A.P.; Chernov, V.A.; Revyakin, V.V.; Varykhanov, V.S.; Revnov, V.N.

    2000-01-01

    During last years the problem of commercial MOX fuel fabrication for nuclear reactors in Russia was solved in a number of directions. The paper deals with the solution of the problem of creating a continuous pilot plant for the production of MOX fuel powders on the basis of the home technology 'Granat', that was tested before on a small-scale pilot-commercial batch-operated plant of the same name and confirmed good results. (authors)

  8. Tests of experimental fuel elements by the method of nuclear-thermal pulse loadings in 'HYDRA' reactor

    International Nuclear Information System (INIS)

    Nastoyashchaya, O.V.; Lebedev, Yu. M.; Chechurov, A.M.; Khvostionov, Ye

    1997-01-01

    The results of tests of experimental fuel elements with uranium dioxide fuel composition embedded in Al and Zr matrix with the enrichment from 90% to 36% in respect to U-235 performed at the pulse 'HYDRA' reactor are presented in this paper. Testing is performed in the frame-work of extensive research program studying the behavior of fuel elements (FE) of research and mini nuclear power systems in case of practically immediate energy release in the fuel taking place during the RIA-type accidents. Duration of the neutron pulse when testing in 'HYDRA' reactor is from 7 to 20 ms. The methods of diagnostics of the state of FE prior to and after testing in the reactor are developed and verified. Mathematical model describing temperature fields inside the FE in the process of testing. and accounting for non-uniformity of fuel composition has been developed in order to summarize experimental results. Experimental data on the limiting values of the energy density leading to deformation and degradation of FE depending on the type of fuel composition have been obtained and the mechanisms for the development of these processes have been determined. The nature of physical-chemical processes taking place in the fuel composition and fuel cladding depending on material composition under different levels of energy deposition is demonstrated. The data on hydrogen generation and radioactive product release out of fuel after failure of FE are presented. (author)

  9. Welding of metallic fuel elements for the irradiation test in JOYO. Preliminary tests and welding execution tests (Joint research)

    International Nuclear Information System (INIS)

    Kikuchi, Hironobu; Nakamura, Kinya; Iwai, Takashi; Arai, Yasuo

    2009-10-01

    Irradiation tests of metallic fuels elements in fast test reactor JOYO are planned under the joint research of Japan Atomic Energy Agency (JAEA) and Central Research Institute of Electric Power Industry (CRIEPI). Six U-Pu-Zr fuel elements clad with ferritic martensitic steel are fabricated in Plutonium Fuel Research Facility (PFRF) of JAEA-Oarai for the first time in Japan. In PFRF, the procedures of fabrication of the fuel elements were determined and the test runs of the equipments were carried out before the welding execution tests for the fuel elements. Test samples for confirming the welding condition between the cladding tube and top and bottom endplugs were prepared, and various test runs were carried out before the welding execution tests. As a result, the welding conditions were finalized by passing the welding execution tests. (author)

  10. Biomass gasification for liquid fuel production

    International Nuclear Information System (INIS)

    Najser, Jan; Peer, Václav; Vantuch, Martin

    2014-01-01

    In our old fix-bed autothermal gasifier we tested wood chips and wood pellets. We make experiments for Czech company producing agro pellets - pellets made from agricultural waste and fastrenewable natural resources. We tested pellets from wheat and rice straw and hay. These materials can be very perspective, because they dońt compete with food production, they were formed in sufficient quantity and in the place of their treatment. New installation is composed of allothermal biomass fixed bed gasifier with conditioning and using produced syngas for Fischer - Tropsch synthesis. As a gasifying agent will be used steam. Gas purification will have two parts - separation of dust particles using a hot filter and dolomite reactor for decomposition of tars. In next steps, gas will be cooled, compressed and removed of sulphur and chlorine compounds and carbon dioxide. This syngas will be used for liquid fuel synthesis

  11. Biomass gasification for liquid fuel production

    Energy Technology Data Exchange (ETDEWEB)

    Najser, Jan, E-mail: jan.najser@vsb.cz, E-mail: vaclav.peer@vsb.cz; Peer, Václav, E-mail: jan.najser@vsb.cz, E-mail: vaclav.peer@vsb.cz [VSB - Technical university of Ostrava, Energy Research Center, 17. listopadu 15/2172, 708 33 Ostrava-Poruba (Czech Republic); Vantuch, Martin [University of Zilina, Faculty of Mechanical Engineering, Department of Power Engineering, Univerzitna 1, 010 26 Zilina (Slovakia)

    2014-08-06

    In our old fix-bed autothermal gasifier we tested wood chips and wood pellets. We make experiments for Czech company producing agro pellets - pellets made from agricultural waste and fastrenewable natural resources. We tested pellets from wheat and rice straw and hay. These materials can be very perspective, because they dońt compete with food production, they were formed in sufficient quantity and in the place of their treatment. New installation is composed of allothermal biomass fixed bed gasifier with conditioning and using produced syngas for Fischer - Tropsch synthesis. As a gasifying agent will be used steam. Gas purification will have two parts - separation of dust particles using a hot filter and dolomite reactor for decomposition of tars. In next steps, gas will be cooled, compressed and removed of sulphur and chlorine compounds and carbon dioxide. This syngas will be used for liquid fuel synthesis.

  12. ACRR fission product release tests: ST-1 and ST-2

    International Nuclear Information System (INIS)

    Allen, M.D.; Stockman, H.W.; Reil, K.O.; Grimley, A.J.; Camp, W.J.

    1988-01-01

    Two experiments (ST-1 and ST-2) have been performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNLA) to obtain time-resolved data on the release of fission products from irradiated fuels under light water reactor (LWR) severe accident conditions. Both experiments were conducted in a highly reducing environment at maximum fuel temperatures of greater than 2400 K. These experiments were designed specifically to investigate the effect of increased total pressure on fission product release; ST-1 was performed at approximately 0.16 MPa and ST-2 was run at 1.9 MPa, whereas other parameters were matched as closely as possible. Release rate data were measured for Cs, I, Ba, Sr, Eu, Te, and U. The release rates were higher than predicted by existing codes for Ba, Sr, Eu, and U. Te release was very low, but Te did not appear to be sequestered by the zircaloy cladding; it was evenly distributed in the fuel. In addition, in posttest analysis a unique fuel morphology (fuel swelling) was observed which may have enhanced fission product release, especially in the high pressure test (ST-2). These data are compared with analytical results from the CORSOR correlation and the VICTORIA computer model

  13. Leaching characteristics of the metal waste form from the electrometallurgical treatment process: Product consistency testing

    International Nuclear Information System (INIS)

    Johnson, S. G.; Keiser, D. D.; Frank, S. M.; DiSanto, T.; Noy, M.

    1999-01-01

    Argonne National Laboratory is developing an electrometallurgical treatment for spent fuel from the experimental breeder reactor II. A product of this treatment process is a metal waste form that incorporates the stainless steel cladding hulls, zirconium from the fuel and the fission products that are noble to the process, i.e., Tc, Ru, Nb, Pd, Rh, Ag. The nominal composition of this waste form is stainless steel/15 wt% zirconium/1--4 wt% noble metal fission products/1--2 wt % U. Leaching results are presented from several tests and sample types: (1) 2 week monolithic immersion tests on actual metal waste forms produced from irradiated cladding hulls, (2) long term (>2 years) pulsed flow tests on samples containing technetium and uranium and (3) crushed sample immersion tests on cold simulated metal waste form samples. The test results will be compared and their relevance for waste form product consistency testing discussed

  14. 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

  15. 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)

  16. Evaluation of fuel rods behavior - under irradiation test

    International Nuclear Information System (INIS)

    Lameiras, F.S.; Terra, J.L.; Pinto, L.C.M.; Dias, M.S.; Pinheiro, R.B.

    1981-04-01

    By the accompanying of the irradiation of instrumented test fuel rods simulating the operational conditions in reactors, plus the results of post - irradiation exams, tests, evaluation and calibration of analitic modelling of such fuel rods is done. (E.G.) [pt

  17. Severe fuel-damage scoping test performance

    International Nuclear Information System (INIS)

    Gruen, G.E.; Buescher, B.J.

    1983-01-01

    As a result of the Three Mile Island Unit-2 (TMI-2) accident, the Nuclear Regulatory Commission has initiated a severe fuel damage test program to evaluate fuel rod and core response during severe accidents similar to TMI-2. The first test of Phase I of this series has been successfully completed in the Power Burst Facility at the Idaho National Engineering Laboratory. Following the first test, calculations were performed using the TRAC-BD1 computer code with actual experimental boundary conditions. This paper discusses the test conduct and performance and presents the calculated and measured test bundle results. The test resulted in a slow heatup to 2000 K over about 4 h, with an accelerated reaction of the zirconium cladding at temperatures above 1600 K in the lower part or the bundle and 2000 K in the upper portion of the bundle

  18. Fission product release from nuclear fuel II. Validation of ASTEC/ELSA on analytical and large scale experiments

    International Nuclear Information System (INIS)

    Brillant, G.; Marchetto, C.; Plumecocq, W.

    2013-01-01

    Highlights: • A wide range of experiments is presented for the ASTEC/ELSA code validation. • Analytical tests such as AECL, ORNL and VERCORS are considered. • A large-scale experiment, PHEBUS FPT1, is considered. • The good agreement with measurements shows the efficiency of the ASTEC modelling. • Improvements concern the FP release modelling from MOX and high burn-up UO 2 fuels. - Abstract: This article is the second of two articles dedicated to the mechanisms of fission product release from a degraded core. The models of fission product release from nuclear fuel in the ASTEC code have been described in detail in the first part of this work (Brillant et al., this issue). In this contribution, the validation of ELSA, the module of ASTEC that deals with fission product and structural material release from a degraded core, is presented. A large range of experimental tests, with various temperature and conditions for the fuel surrounding atmosphere (oxidising and reducing), is thus simulated with the ASTEC code. The validation database includes several analytical experiments with both bare fuel (e.g. MCE1 experiments) and cladded fuel (e.g. HCE3, VERCORS). Furthermore, the PHEBUS large-scale experiments are used for the validation of ASTEC. The rather satisfactory comparison between ELSA calculations and experimental measurements demonstrates the efficiency of the analytical models to describe fission product release in severe accident conditions

  19. Reusable fuel test assembly for the FFTF

    International Nuclear Information System (INIS)

    Pitner, A.L.; Dittmer, J.O.

    1992-01-01

    A fuel test assembly that provides re-irradiation capability after interim discharge and reconstitution of the test pin bundle has been developed for use in the Fast Flux Test Facility (FFTF). This test vehicle permits irradiation test data to be obtained at multiple exposures on a few select test pins without the substantial expense of fabricating individual test assemblies as would otherwise be required. A variety of test pin types can be loaded in the reusable test assembly. A reusable test vehicle for irradiation testing in the FFTF has long been desired, but a number of obstacles previously prevented the implementation of such an experimental rig. The MFF-8A test assembly employs a 169-pin bundle using HT-9 alloy for duct and cladding material. The standard driver pins in the fuel bundle are sodium-bonded metal fuel (U-10 wt% Zr). Thirty-seven positions in the bundle are replaceable pin positions. Standard MFF-8A driver pins can be loaded in any test pin location to fill the bundle if necessary. Application of the MFF-8A reusable test assembly in the FFTF constitutes a considerable cost-saving measure with regard to irradiation testing. Only a few well-characterized test pins need be fabricated to conduct a test program rather than constructing entire test assemblies

  20. Three generation production biotechnology of biomass into bio-fuel

    Science.gov (United States)

    Zheng, Chaocheng

    2017-08-01

    The great change of climate change, depletion of natural resources, and scarcity of fossil fuel in the whole world nowadays have witnessed a sense of urgency home and abroad among scales of researchers, development practitioners, and industrialists to search for completely brand new sustainable solutions in the area of biomass transforming into bio-fuels attributing to our duty-that is, it is our responsibility to take up this challenge to secure our energy in the near future with the help of sustainable approaches and technological advancements to produce greener fuel from nature organic sources or biomass which comes generally from organic natural matters such as trees, woods, manure, sewage sludge, grass cuttings, and timber waste with a source of huge green energy called bio-fuel. Biomass includes most of the biological materials, livings or dead bodies. This energy source is ripely used industrially, or domestically for rather many years, but the recent trend is on the production of green fuel with different advance processing systems in a greener. More sustainable method. Biomass is becoming a booming industry currently on account of its cheaper cost and abundant resources all around, making it fairly more effective for the sustainable use of the bio-energy. In the past few years, the world has witnessed a remarkable development in the bio-fuel production technology, and three generations of bio-fuel have already existed in our society. The combination of membrane technology with the existing process line can play a vital role for the production of green fuel in a sustainable manner. In this paper, the science and technology for sustainable bio-fuel production will be introduced in detail for a cleaner world.

  1. Spent fuel handling system for a geologic storage test at the Nevada Test Site

    International Nuclear Information System (INIS)

    Duncan, J.E.; House, P.A.; Wright, G.W.

    1980-01-01

    The Lawrence Livermore Laboratory is conducting a test of the geologic storage of encapsulated spent commercial reactor fuel assemblies in a granitic rock at the Nevada Test Site. The test, known as the Spent Fuel Test-Climax (SFT-C), is sponsored by the US Department of Energy, Nevada Operations Office. Eleven pressurized-water-reactor spent fuel assemblies are stored retrievably for three to five years in a linear array in the Climax stock at a depth of 420 m

  2. Accident Testing of High Temperature Reactor Fuel Elements with the KueFA Device

    International Nuclear Information System (INIS)

    Seeger, O.; Laurie, M.; Bottomley, P.D.W.; Ferreira-Teixeira, A.E.; Van Winckel, S.; Rondinella, V.V.; Allelein, H.J.

    2013-06-01

    The High Temperature Reactor (HTR) is characterised by an advanced design with passive safety features. Fuel elements are constituted by a graphite matrix containing sub-mm-sized fuel particles with Tri-Isotropic (TRISO) coating, designed to provide high fission product retention. During a loss of coolant accident scenario in a HTR the maximum temperature is foreseen to be in the range of 1600-1650 deg. C, remaining well below the melting point of the fuel. The Cold Finger Apparatus (KueFA) is used to observe the combined effects of Depressurization and Loss of Forced Circulation (DLOFC) accident scenarios on HTR fuel. Originally designed at the Forschungszentrum Juelich (FZJ), an adapted KueFA operates on irradiated fuel in hot cell at JRC-ITU. A fuel pebble is heated in He atmosphere for several hundred hours, mimicking accident temperatures up to 1800 deg. C and realistic temperature transients. Non-gaseous volatile fission products released from the fuel condense on a water cooled stainless steel plate dubbed 'Cold Finger'. Exchanging plates frequently during the experiment and analysing plate deposits by means of HPGe gamma spectroscopy allows a reconstruction of the fission product release as a function of time and temperature. In order to achieve a good quantification of the release, a careful calibration of the setup is mandatory. An especially tailored collimator was designed to perform plate scanning with high spatial resolution, thus yielding information about the fission product distribution on the condensation plates. The analysis of condensation plates from recent KueFA tests shows that fission product release quantification is possible at high and low activity levels. Chemical dissolution has been performed for some condensation plates in order to assess beta nuclides of interest such as 90 Sr and possibly 129 I using an Inductively Coupled Plasma - Mass Spectrometer (ICP-MS) and to cross check the HPGe gamma spectroscopy measurements

  3. Test of fuel handling machine for Monju in sodium

    International Nuclear Information System (INIS)

    Ishii, Yoichiro; Masuda, Yoichi; Kataoka, Hajime

    1980-01-01

    Various types of fuel handling machines were studied, and under-the-plug method of fuel exchange and the fuel handling machine of single turning plug, fixed arm type were selected for the prototype reactor ''Monju'', because the turning plug is relatively small, and the rate of operation, safety, operational ability, maintainability and reliability required for the reactor are satisfied, moreover, the extrapolation to the demonstration reactor was considered. Attention must be paid to the points that the fuel handling machine is very long and invisible from outside, and the smooth operation and endurance in sodium are required for it. The full mock-up testing facility of single turning plug, fixed arm type was installed in 1974, and the full mock-up test has been carried out since 1975 in Oarai. Fuel exchange is carried out at about 6 months intervals in Monju, and about 20 to 30% of core and blanket fuels are exchanged for about one month period. The functions required for the fuel handling machine for Monju, the outline of the testing facility, the schedule of the testing, the items of testing and the results, and the matters to be specially written are described. The full mock-up test in sodium has been carried out for 5 years, and the functions and the endurance have been proved sufficiently. (Kako, I.)

  4. Cadmium depletion impacts on hardening neutron spectrum for advanced fuel testing in ATR

    International Nuclear Information System (INIS)

    Chang, Gray S.

    2011-01-01

    For transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products effectively is in a fast neutron spectrum reactor. In the absence of a fast spectrum test reactor in the United States of America (USA), initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. A test region is achieved with a Cadmium (Cd) filter which can harden the neutron spectrum to a spectrum similar (although still somewhat softer) to that of the liquid metal fast breeder reactor (LMFBR). A fuel test loop with a Cd-filter has been installed within the East Flux Trap (EFT) of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). A detailed comparison analyses between the cadmium (Cd) filter hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum have been performed using MCWO. MCWO is a set of scripting tools that are used to couple the Monte Carlo transport code MCNP with the isotope depletion and buildup code ORIGEN-2.2. The MCWO-calculated results indicate that the Cd-filter can effectively flatten the Rim-Effect and reduce the linear heat rate (LHGR) to meet the advanced fuel testing project requirements at the beginning of irradiation (BOI). However, the filtering characteristics of Cd as a strong absorber quickly depletes over time, and the Cd-filter must be replaced for every two typical operating cycles within the EFT of the ATR. The designed Cd-filter can effectively depress the LHGR in experimental fuels and harden the neutron spectrum enough to adequately flatten the Rim-Effect in the test region. (author)

  5. Diesel fuel lubricity testing revisited : Tests von Dieselkraftstoffschmierfähigkeit erneut betrachtet

    NARCIS (Netherlands)

    van Leeuwen, H.J.

    2017-01-01

    Fuel is used as a lubricant in several engine components. Diesel fuel is known for its good lubrication properties, better than gasoline. These properties are examined in standard tests, as prescribed by ASTM. Good lubrication properties are designated as a good lubricity. Most commonly, fuel

  6. Feasibility study on the transient fuel test loop installation

    International Nuclear Information System (INIS)

    Kim, J. Y.; Lee, C. Y.

    1997-02-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. The objective of this study is to investigate and analyze the test capsules and loops in research reactors of the other countries and to design preliminarily the eligible transient fuel test facility to be installed in HANARO. The principle subjects of this study are to analyze the contents, kinds and scopes of the irradiation test facilities for nuclear technology development. The guidances for the basic and detail design of the transient fuel test facility in the future are presented. The investigation and analysis of various kinds of test facilities that are now in operation at the research reactors of nuclear advanced countries are carried out. Based on the design data of HANARO the design materials for an eligible transient fuel test facility comprises two pacts : namely, in pile test fuel in reactor core site, and out of pile system regulates the experimental conditions in the in pile test section. Especially for power ramping and cycling selection of the eligible power variation equipment in HANARO is carried out. (author). 13 refs., 4 tabs., 46 figs

  7. Spent fuel metal storage cask performance testing and future spent fuel concrete module performance testing

    International Nuclear Information System (INIS)

    McKinnon, M.A.; Creer, J.M.

    1988-10-01

    REA-2023 Gesellshaft fur Nuklear Service (GNS) CASTOR-V/21, Transnuclear TN-24P, and Westinghouse MC-10 metal storage casks, have been performance tested under the guidance of the Pacific Northwest Laboratory to determine their thermal and shielding performance. The REA-2023 cask was tested under Department of Energy (DOE) sponsorship at General Electric's facilities in Morris, Illinois, using BWR spent fuel from the Cooper Reactor. The other three casks were tested under a cooperative agreement between Virginia Power Company and DOE at the Idaho National Engineering Laboratory (INEL) by EGandG Idaho, Inc., using intact spent PWR fuel from the Surry reactors. The Electric Power Research Institute (EPRI) made contributions to both programs. A summary of the various cask designs and the results of the performance tests is presented. The cask designs include: solid and liquid neutron shields; lead, steel, and nodular cast iron gamma shields; stainless steel, aluminum, and copper baskets; and borated materials for criticality control. 4 refs., 8 figs., 6 tabs

  8. Modular, High-Volume Fuel Cell Leak-Test Suite and Process

    Energy Technology Data Exchange (ETDEWEB)

    Ru Chen; Ian Kaye

    2012-03-12

    Fuel cell stacks are typically hand-assembled and tested. As a result the manufacturing process is labor-intensive and time-consuming. The fluid leakage in fuel cell stacks may reduce fuel cell performance, damage fuel cell stack, or even cause fire and become a safety hazard. Leak check is a critical step in the fuel cell stack manufacturing. The fuel cell industry is in need of fuel cell leak-test processes and equipment that is automatic, robust, and high throughput. The equipment should reduce fuel cell manufacturing cost.

  9. Fission product release from fuel of water-cooled reactors

    International Nuclear Information System (INIS)

    Strupczewski, A.; Marks, P.; Klisinska, M.

    1997-01-01

    The report contains a review of theoretical models and experimental works of gaseous and volatile fission products from uranium dioxide fuel. The experimental results of activity release at low burnup and the model of fission gas behaviour at initial stage of fuel operational cycle are presented. Empirical models as well as measured results of transient fission products release rate in the temperature up to UO 2 melting point, with consideration of their chemical reactions with fuel and cladding, are collected. The theoretical and experimental data were used for calculations of gaseous and volatile fission products release, especially iodine and caesium, to the gas volume of WWER-1000 and WWER-440 type fuel rods at low and high burnup and their further release from defected rods at the assumed loss-of-coolant accident. (author)

  10. Equipment for nondestructive testing of the PWR and BWR spept fUel elements and assemblies in the NPP storage pools

    International Nuclear Information System (INIS)

    Gorskij, V.V.

