WorldWideScience

Sample records for unirradiated fuels required

  1. Transport insurance of unirradiated nuclear fuels

    International Nuclear Information System (INIS)

    Matto, H.

    1985-01-01

    Special conditions must be taken into account in transport insurance for nuclear materials even if the nuclear risk involved is negligible, as in shipments of unirradiated nuclear fuels. The shipwreck of the 'Mont Louis' has raised a number of open points which must be solved pragmatically within the framework of transport insurance. Some proposals are outlined in the article. (orig.) [de

  2. Truck accident involving unirradiated nuclear fuel

    International Nuclear Information System (INIS)

    Carlson, R.W.; Fischer, L.E.

    1993-01-01

    In the early morning of Dec. 16, 1991, a severe accident occurred when a passenger vehicle traveling in the wrong direction collided with a tractor trailer carrying 24 unirradiated nuclear fuel assemblies in 12 containers on Interstate I-91 in Springfield, Massachusetts. This paper documents the mechanical circumstances of the accident and assesses the physical environment to which the containers were exposed and the response of the containers and their contents. The accident involved four impacts where the truck was struck by the car, impacted on the center guardrail, impacted on the outer concrete barrier and came to rest against the center guardrail. The impacts were followed by a fire that began in the engine compartment, spread to the tractor and cab, and eventually spread to the trailer and payload. The fire lasted for about three hours and the packages were involved in the fire for about two hours. As a result of the fire, the tractor-trailer was completely destroyed and the packages were exposed to flames with temperatures between 1,300 F and 1,800 F. The fuel assemblies remained intact during the accident and there was no release of any radioactive material during the accident. This was a very severe accident; however, the injuries were minor and at no time was the public health and safety at risk

  3. Truck accident involving unirradiated nuclear fuel

    International Nuclear Information System (INIS)

    Carlson, R.W.; Fischer, L.E.

    1992-07-01

    In the early morning of Dec. 16, 1991, a severe accident occurred when a passenger vehicle traveling in the wrong direction collided with a tractor trailer carrying 24 nuclear fuel assemblies in 12 containers on Interstate 1-91 in Springfield, Massachusetts. This paper documents the mechanical circumstances of the accident and the physical environment to which the containers were exposed and the response of the containers and their contents. The accident involved four impacts where the truck was struck by the car, impacted on the center guardrail, impacted on the outer concrete barrier and came to rest against the center guardrail. The impacts were followed by a fire that began in the engine compartment, spread to the.tractor and cab, and eventually spread to the trailer and payload. The fire lasted for about three hours and the packages were involved in the fire for about two hours. As a result of the fire, the tractor-trailer was completely destroyed and the packages were exposed to flames with temperatures between 1300 degrees F and 1800 degrees F. The fuel assemblies remained intact during the accident and there was no release of any radioactive material during the accident. This was a very severe accident; however, the injuries were minor and at no time was the public health and safety at risk

  4. Plant-scale anodic dissolution of unirradiated N-Reactor fuel

    International Nuclear Information System (INIS)

    Gay, E.C.; Miller, W.E.; Laidler, J.J.

    1995-01-01

    Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the fuel segment length, diameter, and shape required for high throughput electrorefiner treatment for ultimate disposal in a geologic repository. Based on these tests, a conceptual design was produced of an electrorefiner for a full-scale plant to process N-Reactor spent fuel. In this design, the diameter of an electrode assembly is about 0.6 m (25 in.). Eight of these assemblies in an electrorefiner would accommodate a 1.333-metric-ton batch of N-Reactor fuel. Electrorefining would proceed at a rate of 40 kg uranium per hour

  5. Reaction of unirradiated high-density fuel with aluminum

    International Nuclear Information System (INIS)

    Wiencek, T.C.; Meyer, M.K.; Prokofiev, I.G.; Keiser, D.D.

    1997-01-01

    Excellent dispersion fuel performance requires that fuel particles remain stable and do not react significantly with the surrounding aluminum matrix. A series of high-density fuels, which contain uranium densities >12 g/cm 3 , have been fabricated into plates. As part of standard processing, all of these fuels were subjected to a blister anneal of 1 h at 485 deg. C. Changes in plate thickness were measured and evaluated. From these results, suppositions about the probable irradiation properties of these fuels have been proposed. In addition, two fuels, U-10 wt% Mo and U 2 Mo, were subjected to various heat treatments and were found to be very stable in an aluminum matrix. On the basis of the experimental data, hypotheses of the irradiation behavior of these fuels are presented. (author)

  6. Plant-scale anodic dissolution of unirradiated N-Reactor fuel

    International Nuclear Information System (INIS)

    Gay, E.C.; Miller, W.E.; Laidler, J.J.

    1995-01-01

    Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the fuel segment length, diameter, and shape required for high throughput electro-refiner treatment for ultimate disposal in a geologic repository. Based on these tests, a conceptual design was produced of an electro-refiner for a full-scale plant to process N-Reactor spent fuel. In this design, the diameter of an electrode assembly is about 0.6 m (25 in.). Eight of these assemblies in an electro-refiner would accommodate a 1.333-metric-ton batch of N-Reactor fuel. Electrorefining would proceed at a rate of 40 kg uranium per hour. (author)

  7. Safety analysis report for packaging: the ORNL HFIR unirradiated fuel element shipping container

    International Nuclear Information System (INIS)

    Evans, J.H.; Boulet, J.A.M.; Eversole, R.E.

    1977-11-01

    The ORNL HFIR unirradiated fuel element shipping container was designed and fabricated at the Oak Ridge National Laboratory for the transport of HFIR unirradiated fuel elements. The container was evaluated analytically and experimentally to determine its compliance with the applicable regulations governing containers in which radioactive and fissile materials are transported, and the evaluation is the subject of this report. Computational and test procedures were used to determine the structural integrity and thermal behavior of the cask relative to the general standards for normal conditions of transport and the standards for the hypothetical accident conditions. The results of the evaluation demonstrate that the container is in compliance with the applicable regulations

  8. Damage and failure of unirradiated and irradiated fuel rods tested under film boiling conditions

    International Nuclear Information System (INIS)

    Mehner, A.S.; Hobbins, R.R.; Seiffert, S.L.; MacDonald, P.E.; McCardell, R.K.

    1979-01-01

    Power-cooling-mismatch experiments are being conducted as part of the Thermal Fuels Behavior Program in the Power Burst Facility at the Idaho National Engineering Laboratory to evaluate the behavior of unirradiated and previously irradiated light water reactor fuel rods tested under stable film boiling conditions. The observed damage that occurs to the fuel rod cladding and the fuel as a result of film boiling operation is reported. Analyses performed as a part of the study on the effects of operating failed fuel rods in film boiling, and rod failure mechanisms due to cladding embrittlement and cladding melting upon being contacted by molten fuel are summarized

  9. Plant-scale anodic dissolution of unirradiated IFR fuel pins

    International Nuclear Information System (INIS)

    Gay, E.C.; Tomczuk, Z.; Miller, W.E.

    1993-01-01

    This report discusses anodic dissolution which is a major operation in the pyrometallurgical process for recycling spent metal fuels from the Integral Fast Reactor (IFR), an advanced reactor design developed at Argonne National Laboratory. This process involves electrorefining the heavy metals (uranium and plutonium) from chopped, steel-clad fuel segments. The heavy metals are electrotransported from anodic dissolution baskets to solid and liquid cathodes in a molten salt electrolyte (LiCl-KCI) at 500 degrees C. Uranium is recovered on a solid cathode mandrel, while a uranium-plutonium mixture is recovered in a liquid cadmium cathode. The anode configuration consists of four baskets mounted on an anode shaft. These baskets provide parallel circuits in the electrolyte and salt flow through the chopped fuelbed as the baskets are rotated. The baskets for the engineering-scale tests were sized to contain up to 2.5 kg of heavy metal. Anodic dissolution of 10 kg batches of chopped, steel-clad simulated tuel (U-10% Zr and U-Zr-Fs alloy) was demonstrated

  10. A highway accident involving unirradiated nuclear fuel in Springfield, Massachusetts, on December 16, 1991

    International Nuclear Information System (INIS)

    Carlson, R.W.; Fischer, L.E.

    1992-06-01

    In the early morning of Dec. 16, 1991, a severe accident occurred when a passenger vehicle traveling in the wrong direction collided with a tractor trailer carrying 24 unirradiated nuclear fuel assemblies in 12 containers on Interstate I-91 in Springfield, Massachusetts. The purpose of this report is to document the mechanical circumstances of the severe accident, confirm the nature and quantity of the radioactive materials involved, and assess the physical environment to which the containers were exposed and the response of the containers and their contents. The report consists of five major sections. The first section describes the circumstances and conditions of the accident and the finding of facts. The second describes the containers, the unirradiated nuclear fuel assemblies, and the tie down arrangement used for the trailer. The third describes the damage sustained during the accident to the tractor, trailer, containers, and unirradiated nuclear fuel assemblies. The fourth evaluates the accident environment and its effects on the containers and their contents. The final section gives conclusions derived from the analysis and fact finding investigation. During this severe accident, only minor injuries occurred, and at no time was the public health and safety at risk

  11. Storage for the Fast Flux Test Facility unirradiated fuel in the Plutonium Finishing Plant Complex, Hanford Site, Richland, Washington

    International Nuclear Information System (INIS)

    1992-01-01

    This Environmental Assessment evaluates the proposed action to relocate and store unirradiated Fast Flux Test Facility fuel in the Plutonium Finishing Plant Complex on the Hanford Site, Richland, Washington. The US Department of Energy has decided to cease fuel fabrication activities in the 308 Building in the 300 Area. This decision was based on a safety concern over the ability of the fuel fabrication portion of the 308 Building to withstand a seismic event. The proposed action to relocate and store the fuel is based on the savings that could be realized by consolidating security costs associated with storage of the fuel. While the 308 Building belowgrade fuel storage areas are not at jeopardy by a seismic event, the US Department of Energy is proposing to cease storage operations along with the related fabrication operations. The US Department of Energy proposes to remove the unirradiated fuel pins and fuel assemblies from the 308 Building and store them in Room 192A, within the 234-5Z Building, a part of the Plutonium Finishing Plant Complex, located in the 200 West Area. Minor modifications to Room 192A would be required to accommodate placement of the fuel. The US Department of Energy estimates that removing all of the fuel from the 308 Building would save $6.5 million annually in security expenditures for the Fast Flux Test Facility. Environmental impacts of construction, relocation, and operation of the proposed action and alternatives were evaluated. This evaluation concluded that the proposed action would have no significant impacts on the human environment

  12. Safety Analysis Report for Packaging: The unirradiated fuel shipping container USA/9853/AF

    International Nuclear Information System (INIS)

    1991-01-01

    The HFBR Unirradiated Fuel Shipping Container was designed and fabricated at the Oak Ridge National Laboratory in 1978 for the transport of fuel for the High Flux Beam Reactor (HFBR) for Brookhaven National Laboratory. The package has been evaluated analytically, as well as the comparison to tests on similar packages, to demonstrate compliance with the applicable regulations governing packages in which radioactive and fissile materials are transported. The contents of this Safety Analysis Report for Packaging (SARP) are based on Regulatory Guide 7.9 (proposed Revision 2 - May 1986), 10 CFR Part 71, DOE Order 1540.2, DOE Order 5480.3, and 49 CFR Part 173

  13. A study on density, porosity and grain size of unirradiated ROX fuels and simulated ROX fuels

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Shirasu, Noriko; Ohmichi, Toshihiko

    1999-05-01

    The unirradiated ROX fuels made of (1) 23wt%PuO 2 -17wt%SZR-56wt%Al 2 O 3 -4wt%MgO (ROX-SZR) and (2) 20wt%PuO 2 -29wt%ThO 2 -48wt%Al 2 O 3 -3wt%MgO (ROX-ThO 2 ) were fabricated by JAERI. Additionally, 9 simulated ROX fuels (used UO 2 instead of PuO 2 ) were fabricated by NFI. Densities, porosity and grain sizes of those were studied. Obtained results are: (1) ROX fuels: The estimated theoretical density (TD) was 5.6g/cc for the ROX-SZR and 6.2g/cc for the ROX-ThO 2 . Those were about half of that in UO 2 (10.96g/cc). As-fabricated density determined was 4.6g/cc (82%TD) for the former and 5.2g/cc (83%TD) for the latter. The %TD of those was low than that of UO 2 (95%TD). One of main reason is that the ROX fuels were sintered at temperature of 1,400degC while UO 2 was sintered at temperature of 1,700degC. The average pore diameter was about 3 μm and the porosity was 17-18%, implying that the ROX fuels were resulted in containing a lot of small pores 2 . As-fabricated density determined in the present study was about 4.5-5.5g/cc. The %TD of the simulated ROX fuels ranged about 94-98%TD. They were in the same as that of UO 2 because as high sintering temperature of 1,750degC as UO 2 . The average pore diameter was about 4-8 μm and the porosity was <6%. Grain size was revealed to be about 1-4 μm. (author)

  14. Results from In-pile experiments on LWR fuel rod behavior under LOCA conditions with unirradiated rods

    International Nuclear Information System (INIS)

    Sepold, L.; Karb, E.H.; Pruessmann, M.

    1981-06-01

    This report summarizes the results of the FR2-in-pile tests at KfK (Kernforschungszentrum Karlsruhe) with unirradiated test rods. The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechanisms of fuel rod failure were being performed with irradiated and unirradiated single rods of a PWR design in the DK loop of the FR2 reactor. The main parameter of the test program was the burnup, ranging from 2.500 to 35.000 MWd/t. The program with unirradiated specimens comprised the series A and B with a total of 14 tests. (orig.) [de

  15. Precession electron diffraction for SiC grain boundary characterization in unirradiated TRISO fuel

    International Nuclear Information System (INIS)

    Lillo, T.M.; Rooyen, I.J. van; Wu, Y.Q.

    2016-01-01

    Highlights: • SiC grain orientation determined by TEM-based precession electron diffraction. • Orientation data improved with increasing TEM sample thickness. • Fraction of low angle grain boundaries lower from PED data than EBSD data. • Fractions of high angle and CSL-related boundaries similar to EBSD data. - Abstract: Precession electron diffraction (PED), a transmission electron microscopy-based technique, has been evaluated for the suitability for evaluating grain boundary character in the SiC layer of tristructural isotropic (TRISO) fuel. This work reports the effect of transmission electron microscope (TEM) lamella thickness on the quality of data and establishes a baseline comparison to SiC grain boundary characteristics, in an unirradiated TRISO particle, determined previously using a conventional electron backscatter diffraction (EBSD) scanning electron microscope (SEM)-based technique. In general, it was determined that the lamella thickness produced using the standard focused ion beam (FIB) fabrication process (∼80 nm), is sufficient to provide reliable PED measurements, although thicker lamellae (∼120 nm) were found to produce higher quality orientation data. Also, analysis of SiC grain boundary character from the TEM-based PED data showed a much lower fraction of low-angle grain boundaries compared to SEM-based EBSD data from the SiC layer of a TRISO-coated particle made using the same fabrication parameters and a SiC layer deposited at a slightly lower temperature from a surrogate TRISO particle. However, the fractions of high-angle and coincident site lattice (CSL)-related grain boundaries determined by PED are similar to those found using SEM-based EBSD. Since the grain size of the SiC layer of TRSIO fuel can be as small as 250 nm (Kirchhofer et al., 2013), depending on the fabrication parameters, and since grain boundary fission product precipitates in irradiated TRISO fuel can be nano-sized, the TEM-based PED orientation data

  16. Response of unirradiated and irradiated PWR fuel rods tested under power-cooling-mismatch conditions

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Quapp, W.J.; Martinson, Z.R.; McCardell, R.K.; Mehner, A.S.

    1978-01-01

    This report summarizes the results from the single-rod power-cooling-mismatch (PCM) and irradiation effects (IE) tests conducted to date in the Power Burst Facility (PBF) at the U.S. DOE Idaho National Engineering Laboratory. This work was performed for the U.S. NRC under contact to the Department of Energy. These tests are part of the NRC Fuel Behavior Program, which is designed to provide data for the development and verification of analytical fuel behavior models that are used to predict fuel response to abnormal or postulated accident conditions in commercial LWRs. The mechanical, chemical and thermal response of both previously unirradiated and previously irradiated LWR-type fuel rods tested under power-cooling-mismatch condition is discussed. A brief description of the test designs is presented. The results of the PCM thermal-hydraulic studies are summarized. Primary emphasis is placed on the behavior of the fuel and cladding during and after stable film boiling. (orig.) [de

  17. The single SNR fuel assembly container (ESBB) to transport unirradiated SNR 300 fuel assemblies

    International Nuclear Information System (INIS)

    Hilbert, F.; Hottenrott, G.

    1998-01-01

    In this paper a new type B(U) package design is presented. The Single SNR Fuel Assembly Container (ESBB) is designed for the transport and storage of a single SNR 300 fuel assembly. This package is the main component for the future interim storage of the fuel assemblies in heavy storage casks. Its benefits are that it is compatible with the Category I transport system of Nuclear Cargo + Service NCS) used in Germany and that it can be easily handled at the current storage locations as well as in an interim storage facility. In total 205 fuel assemblies are currently stored in Hanau, Germany and Dounreay, U.K. Former studies have shown, that heavy transport and storage casks can be handled there only with considerable efforts. But the required category I transport to an interim storage is not reasonably feasible. To overcome these problems the ESBB was designed. It consists of a stainless steel tube with welded bottom, a welded plug as closure system and shock absorbers 26 packages at maximum can be transported in one batch with the NCS security vehicle. The safety analysis shows that the package complies with IAEA 1996. Standard calculations methods and computer codes like HEATING 7.2 (Childs 1993) have been used for the analysis. Criticality safety assessment is based on conservative assumptions as required in IAEA 1996. Drop tests carried out by BAM will be used to verify the design. These tests are scheduled for mid 1998. For the validation of the design prototypes have already been manufactured. Handling tests show that the design complies with the requirements. Preliminary drop tests show that the certification drop tests will be passed positively. (authors)

  18. Characterization of un-irradiated MIMAS MOX fuel by Raman spectroscopy and EPMA

    Science.gov (United States)

    Talip, Zeynep; Peuget, Sylvain; Magnin, Magali; Tribet, Magaly; Valot, Christophe; Vauchy, Romain; Jégou, Christophe

    2018-02-01

    In this study, Raman spectroscopy technique was implemented to characterize un-irradiated MIMAS (MIcronized - MASter blend) MOX fuel samples with average 7 wt.% Pu content and different damage levels, 13 years after fabrication, one year after thermal recovery and soon after annealing, respectively. The impacts of local Pu content, deviation from stoichiometry and self-radiation damage on Raman spectrum of the studied MIMAS MOX samples were assessed. MIMAS MOX fuel has three different phases Pu-rich agglomerate, coating phase and uranium matrix. In order to distinguish these phases, Raman results were associated with Pu content measurements performed by Electron Microprobe Analysis. Raman results show that T2g frequency significantly shifts from 445 to 453 cm-1 for Pu contents increasing from 0.2 to 25 wt.%. These data are satisfactorily consistent with the calculations obtained with Gruneisen parameters. It was concluded that the position of the T2g band is mainly controlled by Pu content and self-radiation damage. Deviation from stoichiometry does not have a significant influence on T2g band position. Self-radiation damage leads to a shift of T2g band towards lower frequency (∼1-2 cm-1 for the UO2 matrix of damaged sample). However, this shift is difficult to quantify for the coating phase and Pu agglomerates given the dispersion of high Pu concentrations. In addition, 525 cm-1 band, which was attributed to sub-stoichiometric structural defects, is presented for the first time for the self-radiation damaged MOX sample. Thanks to the different oxidation resistance of each phase, it was shown that laser induced oxidation could be alternatively used to identify the phases. It is demonstrated that micro-Raman spectroscopy is an efficient technique for the characterization of heterogeneous MOX samples, due to its low spatial resolution.

  19. Application of fire-retardant treatment to the wood in Type A unirradiated nuclear fuel outer containers

    International Nuclear Information System (INIS)

    Whitlow, J.D.; Luna, R.E.

    1992-01-01

    Packagings for transporting unirradiated nuclear fuel assemblies in the United States are commonly constructed as rectangular boxes consisting of a metal inner container, a wooden outer container, and cushioning material separating the two. The wood in the outer container is a potential source of fuel for fire. Use of a fire-retardant treatment on the wood may reduce or eliminate the damage to nuclear fuel assemblies in some types of accidents involving fire. The applicability of using fire-retardant treatments on the wood of outer containers is addressed. An approximate cost-benefit analysis to determine if fire-retardant treatments are economically justified is presented. (Author)

  20. Flow sheet development for the dissolution of unirradiated Mark 42 fuel tubes in F-Canyon, Part II

    International Nuclear Information System (INIS)

    Murray, A.M.

    1999-01-01

    Two dissolution flow sheets were tested for the desorption of unirradiated Mark 42 fuel tubes. Both the aluminum (from the can, cladding, and fuel core) and the plutonium oxide (PuO 2 ) are dissolved simultaneously, i.e., a co-dissolution flow sheet. In the first series of tests, 0.15 and 0.20 molar (M) potassium fluoride (KF) solutions were used and the dissolution extended over several days. In the other series of tests, solutions with higher concentrations of fluoride (0.25 to 0.30 M) were used. Calcium fluoride (CaF 2 ) was used in those tests as the fluoride source

  1. The dissolution of unirradiated UO2 fuel pellets under simulated disposal conditions

    International Nuclear Information System (INIS)

    Ollila, K.; Leino-Forsman, H.