    1983-01-01

    Design features are considered of units for nondestructive testing of spent fUel elements and fuel assemblies (FA) in the storage pools of NPP with the PWR and BWR reactors. Units for remote viewing control of fuel element cans and FA, for direct measurements of their geometrical dimensions, for FA leak-testing, fuel element can nondestructive testing and gamma scanning, for measuring gaseous fission product pressure and fuel element free volume are described along with units for complex checking of fuel element and FA parameters. The units for nondestructive testing of spent fuel elements and EA are shown to differ both in their designs and a number of checked parameters of fuel elements and FA. The remote viewing and those for measuring the basic FA parameters are most generally employed. Units for complex testing of multiple fuel element parameters, designed in the last few years, are intended for operation with FA disassembled partially or fully and are characteristic of a high degree of computer measuring automation both for the process control and data processing

  11. The use of niobium based catalysts for liquid fuel production

    Directory of Open Access Journals (Sweden)

    Reguera Frank Martin

    2004-01-01

    Full Text Available The catalytic properties of niobium based catalysts were investigated in the conversion of oleic acid to liquid fuels at atmospheric pressure and at 623 K. The catalytic tests were performed in a fixed bed and continuous flow reactor using an acid to catalyst ratio equal to 4 and N2 as carrier gas. The reaction products were analyzed by gas chromatography and acidity measurements. NH3 temperature programmed desorption, N2 adsorption-desorption (BET method and Xray diffraction were also performed in order to determine the structural and acidic properties of the catalysts. From the catalytic tests, it was detected the formation of compounds in the range of gasoline, diesel and lubricant oils. Higher catalytic activity and selectivity for diesel fuel were observed for the catalysts NbOPO4 and H3PO4/Nb2O5 that possesses higher acidities and surface areas.

  12. HTR fuel modelling with the ATLAS code. Thermal mechanical behaviour and fission product release assessment

    International Nuclear Information System (INIS)

    Guillermier, Pierre; Daniel, Lucile; Gauthier, Laurent

    2009-01-01

    To support AREVA NP in its design on HTR reactor and its HTR fuel R and D program, the Commissariat a l'Energie Atomique developed the ATLAS code (Advanced Thermal mechanicaL Analysis Software) with the objectives: - to quantify, with a statistical approach, the failed particle fraction and fission product release of a HTR fuel core under normal and accidental conditions (compact or pebble design). - to simulate irradiation tests or benchmark in order to compare measurements or others code results with ATLAS evaluation. These two objectives aim at qualifying the code in order to predict fuel behaviour and to design fuel according to core performance and safety requirements. A statistical calculation uses numerous deterministic calculations. The finite element method is used for these deterministic calculations, in order to be able to choose among three types of meshes, depending on what must be simulated: - One-dimensional calculation of one single particle, for intact particles or particles with fully debonded layers. - Two-dimensional calculations of one single particle, in the case of particles which are cracked, partially debonded or shaped in various ways. - Three-dimensional calculations of a whole compact slice, in order to simulate the interactions between the particles, the thermal gradient and the transport of fission products up to the coolant. - Some calculations of a whole pebble, using homogenization methods are being studied. The temperatures, displacements, stresses, strains and fission product concentrations are calculated on each mesh of the model. Statistical calculations are done using these results, taking into account ceramic failure mode, but also fabrication tolerances and material property uncertainties, variations of the loads (fluence, temperature, burn-up) and core data parameters. The statistical method used in ATLAS is the importance sampling. The model of migration of long-lived fission products in the coated particle and more

  13. Computational analysis of modern HTGR fuel performance and fission product release during the HFR-EU1 irradiation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl, E-mail: k.verfondern@fz-juelich.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Xhonneux, André, E-mail: xhonneux@lrst.rwth-aachen.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Nabielek, Heinz, E-mail: heinznabielek@me.com [Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Allelein, Hans-Josef, E-mail: h.j.allelein@fz-juelich.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, Chair for Reactor Safety and Reactor Technology, 52072 Aachen (Germany)

    2014-07-01

    Highlights: • HFR-EU1 irradiation test demonstrates high quality of HTGR spherical fuel elements. • Irradiation performance is in good agreement with German fuel performance modeling. • International benchmark exercise expected first particle to fail at ∼13–17% FIMA. • EOL silver release is predicted to be in the percentage range. • EOL cesium and strontium are expected to remain at a low level. - Abstract: Various countries engaged in the development and fabrication of modern HTGR fuel have initiated activities of modeling the fuel and fission product release behavior with the aim of predicting the fuel performance under HTGR operating and accident conditions. Verification and validation studies are conducted by code-to-code benchmarking and code-to-experiment comparisons as part of international exercises. The methodology developed in Germany since the 1980s represents valuable and efficient tools to describe fission product release from spherical fuel elements and TRISO fuel performance, respectively, under given conditions. Continued application to new results of irradiation and accident simulation testing demonstrates the appropriateness of the models in terms of a conservative estimation of the source term as part of interactions with HTGR licensing authorities. Within the European irradiation testing program for HTGR fuel and as part of the former EU RAPHAEL project, the HFR-EU1 irradiation experiment explores the potential for high performance of the presently existing German and newly produced Chinese fuel spheres under defined conditions up to high burnups. The fuel irradiation was completed in 2010. Test samples are prepared for further postirradiation examinations (PIE) including heatup simulation testing in the KÜFA-II furnace at the JRC-ITU, Karlsruhe, to be conducted within the on-going ARCHER Project of the European Commission. The paper will describe the application of the German computer models to the HFR-EU1 irradiation test and

  14. Research reactors for power reactor fuel and materials testing - Studsvik's experience

    International Nuclear Information System (INIS)

    Grounes, M.

    1998-01-01

    Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

  15. Spent fuel's behavior under dynamic drip tests

    International Nuclear Information System (INIS)

    Finn, P.A.; Buck, E.C.; Hoh, J.C.; Bates, J.K.

    1995-01-01

    In the potential repository at Yucca Mountain, failure of the waste package container and the cladding of the spent nuclear fuel would expose the fuel to water under oxidizing conditions. To simulate the release behavior of radionuclides from spent fuel, dynamic drip and vapor tests with spent nuclear fuel have been ongoing for 2.5 years. Rapid alteration of the spent fuel has been noted with concurrent release of radionuclides. Colloidal species containing americium and plutonium have been found in the leachate. This observation suggests that colloidal transport of radionuclides should be included in the performance assessment of a potential repository

  16. Feasibility study of the Dragon reactor for HTGR fuel testing

    International Nuclear Information System (INIS)

    Wallroth, C.F.

    1975-01-01

    The Organization of European Community Development (OECD) Dragon high-temperature reactor project has performed HTGR fuel and fuel element testing for about 10 years. To date, a total of about 250 fuel elements have been irradiated and the test program continues. The feasibility of using this test facility for HTGR fuel testing, giving special consideration to U. S. needs, is evaluated. A detailed description for design, preparation, and data acquisition of a test experiment is given together with all possible options on supporting work, which could be carried out by the experienced Dragon project staff. 11 references. (U.S.)

  17. Commercial Aspect of Research Reactor Fuel Element Production

    International Nuclear Information System (INIS)

    Susanto, B.G; Suripto, A

    1998-01-01

    Several aspects affecting the commercialization of the Research Reactor Fuel Element Production Installation (RR FEPI) under a BUMN (state-owned company)have been studied. The break event point (BEP) value based on total production cost used is greatly depending upon the unit selling price of the fuel element. At a selling price of USD 43,500/fuel element, the results of analysis shows that the BEP will be reached at 51% of minimum available capacity. At a selling price of US$ 43.500/fuel element the total income (after tax) for 7 years ahead is US $ 4.620.191,- The net present value in this study has a positive value is equal to US $ 2.827.527,- the internal rate of return will be 18% which is higher than normal the bank interest rare (in US dollar) at this time. It is concluded therefore that the nuclear research reactor fuel element produced by state-owned company BUMN has a good prospect to be sold commercially

  18. Behavior of metallic uranium-fissium fuel in TREAT transient overpower tests

    International Nuclear Information System (INIS)

    Bauer, T.H.; Klickman, A.E.; Lo, R.K.; Rhodes, E.A.; Robinson, W.R.; Stanford, G.S.; Wright, A.E.

    1986-01-01

    TREAT tests M2, M3, and M4 were performed to obtain information on two key behavior characteristics of fuel under transient overpower accident conditions in metal-fueled fast reactors: the prefailure axial self-extrusion (elongation beyond thermal expansion) of fuel within intact cladding and the margin to cladding breach. Uranium-5 wt% fissium Experimental Breeder Reactor-II driver fuel pins were used for the tests since they were available as suitable stand-ins for the uranium-plutonium-zirconium ternary fuel, which is the reference fuel of the integral fast reactor (IFR) concept. The ternary fuel will be used in subsequent TREAT tests. Preliminary results from tests M2 and M3 were presented earlier. The present report includes significant advances in analysis as well as additional data from test M4. Test results and analysis have led to the development and validation of pin cladding failure and fuel extrusion models for metallic fuel, within reasonable uncertainties for the uranium-fissium alloy. Concepts involved are straightforward and readily extendable to ternary alloys and behavior in full-size reactors

  19. 33 CFR 183.580 - Static pressure test for fuel tanks.

    Science.gov (United States)

    2010-07-01

    ... pressure test for fuel tanks. A fuel tank is tested by performing the following procedures in the following... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Static pressure test for fuel tanks. 183.580 Section 183.580 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND...

  20. Production of chemicals and fuels from biomass

    Science.gov (United States)

    Qiao, Ming; Woods, Elizabeth; Myren, Paul; Cortright, Randy; Kania, John

    2018-01-23

    Methods, reactor systems, and catalysts are provided for converting in a continuous process biomass to fuels and chemicals, including methods of converting the water insoluble components of biomass, such as hemicellulose, cellulose and lignin, to volatile C.sub.2+O.sub.1-2 oxygenates, such as alcohols, ketones, cyclic ethers, esters, carboxylic acids, aldehydes, and mixtures thereof. In certain applications, the volatile C.sub.2+O.sub.1-2 oxygenates can be collected and used as a final chemical product, or used in downstream processes to produce liquid fuels, chemicals and other products.

  1. Power ramp testing method for PWR fuel rod at research reactor

    International Nuclear Information System (INIS)

    Zhou Yidong; Zhang Peisheng; Zhang Aimin; Gao Yongguang; Wang Huarong

    2003-01-01

    A tentative power ramp test for short PWR fuel rod has been conducted at the Heavy Water Research Reactor (HWRR) in China Institute of Atomic Energy (CIAE). The test fuel rod was cooled by the circulating water in the test loop. The power ramp was realized by moving solid neutron-absorbing screen around the fuel rod. The linear power of the fuel rod increased from 220 W/cm to 340 W/cm with a power ramp rate of 20 W/cm/min. The power of the fuel rod was monitored by both in-core thermal and nuclear measurement sensors in the test rig. This test provides experiences for further developing the power ramp test methods for PWR fuel rods at research reactor. (author)

  2. Irradiated fuel by-product separation research in Canada

    International Nuclear Information System (INIS)

    Burston, M.

    1984-01-01

    Although no decision has been made to reprocess irradiated CANDU fuel, by-product separation research has recently been initiated in Canada because of its potential importance to Canadian research programs in advanced fuel cycles (especially U/Pu cycle development in the near term) and nuclear waste management. In addition, separated by-products could have a significant commercial potential. Demonstrated applications include: heat sources, gamma radiation sources, light sources, new materials for productions of other useful isotopes, etc. For illustrative purposes the calculated market value of by-products currently stored in irradiated CANDU fuel is approximately $210/kgU. Ontario Hydro has initiated a program to study the application of new separation technolgies, such as laser-based techniques and the plasma ion cyclotron resonance separation technique, to either augment and/or supplant the chemical extraction methods. The main goal is to develop new, more economical extraction methods in order to increase the magnitude of the advantages resulting from this approach to reprocessing. (author)

  3. Drop testing of the Westinghouse fresh nuclear fuel package

    International Nuclear Information System (INIS)

    Shappert, L.B.; Sanders, C.F.

    1993-01-01

    The Westinghouse Columbia Fuel Fabrication Facility has decided to develop and certify a new fresh fuel package design (type A, fissile) that has the capability to transport more highly enriched fuel than was previously possible. A prototype package was tested in support of the Safety Analysis Report of the Packaging (SARP). This paper provides detailed information on the tests and test results. A first prototype test was carried out at the STF, and the design did not give the safety margin that Westinghouse wanted for their containers. The data from the test were used to redesign the connection between the clamping frame and the pressure pad, and the tests were reinitiated. Three packages were then tested at the STF. All packages met the acceptance criteria and acceleration information was obtained that provided an indication of the behavior of the cradle and strongback which holds the fuel assemblies and nuclear poison in place. (J.P.N.)

  4. Spent fuel drying system test results (second dry-run)

    International Nuclear Information System (INIS)

    Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

    1998-07-01

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks have been detected in the basins and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 7.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the second dry-run test, which was conducted without a fuel element. With the concurrence of project management, the test protocol for this run, and subsequent drying test runs, was modified. These modifications were made to allow for improved data correlation with drying procedures proposed under the IPS. Details of these modifications are discussed in Section 3.0

  5. Reducing Actinide Production Using Inert Matrix Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Deinert, Mark [Colorado School of Mines, Golden, CO (United States)

    2017-08-23

    The environmental and geopolitical problems that surround nuclear power stem largely from the longlived transuranic isotopes of Am, Cm, Np and Pu that are contained in spent nuclear fuel. New methods for transmuting these elements into more benign forms are needed. Current research efforts focus largely on the development of fast burner reactors, because it has been shown that they could dramatically reduce the accumulation of transuranics. However, despite five decades of effort, fast reactors have yet to achieve industrial viability. A critical limitation to this, and other such strategies, is that they require a type of spent fuel reprocessing that can efficiently separate all of the transuranics from the fission products with which they are mixed. Unfortunately, the technology for doing this on an industrial scale is still in development. In this project, we explore a strategy for transmutation that can be deployed using existing, current generation reactors and reprocessing systems. We show that use of an inert matrix fuel to recycle transuranics in a conventional pressurized water reactor could reduce overall production of these materials by an amount that is similar to what is achievable using proposed fast reactor cycles. Furthermore, we show that these transuranic reductions can be achieved even if the fission products are carried into the inert matrix fuel along with the transuranics, bypassing the critical separations hurdle described above. The implications of these findings are significant, because they imply that inert matrix fuel could be made directly from the material streams produced by the commercially available PUREX process. Zirconium dioxide would be an ideal choice of inert matrix in this context because it is known to form a stable solid solution with both fission products and transuranics.

  6. Reactivity initiated accident test series Test RIA 1-4 fuel behavior report

    International Nuclear Information System (INIS)

    Cook, B.A.; Martinson, Z.R.

    1984-09-01

    This report presents and discusses results from the final test in the Reactivity Initiated Accident (RIA) Test Series, Test RIA 1-4, conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. Nine preirradiated fuel rods in a 3 x 3 bundle configuration were subjected to a power burst while at boiling water reactor hot-startup system conditions. The test resulted in estimated axial peak, radial average fuel enthalpies of 234 cal/g UO 2 on the center rod, 255 cal/g UO 2 on the side rods, and 277 cal/g UO 2 on the corner rods. Test RIA 1-4 was conducted to investigate fuel coolability and channel blockage within a bundle of preirradiated rods near the present enthalpy limit of 280 cal/g UO 2 established by the US Nuclear Regulatory Commission. The test design and conduct are described, and the bundle and individual rod thermal and mechanical responses are evaluated. Conclusions from this final test and the entire PBF RIA Test Series are presented

  7. Fuel cycle related parametric study considering long lived actinide production, decay heat and fuel cycle performances

    International Nuclear Information System (INIS)

    Raepsaet, X.; Damian, F.; Lenain, R.; Lecomte, M.

    2001-01-01

    One of the very attractive HTGR reactor characteristics is its highly versatile and flexible core that can fulfil a wide range of diverse fuel cycles. Based on a GTMHR-600 MWth reactor, analyses of several fuel cycles were carried out without taking into account common fuel particle performance limits (burnup, fast fluence, temperature). These values are, however, indicated in each case. Fuel derived from uranium, thorium and a wide variety of plutonium grades has been considered. Long-lived actinide production and total residual decay heat were evaluated for the various types of fuel. The results presented in this papers provide a comparison of the potential and limits of each fuel cycle and allow to define specific cycles offering lowest actinide production and residual heat associated with a long life cycle. (author)

  8. Grain boundary sweeping and dissolution effects on fission product behavior under severe fuel damage accident conditions

    International Nuclear Information System (INIS)

    Rest, J.

    1985-10-01

    The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, tellurium, and cesium release from severe-fuel-damage (SFD) tests performed in the PBF reactor in Idaho. A theory of grain boundary sweeping of gas bubbles, gas bubble behavior during fuel liquefaction (destruction of grain boundaries due to formation of a U-rich melt phase), and during U-Zr eutectic melting has been included within the FASTGRASS-VFP formalism. The grain-boundary-sweeping theory considers the interaction between the moving grain boundary and two distinct size classes of bubbles, those on grain faces and on grain edges. The theory of the effects of fuel liquefaction and U-Zr eutectic melting on fission product behavior considers the migration and coalescence of fission gas bubbles in either molten uranium, or a zircaloy-uranium eutectic melt. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in normally irradiated fuel are highlighted

  9. Concepts for Small-Scale Testing of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Marschman, Steven Craig [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Philip Lon [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    This report documents a concept for a small-scale test involving between one and three Boiling Water Rector (BWR) high burnup (HBU) fuel assemblies. This test would be similar to the DOE funded High Burn-Up (HBU) Confirmatory Data Project to confirm the behavior of used high burn-up fuel under prototypic conditions, only on a smaller scale. The test concept proposed would collect data from fuel stored under prototypic dry storage conditions to mimic, as closely as possible, the conditions HBU UNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage.

  10. Scoping study of flowpath of simulated fission products during secondary burning of crushed HTGR fuel in a quartz fluidized-bed burner

    International Nuclear Information System (INIS)

    Rindfleisch, J.A.; Barnes, V.H.

    1976-04-01

    The results of four experimental runs in which isotopic tracers were used to simulate fission products during fluidized bed secondary burning of HTGR fuel were studied. The experimental tests provided insight relative to the flow path of fission products during fluidized-bed burning of HTGR fuel

  11. Modelling of fission product release behavior from HTR spherical fuel elements under accident conditions

    International Nuclear Information System (INIS)

    Verfondern, K.; Mueller, D.

    1991-01-01

    Computer codes for modelling the fission product release behavior of spherical fuel elements for High Temperature Reactors (HTR) have been developed for the purpose of being used in risk analyses for HTRs. An important part of the validation and verification procedure for these calculation models is the theoretical investigation of accident simulation experiments which have been conducted in the KueFA test facility in the Hot Cells at KFA. The paper gives a presentation of the basic modeling and the calculational results of fission product release from modern German HTR fuel elements in the temperature range 1600-1800 deg. C using the TRISO coated particle failure model PANAMA and the diffusion model FRESCO. Measurements of the transient release behavior for cesium and strontium and of their concentration profiles after heating have provided informations about diffusion data in the important retention barriers of the fuel: silicon carbide and matrix graphite. It could be shown that the diffusion coefficients of both cesium and strontium in silicon carbide can significantly be reduced using a factor in the range of 0.02 - 0.15 compared to older HTR fuel. Also in the development of fuel element graphite, a tendency towards lower diffusion coefficients for both nuclides can be derived. Special heating tests focussing on the fission gases and iodine release from the matrix contamination have been evaluated to derive corresponding effective diffusion data for iodine in fuel element graphite which are more realistic than the iodine transport data used so far. Finally, a prediction of krypton and cesium release from spherical fuel elements under heating conditions will be given for fuel elements which at present are irradiated in the FRJ2, Juelich, and which are intended to be heated at 1600/1800 deg. C in the KueFA furnace in near future. (author). 7 refs, 11 figs

  12. Fission product released experiment of coated fuel particles

    Energy Technology Data Exchange (ETDEWEB)

    Shijiang, Xu; Bing, Yang; Chunhe, Tang; Junguo, Zhu; Jintao, Huang; Binzhong, Zhang [Inst. of Nucl. Energy Technology, Tsinghua Univ., Beijing (China); Jinghan, Luo [Inst. of Atomic Energy, Beijing (China)

    1992-01-15

    Four samples of coated fuel particles were irradiated in the Heavy-Water Research Reactor of the Institute of Atomic Energy. Each of them was divided into two groups and irradiated to the burn up of 0.394% fima and 0.788% fima in two static capsules, respectively. After irradiation and cooling, post irradiation annealing experiment was carried out, the release ratios of the fission product {sup 133}Xe and {sup 131}I were measured, they are in the order of 10{sup -6}{approx}10{sup -7}. The fission product release ratio of naked kernel was also measured under the same conditions as for the coated fuel particles, the ratio of the fission product release of the coated fuel particles and of the naked kernel was in the order of 10{sup -5}{approx}10{sup -4}.

  13. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    International Nuclear Information System (INIS)

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

    2013-01-01

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses

  14. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses.

  15. Fission product monitoring of TRISO coated fuel for the advanced gas reactor-1 experiment

    International Nuclear Information System (INIS)

    Scates, Dawn M.; Hartwell, John K.; Walter, John B.; Drigert, Mark W.; Harp, Jason M.

    2010-01-01

    The US Department of Energy has embarked on a series of tests of TRISO coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burnup of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B's) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  16. Certification test for safety of new fuel transportation package

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Sugawa, Osami; Suga, Masao.