    1993-03-01

    The dissolution behaviour of unirradiated UO 2 pellets was studied as a function of water composition under oxidizing and reducing conditions at 25 deg C. The waters included deionized water as the reference water, sodium bicarbonate solutions with varying bicarbonate content, and two different synthetic groundwaters. The release of uranium was measured during static batch dissolution experiments of long duration (3-4 years)

  2. Dissolution of unirradiated UO{sub 2} fuel in synthetic groundwater. Final report (1996-1998)

    Energy Technology Data Exchange (ETDEWEB)

    Ollila, K. [VTT Chemical Technology, Espoo (Finland)

    1999-05-01

    This study was a part of the EU R and D programme 1994-1998: Nuclear Fission Safety, entitled `Source term for performance assessment of spent fuel as a waste form`. The research carried out at VTT Chemical Technology was focused on the effects of granitic groundwater composition and redox conditions on UO{sub 2} solubility and dissolution mechanisms. The synthetic groundwater compositions simulated deep granitic fresh and saline groundwaters, and the effects of the near-field material, bentonite, on very saline groundwater. Additionally, the Spanish granite/bentonite water was used. The redox conditions (Eh), which are obviously the most important factors that influence on UO{sub 2} solubility under the disposal conditions of spent fuel, varied from strongly oxidising (air-saturated), anaerobic (N{sub 2}, O{sub 2} < l ppm) to reducing (N{sub 2}, low Eh). The objective of the air-saturated dissolution experiments was to yield the maximum solution concentrations of U, and information on the formation of secondary phases that control the concentrations, with different groundwater compositions. The static batch solubility experiments of long duration (up to 1-2 years) were performed using unirradiated UO{sub 2} pellets and powder. Under anaerobic and reducing conditions, the solubilities were also approached from oversaturation. The results of the oxic, air-saturated dissolution experiments with UO{sub 2} powder showed that the increase in the salinity (< 1.7 M) had a minor effect on the measured steady-state concentrations of U. The concentrations, (1.2 ...2.5) x 10{sup -5} M, were at the level of the theoretical solubility of schoepite or another uranyl oxide hydrate, e.g. becquerelite (possibly Na-polyuranate). The higher alkalinity of the fresh (Allard) composition increased the aqueous U concentration. Only some kind of oxidised U-phase (U{sub 3}O{sub 8}-UO{sub 3}) was identified with XRD when studying possible secondary phases after the contact time of one year

  3. New method to calculate the mechanical properties of unirradiated fuel cladding from ring tensile tests

    Energy Technology Data Exchange (ETDEWEB)

    Martin-Rengel, M.A. [Departamento de Ciencia de Materiales, UPM, E.T.S.I. Caminos, Canales y Puertos, Profesor Aranguren s/n, E-28040 Madrid (Spain); Consejo de Seguridad Nuclear (CSN), Justo Dorado 11, E-28040 Madrid (Spain); Gomez, F.J.; Ruiz-Hervias, J.; Caballero, L.; Valiente, A. [Departamento de Ciencia de Materiales, UPM, E.T.S.I. Caminos, Canales y Puertos, Profesor Aranguren s/n, E-28040 Madrid (Spain)

    2009-06-15

    Nuclear fuel cladding is the first barrier used to confine the fuel and the fission products produced during irradiation. Zirconium alloys are used for this purpose due to their remarkable neutron transparency, together with their good mechanical properties at operational temperatures. Consequently, it is very important to be able to characterize the mechanical response of the irradiated cladding. The mechanical behaviour of the material can be modelled as elastoplastic with different stress-strain curves depending on the direction: radial, hoop or longitudinal direction. The ring tensile test has been proposed to determine the mechanical properties of the cladding along the hoop direction. The initial test consisted of applying a force inside the tube, by means of two half cylinders. Later Arsene and Bai [1,2] modified the experimental device to avoid tube bending at the beginning of the test. The same authors proposed a numerical method to obtain the stress-strain curve in the hoop direction from the experimental load versus displacement results and a given friction coefficient between the loading pieces and the sample [3]. This method has been used by different authors [4] with slight modifications. It is based on the existence of two universal curves under small strain hypothesis: the first correlating the hoop strain and the displacement of the loading piece and the second one correlating the hoop stress and the applied load. In this work, a new method to determine the mechanical properties of the cladding from the ring tensile test results is proposed. Non-linear geometry is considered and an iterative procedure is proposed so universal curves are not needed. A stress-strain curve is determined by combining numerical calculations with experimental results in a convergent loop. The two universal curves proposed by Arsene and Bai [3] are substituted by two relationships, one between the equivalent plastic strain in the centre of the specimen ligament and the

  4. Dissolution of unirradiated UO2 fuel in synthetic groundwater. Final report (1996-1998)

    International Nuclear Information System (INIS)

    Ollila, K.

    1999-05-01

    This study was a part of the EU R and D programme 1994-1998: Nuclear Fission Safety, entitled 'Source term for performance assessment of spent fuel as a waste form'. The research carried out at VTT Chemical Technology was focused on the effects of granitic groundwater composition and redox conditions on UO 2 solubility and dissolution mechanisms. The synthetic groundwater compositions simulated deep granitic fresh and saline groundwaters, and the effects of the near-field material, bentonite, on very saline groundwater. Additionally, the Spanish granite/bentonite water was used. The redox conditions (Eh), which are obviously the most important factors that influence on UO 2 solubility under the disposal conditions of spent fuel, varied from strongly oxidising (air-saturated), anaerobic (N 2 , O 2 2 , low Eh). The objective of the air-saturated dissolution experiments was to yield the maximum solution concentrations of U, and information on the formation of secondary phases that control the concentrations, with different groundwater compositions. The static batch solubility experiments of long duration (up to 1-2 years) were performed using unirradiated UO 2 pellets and powder. Under anaerobic and reducing conditions, the solubilities were also approached from oversaturation. The results of the oxic, air-saturated dissolution experiments with UO 2 powder showed that the increase in the salinity ( -5 M, were at the level of the theoretical solubility of schoepite or another uranyl oxide hydrate, e.g. becquerelite (possibly Na-polyuranate). The higher alkalinity of the fresh (Allard) composition increased the aqueous U concentration. Only some kind of oxidised U-phase (U 3 O 8 -UO 3 ) was identified with XRD when studying possible secondary phases after the contact time of one year with all groundwater compositions. Longer contact times are needed to identify secondary phases predicted by modelling (EQ3/6). In the anoxic dissolution experiments with UO 2 pellets, the

  5. Properties of unirradiated fuel element graphites H-451 and SO818. [Bulk density, tensile properties, thermal expansion, thermal conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Engle, G.B.; Johnson, W.R.

    1976-10-08

    Nuclear graphites H-451, lot 440 (Great Lakes Carbon Corporation (GLCC)), and SO818 (Airco Speer Division, Air Reduction Corporation (AS)) are described, and physical, mechanical, and chemical property data are presented for the graphites in the unirradiated state. A summary of the mean values of the property data and of data on TS-1240 and H-451, lot 426, is tabulated. A direct comparison of H-451, lot 426, chosen for Fort St. Vrain (FSV) fuel reload production, TS-1240, and SO818 may be made from the table. (auth)

  6. Study on the behavior of unirradiated light water reactor fuel with iodine-127 under the reactivity initiated accident (RIA) conditions

    International Nuclear Information System (INIS)

    Sasajima, Hideo; Yanagisawa, Kazuaki; Kanazawa, Hiroyuki

    1988-07-01

    In a light water reactor fuel, a stress arised from pellet-cladding interaction (PCI) will have possibility to cause stress corrosion claddling (PCI failure) under an environment of corrosive fission product ; line iodine. A pulse irradiation experiment by NSRR was carried out to examine whether or not an unirradiated NSRR standard fuel rod in which 1.3 mg (33 x 10 -6 g/cm 2 ) of iodine was artificially filled could cause the PCI failure. Obtained results are: (1) The fuel rods with iodine did not fail both at deposited energy levels of 268 and 280 cal/g · UO 2 . On the other hand, the fuel rods without iodine failed at the same energy levels due to thinning of the cladding wall thickness. Within this experimental scope, PCI-failure did not occur on iodine filled fuel rods. (2) At a periphery of the fuel pellet of iodine filled rod, an uniform torus ring was formed. The torus ring consisted of an equi-axed large grains at 268 cal/g · UO 2 and a columnar ones at 280 cal/g · UO 2 . The torus ring was not formed in the fuel without iodine. (author)

  7. Lift-based up-ender and methods using same to manipulate a shipping container containing unirradiated nuclear fuel

    Science.gov (United States)

    Nilles, Michael J.

    2017-08-01

    A shipping container containing an unirradiated nuclear fuel assembly is lifted off the ground by operating a crane to raise a lifting tool comprising a winch. The lifting tool is connected with the shipping container by a rigging line connecting with the shipping container at a lifting point located on the shipping container between the top and bottom of the shipping container, and by winch cabling connecting with the shipping container at the top of the shipping container. The shipping container is reoriented by operating the winch to adjust the length of the winch cabling so as to rotate the shipping container about the lifting point. Shortening the winch cabling rotates the shipping container about the lifting point from a horizontal orientation to a vertical orientation, while lengthening the winch cabling rotates the shipping container about the lifting point from the vertical orientation to the horizontal orientation.

  8. Storage of unirradiated fuel in borated concrete at the Savannah River Plant

    International Nuclear Information System (INIS)

    Honkonen, D.L.

    1979-06-01

    At the Savannah River Plant (SRP), more than 3000 enriched uranium fuel elements can be stored in horizontal holes in borated concrete racks. This method of storage was selected. This paper describes the largest of these racks and the reactivity calculations and measurements which confirmed that SRP fuel may be safely stored in them

  9. Unirradiated characteristics of U-Si alloys as dispersed-phase fuels

    International Nuclear Information System (INIS)

    Domagala, R.F.; Wiencek, T.C.

    1987-06-01

    To satisfy the power demands of many research reactors, a new LEU fuel with a high density and U content was needed. Any fuel must be compatible with Al and its alloys so that it may be fabricable as a dispersed-phase in Al alloy and Al matrix plate-type elements following, as nearly as possible, established commercial manufacturing techniques. U-Si and U-Si-Al alloys at or near the composition of U 3 Si were immediately attractive because of work documented by the Canadians. 8 refs., 2 figs

  10. Interaction between UO2 kernel and pyrocarbon coating in irradiated and unirradiated HTR fuel particles

    International Nuclear Information System (INIS)

    Drago, A.; Klersy, R.; Simoni, O.; Schrader, K.H.

    1975-08-01

    Experimental observations on unidirectional UO 2 kernel migration in TRISO type coated particle fuels are reported. An analysis of the experimental results on the basis of data and models from the literature is reported. The stoichiometric composition of the kernel is considered the main parameter that, associated with a temperature gradient, controls the unidirectional kernel migration

  11. Uranium and plutonium distribution in unirradiated mixed oxide fuel from industrial fabrication

    International Nuclear Information System (INIS)

    Hanus, D.; Kleykamp, H.

    1982-01-01

    Different process variants developed in the last few years by the firm ALKEM to manufacture FBR and LWR mixed oxide fuel are given. The uranium and plutonium distribution is determined on the pellets manufactured with the help of the electron beam microprobe. The stepwise improvement of the uranium-plutonium homogeneity in the short-term developed granulate variants and in the long-term developed new processes are illustrated starting with early standard processes for FBR fuel. An almost uniform uranium-plutonium distribution could be achieved for the long-term developed new processes (OKOM, AuPuC). The uranium-plutonium homogeneity are quantified in the pellets manufactured according to the considered process variants with a newly defined quality number. (orig.)

  12. Oxide particle size distribution from shearing irradiated and unirradiated LWR fuels in Zircaloy and stainless steel cladding: significance for risk assessment

    Energy Technology Data Exchange (ETDEWEB)

    Davis, W. Jr.; West, G.A.; Stacy, R.G.

    1979-03-22

    Sieve fractionation was performed with oxide particles dislodged during shearing of unirradiated or irradiated fuel bundles or single rods of UO/sub 2/ or 96 to 97% ThO/sub 2/--3 to 4% UO/sub 2/. Analyses of these data by nonlinear least-squares techniques demonstrated that the particle size distribution is lognormal. Variables involved in the numerical analyses include lognormal median size, lognormal standard deviation, and shear cut length. Sieve-fractionation data are presented for unirradiated bundles of stainless-steel-clad or Zircaloy-2-clad UO/sub 2/ or ThO/sub 2/--UO/sub 2/ sheared into lengths from 0.5 to 2.0 in. Data are also presented for irradiated single rods (sheared into lengths of 0.25 to 2.0 in.) of Zircaloy-2-clad UO/sub 2/ from BWRs and of Zircaloy-4-clad UO/sub 2/ from PWRs. Median particle sizes of UO/sub 2/ from shearing irradiated stainless-steel-clad fuel ranged from 103 to 182 ..mu..m; particle sizes of ThO/sub 2/--UO/sub 2/, under these same conditions, ranged from 137 to 202 ..mu..m. Similarly, median particle sizes of UO/sub 2/ from shearing unirradiated Zircaloy-2-clad fuel ranged from 230 to 957 ..mu..m. Irradiation levels of fuels from reactors ranged from 9,000 to 28,000 MWd/MTU. In general, particle sizes from shearing these irradiated fuels are larger than those from the unirradiated fuels; however, unirradiated fuel from vendors was not available for performing comparative shearing experiments. In addition, variations in particle size parameters pertaining to samples of a single vendor varied as much as those between different vendors. The fraction of fuel dislodged from the cladding is nearly proportional to the reciprocal of the shear cut length, until the cut length attains some minimum value below which all fuel is dislodged. Particles of fuel are generally elongated with a long-to-short axis ratio usually less than 3. Using parameters of the lognormal distribution estimates can be made of fractions of dislodged fuel having

  13. Oxide particle size distribution from shearing irradiated and unirradiated LWR fuels in Zircaloy and stainless steel cladding: significance for risk assessment

    International Nuclear Information System (INIS)

    Davis, W. Jr.; West, G.A.; Stacy, R.G.

    1979-01-01

    Sieve fractionation was performed with oxide particles dislodged during shearing of unirradiated or irradiated fuel bundles or single rods of UO 2 or 96 to 97% ThO 2 --3 to 4% UO 2 . Analyses of these data by nonlinear least-squares techniques demonstrated that the particle size distribution is lognormal. Variables involved in the numerical analyses include lognormal median size, lognormal standard deviation, and shear cut length. Sieve-fractionation data are presented for unirradiated bundles of stainless-steel-clad or Zircaloy-2-clad UO 2 or ThO 2 --UO 2 sheared into lengths from 0.5 to 2.0 in. Data are also presented for irradiated single rods (sheared into lengths of 0.25 to 2.0 in.) of Zircaloy-2-clad UO 2 from BWRs and of Zircaloy-4-clad UO 2 from PWRs. Median particle sizes of UO 2 from shearing irradiated stainless-steel-clad fuel ranged from 103 to 182 μm; particle sizes of ThO 2 --UO 2 , under these same conditions, ranged from 137 to 202 μm. Similarly, median particle sizes of UO 2 from shearing unirradiated Zircaloy-2-clad fuel ranged from 230 to 957 μm. Irradiation levels of fuels from reactors ranged from 9,000 to 28,000 MWd/MTU. In general, particle sizes from shearing these irradiated fuels are larger than those from the unirradiated fuels. In addition, variations in particle size parameters pertaining to samples of a single vendor varied as much as those between different vendors. The fraction of fuel dislodged from the cladding is nearly proportional to the reciprocal of the shear cut length, until the cut length attains some minimum value below which all fuel is dislodged. Particles of fuel are generally elongated with a long-to-short axis ratio usually less than 3. Using parameters of the lognormal distribution deduced from experimental data, realistic estimates can be made of fractions of dislodged fuel having dimensions less than specified values

  14. Oxidative dissolution of unirradiated Mimas MOX fuel (U/Pu oxides) in carbonated water under oxic and anoxic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Odorowski, Mélina [CEA/DEN/DTCD/SECM/LMPA, BP 17171, 30207 Bagnols-sur-Cèze Cedex (France); MINES ParisTech, PSL Research University, Centre de Géosciences, 35 rue St Honoré, 77305 Fontainebleau (France); Jégou, Christophe, E-mail: christophe.jegou@cea.fr [CEA/DEN/DTCD/SECM/LMPA, BP 17171, 30207 Bagnols-sur-Cèze Cedex (France); De Windt, Laurent [MINES ParisTech, PSL Research University, Centre de Géosciences, 35 rue St Honoré, 77305 Fontainebleau (France); Broudic, Véronique; Peuget, Sylvain; Magnin, Magali; Tribet, Magaly [CEA/DEN/DTCD/SECM/LMPA, BP 17171, 30207 Bagnols-sur-Cèze Cedex (France); Martin, Christelle [Agence nationale pour la gestion des déchets radioactifs (Andra), DRD/CM, 1-7 rue Jean-Monnet, 92298 Châtenay-Malabry Cedex (France)

    2016-01-15

    Few studies exist concerning the alteration of Mimas Mixed-OXide (MOX) fuel, a mixed plutonium and uranium oxide, and data is needed to better understand its behavior under leaching, especially for radioactive waste disposal. In this study, two leaching experiments were conducted on unirradiated MOX fuel with a strong alpha activity (1.3 × 10{sup 9} Bq.g{sub MOX}{sup −1} reproducing the alpha activity of spent MOX fuel with a burnup of 47 GWd·t{sub HM}{sup −1} after 60 years of decay), one under air (oxic conditions) for 5 months and the other under argon (anoxic conditions with [O{sub 2}] < 1 ppm) for one year in carbonated water (10{sup −2} mol L{sup −1}). For each experiment, solution samples were taken over time and Eh and pH were monitored. The uranium in solution was assayed using a kinetic phosphorescence analyzer (KPA), plutonium and americium were analyzed by a radiochemical route, and H{sub 2}O{sub 2} generated by the water radiolysis was quantified by chemiluminescence. Surface characterizations were performed before and after leaching using Scanning Electron Microscopy (SEM), Electron Probe Microanalyzer (EPMA) and Raman spectroscopy. Solubility diagrams were calculated to support data discussion. The uranium releases from MOX pellets under both oxic and anoxic conditions were similar, demonstrating the predominant effect of alpha radiolysis on the oxidative dissolution of the pellets. The uranium released was found to be mostly in solution as carbonate species according to modeling, whereas the Am and Pu released were significantly sorbed or precipitated onto the TiO{sub 2} reactor. An intermediate fraction of Am (12%) was also present as colloids. SEM and EPMA results indicated a preferential dissolution of the UO{sub 2} matrix compared to the Pu-enriched agglomerates, and Raman spectroscopy showed the Pu-enriched agglomerates were slightly oxidized during leaching. Unlike Pu-enriched zones, the UO{sub 2} grains were much more

  15. Solubility of unirradiated UO2 fuel in aqueous solutions. Comparison between experimental and calculated (EQ3/6) data

    International Nuclear Information System (INIS)

    Ollila, K.

    1995-11-01

    The solubility behaviour of unirradiated UO 2 pellets was studied under oxic (air-saturated) and anoxic (N 2 ) conditions in deionized water, in sodium bicarbonate solutions with varying bicarbonate content (60 - 600 ppm), in Allard groundwater simulating granitic fresh groundwater conditions, and in bentonite water simulating the effects of bentonite on granitic fresh groundwater (25 deg C). The release of uranium was measured during static batch dissolution experiments of long duration (2-6 years). A comparison was made with the theoretical solubility data calculated with the geochemical code EQ3/6 in order to evaluate solubility (steady state) limiting factors. (orig.) (26 refs., 32 figs., 13 tabs.)

  16. Dissolution rates of unirradiated UO2, UO2 doped with 233U, and spent fuel under normal atmospheric conditions and under reducing conditions using an isotope dilution method

    International Nuclear Information System (INIS)

    Ollila, Kaija; Albinsson, Yngve; Oversby, Virginia; Cowper, Mark

    2003-10-01

    The experimental results given in this report allow us to draw the following conclusions. 1) Tests using unirradiated fuel pellet materials from two different manufacturers gave very different dissolution rates under air atmosphere testing. Tests for fragments of pellets from different pellets made by the same manufacturer gave good agreement. This indicates that details of the manufacturing process have a large effect on the behavior of unirradiated UO 2 in dissolution experiments. Care must be taken in interpreting differences in results obtained in different laboratories because the results may be affected by manufacturing effects. 2) Long-term tests under air atmosphere have begun to show the effects of precipitation. Further testing will be needed before the samples reach steady state. 3) Testing of unirradiated UO 2 in systems containing an iron strip to produce reducing conditions gave [U] less than detection limits ( 235 U added as spike was recovered, indicating that 90% of the spike had precipitated onto the solid sample or the iron strip. 9) Tests of UO 2 pellet materials containing 233 U to provide an alpha decay activity similar to that expected for spent fuel 3000 and 10,000 years after disposal showed that the pellet materials behaved as expected under air atmosphere conditions, showing that the manufacturing method was successful. 10) Early testing of the 233 U-doped materials under reducing conditions showed relatively rapid (30 minute) dissolution of small amounts of U at the start of the puff test procedure. Results of analyses of an acidified fraction of the same solutions after 1 or 2 weeks holding indicate that the solutions were inhomogeneous, indicating the presence of colloidal material or small grains of solid. 11) Samples from the 233 U-doped tests initially indicated dissolution of solid during the first week of testing, with some indication of more rapid dissolution of the material with the higher doping. 12) The second cycle of testing

  17. 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

  18. Nanoindentation measurements of the mechanical properties of zirconium matrix and hydrides in unirradiated pre-hydrided nuclear fuel cladding

    International Nuclear Information System (INIS)

    Rico, A.; Martin-Rengel, M.A.; Ruiz-Hervias, J.; Rodriguez, J.; Gomez-Sanchez, F.J.