    1993-01-01

    The objective of this certification test is to prove the safety of new fuel transportation package against a fire of actual size caused by traffic accidents. After the fire test, the fuel assemblies were covered with coal-tar like material vaporized from anti-shock material used in the container. Surface color of BWR-type fuel assembly was dark grey that is supposed to be the color of oxide of Zircaloy. As for PWR-type fuel assembly, the condition encountered during fire test caused no change to the outlook of the rod element. Both the BWR and PWR type fuel rod elements showed no deformation and were completely sound. Therefore it may be concluded that the container protected the mimic fuel assemblies against fire of 30 minutes duration and caused no damage. This report is the result of the above experiments and examinations, and we appreciate the cooperation of those who are concerned. (J.P.N.)

  17. Preliminary test results for post irradiation examination on the HTTR fuel

    International Nuclear Information System (INIS)

    Ueta, Shohei; Umeda, Masayuki; Sawa, Kazuhiro; Sozawa, Shizuo; Shimizu, Michio; Ishigaki, Yoshinobu; Obata, Hiroyuki

    2007-01-01

    The future post-irradiation program for the first-loading fuel of the HTTR is scheduled using the HTTR fuel handling facilities and the Hot Laboratory in the Japan Materials Testing Reactor (JMTR) to confirm its irradiation resistance and to obtain data on its irradiation characteristics in the core. This report describes the preliminary test results and the future plan for a post-irradiation examination for the HTTR fuel. In the preliminary test, fuel compacts made with the same SiC-coated fuel particle as the first loading fuel were used. In the preliminary test, dimension, weight, fuel failure fraction, and burnup were measured, and X-ray radiograph, SEM, and EPMA observations were carried out. Finally, it was confirmed that the first-loading fuel of the HTTR showed good quality under an irradiation condition. The future plan for the post-irradiation tests was described to confirm its irradiation performance and to obtain data on its irradiation characteristics in the HTTR core. (author)

  18. Investigation of fission products release and structural changes of WWER spent fuel in inert and oxidizing environment

    International Nuclear Information System (INIS)

    Kungurtsev, I.A.; Smirnov, V.P.; Kuzmin, I.V.; Lebeduk, I.V.; Pimonov, Y.I.; Sohcilin, G.I.; Stupina, L.N.; Chesanov, V.V.; Shtuckert, Y.A.; Zvir, E.A.

    1996-01-01

    At the Research Institute of Atomic Reactors in-cell experiments were carried out which were aimed at investigation of WWER spent fuel behaviour under accident conditions. Gaseous and volatile fission products release and the influence of gaseous swelling, fuel interaction with the cladding and oxidation on it have been investigated. At the present time, series of experiments in inert and air environments have been finished and the tests in steam environment have been carrying out. In all series the samples in the form of fuel pellets fragments and fuel elements pieces were used. This report presents some results of annealing tests and investigations of the sample microstructure after annealing. (author). 4 refs, 15 figs, 2 tabs

  19. Iowa Central Quality Fuel Testing Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Heach, Don; Bidieman, Julaine

    2013-09-30

    The objective of this project is to finalize the creation of an independent quality fuel testing laboratory on the campus of Iowa Central Community College in Fort Dodge, Iowa that shall provide the exploding biofuels industry a timely and cost-effective centrally located laboratory to complete all state and federal fuel and related tests that are required. The recipient shall work with various state regulatory agencies, biofuel companies and state and national industry associations to ensure that training and testing needs of their members and American consumers are met. The recipient shall work with the Iowa Department of Ag and Land Stewardship on the development of an Iowa Biofuel Quality Standard along with the Development of a standard that can be used throughout industry.

  20. Production of biosolid fuels from municipal sewage sludge: Technical and economic optimisation.

    Science.gov (United States)

    Wzorek, Małgorzata; Tańczuk, Mariusz

    2015-08-01

    The article presents the technical and economic analysis of the production of fuels from municipal sewage sludge. The analysis involved the production of two types of fuel compositions: sewage sludge with sawdust (PBT fuel) and sewage sludge with meat and bone meal (PBM fuel). The technology of the production line of these sewage fuels was proposed and analysed. The main objective of the study is to find the optimal production capacity. The optimisation analysis was performed for the adopted technical and economic parameters under Polish conditions. The objective function was set as a maximum of the net present value index and the optimisation procedure was carried out for the fuel production line input capacity from 0.5 to 3 t h(-1), using the search step 0.5 t h(-1). On the basis of technical and economic assumptions, economic efficiency indexes of the investment were determined for the case of optimal line productivity. The results of the optimisation analysis show that under appropriate conditions, such as prices of components and prices of produced fuels, the production of fuels from sewage sludge can be profitable. In the case of PBT fuel, calculated economic indexes show the best profitability for the capacity of a plant over 1.5 t h(-1) output, while production of PBM fuel is beneficial for a plant with the maximum of searched capacities: 3.0 t h(-1). Sensitivity analyses carried out during the investigation show that influence of both technical and economic assessments on the location of maximum of objective function (net present value) is significant. © The Author(s) 2015.

  1. Measurement of fuel corrosion products using planar laser-induced fluorescence

    International Nuclear Information System (INIS)

    Wantuck, P.J.; Sappey, A.D.; Butt, D.P.

    1993-01-01

    Characterizing the corrosion behavior of nuclear fuel material in a high-temperature hydrogen environment is critical for ascertaining the operational performance of proposed nuclear thermal propulsion (NTP) concepts. In this paper, we describe an experimental study undertaken to develop and test non-intrusive, laser-based diagnostics for ultimately measuring the distribution of key gas-phase corrosion products expected to evolve during the exposure of NTP fuel to hydrogen. A laser ablation technique is used to produce high temperature, vapor plumes from uranium-free zirconium carbide (ZrC) and niobium carbide (NbC) forms for probing by various optical diagnostics including planar laser-induced fluorescence (PLIF). We discuss the laser ablation technique, results of plume emission measurements, and we describe both the actual and proposed planar LIF schemes for imaging constituents of the ablated ZrC and NbC plumes. Envisioned testing of the laser technique in rf-heated, high temperature gas streams is also discussed

  2. Water Footprint and Land Requirement of Solar Thermochemical Jet-Fuel Production.

    Science.gov (United States)

    Falter, Christoph; Pitz-Paal, Robert

    2017-11-07

    The production of alternative fuels via the solar thermochemical pathway has the potential to provide supply security and to significantly reduce greenhouse gas emissions. H 2 O and CO 2 are converted to liquid hydrocarbon fuels using concentrated solar energy mediated by redox reactions of a metal oxide. Because attractive production locations are in arid regions, the water footprint and the land requirement of this fuel production pathway are analyzed. The water footprint consists of 7.4 liters per liter of jet fuel of direct demand on-site and 42.4 liters per liter of jet fuel of indirect demand, where the dominant contributions are the mining of the rare earth oxide ceria, the manufacturing of the solar concentration infrastructure, and the cleaning of the mirrors. The area-specific productivity is found to be 33 362 liters per hectare per year of jet fuel equivalents, where the land coverage is mainly due to the concentration of solar energy for heat and electricity. The water footprint and the land requirement of the solar thermochemical fuel pathway are larger than the best power-to-liquid pathways but an order of magnitude lower than the best biomass-to-liquid pathways. For the production of solar thermochemical fuels arid regions are best-suited, and for biofuels regions of a moderate and humid climate.

  3. Economic Efficiency of Establishing Domestic Production of Synthetic Liquid Fuel

    Directory of Open Access Journals (Sweden)

    Kyzym Mykola O.

    2017-06-01

    Full Text Available The article notes a stable tendency to increasing the oil dependence of Ukraine, which creates a threat to the national economic security, and proves an expediency of establishing domestic production of synthetic liquid fuel. The technical, organizational and economic features of establishing synthetic liquid fuel production in Ukraine are presented. There proved a hypothesis on the expediency of organizing the production of synthetic liquid fuels based on steam-plasma coal gasification technology. The forecast resource cycle of the country until 2020 under conditions of developing this technology is modeled.

  4. Genetically Modified Bacteria for Fuel Production: Development of Rhodobacteria as a Versatile Platform for Fuels Production

    Energy Technology Data Exchange (ETDEWEB)

    None

    2010-07-01

    Electrofuels Project: Penn State is genetically engineering bacteria called Rhodobacter to use electricity or electrically generated hydrogen to convert carbon dioxide into liquid fuels. Penn State is taking genes from oil-producing algae called Botryococcus braunii and putting them into Rhodobacter to produce hydrocarbon molecules, which closely resemble gasoline. Penn State is developing engineered tanks to support microbial fuel production and determining the most economical way to feed the electricity or hydrogen to the bacteria, including using renewable sources of power like solar energy.

  5. 14 CFR 25.952 - Fuel system analysis and test.

    Science.gov (United States)

    2010-01-01

    ... using the airplane fuel system or a test article that reproduces the operating characteristics of the... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.952 Fuel system...

  6. Fuel leak testing performance at NPP Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Slugen, V.; Krnac, S.; Smiesko, I.

    1995-01-01

    The NPP Bohunice VVER-440 fuel leak testing experience are relatively extensive in comparison with other VVER-440 users. As the first Europe NPP was adapted Siemens (KWU) in core-sipping equipment to VVER-440 units and since this time were have done these tests also for NPP Paks (Hungary) and NPP Dukovany (Czech Republic). The occurrence of leaking fuel assemblies in NPP is in the last 5 years relatively stabilised and low. A significant difference can be observed between type V-230 (31 leaks) and type V-213 (1 leak). None of of the indicated leaking fuel assemblies has been investigated in the hot cell. Therefore cannot be confirm the effective causes of leak occurrence. Nevertheless, the fuel failure rate and the performance of leak testing in NPP Bohunice are comparable to the world standard at PWR's. 1 tab., 2 figs., 3 refs

  7. Fuel leak testing performance at NPP Jaslovske Bohunice

    Energy Technology Data Exchange (ETDEWEB)

    Slugen, V; Krnac, S [Slovak Technical Univ., Bratislava (Slovakia); Smiesko, I [Nuclear Powr Plant EBO, Jaslovske Bohuce (Slovakia)

    1996-12-31

    The NPP Bohunice VVER-440 fuel leak testing experience are relatively extensive in comparison with other VVER-440 users. As the first Europe NPP was adapted Siemens (KWU) in core-sipping equipment to VVER-440 units and since this time were have done these tests also for NPP Paks (Hungary) and NPP Dukovany (Czech Republic). The occurrence of leaking fuel assemblies in NPP is in the last 5 years relatively stabilised and low. A significant difference can be observed between type V-230 (31 leaks) and type V-213 (1 leak). None of of the indicated leaking fuel assemblies has been investigated in the hot cell. Therefore cannot be confirm the effective causes of leak occurrence. Nevertheless, the fuel failure rate and the performance of leak testing in NPP Bohunice are comparable to the world standard at PWR`s. 1 tab., 2 figs., 3 refs.

  8. Irradiation test plan of the simulated DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Ki Kwang; Yang, M. S.; Kim, B. K. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-11-01

    Simulated DUPIC fuel had been irradiated from Aug. 4, 1999 to Oct. 4 1999, in order to produce the data of its in-core behavior, to verify the design of DUPIC non-instrumented capsule developed, and to ensure the irradiation requirements of DUPIC fuel at HANARO. The welding process was certified for manufacturing the mini-element, and simulated DUPIC fuel rods were manufactured with simulated DUPIC pellets through examination and test. The non-instrumented capsule for a irradiation test of DUPIC fuel has been designed and manufactured referring to the design specification of the HANARO fuel. This is to be the design basis of the instrumented capsule under consideration. The verification experiment, whether the capsule loaded in the OR4 hole meet the HANARO requirements under the normal operation condition, as well as the structural analysis was carried out. The items for this experiment were the pressure drop test, vibration test, integrity test, et. al. It was noted that each experimental result meet the HANARO operational requirements. For the safety analysis of the DUPIC non-instrumented capsule loaded in the HANARO core, the nuclear/mechanical compatibility, thermodynamic compatibility, integrity analysis of the irradiation samples according to the reactor condition as well as the safety analysis of the HANARO were performed. Besides, the core reactivity effects were discussed during the irradiation test of the DUPIC capsule. The average power of each fuel rod in the DUPIC capsule was calculated, and maximal linear power reflecting the axial peaking power factor from the MCNP results was evaluated. From these calculation results, the HANARO core safety was evaluated. At the end of this report, similar overseas cases were introduced. 9 refs., 16 figs., 10 tabs. (Author)

  9. Production of chemicals and fuels from biomass

    Energy Technology Data Exchange (ETDEWEB)

    Woods, Elizabeth; Qiao, Ming; Myren, Paul; Cortright, Randy D.; Kania, John

    2015-12-15

    Described are methods, reactor systems, and catalysts for converting biomass to fuels and chemicals in a batch and/or continuous process. The process generally involves the conversion of water insoluble components of biomass, such as hemicellulose, cellulose and lignin, to volatile C.sub.2+O.sub.1-2 oxygenates, such as alcohols, ketones, cyclic ethers, esters, carboxylic acids, aldehydes, and mixtures thereof. In certain applications, the volatile C.sub.2+O.sub.1-2 oxygenates can be collected and used as a final chemical product, or used in downstream processes to produce liquid fuels, chemicals and other products.

  10. A crop production ecology (CPE) approach to sustainable production of biomass for food, feed and fuel

    NARCIS (Netherlands)

    Haverkort, A.J.; Bindraban, P.S.; Conijn, J.G.; Ruijter, de F.J.

    2009-01-01

    With the rapid increase in demand for agricultural products for food, feed and fuel, concerns are growing about sustainability issues. Can agricultural production meet the needs of increasing numbers of people consuming more animal products and using a larger share of crops as fuel for transport,

  11. Space reactor fuel element testing in upgraded TREAT

    International Nuclear Information System (INIS)

    Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

    1993-01-01

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ∼60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ∼100 MW/L may be achievable

  12. Space reactor fuel element testing in upgraded TREAT

    Science.gov (United States)

    Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

    1993-01-01

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

  13. Tests on CANDU fuel elements sheath samples

    International Nuclear Information System (INIS)

    Ionescu, S.; Uta, O.; Mincu, M.; Prisecaru, I.

    2016-01-01

    This work is a study of the behavior of CANDU fuel elements after irradiation. The tests are made on ring samples taken from fuel cladding in INR Pitesti. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory. By metallographic and ceramographic examination we determinate that the hydride precipitates are orientated parallel to the cladding surface. A content of hydrogen of about 120 ppm was estimated. After the preliminary tests, ring samples were cut from the fuel rod, and were subject of tensile test on an INSTRON 5569 model machine in order to evaluate the changes of their mechanical properties as consequence of irradiation. Scanning electron microscopy was performed on a microscope model TESCAN MIRA II LMU CS with Schottky FE emitter and variable pressure. The analysis shows that the central zone has deeper dimples, whereas on the outer zone, the dimples are tilted and smaller. (authors)

  14. Improvement of test methodology for evaluating diesel fuel stability

    Energy Technology Data Exchange (ETDEWEB)

    Gutman, M.; Tartakovsky, L.; Kirzhner, Y.; Zvirin, Y. [Internal Combustion Engines Lab., Haifa (Israel); Luria, D. [Fuel Authority, Tel Aviv (Israel); Weiss, A.; Shuftan, M. [Israel Defence Forces, Tel Aviv (Israel)

    1995-05-01

    The storage stability of diesel fuel has been extensively investigated for many years under laboratory conditions. Although continuous efforts have been made to improve testing techniques, there does not yet exist a generally accepted correlation between laboratory methods (such as chemical analysis of the fuel) and actual diesel engine tests. A testing method was developed by the Technion Internal Combustion Engines Laboratory (TICEL), in order to address this problem. The test procedure was designed to simulate diesel engine operation under field conditions. It is based on running a laboratory-modified single cylinder diesel engine for 50 h under cycling operating conditions. The overall rating of each test is based on individual evaluation of the deposits and residue formation in the fuel filter, nozzle body and needle, piston head, piston rings, exhaust valve, and combustion chamber (six parameters). Two methods for analyzing the test results were used: objective, based on measured data, and subjective, based on visual evaluation results of these deposits by a group of experts. Only the residual level in the fuel filter was evaluated quantitatively by measured results. In order to achieve higher accuracy of the method, the test procedure was improved by introducing the measured results of nozzle fouling as an additional objective evaluating (seventh) parameter. This factor is evaluated on the basis of the change in the air flow rate through the nozzle before and after the complete engine test. Other improvements in the method include the use of the nozzle assembly photograph in the test evaluation, and representation of all seven parameters on a continuous scale instead of the discrete scale used anteriorly, in order to achieve higher accuracy. This paper also contains the results obtained by application of this improved fuel stability test for a diesel fuel stored for a five-year period.

  15. 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

  16. 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.

  17. The development of flow test technology for PWR fuel assembly

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Cha, Chong Hee; Chung, Chang Hwan; Chun, Se Young; Song, Chul Hwa; Chung, Heung Joon; Won, Soon Yeun; Cho, Yeong Rho; Kim, Bok Deuk

    1988-05-01

    KAERI has an extensive program to develope PWR fuel assembly. In relation to the program, development of flow test technology is needed to evaluate the thermal hydraulic compactibility and mechanical integrity of domestically fabricated nuclear fuels. A high-pressure and high-temperature flow test facility was designed to test domestically fabricated fuel assembly. The test section of the facility has capacity of a 6x6 full length PWR fuel assembly. A flow test rig was designed and installed at Cold Test Loop to carry out model experiments with 5x5 rod assembly under atmosphere pressure to get information about the characteristics of pressure loss of spacer grids and velocity distribution in the subchannels. LDV measuring technology was established using TSI's Laser Dopper Velocimeter 9100-3 System

  18. Irradiation testing of miniature fuel plates for the RERTR program

    Energy Technology Data Exchange (ETDEWEB)

    Senn, R L; Martin, M M [Oak Ridge National Laboratory, Oak Ridge, TN 37830 (United States)

    1983-08-01

    An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. The objective of these tests is to screen various candidate fuel materials as to their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% {sup 235}U in place of highly enriched fuel for these reactors would reduce the potential for {sup 235}U diversion. Fuel materials currently being evaluated in this first phase of these screening tests include aluminum-base dispersion-type fuel plates with fuel cores of 1) high uranium content U{sup 3}){sup 8}-Al being developed by ORNL, 2) high uranium content UAI{sub x}-Al being developed by EG and G Idaho, Inc., and 3) very high uranium content U{sub 3}Si-Al- being developed by ANL. The miniplates are 115-mm long by 50-mm wide with overall plate thicknesses of 1.27 or 1.52 mm. The fuel core dimensions vary according to overall plate thicknesses with a minimal clad thickness requirement of 0.20 mm. Sixty such miniplates (thirty of each thickness) can be irradiated in one test facility. The irradiation test facility, designated as HFED-1 is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The peak neutron flux measured for this experiment is 1.96 x 10{sup 18} neutrons m{sub -2} s{sub -1}. The various types of miniplates will achieve burnups of up to approximately 2.2x10{sup 27} fissions/m{sup 3} of fuel, which will require approximately eight full power months of irradiation. During reactor shutdown periods, the experiment is removed from the reactor, moved to a special poolside station, disassembled, and inspected

  19. Methods for conversion of lignocellulosic-derived products to transportation fuel precursors

    Science.gov (United States)

    Lilga, Michael A.; Padmaperuma, Asanga B.

    2017-10-03

    Methods are disclosed for converting a biomass-derived product containing levulinic acid and/or gamma-valerolactone to a transportation fuel precursor product containing diesel like hydrocarbons. These methods are expected to produce fuel products at a reduced cost relative to conventional approaches.

  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. FAST FLUX TEST FACILITY DRIVER FUEL MEETING

    Energy Technology Data Exchange (ETDEWEB)

    None,

    1966-06-01

    The Pacific Northwest Laboratory has convened this meeting to enlist the best talents of our laboratories and industry in soliciting factual, technical information pertinent to the Pacific Northwest's Laboratory's evaluation of the potential fuel systems for the Fast Flux Test Facility. The particular factors emphasized for these fuel systems are those associated with safety, ability to meet testing objectives, and economics. The proceedings includes twenty-three presentations, along with a transcript of the discussion following each, as well as a summary discussion.

  2. FREVAP-6, Metal Fission Products Release from HTGR Fuel Elements

    International Nuclear Information System (INIS)

    Pierce, V.H.

    2005-01-01

    1 - Description of problem or function: The FREVAP type of code for estimating the release of longer-lived metallic fission products from HTGR fuel elements has been developed to take into account the combined effects of the retention of metallic fission products by fuel particles and the rather strong absorption of these fission products by the graphite of the fuel elements. Release calculations are made on the basis that the loss of fission product nuclides such as strontium, cesium, and barium is determined by their evaporation from the graphite surfaces and their transpiration induced by the flowing helium coolant. The code is devised so that changes of fission rate (fuel element power), fuel temperature, and graphite temperature may be incorporated into the calculation. Temperature is quite important in determining release because, in general, both release from fuel particles and loss by evaporation (transpiration) vary exponentially with the reciprocal of the absolute temperature. NESC0301/02: This version differs from the previous one in the following points: The source and output files were converted from BCD to ASCII coding. 2 - Method of solution: A problem is defined as having a one-dimensional segment made up of three parts - (1) the fission product source (fuel particles) in series with, (2) a non-source and absorption part (element graphite) and (3) a surface for evaporation to the coolant (graphite-helium interface). More than one segment may be connected (possibly segments stacked axially) by way of the coolant. At any given segment, a continuity equation is solved assuming equilibrium between the source term, absorption term, evaporation at coolant interface and the partial pressure of the fission product isotope in the coolant. 3 - Restrictions on the complexity of the problem - Maxima of: 5 isotopes; 10 time intervals for time-dependent variable; 49 segments (times number of isotopes); 5 different output print time-steps

  3. Characterization and testing of monolithic RERTR fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, D.D.; Jue, J.F.; Burkes, D.E. [Idaho National Lab., Idaho Falls, ID (United States)

    2007-07-01

    Monolithic fuel plates are being developed as a LEU (low enrichment uranium) fuel for application in research reactors throughout the world. These fuel plates are comprised of a U-Mo alloy foil encased in aluminum alloy cladding. Three different fabrication techniques have been looked at for producing monolithic fuel plates: hot isostatic pressing (HIP), transient liquid phase bonding (TLPB), and friction stir welding (FSW). Of these three techniques, HIP and FSW are currently being emphasized. As part of the development of these fabrication techniques, fuel plates are characterized and tested to determine properties like hardness and the bond strength at the interface between the fuel and cladding. Testing of HIP-made samples indicates that the foil/cladding interaction behavior depends on the Mo content in the UMo foil, the measured hardness values are quite different for the fuel, cladding, and interaction zone phase and Ti, Zr and Nb are the most effective diffusion barriers. For FSW samples, there is a dependence of the bond strength at the foil/cladding interface on the type of tool that is employed for performing the actual FSW process. (authors)

  4. Metallographic analysis of irradiated RERTR-3 fuel test specimens

    International Nuclear Information System (INIS)

    Meyer, M. K.; Hofman, G. L.; Strain, R. V.; Clark, C. R.; Stuart, J. R.

    2000-01-01

    The RERTR-3 irradiation test was designed to investigate the irradiation behavior of aluminum matrix U-MO alloy dispersion fuels under high-temperature, high-fission-rate conditions. Initial postirradiation examination of RERTR-3 fuel specimens has concentrated on binary U-MO atomized fuels. The rate of matrix aluminum depletion was found to be higher than predictions based on low temperature irradiation data. Wavelength Dispersive X-ray Spectroscopy (WDS) indicates that aluminum is present in the interior of the fuel particles. WDS data is supported by a mass and volume balance calculation performed on the basis of image analysis results. The depletion of matrix aluminum seems to have no detrimental effects on fuel performance under the conditions tested to date

  5. Drop testing of the Westinghouse fresh nuclear fuel package

    International Nuclear Information System (INIS)

    Shappert, L.B.; Sanders, C.F.

    1992-01-01

    In recent years, the Westinghouse Columbia Fuel Fabrication Facility has been faced with increasing pressure from utilities that wished to take the fuel in their nuclear power plants to higher burnups. To help accommodate this trend, Westinghouse has determined that it needs the ability to increase the enrichment of the fresh fuel it delivers to its customers. One critical step in this process is to certify a new (Type A, fissile) fresh fuel package design that has the capability to transport fuel with a higher enrichment than was previously available. A prototype package was tested in support of the Safety Analysis Report of the Packaging. This paper provides detailed information on those tests and their results

  6. Biodiesel production from tall oil with synthesized Mn and Ni based additives: effects of the additives on fuel consumption and emissions

    Energy Technology Data Exchange (ETDEWEB)

    Ali Keskin; Metin Guru; Duran Altiparmak [Mersin University, Mersin (Turkey). Technical Education Faculty

    2007-05-15

    In this study, biodiesel fuel and fuel additives were produced from crude tall oil that is a by-product in the pulp manufacturing by craft or sulphate pulping process. Fatty acids and resinic acids were obtained from crude tall oil by distillation method. Tall oil methyl ester (biodiesel) was produced from fatty acids. Resinic acids were reacted with NiO and MnO{sub 2} stoichiometrically for production of metallic fuel additives. Each metallic fuel additive was added at the rate of 8 {mu}mol/l and 12 {mu}mol/l to make mixtures of 60% tall oil methyl ester/40% diesel fuel (TE60) for preparing test fuels. Metallic fuel additives improved properties of biodiesel fuels, such as pour point and viscosity values. Biodiesel fuels were tested in an unmodified direct injection diesel engine at full load condition. Specific fuel consumption of biodiesel fuels increased by 6.00%, however, in comparison with TE60, it showed trend of decreasing with adding of additives. Exhaust emission profile of biodiesel fuels improved. CO emissions and smoke opacity decreased up to 64.28% and 30.91% respectively. Low NOx emission was also observed in general for the biodiesel fuels. 28 refs., 4 figs., 3 tabs.

  7. Hydrogen Fuel Cell Vehicle Fuel Economy Testing at the U.S. EPA National Vehicle and Fuel Emissions Laboratory (SAE Paper 2004-01-2900)

    Science.gov (United States)

    The introduction of hydrogen fuel cell vehicles and their new technology has created the need for development of new fuel economy test procedures and safety procedures during testing. The United States Environmental Protection Agency-National Vehicle Fuels and Emissions Laborato...

  8. Small-scale irradiated fuel electrorefining

    International Nuclear Information System (INIS)

    Benedict, R.W.; Krsul, J.R.; Mariani, R.D.; Park, K.; Teske, G.M.

    1993-01-01

    In support of the metallic fuel cycle development for the Integral Fast Reactor (IFR), a small scale electrorefiner was built and operated in the Hot Fuel Examination Facility (HFEF) at Argonne National Laboratory-West. The initial purpose of this apparatus was to test the single segment dissolution of irradiated metallic fuel via either direct dissolution in cadmium or anodic dissolution. These tests showed that 99.95% of the uranium and 99.99% of the plutonium was dissolved and separated from the fuel cladding material. The fate of various fission products was also measured. After the dissolution experiments, the apparatus was upgraded to stady fission product behavior during uranium electrotransport. Preliminary decontamination factors were estimated for different fission products under different processing conditions. Later modifications have added the following capabilities: Dissolution of multiple fuel segments simultaneously, electrotransport to a solid cathode or liquid cathode and actinide recovery with a chemical reduction crucible. These capabilities have been tested with unirradiated uranium-zirconium fuel and will support the Fuel Cycle Demonstration program

  9. Quantification of effluents in the production of nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sakai, Mayara C.C.B.; Riella, Humberto G.; Carvalho, Elita F.U. de, E-mail: mcostac@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-07-01

    At the Instituto de Pesquisa Energéticas e Nucleares (IPEN), the Centro de Combustível Nuclear (CCN), Sao Paulo, SP, Brazil, is responsible for manufacturing fuels for the IEA-R1 reactor and, possibly, the multipurpose reactor fuels. In order to meet the demand for both reactors, the CCN developed a new plant. The production process of the fuel generates several types of effluents - containing uranium or not - being solid, liquid and gaseous with varied physical and chemical characteristics. The objective of this work is to follow the nuclear fuel production process and to identify, quantify and characterize the effluents, especially the liquid ones, to later elaborate a plan of management of these and eventually dispose in a responsible way in the environment. (author)

  10. Quantification of effluents in the production of nuclear fuel

    International Nuclear Information System (INIS)

    Sakai, Mayara C.C.B.; Riella, Humberto G.; Carvalho, Elita F.U. de

    2017-01-01

    At the Instituto de Pesquisa Energéticas e Nucleares (IPEN), the Centro de Combustível Nuclear (CCN), Sao Paulo, SP, Brazil, is responsible for manufacturing fuels for the IEA-R1 reactor and, possibly, the multipurpose reactor fuels. In order to meet the demand for both reactors, the CCN developed a new plant. The production process of the fuel generates several types of effluents - containing uranium or not - being solid, liquid and gaseous with varied physical and chemical characteristics. The objective of this work is to follow the nuclear fuel production process and to identify, quantify and characterize the effluents, especially the liquid ones, to later elaborate a plan of management of these and eventually dispose in a responsible way in the environment. (author)

  11. Imminent: Irradiation Testing of (Th,Pu)O{sub 2} Fuel - 13560

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Julian F. [Thor Energy AS, Sommerrogaten 13-15, Oslo 0255 (Norway); Franceschini, Fausto [Westinghouse Electric Company LLC, 1000 Cranberry Woods Drive, Cranberry Township, PA 16066 (United States)

    2013-07-01

    Commercial-prototype thorium-plutonium oxide (Th-MOX) fuel pellets have been loaded into the material test reactor in Halden, Norway. The fuel is being operated at full power - with instrumentation - in simulated LWR / PHWR conditions and its behaviour is measured 'on-line' as it operates to high burn-up. This is a vital test on the commercialization pathway for this robust new thoria-based fuel. The performance data that is collected will support a fuel modeling effort to support its safety qualification. Several different samples of Th-MOX fuel will be tested, thereby collecting information on ceramic behaviours and their microstructure dependency. The fuel-cycle reasoning underpinning the test campaign is that commercial Th- MOX fuels are an achievable intermediate / near-term SNF management strategy that integrates well with a fast reactor future. (authors)

  12. Material test reactor fuel research at the BR2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

    2012-03-15

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

  13. Steam Methane Reformation Testing for Air-Independent Solid Oxide Fuel Cell Systems

    Science.gov (United States)

    Mwara, Kamwana N.

    2015-01-01

    Recently, NASA has been looking into utilizing landers that can be propelled by LOX-CH (sub 4), to be used for long duration missions. Using landers that utilize such propellants, also provides the opportunity to use solid oxide fuel cells as a power option, especially since they are able to process methane into a reactant through fuel reformation. One type of reformation, called steam methane reformation, is a process to reform methane into a hydrogen-rich product by reacting methane and steam (fuel cell exhaust) over a catalyst. A steam methane reformation system could potentially use the fuel cell's own exhaust to create a reactant stream that is hydrogen-rich, and requires less internal reforming of the incoming methane. Also, steam reformation may hold some advantages over other types of reforming, such as partial oxidation (PROX) reformation. Steam reformation does not require oxygen, while up to 25 percent can be lost in PROX reformation due to unusable CO (sub 2) reformation. NASA's Johnson Space Center has conducted various phases of steam methane reformation testing, as a viable solution for in-space reformation. This has included using two different types of catalysts, developing a custom reformer, and optimizing the test system to find the optimal performance parameters and operating conditions.

  14. Blunt impact tests of retired passenger locomotive fuel tanks

    Science.gov (United States)

    2017-08-01

    The Transportation Technology Center, Inc. conducted impact tests on three locomotive fuel tanks as part of the Federal Railroad Administrations locomotive fuel tank crashworthiness improvement program. Three fuel tanks, two from EMD F40PH locomot...

  15. Regional analysis of renewable transportation fuels - production and consumption

    Science.gov (United States)

    Liu, Xiaoshuai

    The transportation sector contributes more than a quarter of total U.S. greenhouse gas emissions. Replacing fossil fuels with renewable fuels can be a key solution to mitigate GHG emissions from the transportation sector. Particularly, we have focused on land-based production of renewable fuels from landfills and brownfield in the southeastern region of the United States. These so call marginal lands require no direct land-use change to avoid environmental impact and, furthermore, have rendered opportunities for carbon trading and low-carbon intensity business. The resources potential and production capacity were derived using federal and state energy databases with the aid of GIS techniques. To maximize fuels production and land-use efficiency, a scheme of co-location renewable transportation fuels for production on landfills was conducted as a case study. Results of economic modeling analysis indicate that solar panel installed on landfill sites could generate a positive return within the project duration, but the biofuel production within the landfill facility is relatively uncertain, requiring proper sizing of the onsite processing facility, economic scale of production and available tax credits. From the consumers' perspective, a life-cycle cost analysis has been conducted to determine the economic and environmental implications of different transportation choices by consumers. Without tax credits, only the hybrid electric vehicles have lifetime total costs equivalent to a conventional vehicles differing by about 1 to 7%. With tax credits, electric and hybrid electric vehicles could be affordable and attain similar lifetime total costs as compared to conventional vehicles. The dissertation research has provided policy-makers and consumers a pathway of prioritizing investment on sustainable transportation systems with a balance of environmental benefits and economic feasibility.

  16. Municipal solid waste combustion: Fuel testing and characterization

    Energy Technology Data Exchange (ETDEWEB)

    Bushnell, D.J.; Canova, J.H.; Dadkhah-Nikoo, A.

    1990-10-01

    The objective of this study is to screen and characterize potential biomass fuels from waste streams. This will be accomplished by determining the types of pollutants produced while burning selected municipal waste, i.e., commercial mixed waste paper residential (curbside) mixed waste paper, and refuse derived fuel. These materials will be fired alone and in combination with wood, equal parts by weight. The data from these experiments could be utilized to size pollution control equipment required to meet emission standards. This document provides detailed descriptions of the testing methods and evaluation procedures used in the combustion testing and characterization project. The fuel samples will be examined thoroughly from the raw form to the exhaust emissions produced during the combustion test of a densified sample.

  17. Fission product release from HTGR fuel under core heatup accident conditions - HTR2008-58160

    International Nuclear Information System (INIS)

    Verfondern, K.; Nabielek, H.

    2008-01-01

    Various countries engaged in the development and fabrication of modern fuel for the High Temperature Gas-Cooled Reactor (HTGR) have initiated activities of modeling the fuel and fission product release behavior with the aim of predicting the fuel performance under operating and accidental conditions of future HTGRs. Within the IAEA directed Coordinated Research Project CRP6 on 'Advances in HTGR Fuel Technology Development' active since 2002, the 13 participating Member States have agreed upon benchmark studies on fuel performance during normal operation and under accident conditions. While the former has been completed in the meantime, the focus is now on the extension of the national code developments to become applicable to core heatup accident conditions. These activities are supported by the fact that core heatup simulation experiments have been resumed recently providing new, highly valuable data. Work on accident performance will be - similar to the normal operation benchmark - consisting of three essential parts comprising both code verification that establishes the correspondence of code work with the underlying physical, chemical and mathematical laws, and code validation that establishes reasonable agreement with the existing experimental data base, but including also predictive calculations for future heating tests and/or reactor concepts. The paper will describe the cases to be studied and the calculational results obtained with the German computer model FRESCO. Among the benchmark cases in consideration are tests which were most recently conducted in the new heating facility KUEFA. Therefore this study will also re-open the discussion and analysis of both the validity of diffusion models and the transport data of the principal fission product species in the HTGR fuel materials as essential input data for the codes. (authors)

  18. Techniques and results of examination of fission product release from VVER fuel rods with artificial defects and a burnup of ∼60 MWd/kgU at the MIR loop facility

    International Nuclear Information System (INIS)

    Burukin, A.; Goryachev, A.; Ilyenko, S.; Izhutov, A.; Konyashov, V.; Shishin, V.; Shulimov, V.; Luzanova, L.; Miglo, V.

    2009-01-01

    Complex of equipment and several techniques for examination of radioactive fission product release from defective fuel rods were developed, prepared and tested at the PV-1 loop facility of the MIR reactor. During the first test, which was conducted at the PV-1 loop facility and aimed at testing of developed equipment and techniques, measurement of radioactive fission product release from an experimental re-fabricated fuel rod with a burnup of ∼60 MWd/kgU and an artificial defect was performed under design-basis steady-state operating conditions of the VVER-1000 reactor. PIE of all main parameters of the experimental defective fuel rod did not reveal any state peculiarities which could be caused by the artificial defect, i.e. fuel and cladding characteristics in the defect area did not differ from the initial ones (before testing) as well as their characteristics in areas distant from the defect; they are typical for fuel rods with a similar irradiation history in the VVER NPP. The gap in the experimental fuel rod was bridged due to close contact between fuel and cladding at increased fuel burnup; it can appreciable reduce release of radioactive fission products into the PV-1 primary coolant. This suggestion and quantitative characteristics of effect of gap bridging in a high-burnup fuel rod on radioactive fission product release should be investigated during the next tests performed at the PV-1 loop facility. Values of radioactive fission product release measured during the first test at the PV-1 loop facility in the MIR reactor will be used for development of an empirical engineering model in order to take into account high burnup effects and their impact on fission product release from fuel and defective fuel rods

  19. Results of the production of wood derived fuels; Puupolttoaineiden tuotantotekniikka - tutkimusalueen katsaus

    Energy Technology Data Exchange (ETDEWEB)

    Korpilahti, A [Metsaeteho, Helsinki (Finland)

    1997-12-31

    During the year 1995 there were over 30 projects concerning the production of wood derived fuels going on. Nearly half of them focused on integrated production of pulp wood and wood fuel. About in ten projects work was carried out to promote wood fuel production from logging residues. Other topics were fire wood production, production logistics and wood fuel resources. For production of fuel chips from logging residues, a new chipper truck, MOHA-SISU, was introduced. Having ability to move on terrain, and equipped with drum chipper, hook technic for interchangeable containers and a trailer, the whole production chain can be carried out by the same machine. In Mikkeli region three years of active work promoted the usage of wood fuel in a district power plant to the level of over 110 000 cubic metres of fuel chips. The production costs tend to be a little high in average, and the production chain still needs to be improved. In the field of integrated production a great stride was taken when the first pilot plant using the MASSAHAKE-method started up. Components of the production line and knowledge to operate the process have increased resulting in good performance of the plant. And even another concept for integrated production was introduced. In order to fully control the debarking of small sized trees, a production line of chain flail equipment and debarking drum followed by a chipper and screening facilities was built up. Equipment and machines for harvesting young stands in a way that increases substantially the yield of energy component are still mostly first prototypes. The development of them into well functioning, efficient tools is the most important task in integrated production

  20. Results of the production of wood derived fuels; Puupolttoaineiden tuotantotekniikka - tutkimusalueen katsaus

    Energy Technology Data Exchange (ETDEWEB)

    Korpilahti, A. [Metsaeteho, Helsinki (Finland)

    1996-12-31

    During the year 1995 there were over 30 projects concerning the production of wood derived fuels going on. Nearly half of them focused on integrated production of pulp wood and wood fuel. About in ten projects work was carried out to promote wood fuel production from logging residues. Other topics were fire wood production, production logistics and wood fuel resources. For production of fuel chips from logging residues, a new chipper truck, MOHA-SISU, was introduced. Having ability to move on terrain, and equipped with drum chipper, hook technic for interchangeable containers and a trailer, the whole production chain can be carried out by the same machine. In Mikkeli region three years of active work promoted the usage of wood fuel in a district power plant to the level of over 110 000 cubic metres of fuel chips. The production costs tend to be a little high in average, and the production chain still needs to be improved. In the field of integrated production a great stride was taken when the first pilot plant using the MASSAHAKE-method started up. Components of the production line and knowledge to operate the process have increased resulting in good performance of the plant. And even another concept for integrated production was introduced. In order to fully control the debarking of small sized trees, a production line of chain flail equipment and debarking drum followed by a chipper and screening facilities was built up. Equipment and machines for harvesting young stands in a way that increases substantially the yield of energy component are still mostly first prototypes. The development of them into well functioning, efficient tools is the most important task in integrated production

  1. 40 CFR 86.340-79 - Gasoline-fueled engine dynamometer test run.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 18 2010-07-01 2010-07-01 false Gasoline-fueled engine dynamometer... Emission Regulations for New Gasoline-Fueled and Diesel-Fueled Heavy-Duty Engines; Gaseous Exhaust Test Procedures § 86.340-79 Gasoline-fueled engine dynamometer test run. (a) This section applies to gasoline...

  2. Design, Fabrication, and Testing of a Laboratory-Scale Voloxidation System for Removal of Tritium and Other Volatile Fission Products from Used Nuclear Fuel

    International Nuclear Information System (INIS)

    Spencer, Barry B; DelCul, Guillermo D; Bradley, Eric Craig; Jubin, Robert Thomas; Hylton, Tom D; Collins, Emory D

    2008-01-01

    Advanced nuclear fuel processing methodologies are being demonstrated at the Oak Ridge National Laboratory (ORNL) as part of the Global Nuclear Energy Partnership (GNEP) program. A coupled end-to-end (CETE) research and development (R and D) capability is being installed to provide all primary processing operations, ranging from spent fuel receipt to production of products and waste forms. This R and D capability is designed for small, laboratory-scale throughput and will permit conduct of experiments in the range of 20 kg of spent fuel per year. The head-end processing segment includes single-pin shearing, voloxidation to remove tritium from the fuel before it enters the aqueous based separations systems, cleanup of the cladding hulls for disposition, and transfer of the fuel powder to the dissolution process. This paper describes the voloxidation system design and presents results from the cold checkout of the hardware. Preliminary results of the initial processing campaign with spent fuel is presented as well

  3. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    International Nuclear Information System (INIS)

    Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

    1997-01-01

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs

  4. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    Energy Technology Data Exchange (ETDEWEB)

    Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

    1997-01-01

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs.

  5. Formate Formation and Formate Conversion in Biological Fuels Production

    Directory of Open Access Journals (Sweden)

    Bryan R. Crable

    2011-01-01

    Full Text Available Biomethanation is a mature technology for fuel production. Fourth generation biofuels research will focus on sequestering CO2 and providing carbon-neutral or carbon-negative strategies to cope with dwindling fossil fuel supplies and environmental impact. Formate is an important intermediate in the methanogenic breakdown of complex organic material and serves as an important precursor for biological fuels production in the form of methane, hydrogen, and potentially methanol. Formate is produced by either CoA-dependent cleavage of pyruvate or enzymatic reduction of CO2 in an NADH- or ferredoxin-dependent manner. Formate is consumed through oxidation to CO2 and H2 or can be further reduced via the Wood-Ljungdahl pathway for carbon fixation or industrially for the production of methanol. Here, we review the enzymes involved in the interconversion of formate and discuss potential applications for biofuels production.

  6. Actinide production in different HTR-fuel cycle concepts

    International Nuclear Information System (INIS)

    Filges, D.; Hecker, R.; Mirza, N.; Rueckert, M.