    2014-01-01

    It is well known that the mechanical properties of the nuclear fuel cladding may be affected by the presence of hydrides. The average mechanical properties of hydrided cladding have been extensively investigated from a macroscopic point of view. In addition, the mechanical and fracture properties of bulk hydride samples fabricated from zirconium plates have also been reported. In this paper, Young’s modulus, hardness and yield stress are measured for each phase, namely zirconium hydrides and matrix, of pre-hydrided nuclear fuel cladding. To this end, nanoindentation tests were performed on ZIRLO samples in as-received state, on a hydride blister and in samples with 150 and 1200 ppm of hydrogen homogeneously distributed along the hoop direction of the cladding. The results show that the measured mechanical properties of the zirconium hydrides and ZIRLO matrix (Young’s modulus, hardness and yield stress) are rather similar. From the experimental data, the hydride volume fraction in the cladding samples with 150 and 1200 ppm was estimated and the average mechanical properties were calculated by means of the rule of mixtures. These values were compared with those obtained from ring compression tests. Good agreement between the results obtained by both methods was found

  19. Spent fuel storage requirements

    International Nuclear Information System (INIS)

    Fletcher, J.

    1982-06-01

    Spent fuel storage requirements, as projected through the year 2000 for U.S. LWRs, were calculated using information supplied by the utilities reflecting plant status as of December 31, 1981. Projections through the year 2000 combined fuel discharge projections of the utilities with the assumed discharges of typical reactors required to meet the nuclear capacity of 165 GWe projected by the Energy Information Administration (EIA) for the year 2000. Three cases were developed and are summarized. A reference case, or maximum at-reactor (AR) capacity case, assumes that all reactor storage pools are increased to their maximum capacities as estimated by the utilities for spent fuel storage utilizing currently licensed technologies. The reference case assumes no transshipments between pools except as currently licensed by the Nuclear Regulatory Commission (NRC). This case identifies an initial requirement for 13 MTU of additional storage in 1984, and a cumulative requirement for 14,490 MTU additional storage in the year 2000. The reference case is bounded by two alternative cases. One, a current capacity case, assumes that only those pool storage capacity increases currently planned by the operating utilities will occur. The second, or maximum capacity with transshipment case, assumes maximum development of pool storage capacity as described above and also assumes no constraints on transshipment of spent fuel among pools of reactors of like type (BWR, PWR) within a given utility. In all cases, a full core discharge capability (full core reserve or FCR) is assumed to be maintained for each reactor, except that only one FCR is maintained when two reactors share a common pool. For the current AR capacity case the indicated storage requirements in the year 2000 are indicated to be 18,190 MTU; for the maximum capacity with transshipment case they are 11,320 MTU

  20. 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)

  1. Stress corrosion crack growth in unirradiated zircaloy

    International Nuclear Information System (INIS)

    Pettersson, K.

    1978-10-01

    Experimental techniques suitable for the determination of stress corrosion crack growth rates in irradiated Zircaloy tube have been developed. The techniques have been tested on unirradiated. Zircaloy and it was found that the results were in good agreement with the results of other investigations. Some of the results were obtained at very low stress intensities and the crack growth rates observed, gave no indication of the existance of a K sub(ISCC) for iodine induced stress corrosion cracking in Zircaloy. This is of importance both for fuel rod behavior after a power ramp and for long term storage of spent Zircaloy-clad fuel. (author)

  2. Programmatic and technical requirements for the FMDP fresh MOX fuel transport package

    International Nuclear Information System (INIS)

    Ludwig, S.B.; Michelhaugh, R.D.; Pope, R.B.

    1997-12-01

    This document is intended to guide the designers of the package to all pertinent regulatory and other design requirements to help ensure the safe and efficient transport of the weapons-grade (WG) fresh MOX fuel under the Fissile Materials Disposition Program. To accomplish the disposition mission using MOX fuel, the unirradiated MOX fuel must be transported from the MOX fabrication facility to one or more commercial reactors. Because the unirradiated fuel contains large quantities of plutonium and is not sufficient radioactive to create a self-protecting barrier to deter the material from theft, DOE intends to use its fleet of safe secure trailers (SSTs) to provide the necessary safeguards and security for the material in transit. In addition to these requirements, transport of radioactive materials must comply with regulations of the Department of Transportation and the Nuclear Regulatory Commission (NRC). In particular, NRC requires that the packages must meet strict performance requirements. The requirements for shipment of MOX fuel (i.e., radioactive fissile materials) specify that the package design is certified by NRC to ensure the materials contained in the packages are not released and remain subcritical after undergoing a series of hypothetical accident condition tests. Packages that pass these tests are certified by NRC as a Type B fissile (BF) package. This document specifies the programmatic and technical design requirements a package must satisfy to transport the fresh MOX fuel assemblies

  3. Unirradiated cladding rip-propagation tests

    International Nuclear Information System (INIS)

    Hu, W.L.; Hunter, C.W.

    1981-04-01

    The size of cladding rips which develop when a fuel pin fails can affect the subassembly cooling and determine how rapidly fuel escapes from the pin. The object of the Cladding Rip Propagation Test (CRPT) was to quantify the failure development of cladding so that a more realistic fuel pin failure modeling may be performed. The test results for unirradiated 20% CS 316 stainless steel cladding show significantly different rip propagation behavior at different temperatures. At room temperature, the rip growth is stable as the rip extension increases monotonically with the applied deformation. At 500 0 C, the rip propagation becomes unstable after a short period of stable rip propagation. The rapid propagation rate is approximately 200 m/s, and the critical rip length is 9 mm. At test temperatures above 850 0 C, the cladding exhibits very high failure resistances, and failure occurs by multiple cracking at high cladding deformation. 13 figures

  4. Special Analysis for the Disposal of the Idaho National Laboratory Unirradiated Light Water Breeder Reactor Rods and Pellets Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-08-31

    The purpose of this special analysis (SA) is to determine if the Idaho National Laboratory (INL) Unirradiated Light Water Breeder Reactor (LWBR) Rods and Pellets waste stream (INEL103597TR2, Revision 2) is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS). The INL Unirradiated LWBR Rods and Pellets waste stream consists of 24 containers with unirradiated fabricated rods and pellets composed of uranium oxide (UO2) and thorium oxide (ThO2) fuel in zirconium cladding. The INL Unirradiated LWBR Rods and Pellets waste stream requires an SA because the 229Th, 230Th, 232U, 233U, and 234U activity concentrations exceed the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC) Action Levels.

  5. World nuclear fuel cycle requirements 1991

    Energy Technology Data Exchange (ETDEWEB)

    1991-10-10

    The nuclear fuel cycle consists of mining and milling uranium ore, processing the uranium into a form suitable for generating electricity, burning'' the fuel in nuclear reactors, and managing the resulting spent nuclear fuel. This report presents projections of domestic and foreign requirements for natural uranium and enrichment services as well as projections of discharges of spent nuclear fuel. These fuel cycle requirements are based on the forecasts of future commercial nuclear power capacity and generation published in a recent Energy Information Administration (EIA) report. Also included in this report are projections of the amount of spent fuel discharged at the end of each fuel cycle for each nuclear generating unit in the United States. The International Nuclear Model is used for calculating the projected nuclear fuel cycle requirements. 14 figs., 38 tabs.

  6. World nuclear fuel cycle requirements 1991

    International Nuclear Information System (INIS)

    1991-01-01

    The nuclear fuel cycle consists of mining and milling uranium ore, processing the uranium into a form suitable for generating electricity, ''burning'' the fuel in nuclear reactors, and managing the resulting spent nuclear fuel. This report presents projections of domestic and foreign requirements for natural uranium and enrichment services as well as projections of discharges of spent nuclear fuel. These fuel cycle requirements are based on the forecasts of future commercial nuclear power capacity and generation published in a recent Energy Information Administration (EIA) report. Also included in this report are projections of the amount of spent fuel discharged at the end of each fuel cycle for each nuclear generating unit in the United States. The International Nuclear Model is used for calculating the projected nuclear fuel cycle requirements. 14 figs., 38 tabs

  7. Design requirement on HYPER blanket fuel assembly

    International Nuclear Information System (INIS)

    Hwang, Woan; Lee, B. O.; Nam, C.; Ryu, W. S.; Lee, B. S.; Park, W. S.

    2000-07-01

    This document describes design requirements which are needed for designing the blanket assembly of the HYPER as design guidance. The blanket assembly of the HYPER consists of blanket fuel rods, mounting rail, spacer, upper nozzle with handling socket, bottom nozzle with mounting rail and skeleton structure. The blanket fuel rod consists of top end plug, bottom end plug with key way, blanket fuel slug, and cladding. In the assembly, the rods are in a triangular pitch array. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements for the blanket fuel assembly of the HYPER

  8. Dissolution rates of unirradiated UO{sub 2}, UO{sub 2} doped with {sup 233}U, and spent fuel under normal atmospheric conditions and under reducing conditions using an isotope dilution method

    Energy Technology Data Exchange (ETDEWEB)

    Ollila, Kaija [VTT Processes, Helsinki (Finland); Albinsson, Yngve [Chalmers Univ. of Technology, Goeteborg (Sweden); Oversby, Virginia [VMO Konsult, Stockholm (Sweden); Cowper, Mark [AEA Technology, Harwell (United Kingdom)