    1978-01-01

    At the 'Institut fuer Reaktorentwicklung der Kernforschungsanlage Juelich' the production of α-activities in the following HTR-OTTO cycle concepts were studied: 1. standard HTR cycle (U-Th); 2. low enriched HTR cycle (U-Pu); 3. near breeder HTR cycle (U-Th); 4. combined system (conventional and near breeder HTR). The production of α-activity in HTR Uranium-Thorium fuel cycles has been investigated and compared with the standard LWR cycles. The production of α-activity in HTR Uranium-Thorium fuel cycles has been investigated and compared with the standard LWR cycles. The calculations were performed by the short depletion code KASCO and the well-known ORIGEN program

  7. REFINERY INTEGRATION OF BY-PRODUCTS FROM COAL-DERIVED JET FUELS

    Energy Technology Data Exchange (ETDEWEB)

    Leslie R. Rudnick; Andre Boehman; Chunshan Song; Bruce Miller; Gareth Mitchell

    2005-05-18

    This report summarizes the accomplishments toward project goals during the first six months of the second year of the project to assess the properties and performance of coal based products. These products are in the gasoline, diesel and fuel oil range and result from coal based jet fuel production from an Air Force funded program. Specific areas of progress include generation of coal based material that has been fractionated into the desired refinery cuts, acquisition and installation of a research gasoline engine, and modification of diesel engines for use in evaluating diesel produced in the project. The desulfurization of sulfur containing components of coal and petroleum is being studied so that effective conversion of blended coal and petroleum streams can be efficiently converted to useful refinery products. Equipment is now in place to begin fuel oil evaluations to assess the quality of coal based fuel oil. Coal samples have procured and are being assessed for cleaning prior to use in coking studies.

  8. Refinery Integration of By-Products from Coal-Derived Jet Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Leslie R. Rudnick; Andre Boehman; Chunshan Song; Bruce Miller; John Andresen

    2004-09-17

    This report summarizes the accomplishments toward project goals during the first twelve months of the project to assess the properties and performance of coal based products. These products are in the gasoline, diesel and fuel oil range and result from coal based jet fuel production from an Air Force funded program. Specific areas of progress include generation of coal based material that has been fractionated into the desired refinery cuts, acquisition and installation of a research gasoline engine, and modification of diesel engines for use in evaluating diesel produced in the project. The desulfurization of sulfur containing components of coal and petroleum is being studied so that effective conversion of blended coal and petroleum streams can be efficiently converted to useful refinery products. Equipment is now in place to begin fuel oil evaluations to assess the quality of coal based fuel oil. Coal samples have procured and are being assessed for cleaning prior to use in coking studies.

  9. Electronuclear fissile fuel production. Linear accelerator fuel regenerator and producer LAFR and LAFP

    International Nuclear Information System (INIS)

    Steinberg, M.; Powell, J.R.; Takahashi, H.; Grand, P.; Kouts, H.J.C.

    1978-04-01

    A linear accelerator fuel generator is proposed to enrich naturally occurring fertile U-238 or thorium 232 with fissile Pu-239 or U-233 for use in LWR power reactors. High energy proton beams in the range of 1 to 3 GeV energy are made to impinge on a centrally located dispersed liquid lead target producing spallation neutrons which are then absorbed by a surrounding assembly of fabricated LWR fuel elements. The accelerator-target design is reviewed and a typical fuel cycle system and economic analysis is presented. One 300 MW beam (300 ma-1 GeV) linear accelerator fuel regenerator can provide fuel for 3 to 1000 MW(e) LWR power reactors over its 30-year lifetime. There is a significant saving in natural uranium requirement which is a factor of 4.5 over the present LWR fuel requirement assuming the restraint of no fissile fuel recovery by reprocessing. A modest increase (approximately 10%) in fuel cycle and power production cost is incurred over the present LWR fuel cycle cost. The linear accelerator fuel regenerator and producer assures a long-term supply of fuel for the LWR power economy even with the restraint of the non-proliferation policy of no reprocessing. It can also supply hot-denatured thorium U-233 fuel operating in a secured reprocessing fuel center

  10. Characterization and chemistry of fission products released from LWR fuel under accident conditions

    International Nuclear Information System (INIS)

    Norwood, K.S.; Collins, J.L.; Osborne, M.F.; Lorenz, R.A.; Wichner, R.P.

    1984-01-01

    Segments from commercial LWR fuel rods have been tested at temperatures between 1400 and 2000 0 C in a flowing steam-helium atmosphere to simulate severe accident conditions. The primary goals of the tests were to determine the rate of fission product release and to characterize the chemical behavior. This paper is concerned primarily with the identification and chemical behavior of the released fission products with emphasis on antimony, cesium, iodine, and silver. The iodine appeared to behave primarily as cesium iodide and the antimony and silver as elements, while cesium behavior was much more complex. 17 refs., 7 figs., 1 tab

  11. Testing Systems and Results for Advanced Nuclear Fuel Materials

    International Nuclear Information System (INIS)

    Rooyen, I.J. van; Griffith, G.W.; Garnier, J.E.

    2012-01-01

    Light Water Reactor Sustainability (LWRS) Program Advanced LWR Nuclear Fuel Development (ALFD) Pathway. Development and testing of high performance fuel cladding identified as high priority to support: enhancement of fuel performance, reliability, and reactor safety. One of the technologies being examined is an advanced fuel cladding made from ceramic matrix composites (CMC) utilizing silicon carbide (SiC) as a structural material supplementing a commercial Zircaloy-4 (Zr-4) tube. A series of out-of-pile tests to fully characterize the SiC CMC hybrid design to produce baseline data. The planned tests are intended to either produce quantitative data or to demonstrate the properties required to achieve two initial performance conditions relative to standard zircaloybased cladding: decreased hydrogen uptake (corrosion) and decreased fretting of the cladding tube under normal operating and postulated accident conditions. These two failure mechanisms account for approximately 70% of all in-pile failures of LWR commercial fuel assemblies

  12. Experimental data report for Test TS-2 reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo

    1993-02-01

    This report presents experimental data for Test TS-2 which was the second test in a series of Reactivity Initiated Accident (RIA) condition test using pre-irradiated BWR fuel rods, performed at the Nuclear Safety Research Reactor (NSRR) in February, 1990. Test fuel rod used in the Test TS-2 was a short sized BWR (7x7) type rod which was fabricated from a commercial rod irradiated at Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79% and a burnup of 21.3Gwd/tU (bundle average). A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 72±5cal/g·fuel (66±5cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and, results of pre and post pulse irradiation examinations are described in this report. (author)

  13. Fuel elements and fuel element materials. Experimental facilities for fission products lift-off tests

    International Nuclear Information System (INIS)

    Blanchard, R.J.; Veyrat, J.F.

    1978-01-01

    One of the hypothetical accidents on the HTGR primary cooling circuits is the failure of a circuit resulting in a depressurization in the primary loops of the reactor. There is a risk of release of fission products in relation to the size of the failure. Experimental facilities for HTGR tests were developed: an in pile helium loop Comedie and an out of pile helium loop

  14. Fuel performance analysis for the HAMP-1 mini plate test

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Byoung Jin; Tahka, Y. W.; Yim, J. S.; Lee, B. H. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    U-7wt%Mo/Al- 5wt%Si dispersion fuel with 8gU/cm{sup 3} is chosen to achieve more efficiency and higher performance than the conventional U{sub 3}Si{sub 2} fuel. As part of the fuel qualification program for the KiJang research reactor (KJRR), three irradiation tests with mini-plates are on the way at the High-flux Advanced Neutron Application Reactor (HANARO). The first test among three HANARO Mini-Plate Irradiation tests (HAMP-1, 2, 3) has completed. PLATE code has been initially developed to analyze the thermal performance of high density U-Mo/Al dispersion fuel plates during irradiation [1]. We upgraded the PLATE code with the latest irradiation results which were implemented by corrosion, thermal conductivity and swelling model. Fuel performance analysis for HAMP-1 was conducted with updated PLATE. This paper presents results of performance evaluation of the HAMP-1. Maximum fuel temperature was obtained 136 .deg., which is far below the preset limit of 200 .deg. for the irradiation test. The meat swelling and corrosion thickness was also confirmed that the developed fuel would behave as anticipated.

  15. Production and quality control of fuel cladding tubes for LWRs

    International Nuclear Information System (INIS)

    Matsuda, Katsuhiko; Hagi, Shigeki; Anada, Hiroyuki; Abe, Hideaki; Hyodo, Shigetoshi

    1994-01-01

    This paper reviews the recent fabrication technology and corrosion resistance study of fuel cladding tubes for LWRs conducted by Sumitomo Metal Industries Ltd. started the research on zircaloy in 1957. In 1980, the factory exclusively for the production of cladding tubes was founded, and the mass production system on full scale was established. Thereafter, the various improvement of the production technology, the development of new products, and the heightening of the performance mainly on the corrosion resistance have been tested and studied. Recently, the works in the production processes were almost automated, and the installation of the production lines advanced, and the stabilization of product quality and the rationalization of costs are promoted. Moreover, the development of the zircaloy cladding tubes having high corrosion resistance has been advanced to cope with the long term cycle operation of LWRs hereafter. The features of zircaloy cladding tubes, the manufacturing processes, the improvement of the manufacturing technology, the improvement of the corrosion resistance and so on are reported. (K.I.)

  16. PRODUCTION OF NEW BIOMASS/WASTE-CONTAINING SOLID FUELS

    Energy Technology Data Exchange (ETDEWEB)

    David J. Akers; Glenn A. Shirey; Zalman Zitron; Charles Q. Maney

    2001-04-20

    CQ Inc. and its team members (ALSTOM Power Inc., Bliss Industries, McFadden Machine Company, and industry advisors from coal-burning utilities, equipment manufacturers, and the pellet fuels industry) addressed the objectives of the Department of Energy and industry to produce economical, new solid fuels from coal, biomass, and waste materials that reduce emissions from coal-fired boilers. This project builds on the team's commercial experience in composite fuels for energy production. The electric utility industry is interested in the use of biomass and wastes as fuel to reduce both emissions and fuel costs. In addition to these benefits, utilities also recognize the business advantage of consuming the waste byproducts of customers both to retain customers and to improve the public image of the industry. Unfortunately, biomass and waste byproducts can be troublesome fuels because of low bulk density, high moisture content, variable composition, handling and feeding problems, and inadequate information about combustion and emissions characteristics. Current methods of co-firing biomass and wastes either use a separate fuel receiving, storage, and boiler feed system, or mass burn the biomass by simply mixing it with coal on the storage pile. For biomass or biomass-containing composite fuels to be extensively used in the U.S., especially in the steam market, a lower cost method of producing these fuels must be developed that includes both moisture reduction and pelletization or agglomeration for necessary fuel density and ease of handling. Further, this method of fuel production must be applicable to a variety of combinations of biomass, wastes, and coal; economically competitive with current fuels; and provide environmental benefits compared with coal. Notable accomplishments from the work performed in Phase I of this project include the development of three standard fuel formulations from mixtures of coal fines, biomass, and waste materials that can be used in

  17. LMFBR fuel-design environment for endurance testing, primarily of oxide fuel elements with local faults

    International Nuclear Information System (INIS)

    Warinner, D.K.

    1980-01-01

    The US Department of Energy LMFBR Lines-of-Assurance are briefly stated and local faults are given perspective with an historical review and definition to help define the constraints of LMFBR fuel-element designs. Local-fault-propagation (fuel-element failure-propagation and blockage propagation) perceptions are reviewed. Fuel pin designs and major LMFBR parameters affecting pin performance are summarized. The interpretation of failed-fuel data is aided by a discussion of the effects of nonprototypicalities. The fuel-pin endurance expected in the US, USSR, France, UK, Japan, and West Germany is outlined. Finally, fuel-failure detection and location by delayed-neutron and gaseous-fission-product monitors are briefly discussed to better realize the operational limits

  18. Characterization of spent fuel approved testing material: ATM-106

    International Nuclear Information System (INIS)

    Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thornhill, C.K.

    1988-10-01

    The characterization data obtained to date are described for Approved Testing Material (ATM)-106 spent fuel from Assembly BT03 of pressurized-water reactor Calvert Cliffs No. 1. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well- characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCWRM) program. ATM-106 consists of 20 full-length irradiated fuel rods with rod-average burnups of about 3700 GJ/kgM (43 MWd/kgM) and expected fission gas release of /approximately/10%. Characterization data include (1) as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) calculated nuclide inventories and radioactivities in the fuel and cladding; and (6) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel rod are being conducted and will be included in planned revisions of this report. 12 refs., 110 figs., 81 tabs

  19. Instant release of fission products in leaching experiments with high burn-up nuclear fuels in the framework of the Euratom project FIRST- Nuclides

    Energy Technology Data Exchange (ETDEWEB)

    Lemmens, K., E-mail: klemmens@sckcen.be [Waste and Disposal Expert Group, Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, 2400 Mol (Belgium); González-Robles, E.; Kienzler, B. [Karlsruhe Institute of Technology Institute for Nuclear Waste Disposal (KIT-INE), PO Box 3640, D-76021 Karlsruhe (Germany); Curti, E. [Laboratory for Waste Management, Nuclear Energy and Safety Dept., Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Serrano-Purroy, D. [European Commission, DG Joint Research Centre - JRC, Directorate G - Nuclear Safety & Security, Department G.III, PO Box 2340, D-76125 Karlsruhe (Germany); Sureda, R.; Martínez-Torrents, A. [CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Roth, O. [Studsvik, Nuclear AB, 611 82 Nyköping (Sweden); Slonszki, E. [Magyar Tudományos Akadémia Energiatudományi Kutatóközpont (MTA EK), PO Box 49, H-1525 Budapest (Hungary); Mennecart, T. [Waste and Disposal Expert Group, Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, 2400 Mol (Belgium); Günther-Leopold, I. [Laboratory for Waste Management, Nuclear Energy and Safety Dept., Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Hózer, Z. [Magyar Tudományos Akadémia Energiatudományi Kutatóközpont (MTA EK), PO Box 49, H-1525 Budapest (Hungary)

    2017-02-15

    The instant release of fission products from high burn-up UO{sub 2} fuels and one MOX fuel was investigated by means of leach tests. The samples covered PWR and BWR fuels at average rod burn-up in the range of 45–63 GWd/t{sub HM} and included clad fuel segments, fuel segments with opened cladding, fuel fragments and fuel powder. The tests were performed with sodium chloride – bicarbonate solutions under oxidizing conditions and, for one test, in reducing Ar/H{sub 2} atmosphere. The iodine and cesium release could be partially explained by the differences in sample preparation, leading to different sizes and properties of the exposed surface areas. Iodine and cesium releases tend to correlate with FGR and linear power rating, but the scatter of the data is significant. Although the gap between the fuel and the cladding was closed in some high burn-up samples, fissures still provide possible preferential transport pathways. - Highlights: • Leach tests were performed to study the instant release of fission products from high burn-up UO{sub 2} fuels and one MOX fuel. • In these tests, the fission gas release given by the operator was a pessimistic estimator of the iodine and cesium release. • Iodine and cesium release is proportional to linear power rating beyond 200 W cm{sup −1}. • Closure of the fuel-cladding gap at high burn-up slows down the release. • The release rate decreases following an exponential equation.

  20. Spent fuel test. Climax data acquisition system integration report

    International Nuclear Information System (INIS)

    Nyholm, R.A.; Brough, W.G.; Rector, N.L.

    1982-06-01

    The Spent Fuel Test - Climax (SFT-C) is a test of the retrievable, deep geologic storage of commercially generated, spent nuclear reactor fuel in granitic rock. Eleven spent fuel assemblies, together with 6 electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the Nevada Test Site. On June 2, 1978, Lawrence Livermore National Laboratory (LLNL) secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. This multi-year duration test is located in a remote area and is unattended much of the time. An extensive array of radiological safety and geotechnical instrumentation is deployed to monitor the test performance. A dual minicomputer-based data acquisition system collects and processes data from more than 900 analog instruments. This report documents the design and functions of the hardware and software elements of the Data Acquisition System and describes the supporting facilities which include environmental enclosures, heating/air-conditioning/humidity systems, power distribution systems, fire suppression systems, remote terminal stations, telephone/modem communications, and workshop areas. 9 figures

  1. Spent Fuel Test - Climax data acquisition system operations manual

    International Nuclear Information System (INIS)

    Nyholm, R.A.

    1983-01-01

    The Spent Fuel Test-Climax (SFT-C) is a test of the retrievable, deep geologic storage of commercially generated, spent nuclear reactor fuel in granite rock. Eleven spent fuel assemblies, together with 6 electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the US Department of Energy Nevada Test Site. On June 2, 1978, Lawrence Livermore National Laboratory (LLNL) secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. The multi-year duration test is located in a remote area and is unattended much of the time. An extensive array of radiological safety and geotechnical instrumentation is deployed to monitor the test performance. A dual minicomputer-based data acquisition system (DAS) collects and processes data from more than 900 analog instruments. This report documents the software element of the LLNL developed SFT-C Data Acquisition System. It defines the operating system and hardware interface configurations, the special applications software and data structures, and support software

  2. RECH-1 test fuel irradiation status report

    International Nuclear Information System (INIS)

    Marin, J.; Lisboa, J.; Olivares, L.; Chavez, J.

    2005-01-01

    Since May 2003, one RECH-1 fuel element has been submitted to irradiation at the HFR-Petten, Holland. By November 2004 the irradiation has achieved its pursued goal of 55% burn up. This irradiation qualification service will finish in the year 2005 with PIE tests, as established in a contractual agreement between the IAEA, NRG, and CCHEN. This report presents the objectives and the current results of this fuel qualification under irradiation. Besides, a brief description of CHI/4/021, IAEA's Technical Cooperation Project that has supported this irradiation test, is also presented here. (author)

  3. Development of a non-engine fuel injector deposit test for alternative fuels (ENIAK-project)

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, Hajo; Pohland vom Schloss, Heide [OWI - Oel Waerme Institut GmbH, Herzogenrath (Germany)

    2013-06-01

    Deposit formation in and on the injectors of diesel engines may lead to injector malfunction, resulting in a loss in power, rough engine operation and poor emission levels. Poor Biodiesel quality, contamination with copper and zinc as well as undesired reactions between (several) additives and biodiesel components are known causes for nozzle fouling. Therefore, good housekeeping when using biodiesel is required, and all additives have to pass a no-harm test concerning injector fouling. The standard fouling tests are two engine tests: The XUD9-test (CEC F-23-01) and the DW-10-test (CEC DF 98-08). The XUD9 is a cost efficient, fast and proven testing method. It uses, however, an obsolete indirect injection diesel engine and cannot reproduce internal diesel injector deposits (IDID). The newer DW10 test is complex, costly and designed for high stress. This reduces the engine life and leads to a fuel consumption of approximately 1,000 1 per test, both contributing to the high costs of the test. The ENIAK-Project is funded by the FNR (''Fachagentur Nachwachsende Rohstoffe'', Agency for Renewable Resources) and conducted in cooperation with AGQM, ASG and ERC. Its main goal is the development, assembly, commissioning, and evaluation of a non-engine fuel injector test. It uses a complete common rail system. The injection takes place in a self-designed reactor instead of an engine, and the fuel is not combusted, but re-condensed and pumped in a circle, leading to a low amount of fuel required. If the test method proves to be as reliable as expected, it can be used as an alternative test method for injector fouling with low requirements regarding infrastructure on the testing site and sample volume. (orig.)

  4. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

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

  5. Energy Supply- Production of Fuel from Agricultural and Animal Waste

    Energy Technology Data Exchange (ETDEWEB)

    Gabriel Miller

    2009-03-25

    The Society for Energy and Environmental Research (SEER) was funded in March 2004 by the Department of Energy, under grant DE-FG-36-04GO14268, to produce a study, and oversee construction and implementation, for the thermo-chemical production of fuel from agricultural and animal waste. The grant focuses on the Changing World Technologies (CWT) of West Hempstead, NY, thermal conversion process (TCP), which converts animal residues and industrial food processing biproducts into fuels, and as an additional product, fertilizers. A commercial plant was designed and built by CWT, partially using grant funds, in Carthage, Missouri, to process animal residues from a nearby turkey processing plant. The DOE sponsored program consisted of four tasks. These were: Task 1 Optimization of the CWT Plant in Carthage - This task focused on advancing and optimizing the process plant operated by CWT that converts organic waste to fuel and energy. Task 2 Characterize and Validate Fuels Produced by CWT - This task focused on testing of bio-derived hydrocarbon fuels from the Carthage plant in power generating equipment to determine the regulatory compliance of emissions and overall performance of the fuel. Task 3 Characterize Mixed Waste Streams - This task focused on studies performed at Princeton University to better characterize mixed waste incoming streams from animal and vegetable residues. Task 4 Fundamental Research in Waste Processing Technologies - This task focused on studies performed at the Massachusetts Institute of Technology (MIT) on the chemical reformation reaction of agricultural biomass compounds in a hydrothermal medium. Many of the challenges to optimize, improve and perfect the technology, equipment and processes in order to provide an economically viable means of creating sustainable energy were identified in the DOE Stage Gate Review, whose summary report was issued on July 30, 2004. This summary report appears herein as Appendix 1, and the findings of the report

  6. Autonomy-Enabled Fuel Savings for Military Vehicles: Report on 2016 Aberdeen Test Center Testing

    Energy Technology Data Exchange (ETDEWEB)

    Ragatz, Adam [National Renewable Energy Lab. (NREL), Golden, CO (United States); Prohaska, Robert [National Renewable Energy Lab. (NREL), Golden, CO (United States); Gonder, Jeff [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2017-05-26

    Fuel savings have never been the primary focus for autonomy-enabled military vehicles. However, studies have estimated that autonomy in passenger and commercial vehicles could improve fuel economy by as much as 22%-33% over various drive cycles. If even a fraction of this saving could be realized in military vehicles, significant cost savings could be realized each year through reduced fuel transport missions, reduced fuel purchases, less maintenance, fewer required personnel, and increased vehicle range. Researchers from the National Renewable Energy Laboratory installed advanced data logging equipment and instrumentation on two autonomy-enabled convoy vehicles configured with Lockheed Martin's Autonomous Mobility Applique System to determine system performance and improve on the overall vehicle control strategies of the vehicles. Initial test results from testing conducted at the U.S. Army Aberdeen Test Center at the Aberdeen Proving Grounds are included in this report. Lessons learned from in-use testing and performance results have been provided to the project partners for continued system refinement.

  7. Design criteria of out-pile system of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1997-07-01

    The objective of HANARO aims at the development and localization of nuclear technologies through the engineering tests. Thus it is very important the design and installation of the irradiation test facilities to be installed at the irradiation hole for verification test of the fuel performance are in connection with maximization of the utilization of HANARO. The principle subjects of this study are to presend and informed the detail design criteria and technical specification of out-pile system of HANARO fuel test loop for the developing of the fuel and reactor material. This results will become guidance for the planning of the irradiation testing using the HANARO fuel test loop. (author). 16 refs., 31 tabs., 9 figs.

  8. Characterization of spent fuel approved testing material---ATM-105

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

    1991-12-01

    The characterization data obtained to data are described for Approved Testing Material 105 (ATM-105), which is spent fuel from Bundles CZ346 and CZ348 of the Cooper Nuclear Power Plant, a boiling-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-105 consists of 88 full-length irradiated fuel rods with rod-average burnups of about 2400 GJ/kgM (28 MWd/kgM) and expected fission gas release of about 1%. Characterization data include (1) descriptions of as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel are being conducted and will be included in planned revisions of this report.

  9. Use of MRF residue as alternative fuel in cement production.

    Science.gov (United States)

    Fyffe, John R; Breckel, Alex C; Townsend, Aaron K; Webber, Michael E

    2016-01-01

    Single-stream recycling has helped divert millions of metric tons of waste from landfills in the U.S., where recycling rates for municipal solid waste are currently over 30%. However, material recovery facilities (MRFs) that sort the municipal recycled streams do not recover 100% of the incoming material. Consequently, they landfill between 5% and 15% of total processed material as residue. This residue is primarily composed of high-energy-content non-recycled plastics and fiber. One possible end-of-life solution for these energy-dense materials is to process the residue into Solid Recovered Fuel (SRF) that can be used as an alternative energy resource capable of replacing or supplementing fuel resources such as coal, natural gas, petroleum coke, or biomass in many industrial and power production processes. This report addresses the energetic and environmental benefits and trade-offs of converting non-recycled post-consumer plastics and fiber derived from MRF residue streams into SRF for use in a cement kiln. An experimental test burn of 118 Mg of SRF in the precalciner portion of the cement kiln was conducted. The SRF was a blend of 60% MRF residue and 40% post-industrial waste products producing an estimated 60% plastic and 40% fibrous material mixture. The SRF was fed into the kiln at 0.9 Mg/h for 24h and then 1.8 Mg/h for the following 48 h. The emissions data recorded in the experimental test burn were used to perform the life-cycle analysis portion of this study. The analysis included the following steps: transportation, landfill, processing and fuel combustion at the cement kiln. The energy use and emissions at each step is tracked for the two cases: (1) The Reference Case, where MRF residue is disposed of in a landfill and the cement kiln uses coal as its fuel source, and (2) The SRF Case, in which MRF residue is processed into SRF and used to offset some portion of coal use at the cement kiln. The experimental test burn and accompanying analysis indicate

  10. Logistics of nuclear fuel production for nuclear submarines

    International Nuclear Information System (INIS)

    Guimaraes, Leonam dos Santos

    2000-01-01

    The future acquisition of nuclear attack submarines by Brazilian Navy along next century will imply new requirements on Naval Logistic Support System. These needs will impact all the six logistic functions. Among them, fuel supply could be considered as the one which requires the most important capacitating effort, including not only technological development of processes but also the development of a national industrial basis for effective production of nuclear fuel. This paper presents the technical aspects of the processes involved and an annual production dimensioning for an squadron composed by four units. (author)

  11. Fabrication, inspection, and test plan for the Advanced Test Reactor (ATR) Mixed-Oxide (MOX) fuel irradiation project

    International Nuclear Information System (INIS)

    Wachs, G.W.

    1997-11-01

    The Department of Energy (DOE) Fissile Materials Disposition Materials Disposition Program (FMDP) has announced that reactor irradiation of MOX fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The MOX fuel test will be irradiated in the ATR to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. In addition, the test will contribute experience with irradiation of gallium-containing fuel to the data base required for resolution of generic CLWR fuel design issues (ORNL/MD/LTR-76). This Fabrication, Inspection, and Test Plan (FITP) is a level 2 document as defined in the FMDP LWR MOX Fuel Irradiation Test Project Plan (ORNL/MD/LTR-78)

  12. 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

  13. Examinations of the irradiation behaviour of U3Si2 test fuel plates with low enrichment

    International Nuclear Information System (INIS)

    Muellauer, J.

    1989-01-01

    Five low-enriched (19.7% 235 U), high-density (4.7 gU/cm/ 3 ) U 3 Si 2 -test fuel plates (miniplates) with different fine grain contents have been qualified under irradiation. During the course of irradiation up to burnup of 63% 235 U depletion, no released fractions of gaseous or solid fission products from the fuel plate to the rig coolant were detected. The measured swelling rate of the fuel zone (meat) is less than 0.45% ΔV/10 20 fissions/cm 3 the blister-threshold temperature of the fuel plates is above 520 0 C. The favourable irradiation behavior of the U 3 Si 2 fuel plates was not influenced by using higher amounts of fine grained particles (40% [de

  14. Development of TIG Welding System for a Nuclear Fuel Test Rig

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Changyoung; Ahn, Sungho; Hong, Jintae; Kim, Kahye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rig and rods. To manufacture the nuclear fuel test rig, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rig and rods jointing the various sensors and end caps on a fuel cladding tube, which is charged with fuel pellets and component parts. Thus, we designed and fabricated the precision welding system consisting of an orbital TIG welder, a low-pressure chamber, and a high-pressure chamber. Using this system, the performance tests were performed with the round and seal spot welds for each welding condition. This paper describes not only the contents for the fabrication of precision TIG welding system but also some results from weld tests using the low-pressure and high-pressure chambers to verify the performance of this system. The TIG welding system was developed to manufacture the nuclear fuel test rig and rods. It has been configured to be able to weld the nuclear fuel test rigs and rods by applying the TIG welder using a low-pressure chamber and a high-pressure chamber. The performance tests using this system were performed with the round and seal spot welds for the welding conditions. The soundness of the orbital TIG welding system was confirmed through performance tests in the low-pressure and high-pressure chambers.

  15. Development of TIG Welding System for a Nuclear Fuel Test Rig

    International Nuclear Information System (INIS)

    Joung, Changyoung; Ahn, Sungho; Hong, Jintae; Kim, Kahye

    2013-01-01

    The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rig and rods. To manufacture the nuclear fuel test rig, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rig and rods jointing the various sensors and end caps on a fuel cladding tube, which is charged with fuel pellets and component parts. Thus, we designed and fabricated the precision welding system consisting of an orbital TIG welder, a low-pressure chamber, and a high-pressure chamber. Using this system, the performance tests were performed with the round and seal spot welds for each welding condition. This paper describes not only the contents for the fabrication of precision TIG welding system but also some results from weld tests using the low-pressure and high-pressure chambers to verify the performance of this system. The TIG welding system was developed to manufacture the nuclear fuel test rig and rods. It has been configured to be able to weld the nuclear fuel test rigs and rods by applying the TIG welder using a low-pressure chamber and a high-pressure chamber. The performance tests using this system were performed with the round and seal spot welds for the welding conditions. The soundness of the orbital TIG welding system was confirmed through performance tests in the low-pressure and high-pressure chambers

  16. Experimental data report for Test TS-1 Reactivity Initiated Accident Test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Horiki, Ohichiro; Yamahara, Takeshi; Ichihashi, Yoshinori; Kikuchi, Teruo

    1992-01-01

    This report presents experimental data for Test TS-1 which was the first in a series of tests, simulating Reactivity Initiated Accident (RIA) conditions using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in October, 1989. Test fuel rod used in the Test TS-1 was a short-sized BWR (7 x 7) type rod which was fabricated from a commercial rod provided from Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79 % and burnup of 21.3 GWd/t (bundle average). Pulse irradiation was performed at a condition of stagnant water cooling, atmospheric pressure and ambient temperature using a newly developed double container-type capsule. Energy deposition of the rod in this test was evaluated to be about 61 cal/g·fuel (55 cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, fuel burnup measurements, transient behavior of the test rod during pulse irradiation and results of post pulse irradiation examinations are contained in this report. (author)

  17. Pie technique of LWR fuel cladding fracture toughness test

    International Nuclear Information System (INIS)

    Endo, Shinya; Usami, Koji; Nakata, Masahito; Fukuda, Takuji; Numata, Masami; Kizaki, Minoru; Nishino, Yasuharu

    2006-01-01

    Remote-handling techniques were developed by cooperative research between the Department of Hot Laboratories in the Japan Atomic Energy Research Institute (JAERI) and the Nuclear Fuel Industries Ltd. (NFI) for evaluating the fracture toughness on irradiated LWR fuel cladding. The developed techniques, sample machining by using the electrical discharge machine (EDM), pre-cracking by fatigue tester, sample assembling to the compact tension (CT) shaped test fixture gave a satisfied result for a fracture toughness test developed by NFL. And post-irradiation examination (PIE) using the remote-handling techniques were carried out to evaluate the fracture toughness on BWR spent fuel cladding in the Waste Safety Testing Facility (WASTEF). (author)

  18. Nuclear fuel particle and method of production

    International Nuclear Information System (INIS)

    Wagner-Loffler, M.

    1975-01-01

    The core consisting of fuel oxide (UO 2 or Th or Pu oxide) of a fuel particle coated with carbon-contained material is enriched with a small addition (max 6 wt.%) of a Ba or Sr compound (atomic ratio for nuclear fuel oxide Ba being 5 - 10 : 1) which is to prevent fission products breaking the protective carbon and/or silicon carbide coating; the Ba or Sr molybdate generated is to reduce the pressure of the carbon dioxide produced. Methods to manufacture such nuclear fuel particles are proposed where 1) an agglomerisation and shaping of the spheres in a fast cycling bowle and 2) a formation of drops from a colloidal solution which are made to congeal in a liquid paraffin column, take place followed by the pyrolytic coating of the particles. (UWI/LH) [de

  19. Demonstration tests for HTGR fuel elements and core components with test sections in HENDEL

    Energy Technology Data Exchange (ETDEWEB)

    Miyamoto, Yoshiaki; Hino, Ryutaro; Inagaki, Yoshiyuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1995-03-01

    In the fuel stack test section (T{sub 1}) of the Helium Engineering Demonstration Loop (HENDEL), thermal and hydraulic performances of helium gas flows through a fuel rod channel and a fuel stack have been investigated for the High-Temperature Engineering Test Reactor (HTTR) core thermal design. The test data showed that the turbulent characteristics appearing in the Reynolds number above 2000: no typical behavior in the transition zone, and friction factors and heat transfer coefficients in the fuel channel were found to be higher than those in a smooth annular channel. Heat transfer behavior of gas flow in a fuel element channel with blockage and cross-flow through a gap between upper and lower fuel elements stacked was revealed using the mock-up models. On the other hand, demonstration tests have been performed to verify thermal and hydraulic characteristics and structural integrity related to the core bottom structure using a full-scale test facility named as the in-core structure test section (T{sub 2}). The sealing performance test revealed that the leakage of low-temperature helium gas through gaps between the permanent reflector blocks to the core was very low level compared with the HTTR design value and no change of the leakage flow rate were observed after a long term operation. The heat transfer tests including thermal transient at shutdown of gas circulators verified good insulating performance of core insulation structures in the core bottom structure and the hot gas duct; the temperature of the metal portion of these structure was below the design value. Examination of the thermal mixing characteristics indicated that the mixing of the hot helium gas started at a hot plenum and finished completely at downstream of the outlet hot gas duct. The present results obtained from these demonstration tests have been practically applied to the detailed design works and licensing procedures of the HTTR. (J.P.N.) 92 refs.

  20. Corrosion of spent Advanced Test Reactor fuel

    International Nuclear Information System (INIS)

    Lundberg, L.B.; Croson, M.L.

    1994-01-01

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

  1. Desludging of N Reactor fuel canisters: Analysis, Test, and data requirements

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.

    1996-01-01

    The N Reactor fuel is currently stored in canisters in the K East (KE) and K West (KW) Basins. In KE, the canisters have open tops; in KW, the cans have sealed lids, but are vented to release gases. Corrosion products have formed on exposed uranium metal fuel, on carbon steel basin component surfaces, and on aluminum alloy canister surfaces. Much of the corrosion product is retained on the corroding surfaces; however, large inventories of particulates have been released. Some of the corrosion product particulates form sludge on the basin floors; some particulates are retained within the canisters. The floor sludge inventories are much greater in the KE Basin than in the KW Basin because KE Basin operated longer and its water chemistry was less controlled. Another important factor is the absence of lids on the KE canisters, allowing uranium corrosion products to escape and water-borne species, principally iron oxides, to settle in the canisters. The inventories of corrosion products, including those released as particulates inside the canisters, are only beginning to be characterized for the closed canisters in KW Basin. The dominant species in the KE floor sludge are oxides of aluminum, iron, and uranium. A large fraction of the aluminum and uranium floor sludge particulates may have been released during a major fuel segregation campaign in the 1980s, when fuel was emptied from 4990 canisters. Handling and jarring of the fuel and aluminum canisters seems likely to have released particulates from the heavily corroded surfaces. Four candidate methods are discussed for dealing with canister sludge emerged in the N Reactor fuel path forward: place fuel in multi-canister overpacks (MCOs) without desludging; drill holes in canisters and drain; drill holes in canisters and flush with water; and remove sludge and repackage the fuel

  2. Lignin depolymerization and upgrading via fast pyrolysis and electrocatalysis for the production of liquid fuels and value-added products

    Science.gov (United States)

    Garedew, Mahlet

    The production of liquid hydrocarbon fuels from biomass is needed to replace fossil fuels, which are decreasing in supply at an unsustainable rate. Renewable fuels also address the rising levels of greenhouse gases, an issue for which the Intergovernmental Panel on Climate Change implicated humanity in 2013. In response, the Energy Independence and Security Act (EISA) mandates the production of 21 billion gallons of advanced biofuels by 2022. Biomass fast pyrolysis (BFP) uses heat (400-600 °C) without oxygen to convert biomass to liquids fuel precursors offering an alternative to fossil fuels and a means to meet the EISA mandate. The major product, bio-oil, can be further upgraded to liquid hydrocarbon fuels, while biochar can serve as a solid fuel or soil amendment. The combustible gas co-product is typically burned for process heat. Though the most valuable of the pyrolysis products, the liquid bio-oil is highly oxygenated, corrosive, low in energy content and unstable during storage. As a means of improving bio-oil properties, electrocatalytic hydrogenation (ECH) is employed to reduce and deoxygenate reactive compounds. This work specifically focuses on lignin as a feed material for BFP. As lignin comprises up to 30% of the mass and 40% of the energy stored in biomass, it offers great potential for the production of liquid fuels and value-added products by utilizing fast pyrolysis as a conversion method coupled with electrocatalysis as an upgrading method.

  3. Economic consequences of QA and QC in fuel and fuel assembly production

    International Nuclear Information System (INIS)

    Strasser, A.A.

    1984-01-01

    The planning of quality control and quality assurance programs for fuel fabrication must balance the cost of the programs, their effectiveness, and the economic consequences of failure to meet the product specifications. The cost of fuel failures can be very high in comparison to the cost of quality control, and this provides considerable economic justification for increasing the level of quality control if its effectiveness in reducing failure potential can be demonstrated. Typical costs and examples are discussed. (orig.)

  4. Final report of fuel dynamics Test E7

    International Nuclear Information System (INIS)

    Doerner, R.C.; Murphy, W.F.; Stanford, G.S.; Froehle, P.H.

    1977-04-01

    Test data from an in-pile failure experiment of high-power LMFBR-type fuel pins in a simulated $3/s transient-overpower (TOP) accident are reported and analyzed. Major conclusions are that (1) a series of cladding ruptures during the 100-ms period preceding fuel release injected small bursts of fission gas into the flow stream; (2) gas release influenced subsequent cladding melting and fuel release [there were no measurable FCI's (fuel-coolant interactions), and all fuel motion observed by the hodoscope was very slow]; (3) the predominant postfailure fuel motion appears to be radial swelling that left a spongy fuel crust on the holder wall; (4) less than 4 to 6 percent of the fuel moved axially out of the original fuel zone, and most of this froze within a 10-cm region above the original top of the fuel zone to form the outlet blockage. An inlet blockage approximately 1 cm long was formed and consisted of large interconnected void regions. Both blockages began just beyond the ends of the fuel pellets

  5. New JMTR irradiation test plan on fuels and materials

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Nishiyama, Yutaka; Chimi, Yasuhiro; Sasajima, Hideo; Ogiyanagi, Jin; Nakamura, Jinichi; Suzuki, Masahide; Kawamura, Hiroshi

    2009-01-01

    In order to maintain and enhance safety of light water reactors (LWRs) in long-term and up-graded operations, proper understanding of irradiation behavior of fuels and materials is essentially important. Japanese government and the Japan Atomic Energy Agency (JAEA) have decided to refurbish the Japan Materials Testing Reactor (JMTR) and to install new tests rigs, in order to play an active role for solving irradiation related issues on plant aging and high-duty uses of the current LWRs and on development of next-generation reactors. New tests on fuel integrity under simulated abnormal transients and high-duty irradiation conditions are planned in the JMTR. Power ramp tests of newdesign fuel rods will also be performed in the first stage of the program, which is expected to start in year 2011 after refurbishment of the JMTR. Combination of the JMTR tests with simulated reactivity initiated accident tests in the Nuclear Safety Research Reactor (NSRR) and loss of coolant accident tests in hot laboratories would serve as the integrated fuel safety research on the high performance fuels at extended burnups, covering from the normal to the accident conditions, including abnormal transients. For the materials irradiation, fracture toughness of reactor vessel steels and stress corrosion cracking behavior of stainless steels are being studied in addition to basic irradiation behavior of nuclear materials such as hafnium. The irradiation studies would contribute not only to solve the current problems but also to identify possible seeds of troubles and to make proactive responses. (author)

  6. Spent fuel test-climax: a test of geologic storage of high-level waste in granite

    International Nuclear Information System (INIS)

    Ramspott, L.D.; Ballou, L.B.; Patrick, W.C.

    1981-01-01

    A test of retrievable geologic storage of spent fuel assemblies from an operating commercial nuclear reactor is underway at the Nevada Test Site (NTS) of the US Department of Energy. This generic test is located 420 m below the surface in the Climax granitic stock. Eleven canisters of spent fuel approximately 2.5 years out of reactor core (about 1.6 kW/canister thermal output) were emplaced in a storage drift along with 6 electrical simulator canisters. Two adjacent drifts contain electrical heaters, which are operated to simulate within the test array the thermal field of a large repository. Fuel was loaded during April to May 1980 and initial results of the test will be presented

  7. Technical and economic modelling of processes for liquid fuel production in Europe

    International Nuclear Information System (INIS)

    Bridgwater, A.V.; Double, J.M.

    1991-01-01

    The project which is described had the objective of examining the full range of technologies for liquid fuel production from renewable feedstocks in a technical and economic evaluation in order to identify the most promising technologies. The technologies considered are indirect thermochemical liquefaction (i.e. via gasification) to produce methanol, fuel alcohol or hydrocarbon fuels, direct thermochemical liquefaction or pyrolysis to produce hydrocarbon fuels and fermentation to produce ethanol. Feedstocks considered were wood, refuse derived fuel, straw, wheat and sugar beet. In order to carry out the evaluation, a computer model was developed, based on a unit process approach. Each unit operation is modelled as a process step, the model calculating the mass balance, energy balance and operating cost of the unit process. The results from the process step models are then combined to generate the mass balance, energy balance, capital cost and operating cost for the total process. The results show that the lowest production cost (L7/GJ) is obtained for methanol generated from a straw feedstock, but there is a moderate level of technical uncertainty associated with this result. The lowest production cost for hydrocarbon fuel (L8.6/GJ) is given by the pyrolysis process using a wood feedstock. This process has a high level of uncertainty. Fermentation processes showed the highest production costs, ranging from L14.4/GJ for a simple wood feedstock process to L25.2/GJ for a process based on sugar beet. The important conclusions are as follows: - In every case, the product cost is above current liquid fuel prices; - In most cases the feedstock cost dominates the production cost; -The most attractive products are thermochemically produced alcohol fuels

  8. Technical and economic modelling of processes for liquid fuel production in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Bridgwater, A V; Double, J M [Aston Univ. Birmingham (GB). Dept of Chemical Engineering

    1992-12-31

    The project which is described had the objective of examining the full range of technologies for liquid fuel production from renewable feedstocks in a technical and economic evaluation in order to identify the most promising technologies. The technologies considered are indirect thermochemical liquefaction (i.e. via gasification) to produce methanol, fuel alcohol or hydrocarbon fuels, direct thermochemical liquefaction or pyrolysis to produce hydrocarbon fuels and fermentation to produce ethanol. Feedstocks considered were wood, refuse derived fuel, straw, wheat and sugar beet. In order to carry out the evaluation, a computer model was developed, based on a unit process approach. Each unit operation is modelled as a process step, the model calculating the mass balance, energy balance and operating cost of the unit process. The results from the process step models are then combined to generate the mass balance, energy balance, capital cost and operating cost for the total process. The results show that the lowest production cost (L7/GJ) is obtained for methanol generated from a straw feedstock, but there is a moderate level of technical uncertainty associated with this result. The lowest production cost for hydrocarbon fuel (L8.6/GJ) is given by the pyrolysis process using a wood feedstock. This process has a high level of uncertainty. Fermentation processes showed the highest production costs, ranging from L14.4/GJ for a simple wood feedstock process to L25.2/GJ for a process based on sugar beet. The important conclusions are as follows: - In every case, the product cost is above current liquid fuel prices; - In most cases the feedstock cost dominates the production cost; -The most attractive products are thermochemically produced alcohol fuels.