    2003-10-01

    The experimental results given in this report allow us to draw the following conclusions. 1) Tests using unirradiated fuel pellet materials from two different manufacturers gave very different dissolution rates under air atmosphere testing. Tests for fragments of pellets from different pellets made by the same manufacturer gave good agreement. This indicates that details of the manufacturing process have a large effect on the behavior of unirradiated UO{sub 2} in dissolution experiments. Care must be taken in interpreting differences in results obtained in different laboratories because the results may be affected by manufacturing effects. 2) Long-term tests under air atmosphere have begun to show the effects of precipitation. Further testing will be needed before the samples reach steady state. 3) Testing of unirradiated UO{sub 2} in systems containing an iron strip to produce reducing conditions gave [U] less than detection limits (<0.02 ppb) after a few days to a few weeks of testing. Uranium recovered from the rinsing of reaction vessels and from acid stripping of vessels was shown to be from dissolution of grains of solid dislodged when the samples were handled after the tests were terminated. 4) Batch tests conducted under reducing conditions showed evidence of colloidal material in the early solution samples. 5) In the batch tests, measurements taken at day 3 and day 5 show that precipitation occurs from day 3 to day 5 without any further dissolution of the solid. 6) At termination of the batch tests, all but one sample had [U] in solution less than detection limits (< 0.02 ppb). Materials recovered in test termination samples showed evidence for recovery of small amounts - amounts corresponding to that expected from a few grains of 5 to 10 {mu}m size - in the acidified solution samples. These are interpreted to have been dislodged during sample handling operations. 7) Batch test data show that increasing test duration beyond 2 weeks does not provide

  9. Spent fuel storage requirements, 1988

    International Nuclear Information System (INIS)

    1988-10-01

    Historical inventories of spent fuel and Department of Energy (DOE) estimates of future discharges from US commercial nuclear reactors are presented for the next 20 years, through the year 2007. The eventual needs for additional spent fuel storage capacity are estimated. These estimates are based on the maximum capacities within current and planned at-reactor facilities and on any planned transshipments of fuel to other reactors or facilities. Historical data through December 1987 and projected discharges through the end of reactor life are used in this analysis. The source data was supplied by the utilities to DOE through the 1988 RW-859 data survey and by DOE estimates of future nuclear capacity, generation, and spent fuel discharges. 12 refs., 3 figs., 28 tabs

  10. Spent fuel storage requirements 1987

    International Nuclear Information System (INIS)

    1987-09-01

    Historical inventories of spent fuel and utility estimates of future discharges from US commercial nuclear reactors are presented through the year 2005. The ultimate needs for additional storage capacity are estimated. These estimtes are based on the maximum capacities within current and planned at-reactor facilities and on any planned transshipments of fuel to other reactors or facilities. Historical data through December, 1986, and projected discharges through the end of reactor life are used in this analysis. The source data was supplied by the utilities to the DOE Energy Information Administration (EIA) through the 1987 RW-859 data survey. 14 refs., 4 figs., 9 tabs

  11. Fuel Cycle Requirements Code (FLYER). Summary report

    International Nuclear Information System (INIS)

    Gift, E.H.; Goode, W.D.

    1976-01-01

    Planning for, and the analysis of, the fuel requirements of the nuclear industry requires the ability to evaluate contingencies in many areas of the nuclear fuel cycle. The areas of nuclear fuel utilization, both uranium and plutonium, and of separative work requirements are of particular interest. The Fuel Cycle Requirements (FLYER) model has been developed to provide a flexible, easily managed tool for obtaining a comprehensive analysis of the nuclear fuel cycle. The model allows analysis of the interactions among the nuclear capacity growth rate, reactor technology and mix, and uranium and plutonium recycling capabilities. The model was initially developed as a means of analyzing nuclear growth contingencies with particular emphasis on the uranium feed and separative work requirements. It served to provide the planning group with analyses similar to the OPA's NUFUEL code which has only recently become available for general use. The model has recently been modified to account for some features of the fuel cycle in a more explicit manner than the NUFUEL code. For instance, the uranium requirements for all reactors installed in a given year are calculated for the total lifetime of those reactors. These values are cumulated in order to indicate the total uranium committed for reactors installed by any given year of the campaign. Similarly, the interactions in the back end of the fuel cycle are handled specifically, such as, the impacts resulting from limitations on the industrial capacity for reprocessing and mixed oxide fabrication of both light water reactor and breeder fuels. The principal features of the modified FLYER code are presented in summary form

  12. Requirements for materials of dispersion fuel elements

    International Nuclear Information System (INIS)

    Samojlov, A.G.; Kashtanov, A.I.; Volkov, V.S.

    1982-01-01

    Requirements for materials of dispersion fuel elements are considered. The necessity of structural and fissile materials compatibility at maximum permissible operation temperatures and temperatures arising in a fuel element during manufacture is pointed out. The fuel element structural material must be ductile, possess high mechanical strength minimum neutron absorption cross section, sufficient heat conductivity, good corrosion resistance in a coolant and radiation resistance. The fissile material must have high fissile isotope concentration, radiation resistance, high thermal conductivity, certain porosity high melting temperature must not change the composition under irradiation

  13. Fuel processing requirements and techniques for fuel cell propulsion power

    Science.gov (United States)

    Kumar, R.; Ahmed, S.; Yu, M.

    Fuels for fuel cells in transportation systems are likely to be methanol, natural gas, hydrogen, propane, or ethanol. Fuels other than hydrogen will need to be reformed to hydrogen on-board the vehicle. The fuel reformer must meet stringent requirements for weight and volume, product quality, and transient operation. It must be compact and lightweight, must produce low levels of CO and other byproducts, and must have rapid start-up and good dynamic response. Catalytic steam reforming, catalytic or noncatalytic partial oxidation reforming, or some combination of these processes may be used. This paper discusses salient features of the different kinds of reformers and describes the catalysts and processes being examined for the oxidation reforming of methanol and the steam reforming of ethanol. Effective catalysts and reaction conditions for the former have been identified; promising catalysts and reaction conditions for the latter are being investigated.

  14. World nuclear fuel cycle requirements 1989

    International Nuclear Information System (INIS)

    1989-01-01

    This analysis report presents the projected requirements for uranium concentrate and uranium enrichment services to fuel the nuclear power plants expected to be operating under two nuclear supply scenarios. These two scenarios, the Lower Reference and Upper Reference cases, apply to the United States, Canada, Europe, the Far East, and other countries in the World Outside Centrally Planned Economic Areas (WOCA). A No New Orders scenarios is also presented for the Unites States. This report contains an analysis of the sensitivities of the nuclear fuel cycle projections to different levels and types of projected nuclear capacity, different enrichment tails assays, higher and lower capacity factors, changes in nuclear fuel burnup levels, and other exogenous assumptions. The projections for the United States generally extend through the year 2020, and the WOCA projections, which include the United States, are provided through 2010. The report also presents annual projections of spent nuclear fuel; discharges and inventories of spent fuel. Appendix D includes domestic spent fuel projections through the year 2020 for the Lower and Upper Reference cases and through 2036, the last year in which spent fuel is discharged, for the No New Orders case

  15. World nuclear fuel cycle requirements 1990

    International Nuclear Information System (INIS)

    1990-01-01

    This analysis report presents the projected requirements for uranium concentrate and uranium enrichment services to fuel the nuclear power plants expected to be operating under three nuclear supply scenarios. Two of these scenarios, the Lower Reference and Upper Reference cases, apply to the United States, Canada, Europe, the Far East, and other countries with free market economies (FME countries). A No New Orders scenario is presented only for the United States. These nuclear supply scenarios are described in Commercial Nuclear Power 1990: Prospects for the United States and the World (DOE/EIA-0438(90)). This report contains an analysis of the sensitivities of the nuclear fuel cycle projections to different levels and types of projected nuclear capacity, different enrichment tails assays, higher and lower capacity factors, changes in nuclear fuel burnup levels, and other exogenous assumptions. The projections for the United States generally extend through the year 2020, and the FME projections, which include the United States, are provided through 2010. The report also presents annual projections of spent nuclear fuel discharges and inventories of spent fuel. Appendix D includes domestic spent fuel projections through the year 2030 for the Lower and Upper Reference cases and through 2040, the last year in which spent fuel is discharged, for the No New Orders case. These disaggregated projections are provided at the request of the Department of Energy's Office of Civilian Radioactive Waste Management

  16. Spent fuel storage requirements 1989--2020

    International Nuclear Information System (INIS)

    1989-10-01

    Historical inventories of spent fuel are combined with Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the US to provide estimates of spent fuel storage requirements over the next 32 years, through the year 2020. The needs for storage capacity beyond that presently available in the pools are estimated. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of fuel to other reactors or facilities. Historical data through December 1988 are derived from the 1989 Form RW-859 data survey of nuclear utilities. Projected discharges through the end of reactor life are based on DOE estimates of future nuclear capacity, generation, and spent fuel discharges. 14 refs., 3 figs., 28 tabs

  17. Spent fuel storage requirements, 1991--2040

    International Nuclear Information System (INIS)

    1991-12-01

    Historical inventories of spent fuel are combined with US Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements over the next 50 years, through the year 2040. The needs for storage capacity beyond that presently available in the pools are estimated. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. Historical data through December 1990 are derived from the 1991 Form RW-859 data survey of nuclear utilities. Projected discharges through the end of reactor life are based on DOE estimates of future nuclear capacity, generation, and spent fuel discharges

  18. Spent fuel storage requirements 1993--2040

    International Nuclear Information System (INIS)

    1994-09-01

    Historical inventories of spent fuel are combined with U.S. Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements through the year 2040. The needs are estimated for storage capacity beyond that presently available in the reactor storage pools. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of spent fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. The nuclear utilities provide historical data through December 1992 on the end of reactor life are based on the DOE/Energy Information Administration (EIA) estimates of future nuclear capacity, generation, and spent fuel discharges

  19. Spent fuel storage requirements, 1990--2040

    International Nuclear Information System (INIS)

    Walling, R.; Bierschbach, M.

    1990-11-01

    Historical inventories of spent fuel are combined with US Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements over the next 51 years, through the year 2040. The needs for storage capacity beyond that presently available in the pools are estimated. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. Historical data through December 1989 are derived from the 1990 Form RW-859 data survey of nuclear utilities. Projected discharges through the end of reactor life are based on DOE estimates of future nuclear capacity, generation, and spent fuel discharges. 15 refs., 3 figs., 11 tabs

  20. Nuclear fuels with high burnup: safety requirements

    International Nuclear Information System (INIS)

    Phuc Tran Dai

    2016-01-01

    Vietnam authorities foresees to build 3 reactors from Russian design (VVER AES 2006) by 2030. In order to prepare the preliminary report on safety analysis the Vietnamese Agency for Radioprotection and Safety has launched an investigation on the behaviour of nuclear fuels at high burnups (up to 60 GWj/tU) that will be those of the new plants. This study deals mainly with the behaviour of the fuel assemblies in case of loss of coolant (LOCA). It appears that for an average burnup of 50 GWj/tU and for the advanced design of the fuel assembly (cladding and materials) safety requirements are fulfilled. For an average burnup of 60 GWj/tU, a list of issues remains to be assessed, among which the impact of clad bursting or the hydrogen embrittlement of the advanced zirconium alloys. (A.C.)

  1. Solvent extraction studies of RERTR silicide fuels

    International Nuclear Information System (INIS)

    Gouge, Anthony P.