  9. Thermophysical properties of the products of low-grade fuels thermal recycling

    Directory of Open Access Journals (Sweden)

    Tabakaev Roman B.

    2015-01-01

    Full Text Available The relevance of the work is caused by reorientation of the modern power engineering to use of local low grade fuel resources. Some types of low grade fuels (peat, brown coal, sapropel, wood chips are considered in this work. Thermotechnical characteristics of the investigated fuels and products of their thermal recycling are determined. Thermal recycling process is accompanied by release of fuel dissociation heat (0.33-3.69 MJ/kg. The results of thermal low grade fuel recycling are solid carbonaceous product (semi-coke with a calorific value higher in 1.5-7 times than the value of natural fuels; pyrolysis resin with calorific value 29.4-36.8 MJ/kg; combustible gas with calorific value 15.16-19.06 MJ/m3.

  10. Fuel performance and fission product behaviour in gas cooled reactors

    International Nuclear Information System (INIS)

    1997-11-01

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport

  11. Performance evaluation of UO2-Zr fuel in power ramp tests

    International Nuclear Information System (INIS)

    Knudsen, P.; Bagger, C.

    1977-01-01

    In power reactors using UO 2 -Zr fuel, rapid power increases may lead to failures in fuel pins that have been irradiated at steady or decreasing heat loads. This paper presents results which extend the experience with power ramp performance of high burn-up fuel pins. A test fuel element containing both pellet and vipac UO 2 -Zr fuel pins was irradiated in the HBWR at Halden for effectively 2 1/2 years to an average burn-up of 21,000 MWD/te UO 2 at gradually decreasing power levels. The subsequent non-destructive characterization revealed formation of transverse cracks in the vipac fuel columns. After the HBWR irradiation, five of the fuel pins were power ramp tested individually in the DR 3 Reactor at Riso. The ramp rates in this test series were in the range 3-60 W/cm min. The maximum local heat loads seen in the ramp tests were 20-120% above the highest levels experienced at the same axial positions during the HBWR irradiation. Three pellets and one vipac fuel pin failed, whereas another vipac pin gave no indication of clad penetration. Profilometry after the ramp testing indicated the formation of small ridges for both types of fuel pins. For vipac fuel, the ridges were less regularly distributed along the pin length than for pellet fuel. Neutron radiography revealed the formation of additional transverse and longitudinal fuel cracks during the power ramps for both types of fuel pins. The observed failures seemed to be marginal since little or no indication as to the locations of the clad penetrations could be derived from the non-destructive post-irradiation examinations. The cases have been analyzed by means of the Danish fuel performance codes. The calculations, which are in general agreement with the observations, are discussed. The results of the investigations indicate qualitative similarities in over power performance of the two fuel types

  12. Masonry fireplace emissions test method: Repeatability and sensitivity to fueling protocol.

    Science.gov (United States)

    Stern, C H; Jaasma, D R; Champion, M R

    1993-03-01

    A test method for masonry fireplaces has been evaluated during testing on six masonry fireplace configurations. The method determines carbon monoxide and particulate matter emission rates (g/h) and factors (g/kg) and does not require weighing of the appliance to determine the timing of fuel loading.The intralaboratory repeatability of the test method has been determined from multiple tests on the six fireplaces. For the tested fireplaces, the ratio of the highest to lowest measured PM rate averaged 1.17 and in no case was greater than 1.32. The data suggest that some of the variation is due to differences in fuel properties.The influence of fueling protocol on emissions has also been studied. A modified fueling protocol, tested in large and small fireplaces, reduced CO and PM emission factors by roughly 40% and reduced CO and PM rates from 0 to 30%. For both of these fireplaces, emission rates were less sensitive to fueling protocol than emission factors.

  13. 76 FR 5319 - Regulation of Fuel and Fuel Additives: Alternative Test Method for Olefins in Gasoline

    Science.gov (United States)

    2011-01-31

    ... Regulation of Fuel and Fuel Additives: Alternative Test Method for Olefins in Gasoline AGENCY: Environmental... gasoline. This proposed rule will provide flexibility to the regulated community by allowing an additional... A. Alternative Test Method for Olefins in Gasoline III. Statutory and Executive Order Reviews A...

  14. Fission product release from HTGR coated microparticles and fuel elements

    International Nuclear Information System (INIS)

    Gusev, A.A.; Deryugin, A.I.; Lyutikov, R.A.; Chernikov, A.S.

    1991-01-01

    The article presents the results of the investigation of fission products release from microparticles with UO 2 core and five-layer HII PyC- and SiC base protection layers of TRICO type as well as from spherical fuel elements based thereon. It is shown that relative release of short-lived xenon and crypton from microparticles does not exceed (2-3) 10 -7 . The release of gaseous fission products from fuel elements containing no damaged coated microparticles, is primarily determined by the contamination of matrix graphite with fuel. An analytical dependence is derived, the dependence described the relation between structural parameters of coated microparticles, irradiation conditions and fuel burnup at which depressurization of coated microparticles starts

  15. Conversion of research and test reactors to low enriched uranium fuel: technical overview and program status

    International Nuclear Information System (INIS)

    Roglans-Ribas, J.

    2008-01-01

    Many of the nuclear research and test reactors worldwide operate with high enriched uranium fuel. In response to worries over the potential use of HEU from research reactors in nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel by converting research reactors to low enriched uranium (LEU) fuel. The Reactor Conversion program is currently under the DOE's National Nuclear Security Administration's Global Threat Reduction Initiative (GTRI). 55 of the 129 reactors included in the scope have been already converted to LEU fuel or have shutdown prior to conversion. The major technical activities of the Conversion Program include: (1) the development of advanced LEU fuels; (2) conversion analysis and conversion support; and (3) technology development for the production of Molybdenum-99 (Mo 99 ) with LEU targets. The paper provides an overview of the status of the program, the technical challenges and accomplishments, and the role of international collaborations in the accomplishment of the Conversion Program objectives. Nuclear research and test reactors worldwide have been in operation for over 60 years. Many of these facilities operate with high enriched uranium fuel. In response to increased worries over the potential use of HEU from research reactors in the manufacturing of nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel in research reactors by converting them to low enriched uranium (LEU) fuel. The reactor conversion program was initially focused on U.S.-supplied reactors, but in the early 1990s it expanded and began to collaborate with Russian institutes with the objective of converting Russian supplied reactors to the use of LEU fuel.

  16. Mechanical characterization tests of the X2-Gen fuel assembly and skeleton

    International Nuclear Information System (INIS)

    Kim, Hyung Kyu; Yoon, Kyung Ho; Lee, Kang Hee; Kim, Jae Yong; Lee, Young Ho; Kang, Heung Seok

    2011-01-01

    The KNF (KEPCO Nuclear Fuel) requested mechanical characterization tests of a fuel assembly and a skeleton of the X2-Gen fuel. The tests consisted of the lateral vibration and lateral/axial stiffness, lateral/axial impact and combined deflection tests carried out by using the FAMeCT (Fuel Assembly Mechanical Characterization Tester) in KAERI. The upper and lower core plate simulators were newly designed and manufactured because the fuel geometry of the X2-Gen was different from the KSNP type fuel assembly. In addition to this, the upper carriage was also revised with the LM guide system from the previous two guide rods system. Therefore, the axial and combined deflection tests were soundly executed. Each test was repeated twice to confirm the repeatability. The discrepancy from the repetition was small enough to be neglected. The mechanical characterization tests were accredited with the KOLAS (Korea Laboratory Accreditation Scheme) standard, and the certified test reports (lateral vibration, lateral/axial bending and lateral/axial impact) and the uncertified test report (combined deflection) were issued together with the current test result report

  17. Fission product phases in irradiated carbide fuels

    International Nuclear Information System (INIS)

    Ewart, F.T.; Sharpe, B.M.; Taylor, R.G.

    1975-09-01

    Oxide fuels have been widely adopted as 'first charge' fuels for demonstration fast reactors. However, because of the improved breeding characteristics, carbides are being investigated in a number of laboratories as possible advanced fuels. Irradiation experiments on uranium and mixed uranium-plutonium carbides have been widely reported but the instances where segregate phases have been found and subjected to electron probe analysis are relatively few. Several observations of such segregate phases have now been made over a period of time and these are collected together in this document. Some seven fuel pins have been examined. Two of the irradiations were in thermal materials testing reactors (MTR); the remainder were experimental assemblies of carbide gas bonded oxycarbide and sodium bonded oxycarbide in the Dounreay Fast Reactor (DFR). All fuel pins completed their irradiation without failure. (author)

  18. Development of the advanced PHWR technology -Verification tests for CANDU advanced fuel-

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jang Hwan; Suk, Hoh Chun; Jung, Moon Kee; Oh, Duk Joo; Park, Joo Hwan; Shim, Kee Sub; Jang, Suk Kyoo; Jung, Heung Joon; Park, Jin Suk; Jung, Seung Hoh; Jun, Ji Soo; Lee, Yung Wook; Jung, Chang Joon; Byun, Taek Sang; Park, Kwang Suk; Kim, Bok Deuk; Min, Kyung Hoh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This is the `94 annual report of the CANDU advanced fuel verification test project. This report describes the out-of pile hydraulic tests at CANDU-hot test loop for verification of CANFLEX fuel bundle. It is also describes the reactor thermal-hydraulic analysis for thermal margin and flow stability. The contents in this report are as follows; (1) Out-of pile hydraulic tests for verification of CANFLEX fuel bundle. (a) Pressure drop tests at reactor operation condition (b) Strength test during reload at static condition (c) Impact test during reload at impact load condition (d) Endurance test for verification of fuel integrity during life time (2) Reactor thermal-hydraulic analysis with CANFLEX fuel bundle. (a) Critical channel power sensitivity analysis (b) CANDU-6 channel flow analysis (c) Flow instability analysis. 61 figs, 29 tabs, 21 refs. (Author).

  19. Development of the advanced PHWR technology -Verification tests for CANDU advanced fuel-

    International Nuclear Information System (INIS)

    Jung, Jang Hwan; Suk, Hoh Chun; Jung, Moon Kee; Oh, Duk Joo; Park, Joo Hwan; Shim, Kee Sub; Jang, Suk Kyoo; Jung, Heung Joon; Park, Jin Suk; Jung, Seung Hoh; Jun, Ji Soo; Lee, Yung Wook; Jung, Chang Joon; Byun, Taek Sang; Park, Kwang Suk; Kim, Bok Deuk; Min, Kyung Hoh

    1995-07-01

    This is the '94 annual report of the CANDU advanced fuel verification test project. This report describes the out-of pile hydraulic tests at CANDU-hot test loop for verification of CANFLEX fuel bundle. It is also describes the reactor thermal-hydraulic analysis for thermal margin and flow stability. The contents in this report are as follows; (1) Out-of pile hydraulic tests for verification of CANFLEX fuel bundle. (a) Pressure drop tests at reactor operation condition (b) Strength test during reload at static condition (c) Impact test during reload at impact load condition (d) Endurance test for verification of fuel integrity during life time (2) Reactor thermal-hydraulic analysis with CANFLEX fuel bundle. (a) Critical channel power sensitivity analysis (b) CANDU-6 channel flow analysis (c) Flow instability analysis. 61 figs, 29 tabs, 21 refs. (Author)

  20. Alpha Fuels Environmental Test Facility impact gun

    International Nuclear Information System (INIS)

    Anderson, C.G.

    1978-01-01

    The Alpha Fuels Environmental Test Facility (AFETF) impact gun is a unique tool for impact testing 238 PuO 2 -fueled heat sources of up to 178-mm dia at velocities to 300 m/s. An environmentally-sealed vacuum chamber at the muzzle of the gun allows preheating of the projectile to 1,000 0 C. Immediately prior to impact, the heat source projectile is completely sealed in a vacuum-tight catching container to prevent escape of its radioactive contents should rupture occur. The impact velocity delivered by this gas-powered gun can be regulated to within +-2%

  1. Evaluation of Particle Counter Technology for Detection of Fuel Contamination Detection Utilizing Fuel System Supply Point

    Science.gov (United States)

    2014-06-19

    product used as a diesel product for ground use (1). Free water contamination (droplets) may appear as fine droplets or slugs of water in the fuel...methods and test procedures for the calibration and use of automatic particle counters. The transition of this technology to the fuel industry is...UNCLASSIFIED 6 UNCLASSIFIED Receipt Vehicle Fuel Tank Fuel Injector Aviation Fuel DEF (AUST) 5695B 18/16/13 Parker 18

  2. Behavior of solid fission products in irradiated fuel

    International Nuclear Information System (INIS)

    Song, Ung Sup; Jung, Yang Hong; Kim, Hee Moon; Yoo, Byun Gok; Kim, Do Sik; Choo, Yong Sun; Hong, Kwon Pyo

    2004-01-01

    Many fission products are generated by fission events in UO 2 fuel under irradiation in nuclear reactor. Concentration of each fission product is changed by conditions of neutron energy spectrum, fissile material, critical thermal power, irradiation period and cooling time. Volatile materials such as Cs and I, the fission products, degrade nuclear fuel rod by the decrease of thermal conductivity in pellet and the stress corrosion cracking in cladding. Metal fission products (white inclusion) make pellet be swelled and decrease volume of pellet by densification. It seems that metal fission products are filled in the pore in pellet and placed between UO 2 lattices as interstitial. In addition, metal oxide state may change structural lattice volume. Considering behavior of fission products mentioned above, concentration of them is important. Fission products could be classified as bellows; solid solution in matrix : Sr, Zr, Nb, Y, La, Ce, Pr, Nd, Pm, Sm - metal precipitates : Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sb, Te - oxide precipitates : Ba, Zr, Nb, Mo, (Rb, Cs, Te) - volatile and gases : Kr, Xe, Br, I, (Rb, Cs, Te)

  3. Research on in-pile release of fission products from coated particle fuels

    International Nuclear Information System (INIS)

    Fukuda, K.; Iwamoto, K.

    1985-01-01

    Coated particle fuels fabricated in accordance with VHTR (Very High Temperature gas-cooled Reactor) fuel design have been irradiated by both capsules and an in-pile gas loop (OGL-1), and data on the fission products release under irradiation were obtained for loose coated particles, fuel compacts and fuel rods in the temperature range between 800 deg. C and 1600 deg. C. For the fission gases, temperature- and time dependences of the fractional release(R/B) were measured. Relation between release and failure fraction of the coated particles was elucidated on the VHTR reference fuels. Also measured was tritium concentration in the helium coolant of OGL-1. In-pile release behavior of the metallic fission products was studied by measuring the activities of the fission products adsorbed in the graphite sleeves of the OGL-1 fuel rods and the graphite fuel container of the sweep gas capsules in the PIE. Investigation on palladium interaction with SiC coating layer was included. (author)

  4. Spent nuclear fuel integrity during dry storage - performance tests and demonstrations

    International Nuclear Information System (INIS)

    McKinnon, M.A.; Doherty, A.L.

    1997-06-01

    This report summarizes the results of fuel integrity surveillance determined from gas sampling during and after performance tests and demonstrations conducted from 1983 through 1996 by or in cooperation with the US DOE Office of Commercial Radioactive Waste Management (OCRWM). The cask performance tests were conducted at Idaho National Engineering Laboratory (INEL) between 1984 and 1991 and included visual observation and ultrasonic examination of the condition of the cladding, fuel rods, and fuel assembly hardware before dry storage and consolidation of fuel, and a qualitative determination of the effects of dry storage and fuel consolidation on fission gas release from the spent fuel rods. The performance tests consisted of 6 to 14 runs involving one or two loading, usually three backfill environments (helium, nitrogen, and vacuum backfills), and one or two storage system orientations. The nitrogen and helium backfills were sampled and analyzed to detect leaking spent fuel rods. At the end of each performance test, periodic gas sampling was conducted on each cask. A spent fuel behavior project (i.e., enhanced surveillance, monitoring, and gas sampling activities) was initiated by DOE in 1994 for intact fuel in a CASTOR V/21 cask and for consolidated fuel in a VSC-17 cask. The results of the gas sampling activities are included in this report. Information on spent fuel integrity is of interest in evaluating the impact of long-term dry storage on the behavior of spent fuel rods. Spent fuel used during cask performance tests at INEL offers significant opportunities for confirmation of the benign nature of long-term dry storage. Supporting cask demonstration included licensing and operation of an independent spent fuel storage installation (ISFSI) at the Virginia Power (VP) Surry reactor site. A CASTOR V/21, an MC-10, and a Nuclear Assurance NAC-I28 have been loaded and placed at the VP ISFSI as part of the demonstration program. 13 refs., 14 figs., 9 tabs

  5. Production of leu high density fuels at Babcock and Wilcox

    International Nuclear Information System (INIS)

    Freim, J.B.

    1983-01-01

    A large number of fuel elements of all types are produced for both international and domestic customers by Nuclear Fuel Division of Babcock and Wilcox. A brief history of the division, included previous and present research reactor fuel element fabrication experience is discussed. The manufacturing facilities are briefly described. The fabrication of LEU fuels and economic analysis of the production are included. (A.J.)

  6. Bioenergy Research Programme, Yearbook 1995. Production of wood fuels; Bioenergian tutkimusohjelma, vuosikirja 1995. Puupolttoaineen tuotantotekniikka

    Energy Technology Data Exchange (ETDEWEB)

    Alakangas, E [ed.

    1997-12-31

    Bioenergy Research Programme is one of the energy technology research programmes of the Technology Development Center TEKES. The aim of the Bioenergy Research Programme is to increase, by using technical research and development, the economically profitable and environmentally sound utilisation of bioenergy, to improve the competitiveness of present peat and wood fuels, and to develop new competitive fuels and equipment related to bioenergy. The funding for 1995 was nearly 52 million FIM and the number of projects 66. The main goal of the wood fuels research area is to develop new production methods in order to decrease the production costs to the level of imported fuels. The total potential of the wood fuel use should be at least 1.0 million toe/a (5.5 million m{sup 3}). During the year 1995 There were over 30 projects concerning the production of wood derived fuels going on. Nearly half of them focused on integrated production of pulp wood and wood fuel. About ten projects was carried out to promote the wood fuel production from logging residues. Other topics were firewood production, production logistics and wood fuel resources. For production of fuel chips from logging residues, a new chipper truck, MOHA-SISU, was introduced. The new machine gives a new logistic solution resulting in high productivity and reasonable operating costs. In Mikkeli region three years of active work promoted the usage of wood fuel in a district power plant to the level of over 110 000 m{sup 3} of fuel chips. The production costs tend to be a little high in average, and the production chain still needs to be improved

  7. Bioenergy Research Programme, Yearbook 1995. Production of wood fuels; Bioenergian tutkimusohjelma, vuosikirja 1995. Puupolttoaineen tuotantotekniikka

    Energy Technology Data Exchange (ETDEWEB)

    Alakangas, E. [ed.

    1996-12-31

    Bioenergy Research Programme is one of the energy technology research programmes of the Technology Development Center TEKES. The aim of the Bioenergy Research Programme is to increase, by using technical research and development, the economically profitable and environmentally sound utilisation of bioenergy, to improve the competitiveness of present peat and wood fuels, and to develop new competitive fuels and equipment related to bioenergy. The funding for 1995 was nearly 52 million FIM and the number of projects 66. The main goal of the wood fuels research area is to develop new production methods in order to decrease the production costs to the level of imported fuels. The total potential of the wood fuel use should be at least 1.0 million toe/a (5.5 million m{sup 3}). During the year 1995 There were over 30 projects concerning the production of wood derived fuels going on. Nearly half of them focused on integrated production of pulp wood and wood fuel. About ten projects was carried out to promote the wood fuel production from logging residues. Other topics were firewood production, production logistics and wood fuel resources. For production of fuel chips from logging residues, a new chipper truck, MOHA-SISU, was introduced. The new machine gives a new logistic solution resulting in high productivity and reasonable operating costs. In Mikkeli region three years of active work promoted the usage of wood fuel in a district power plant to the level of over 110 000 m{sup 3} of fuel chips. The production costs tend to be a little high in average, and the production chain still needs to be improved

  8. Evaluation of fuel release rate and mechanism tests under RBCB conditions

    International Nuclear Information System (INIS)

    Adamson, M.G.

    1981-09-01

    This task includes theoretical evaluation of fuel/fission product release behavior from failed LMFBR fuel elements as well as an on-going experimental investigation of the mechanism of oxide fuel dispersal into flowing liquid sodium. The primary objectives of this work are to develop a fuel source term that can be used in predictive models for primary heat transfer system contamination and to understand the separate influences of important system variables (such as flow rate, oxygen impurity level) on this source term. The present report is written in two parts: the first, in condensed form, is an updated evaluation of fuel (U,Pu) and fission product release data, and the second describes the current status of supporting experimental work at General Electric's Vallecitos Laboratory

  9. Compact hydrogen production systems for solid polymer fuel cells

    Science.gov (United States)

    Ledjeff-Hey, K.; Formanski, V.; Kalk, Th.; Roes, J.