    1983-01-01

    Uranium silicide fuels, which are candidate RERTR fuel compositions, may require special considerations in solvent extraction reprocessing. Since Savannah River Plant may be reprocessing RERTR fuels as early as 1985, studies have been conducted at Savannah River Laboratory to demonstrate the solvent extraction behavior of this fuel. Results of solvent extraction studies with both unirradiated and irradiated fuel are presented along with the preliminary RERTR solvent extraction reprocessing flow sheet for Savannah River Plant. (author)

  2. Nuclear fuel cycle requirements in WOCA

    International Nuclear Information System (INIS)

    Klumpp, P.

    1982-02-01

    OECD/NEA will publsih an updated version of its study 'Nuclear Fuel Cycle Requirements and Supply Considerations, Through the Long-Term.' The Nuclear Research Centre Karlsruhe (KfK) was involved in the work necessary to provide this book. Although KfK had only responsiblility for part of the required computations it performed all the calculations for its own documentation interests. This documentation was felt to be a helpful background material for the reader of the second 'Yellow Book'. In this sense the original strategy computer outprints are published now without any discussion of assumptions and results. (orig.) [de

  3. Potential information requirements for spent nuclear fuel

    International Nuclear Information System (INIS)

    Disbrow, J.A.

    1991-01-01

    This paper reports that the Energy Information Administration (EIA) has performed analyses of the requirements for data and information for the management of commercial spent nuclear fuel (SNF) designated for disposal under the Nuclear Waste Policy Act (NWPA). Subsequently, the EIA collected data on the amounts and characteristics of SNF stored at commercial nuclear facilities. Most recently, the EIA performed an analysis of the international and domestic laws and regulations which have been established to ensure the safeguarding, accountability, and safe management of special nuclear materials (SNM). The SNM of interest are those designated for permanent disposal by the NWPA. This analysis was performed to determine what data and information may be needed to fulfill the specific accountability responsibilities of the Department of Energy (DOE) related to SNF handling, transportation, storage and disposal; to work toward achieving a consistency between nuclear fuel assembly identifiers and material weights as reported by the various responsible parties; and to assist in the revision of the Nuclear Fuel Data Form RW-859 used to obtain spent nuclear fuel characteristics data from the nuclear utilities

  4. EPA requirements for the uranium fuel cycle

    International Nuclear Information System (INIS)

    Dunster, H.J.

    1975-01-01

    The draft Environmental Statement issued by the Environmental Protection Agency (EPA) in the United States in preparation for Proposed Rulemaking Action concerning 'Environmental radiation protection requirements for normal operations of activities in the uranium fuel cycle' is summarized and discussed. The standards proposed by the EPA limit the annual dose equivalents to any member of the public, and also the releases of radionuclides to the 'general environment' for each gigawatt year of electrical energy produced. These standards were based on cost effectiveness arguements and levels and correspond to the ICRP recommendation to keep all exposures as low as reasonably achievable, economic and social factors being taken into account. They should be clearly distinguished from dose limits, although the EPA does not make this at all clear. The EPA seems to have shown an unexpected lack of understanding of the recommendations of ICRP Publication 9 (1965) and an apparent unawareness of ICRP Publication 22 (1973), and has therefore wrongly presented the new standards as a significant change in policy. The EPA has reviewed the information on the likely level of dose equivalents to members of the public and the likely cost reductions, thereby quantifying existing principles as applied to the fuel cycle as a whole. The EPA has stated that its proposals could be achieved as a cost in the region of Pound100,000 per death (or major genetic defect). It is pointed out that the EPA's use of the term 'waste' to exclude liquid and gaseous effluents may cause confusion. (U.K.)

  5. World nuclear fuel cycle requirements 1985

    International Nuclear Information System (INIS)

    Moden, R.; O'Brien, B.; Sanders, L.; Steinberg, H.

    1985-01-01

    Projections of uranium requirements (both yellowcake and enrichment services) and spent fuel discharges are presented, corresponding to the nuclear power plant capacity projections presented in ''Commercial Nuclear Power 1984: Prospects for the United States and the World'' (DOE/EIA-0438(85)) and the ''Annual Energy Outlook 1984:'' (DOE/EIA-0383(84)). Domestic projections are provided through the year 2020, with foreign projections through 2000. The domestic projections through 1995 are consistent with the integrated energy forecasts in the ''Annual Energy Outlook 1984.'' Projections of capacity beyond 1995 are not part of an integrated energy foreccast; the methodology for their development is explained in ''Commercial Nuclear Power 1984.'' A range of estimates is provided in order to capture the uncertainty inherent in such forward projections. The methodology and assumptions are also stated. A glossary is provided. Two appendixes present additional material. This report is of particular interest to analysts involved in long-term planning for the disposition of radioactive waste generated from the nuclear fuel cycle. 14 figs., 18 tabs

  6. Dissolution experiments of unirradiated uranium dioxide pellets

    International Nuclear Information System (INIS)

    Ollila, K.

    1985-01-01

    The purpose of this study was to measure the dissolution rate of uranium from unirradiated uranium dioxide pellets in deionized water and natural groundwater. Moreover, the solubility limit of uranium in natural groundwater was measured. Two different temperatures, 25 and 60 deg C were used. The low oxygen content of deep groundwater was simulated. The dissolution rate of uranium varied from 10 -7 to 10 -8 g cm -2 d -1 . The rate in reionized water was one order of magnitude lower than in groundwater. No great difference was observed between the natural groundwaters with different composition. Temperature seems to have effect on the dissolution rate. The solubility limit of uranium in natural groundwater in reducing conditions, at 25 deg C, varied from 20 to 600 μg/l and in oxidizing conditions, at 60 deg C, from 4 to 17 mg/l

  7. 75 FR 26165 - Regulation of Fuels and Fuel Additives: Alternative Affirmative Defense Requirements for Ultra...

    Science.gov (United States)

    2010-05-11

    ... Regulation of Fuels and Fuel Additives: Alternative Affirmative Defense Requirements for Ultra-Low Sulfur... refiners, importers, distributors, and retailers of highway diesel fuel the option to use an alternative affirmative defense if the Agency finds highway diesel fuel samples above the specified sulfur standard at...

  8. A review of the rates of reaction of unirradiated uranium in gaseous atmospheres

    International Nuclear Information System (INIS)

    Pearce, R.J.

    1989-10-01

    The review collates available quantitative rate data for the reaction of unirradiated uranium in dry and moist air, steam and carbon dioxide based atmospheres at temperatures ranging from room temperature to above the melting point of uranium. Reactions in nitrogen and carbon monoxide are also considered. The aim of the review is to provide a compilation of base data for the hazard analysis of fault conditions relating to Magnox fuel. (author)

  9. World nuclear fuel cycle requirements, 1988

    International Nuclear Information System (INIS)

    1988-01-01

    This report contains an analysis of the sensitivities of the nuclear fuel cycle projections to different levels and types of projected nuclear capacity, different enrichment tails assays, higher and lower capacity factors, changes in nuclear fuel burnup levels, and other exogenous assumptions. The projections for the United States generally extend through the year 2020, and the (WOCA) World Outside Centrally Planned Economic Areas projections, which include the United States, are provided through 2010. The report also presents annual projections of spent nuclear fuel discharges and inventories of spent fuel. Appendix E includes aggregated domestic spent fuel projections through the year 2020 for the Lower and Upper References cases and through 2037, the last year in which spent fuel is discharged, for the No New Orders case. Annual projections of spent fuel discharges through the year 2037 for individual US reactors in the No New Orders cases are included for the first time in Appendix H. These disaggregated projections are provided at the request of the Department of Energy's Office of Civilian Radioactive Waste Management

  10. 46 CFR 97.15-55 - Requirements for fuel oil.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Requirements for fuel oil. 97.15-55 Section 97.15-55... OPERATIONS Tests, Drills, and Inspections § 97.15-55 Requirements for fuel oil. (a) It shall be the duty of the chief engineer to cause an entry in the log to be made of each supply of fuel oil received on...

  11. 46 CFR 78.17-75 - Requirements for fuel oil.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 3 2010-10-01 2010-10-01 false Requirements for fuel oil. 78.17-75 Section 78.17-75..., Drills, and Inspections § 78.17-75 Requirements for fuel oil. (a) It shall be the duty of the chief engineer to cause an entry in the log be made of each supply of fuel oil received on board, stating the...

  12. Experimental and theoretical requirements for fuel modelling

    International Nuclear Information System (INIS)

    Gatesoupe, J.P.

    1979-01-01

    From a scientific point of view it may be considered that any event in the life of a fuel pin under irradiation should be perfectly well understood and foreseen from that deterministic point of view, the whole behaviour of the pin maybe analysed and dismantled with a specific function for every component part and each component part related to one basic phenomenon which can be independently studied on pure physical grounds. When extracted from the code structure the subroutine is studied for itself by specialists who try to keep as close as possible to the physics involved in the phenomenon; that often leads to an impressive luxury in details and a subsequent need for many unavailable input data. It might seem more secure to follow that approach since it tries to be firmly based on theoretical grounds. One should think so if the phenomenological situation in the pin were less complex than it is. The codes would not be adequate for off-normal operating conditions since for the accidental transient conditions the key-phenomena would not be the same as for steady-state or slow transient conditions. The orientation given to fuel modelling is based on our two main technological constraints which are: no fuel melting; no cladding failure; no excessive cladding deformation. In this context, the only relevant models are those which have a significant influence on the maximum temperatures in the fuel or on the cladding damage hence the selection between key models and irrelevant models which will next be done. A rather pragmatic view is kept on codification with a special focus on a few determinant aspects of fuel behaviour and no attention to models which are nothing but decorative. Fuel modeling is merely considered as a link between experimental knowledge; it serves as a guide for further improvements in fuel design and as so happens to be quite useful. On this basis the main lacks in of fuel behaviour is described. These are mainly concerning: thermal transfer through

  13. Effects of spent nuclear fuel aging on disposal requirements

    International Nuclear Information System (INIS)

    McKee, R.W.; Johnson, K.I.; Huber, H.D.; Bierschbach, M.C.

    1991-10-01

    This paper describes results of a study to analyze the waste management systems effects of extended spent fuel aging on spent fuel disposal requirements. The analysis considers additional spent fuel aging up to a maximum of 50 years relative to the currently planned 2010 repository startup in the United States. As part of the analysis, an equal energy disposition (EED) methodology was developed for determining allowable waste emplacement densities and waste container loading in a geologic repository. Results of this analysis indicate that substantial benefits of spent fuel aging will already have been achieved by a repository startup in 2010 (spent fuel average age will be 28 years). Even so, further significant aging benefits, in terms of reduced emplacement areas and mining requirements and reduced number of waste containers, will continue to accrue for at least another 50 years when the average spent fuel age would be 78 years, if the repository startup is further delayed

  14. Design requirement on KALIMER blanket fuel assembly duct

    International Nuclear Information System (INIS)

    Hwang, Woan; Kang, H. Y.; Nam, C.; Kim, J. O.