    Generally there are several ways to produce hydrogen gas from carbonaceous fuels like natural gas, oil or alcohols. Most of these processes are designed for large-scale industrial production and are not suitable for a compact hydrogen production system (CHYPS) in the power range of 1 kW. In order to supply solid polymer fuel cells (SPFC) with hydrogen, a compact fuel processor is required for mobile applications. The produced hydrogen-rich gas has to have a low level of harmful impurities; in particular the carbon monoxide content has to be lower than 20 ppmv. Integrating the reaction step, the gas purification and the heat supply leads to small-scale hydrogen production systems. The steam reforming of methanol is feasible at copper catalysts in a low temperature range of 200-350°C. The combination of a small-scale methanol reformer and a metal membrane as purification step forms a compact system producing high-purity hydrogen. The generation of a SPFC hydrogen fuel gas can also be performed by thermal or catalytic cracking of liquid hydrocarbons such as propane. At a temperature of 900°C the decomposition of propane into carbon and hydrogen takes place. A fuel processor based on this simple concept produces a gas stream with a hydrogen content of more than 90 vol.% and without CO and CO2.

  10. 40 CFR 86.335-79 - Gasoline-fueled engine test cycle.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 18 2010-07-01 2010-07-01 false Gasoline-fueled engine test cycle. 86....335-79 Gasoline-fueled engine test cycle. (a) The following test sequence shall be followed in... operating the engine at the higher approved load setting during cycle 1 and at the lower approved load...

  11. Performance test of the I and C system (GSF - 2002) for the irradiation tests using a fuel capsule

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Hwan; Park, S. J.; Kim, B. G.; Ahn, D. H

    2004-12-01

    HANARO is a very important facility in Korea. It offers various types of irradiation tests of nuclear fuels and materials. With the various applications of the HANARO capsule for the academic and industrial applications, new technologies and relevant facilities will become more important especially for the advanced nuclear fuels and materials development. A new I and C system for an irradiation test using an instrumented fuel capsule have been designed and manufactured to provide more qualified data to fuel developer. The performance test which started in 2004, was done to investigate the thermal response of the capsule connected to the gas mixing system of the new I and C system(GSF-2002) in the cold test loop under the HANARO hydraulic operational condition. Main test parameters are mass flow rate of 25, 50 and 100 cc/min of He/Ne gas, gas pressure of 1 to 3 kg/cm{sup 2}, heater power of 1 to 3.4kW and different gas mixing ratios of He to Ne. From the out-pile tests, it was confirmed that the I and C system(GSF-2002) would be feasible for the fuel irradiation tests. Both analytical and test data prepared by this study are directly used for the fuel experiments related to advanced fuel development program.

  12. Simulation of integral local tests with high-burnup fuel

    International Nuclear Information System (INIS)

    Gyori, G.

    2011-01-01

    The behaviour of nuclear fuel under LOCA conditions may strongly depend on the burnup-dependent fuel characteristics, as it has been indicated by recent integral experiments. Fuel fragmentation and the associated fission gas release can influence the integral fuel behaviour, the rod rupture and the radiological release. The TRANSURANUS fuel performance code is a proper tool for the consistent simulation of burnup-dependent phenomena during normal operation and the thermo-mechanical behaviour of the fuel rod in a subsequent accident. The code has been extended with an empirical model for micro-cracking induced FGR and fuel fragmentation and verified against integral LOCA tests of international projects. (author)

  13. Corrosion Tests of LWR Fuels - Nuclide Release

    International Nuclear Information System (INIS)

    P.A. Finn; Y. Tsai; J.C. Cunnane

    2001-01-01

    Two BWR fuels [64 and 71 (MWd)/kgU], one of which contained 2% Gd, and two PWR fuels [30 and 45 (MWd)/kgU], are tested by dripping groundwater on the fuels under oxidizing and hydrologically unsaturated conditions for times ranging from 2.4 to 8.2 yr at 90 C. The 99 Tc, 129 I, 137 Cs, 97 Mo, and 90 Sr releases are presented to show the effects of long reaction times and of gadolinium on nuclide release. This investigation showed that the five nuclides at long reaction times have similar fractional release rates and that the presence of 2% Gd reduced the 99 Tc cumulative release fraction by about an order of magnitude over that of a fuel with a similar burnup

  14. Fuel powder production from ductile uranium alloys

    International Nuclear Information System (INIS)

    Clark, C.R.; Meyer, M.K.

    1998-01-01

    Metallic uranium alloys are candidate materials for use as the fuel phase in very-high-density LEU dispersion fuels. These ductile alloys cannot be converted to powder form by the processes routinely used for oxides or intermetallics. Three methods of powder production from uranium alloys have been investigated within the US-RERTR program. These processes are grinding, cryogenic milling, and hydride-dehydride. In addition, a gas atomization process was investigated using gold as a surrogate for uranium. (author)

  15. Standard test method for analysis of isotopic composition of uranium in nuclear-grade fuel material by quadrupole inductively coupled plasma-mass spectrometry

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2000-01-01

    1.1 This test method is applicable to the determination of the isotopic composition of uranium (U) in nuclear-grade fuel material. The following isotopic weight percentages are determined using a quadrupole inductively coupled plasma-mass spectrometer (Q-ICP-MS): 233U, 234U, 235U, 236U, and 238U. The analysis can be performed on various material matrices after acid dissolution and sample dilution into water or dilute nitric (HNO3) acid. These materials include: fuel product, uranium oxide, uranium oxide alloys, uranyl nitrate (UNH) crystals, and solutions. The sample preparation discussed in this test method focuses on fuel product material but may be used for uranium oxide or a uranium oxide alloy. Other preparation techniques may be used and some references are given. Purification of the uranium by anion-exchange extraction is not required for this test method, as it is required by other test methods such as radiochemistry and thermal ionization mass spectroscopy (TIMS). This test method is also described i...

  16. Methodology of thermalhydraulic tests of fuel assemblies for WWER-1000

    International Nuclear Information System (INIS)

    Archipov, A.; Kolochko, V.N.

    2001-01-01

    At present 11 units with WWER-1000 are in operation in Ukraine. The NPPs are provided with nuclear fuel from Russia. The fuel assemblies are fabricated and delivered to Ukrainian NPPs from Russia. However the contemporary tendencies of nuclear energy development in the world assume a diversification of nuclear fuel vendors. Therefore the creation of the own nuclear fuel cycle of Ukraine is in mind in the strategy of nuclear energy development of Ukraine. As a part of the fuel assemblies fabrication process complex of the thermalhydraulic tests should be carried out to confirm design characteristics of the fuel assemblies before they are loaded in the reactor facility. The experimental basis and scientific infrastructure for the thermalhydraulic tests arrangement and realization of the programs and procedures for the core equipment examination are under consideration. (author)

  17. Fuel performance and fission product behaviour in gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport. Refs, figs, tabs.

  18. Fission product release as a function of chemistry and fuel morphology

    International Nuclear Information System (INIS)

    Hobbins, R.R.; Osetek, D.J.; Petti, D.A.; Hagrman, D.L.

    1989-01-01

    Analysis of the consequences of severe reactor accidents requires knowledge of the location and chemical form of fission products throughout the accident sequence. Two factors that strongly influence the location and chemical form of fission products are the chemistry within the core and the morphology of the fuel or fuel-bearing debris. This paper reviews the current understanding of the these factors garnered from integral and separate effect experiments and the TMI-2 accident, and provides perspective on the significance of contributing phenomena for the analysis of severe accidents, particularly during the in-vessel phase. Information has been obtained recently on phenomena affecting the release of fission products from fuel and the reactor vessel during the in-vessel melt progression phase of a severe accident. The influence of a number of these phenomena will be reviewed, including fuel chemistry, H 2 /H 2 O ratio, fuel liquefaction, molten pools, and debris beds. 13 refs., 1 fig., 1 tab

  19. Experimental data report for test TS-3 Reactivity Initiated Accident test in the NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo; Sobajima, Makoto.

    1993-09-01

    This report presents experimental data for Test TS-3 which was the third test in a series of Reactivity Initiated Accident (RIA) tests using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in September, 1990. Test fuel rod used in the Test TS-3 was a short-sized BWR (7 x 7) type rod which was re-fabricated from a commercial rod irradiated in the Tsuruga Unit 1 power reactor of Japan Atomic Power Co. The fuel had an initial enrichment of 2.79 % and a burnup of 26 Gwd/tU. A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 94 ± 4 cal/g · fuel (88 ± 4 cal/g · fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and results of pre-pulse and post-pulse irradiation examinations are described in this report. (author)

  20. Quality control in the fuel elements production process

    International Nuclear Information System (INIS)

    Katanic-Popovic, J.; Spasic, Z.; Djuricis, Lj.

    1977-01-01

    Recently great attention has been paid at the international level to the analysis of production processes and quality control of fuel and fuel elements with the aim to speed up activity of proposing and accepting standards and measurement methods. IAEA also devoted great interest to these problems appealing to more active participation of all users and producers fuel elements in a general effort to secure successful work of nuclear plants. For adequate and timely participation in future in the establishment and analysis of general requirements and documentation for the control of purchased or self produced fuel elements in out country it is necessary to be well informed and to follow this activity at the international level. (author)

  1. The US Army Foreign Comparative Test fuel cell program

    Science.gov (United States)

    Bostic, Elizabeth; Sifer, Nicholas; Bolton, Christopher; Ritter, Uli; Dubois, Terry

    The US Army RDECOM initiated a Foreign Comparative Test (FCT) Program to acquire lightweight, high-energy dense fuel cell systems from across the globe for evaluation as portable power sources in military applications. Five foreign companies, including NovArs, Smart Fuel Cell, Intelligent Energy, Ballard Power Systems, and Hydrogenics, Inc., were awarded competitive contracts under the RDECOM effort. This paper will report on the status of the program as well as the experimental results obtained from one of the units. The US Army has interests in evaluating and deploying a variety of fuel cell systems, where these systems show added value when compared to current power sources in use. For low-power applications, fuel cells utilizing high-energy dense fuels offer significant weight savings over current battery technologies. This helps reduce the load a solider must carry for longer missions. For high-power applications, the low operating signatures (acoustic and thermal) of fuel cell systems make them ideal power generators in stealth operations. Recent testing has been completed on the Smart Fuel Cell A25 system that was procured through the FCT program. The "A-25" is a direct methanol fuel cell hybrid and was evaluated as a potential candidate for soldier and sensor power applications.

  2. Advanced energy production based on fossil and renewable fuels; Energiantuotannon tehostaminen fossiilisiin ja uusiutuviin polttoaineisiin perustuvassa energiantuotannossa

    Energy Technology Data Exchange (ETDEWEB)

    Hepola, J.; Kurkela, E. [VTT Processes, Espoo (Finland)

    2002-07-01

    The objects of this survey were gasification and combustion techniques, coproduction alternatives of energy, fuels and chemicals, as well as high-temperature fuel cells and hybrid fuel cell systems. The potential and market outlook of pressurised fluidised-bed combustion have been declined, i.a., due to the good competitivity of natural gas combined-cycle technology, tightening emission standards, and the development outlook for gasification combined-cycle technology. Development of a combined-cycle power plant process based on pressurised pulverised combustion of coal is still at an initial stage. The oxygen based IGCC plants in the world have so far been demonstration plants. The IGCC technology is expected to commercialise first in residual oil gasification applications integrated to oil refineries and then in coal powered condensed power plants. In addition, one biomass-based IGCC plant has been constructed. This process is so called simplified IGCC, utilising pressurised air blown gasification and hot gas cleaning. The test trials of this plant were completed in 1999. The process is technically feasible also for large-scale demonstration. Gasification technology has also been developed for black liquor. The ChemRec black liquor gasification process is technically the most advanced process at the moment. The atmospheric process has been demonstrated and the pressurised process demonstration is about to start in Sweden and in USA. In utilising biomass fuels or black liquor, the IGCC process offers the possibility to significantly increase the ratio of electrical power to thermal power with combined cycle. Synthesis gas produced by gasification technology can be used for producing different gaseous or liquid fuels and chemicals and for energy production. In a flexible use of feedstocks and products this coproduction method offers a significantly more feasible alternative to present energy production plants. There are several projects underway in the United States

  3. Conceptual development of a test facility for spent fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs.

  4. Conceptual development of a test facility for spent fuel management

    International Nuclear Information System (INIS)

    Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G.

    1997-01-01

    Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs

  5. Irradiation testing of high-density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-01-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 'microplates'. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U10Mo-0.05Sn, U2Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of approximately 40 and 80 at.% U 235 . Of particular interest are the extent of reaction of the fuel and matrix phases and the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions. (author)

  6. Prototype spent-fuel canister design, analysis, and test

    International Nuclear Information System (INIS)

    Leisher, W.B.; Eakes, R.G.; Duffey, T.A.

    1982-03-01

    Sandia National Laboratories was asked by the US Energy Research and Development Administration (now US Department of Energy) to design the spent fuel shipping cask system for the Clinch River Breeder Reactor Plant (CRBRP). As a part of this task, a canister which holds liquid sodium and the spent fuel assembly was designed, analyzed, and tested. The canister body survived the regulatory Type-B 9.1-m (30-ft) drop test with no apparent leakage. However, the commercially available metal seal used in this design leaked after the tests. This report describes the design approach, analysis, and prototype canister testing. Recommended work for completing the design, when funding is available, is included

  7. Fabrication and testing of uranium nitride fuel for space power reactors

    Science.gov (United States)

    Matthews, R. B.; Chidester, K. M.; Hoth, C. W.; Mason, R. E.; Petty, R. L.

    1988-02-01

    Uranium nitride fuel was selected for previous space power reactors because of its attractive thermal and physical properties; however, all UN fabrication and testing activities were terminated over ten years ago. An accelerated irradiation test, SP-1, was designed to demonstrate the irradiation performance of Nb-1 Zr clad UN fuel pins for the SP-100 program. A carbothermic-reduction/nitriding process was developed to synthesize UN powders. These powders were fabricated into fuel pellets by conventional cold-pressing and sintering. The pellets were loaded into Nb-1 Zr cladding tubes, irradiated in a fast-test reactor, and destructively examined after 0.8 at% burnup. Preliminary postirradiation examination (PIE) results show that the fuel pins behaved as designed. Fuel swelling, fission-gas release, and microstructural data are presented, and suggestions to enhance the reliability of UN fuel pins are discussed.

  8. Fuel Subsystems Flight Test Handbook

    Science.gov (United States)

    1981-12-01

    detailed, accessible .-ltand complete test records for his own protection and for the benefit of his successor in case of promotion, transfer or...and pilot display of fuel quantity, low level warning and a " Bingo " fudl warning. 3.0 TEST OBJECT1VES: ACTION Orriclc. On POUTION•’PHONWE [ DATE PCR...TIS No. 46, paragraph 3.11 3.6 To demonstrate that the low level and bingo warning system are consistent L •and meet the requirements of paragraph

  9. Fission-product retention in HTGR fuels

    International Nuclear Information System (INIS)

    Homan, F.J.; Kania, M.J.; Tiegs, T.N.

    1982-01-01

    Retention data for gaseous and metallic fission products are presented for both Triso-coated and Biso-coated HTGR fuel particles. Performance trends are established that relate fission product retention to operating parameters, such as temperature, burnup, and neutron exposure. It is concluded that Biso-coated particles are not adequately retentive of fission gas or metallic cesium, and Triso-coated particles which retain cesium still lose silver. Design implications related to these performance trends are identified and discussed

  10. Highlights from the IAEA coordinated research programme on fuel performance and fission product data

    International Nuclear Information System (INIS)

    Nabielek, H.; Schenk, W.; Verfondern, K.

    1996-01-01

    Seven countries are cooperating with the objectives (i) to document the status of the experimental data base and of the predictive methods for Gas-Cooled Reactor fuel performance and fission product behaviour; (ii) to verify and validate methods in fuel performance and fission product retention prediction. These countries are China, France, Germany, Japan, Russia, USA and the UK. Duration of the programme is 1993-96. The technology areas addressed in this IAEA Coordinated Research Programme are: Fuel design and manufacture, Normal operation fuel performance and fission product behaviour, Accident condition fuel performance and fission product behaviour, -core heatup, -fast transients, -oxidising conditions (water and air ingress), Plateout, re-entrainment of plateout, fission product behaviour in the reactor building, and Performance of advanced fuels. Work performed so far has generated a 300-page draft document with important information for normal operations (Germany, Japan, China, Russia) and accident conditions (USA, Japan, Germany, Russia) and, additionally, a special chapter on advanced fuels (Japan). (author)

  11. Hydrogen production with fully integrated fuel cycle gas and vapour core reactors

    International Nuclear Information System (INIS)

    Anghaie, S.; Smith, B.

    2004-01-01

    This paper presents results of a conceptual design study involving gas and vapour core reactors (G/VCR) with a combined scheme to generate hydrogen and power. The hydrogen production schemes include high temperature electrolysis as well as two dominant thermochemical hydrogen production processes. Thermochemical hydrogen production processes considered in this study included the calcium-bromine process and the sulphur-iodine processes. G/VCR systems are externally reflected and moderated nuclear energy systems fuelled by stable uranium compounds in gaseous or vapour phase that are usually operated at temperatures above 1500 K. A gas core reactor with a condensable fuel such as uranium tetrafluoride (UF 4 ) or a mixture of UF 4 and other metallic fluorides (BeF 2 , LiF, KF, etc.) is commonly known as a vapour core reactor (VCR). The single most relevant and unique feature of gas/vapour core reactors is that the functions of fuel and coolant are combined into one. The reactor outlet temperature is not constrained by solid fuel-cladding temperature limits. The maximum fuel/working fluid temperature in G/VCR is only constrained by the reactor vessel material limits, which is far less restrictive than the fuel clad. Therefore, G/VCRs can potentially provide the highest reactor and cycle temperature among all existing or proposed fission reactor designs. Gas and vapour fuel reactors feature very low fuel inventory and fully integrated fuel cycle that provide for exceptional sustainability and safety characteristics. With respect to fuel utilisation, there is no fuel burn-up limit for gas core reactors due to continuous recycling of the fuel. Owing to the flexibility in nuclear design characteristics of cavity reactors, a wide range of conversion ratio from completely burner to breeder is achievable. The continuous recycling of fuel in G/VCR systems allow for complete burning of actinides without removing and reprocessing of the fuel. The only waste products at the back

  12. Reactor transients tests for SNR fuel elements in HFR reactor

    International Nuclear Information System (INIS)

    Plitz, H.

    1989-01-01

    In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

  13. Contribution of metallic fission product inclusions to axial fuel motion potential

    International Nuclear Information System (INIS)

    Sasa, P.; Cronenberg, A.; Stevenson, M.

    1979-01-01

    In the analysis of postulated nuclear reactor accidents, axial fuel motion within the fuel pin prior to cladding failure can have an important mitigating effect. The question of primary importance is whether or not metallic inclusions have the potential to vaporize during an overheating event and thus contribute to fuel motion. To assess this potential, two limiting calculations were made: 1) The inclusion constituent assumed insoluble in one another and 2) The constituents assumed totally miscible in one another. Thermodynamic considerations indicate that the metallic fission products found within inclusions of fuel rods irradiated in a fast neutron spectrum, would form homogeneous solutions. Therefore, it is concluded that the metallic fission products would not enhance fuel swelling during an overheating event. 16 refs

  14. Fission Product Release from Spent Nuclear Fuel During Melting

    International Nuclear Information System (INIS)

    Howell, J.P.; Zino, J.F.

    1998-09-01

    The Melt-Dilute process consolidates aluminum-clad spent nuclear fuel by melting the fuel assemblies and diluting the 235U content with depleted uranium to lower the enrichment. During the process, radioactive fission products whose boiling points are near the proposed 850 degrees C melting temperature can be released. This paper presents a review of fission product release data from uranium-aluminum alloy fuel developed from Severe Accident studies. In addition, scoping calculations using the ORIGEN-S computer code were made to estimate the radioactive inventories in typical research reactor fuel as a function of burnup, initial enrichment, and reactor operating history and shutdown time.Ten elements were identified from the inventory with boiling points below or near the 850 degrees C reference melting temperature. The isotopes 137Cs and 85Kr were considered most important. This review serves as basic data to the design and development of a furnace off-gas system for containment of the volatile species

  15. Sustainability assessment of bioethanol and petroleum fuel production in Japan based on emergy analysis

    International Nuclear Information System (INIS)

    Liu, Jin’e; Lin, Bin-Le; Sagisaka, Masayuki

    2012-01-01

    To promote the reduction of greenhouse gas emissions, research and development of bioethanol technologies are encouraged in Japan and a plan to utilize untilled fields to develop rice for bioethanol production as a substitute for petroleum fuel is being devised. This study applies emergy methods to compare the sustainability of petroleum fuel production and two Japanese rice-to-ethanol production scenarios: (a) ethanol from rice grain, while straw and chaff are burned as energy and (b) ethanol from rice+straw+chaff. The major emergy indices, Emergy Yield Ratio (EYR), Environmental Loading Ratio (ELR), Emergy Investment Ratio (EIR), Emergy Sustainability Index (ESI), Environmental Impacts Ratio (EVR) and system transformity (Tr), are analyzed to assess the production processes. The results show that (1) petroleum fuel production presents higher ELR, EIR, EVR and lower EYR, ESI, Tr than rice-to-ethanol production, indicating rice-to-ethanol production makes sense for reduction of greenhouse gases (GHG); (2) scenario (a) performs similarly on major indicators (EYR, ESI, ELR, EIR and EVR) to scenario (b), yet the system efficiency indicator (Tr) of scenario (a, 7.572×10 5 semj/J) is much higher than (b, 4.573×10 5 semj/J), and therefore (b) is a better alternative for policy decisions; (3) both petroleum fuel production and rice-to-ethanol processes are mainly driven by purchased resources and are unsustainable and nonrenewable in the long run. - Highlights: ► We compare petrol fuel and rice-to-ethanol production using emergy indices. ► Rice-to-ethanol reduces green house gas emissions as a substitute for petrol fuel. ► Rice-to-ethanol production has better sustainability than that of petrol fuel. ► Neither petrol fuel nor biofuel production are sustainable in the long term. ► Bioethanol is not a renewable fuel.

  16. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

    2004-07-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

  17. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    International Nuclear Information System (INIS)

    Purcell, P.C.; Dallongeville, M.

    2004-01-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme

  18. Microprobe