    1998-03-01

    This document describes design requirements which are needed for designing the blanket fuel assembly duct of the KALIMER as design guidance. The blanket fuel assembly duct of the KALIMER consists of fuel rods, mounting rail, nosepiece, duct with pad, handling socket with pad. Blanket fuel rod consists of top end plug, bottom end plug with solid ferritic-martensitic steel rod and key way blanket fuel slug, cladding, and wire wrap. In the assembly, the rods are in a triangular pitch array, and the rod bundle is attached to the nosepiece with mounting rails. The bottom end of the assembly duct is formed by a long nosepiece which provides the lower restraint function and the paths for coolant inlet. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. (author). 20 refs., 4 figs

  15. World nuclear capacity and fuel cycle requirements, November 1993

    International Nuclear Information System (INIS)

    1993-01-01

    This analysis report presents the current status and projections of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. Long-term projections of US nuclear capacity, generation, fuel cycle requirements, and spent fuel discharges for three different scenarios through 2030 are provided in support of the Department of Energy's activities pertaining to the Nuclear Waste Policy Act of 1982 (as amended in 1987). The projections of uranium requirements also support the Energy Information Administration's annual report, Domestic Uranium Mining and Milling Industry: Viability Assessment

  16. Safety of Nuclear Fuel Cycle Facilities. Safety Requirements (Arabic Edition)

    International Nuclear Information System (INIS)

    2015-01-01

    This publication covers the broad scope of requirements for fuel cycle facilities that, in light of the experience and present state of technology, must be satisfied to ensure safety for the lifetime of the facility. Topics of specific relevance include aspects of nuclear fuel generation, storage, reprocessing and disposal

  17. Fuel supply shutdown facility interim operational safety requirements

    International Nuclear Information System (INIS)

    Besser, R.L.; Brehm, J.R.; Benecke, M.W.; Remaize, J.A.

    1995-01-01

    These Interim Operational Safety Requirements (IOSR) for the Fuel Supply Shutdown (FSS) facility define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls to ensure safe operation. The IOSRs apply to the fuel material storage buildings in various modes (operation, storage, surveillance)

  18. Background and planning requirements for spent fuel shipments to DOE

    Energy Technology Data Exchange (ETDEWEB)

    Ravenscroft, Norman [Edlow International Company, 1666 Connecticut Avenue, NW, Suite 201, Washington, DC 20009 (United States)

    1996-10-01

    Information is provided on the planning required and the factors that must be included in the planning process for spent fuel shipments to DOE. A summary is also provided on the background concerning renewal of the DOE spent fuel acceptance policy in May 1996. (author)

  19. Fuel utilization experience in Bohunice NPP and regulatory requirements for implementation of progressive fuel management strategies

    Energy Technology Data Exchange (ETDEWEB)

    Patenyi, V [Nuclear Regulatory Authority, Bratislava (Slovakia); Darilek, P; Majercik, J [Vyskumny Ustav Jadrovych Elektrarni, Trnava (Slovakia)

    1994-12-31

    The experience gained in fuel utilization and the basic requirements for fuel licensing in the Slovak NPPs is described. The original project of WWER-440 reactors supposes 3-year fuel cycle with cycle length of about 320 full power days (FPD). Since 1984 it was reduced to 290 FPD. Based on the experience of other countries, a 4-year fuel cycle utilization started in 1987. It is illustrated with data from the Bohunice NPP units. Among 504 fuel assemblies left for the fourth burnup cycle no leakage was observed. The mean burnup achieved in the different units varied from 33.1 to 38.5 Mwd/kg U. The new fuel assemblies used are different from the recent ones in construction, thermohydraulics, water-uranium ratio, enrichment and material design. To meet the safety criteria, regulatory requirements for exploitation of new fuel in WWER-440 were formulated by the Nuclear Regulatory Authority of Slovak Republic. 1 tab., 5 refs.

  20. Safety analysis report for packaging: the unirradiated fuel shipping container

    International Nuclear Information System (INIS)

    Evans, J.H.; Shipley, W.D.; Mouring, R.W.

    1979-09-01

    The container was evaluated analytically to determine its compliance with the applicable regulations governing containers in which radioactive and fissile materials are transported, and the evaluation is the subject of this report. Computational and test procedures were used to determine the structural integrity and thermal behavior of the container relative to the general standards for normal conditions of transport and the standards for hypothetical accident conditions. Results of the evaluation demonstrate that the container is in compliance with the applicable regulations

  1. Fuel Supply Shutdown Facility Interim Operational Safety Requirements

    International Nuclear Information System (INIS)

    BENECKE, M.W.

    2000-01-01

    The Interim Operational Safety Requirements for the Fuel Supply Shutdown (FSS) Facility define acceptable conditions, safe boundaries, bases thereof, and management of administrative controls to ensure safe operation of the facility

  2. Safety of nuclear fuel cycle facilities. Safety requirements

    International Nuclear Information System (INIS)

    2008-01-01

    This publication covers the broad scope of requirements for fuel cycle facilities that, in light of the experience and present state of technology, must be satisfied to ensure safety for the lifetime of the facility. Topics of specific reference include aspects of nuclear fuel generation, storage, reprocessing and disposal. Contents: 1. Introduction; 2. The safety objective, concepts and safety principles; 3. Legal framework and regulatory supervision; 4. The management system and verification of safety; 5. Siting of the facility; 6. Design of the facility; 7. Construction of the facility; 8. Commissioning of the facility; 9. Operation of the facility; 10. Decommissioning of the facility; Appendix I: Requirements specific to uranium fuel fabrication facilities; Appendix II: Requirements specific to mixed oxide fuel fabrication facilities; Appendix III: Requirements specific to conversion facilities and enrichment facilities

  3. Comparison of ring compression testing to three point bend testing for unirradiated ZIRLO cladding

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2015-04-01

    Safe shipment and storage of nuclear reactor discharged fuel requires an understanding of how the fuel may perform under the various conditions that can be encountered. One specific focus of concern is performance during a shipment drop accident. Tests at Savannah River National Laboratory (SRNL) are being performed to characterize the properties of fuel clad relative to a mechanical accident condition such as a container drop. Unirradiated ZIRLO tubing samples have been charged with a range of hydride levels to simulate actual fuel rod levels. Samples of the hydrogen charged tubes were exposed to a radial hydride growth treatment (RHGT) consisting of heating to 400°C, applying initial hoop stresses of 90 to 170 MPa with controlled cooling and producing hydride precipitates. Initial samples have been tested using both a) ring compression test (RCT) which is shown to be sensitive to radial hydride and b) three-point bend tests which are less sensitive to radial hydride effects. Hydrides are generated in Zirconium based fuel cladding as a result of coolant (water) oxidation of the clad, hydrogen release, and a portion of the released (nascent) hydrogen absorbed into the clad and eventually exceeding the hydrogen solubility limit. The orientation of the hydrides relative to the subsequent normal and accident strains has a significant impact on the failure susceptability. In this study the impacts of stress, temperature and hydrogen levels are evaluated in reference to the propensity for hydride reorientation from the circumferential to the radial orientation. In addition the effects of radial hydrides on the Quasi Ductile Brittle Transition Temperature (DBTT) were measured. The results suggest that a) the severity of the radial hydride impact is related to the hydrogen level-peak temperature combination (for example at a peak drying temperature of 400°C; 800 PPM hydrogen has less of an impact/ less radial hydride fraction than 200 PPM hydrogen for the same thermal

  4. Operational requirements of spherical HTR fuel elements and their performance

    International Nuclear Information System (INIS)

    Roellig, K.; Theymann, W.

    1985-01-01

    The German development of spherical fuel elements with coated fuel particles led to a product design which fulfils the operational requirements for all HTR applications with mean gas exit temperatures from 700 deg C (electricity and steam generation) up to 950 deg C (supply of nuclear process heat). In spite of this relatively wide span for a parameter with strong impact on fuel element behaviour, almost identical fuel specifications can be used for the different reactor purposes. For pebble bed reactors with relatively low gas exit temperatures of 700 deg C, the ample design margins of the fuel elements offer the possibility to enlarge the scope of their in-service duties and, simultaneously, to improve fuel cycle economics. This is demonstrated for the HTR-500, an electricity and steam generating 500 MWel eq plant presently proposed as follow-up project to the THTR-300. Due to the low operating temperatures of the HTR-500 core, the fuel can be concentrated in about 70% of the pebbles of the core thus saving fuel cycle costs. Under all design accident conditions fuel temperatures are maintained below 1250 deg C. This allows a significant reduction in the engineered activity barriers outside the primary circuit, in particular for the loss of coolant accident. Furthermore, access to major primary circuit components and the reuse of the fuel elements after any design accident are possible. (author)

  5. Projection of US LWR spent fuel storage requirements

    International Nuclear Information System (INIS)

    Fletcher, J.F.; Cole, B.M.; Purcell, W.L.; Rau, R.G.

    1982-11-01

    The spent fuel storage requirements projection is based on data supplied for each operating or planned nuclear power power plant by the operting utilities. The data supplied by the utilities encompassed details of plant operating history, past records of fuel discharges, current inventories in reactor spent fuel storage pools, and projections of future discharge patterns. Data on storage capacity of storage pools and on characterization of the discharged fuel are also included. The data supplied by the utilities, plus additional data from other appropriate sources, are maintained on a computerized data base by Pacific Northwest Laboratory. The spent fuel requirements projection was based on utility data updated and verified as of December 31, 1981

  6. Coated fuel particles: requirements and status of fabrication technology

    International Nuclear Information System (INIS)

    Huschka, H.; Vygen, P.

    1977-01-01

    Fuel cycle, design, and irradiation performance requirements impose restraints on the fabrication processes. Both kernel and coating fabrication processes are flexible enough to adapt to the needs of the various existing and proposed high-temperature gas-cooled reactors. Extensive experience has demonstrated that fuel kernels with excellent sphericity and uniformity can be produced by wet chemical processes. Similarly experience has shown that the various multilayer coatings can be produced to fully meet design and specification requirements. Quality reliability of coated fuel particles is ensured by quality control and quality assurance programs operated by an aduiting system that includes licensing officials and the customer

  7. TMI-2 fuel canister interface requirements for INEL. Revision 1

    International Nuclear Information System (INIS)

    Wilkins, D.E.; Martz, D.E.; Reno, H.W.

    1984-06-01

    This report focuses on fuel canister interface requirements at INEL which should be incorporated into the canister design criteria. The requirements will ensure compatibility with existing INEL structures and equipment to be used for receipt, unloading, and storage of fuel canisters. INEL can and does receive and store radioactive materials in many different forms, including reactor fuel. INEL requires detailed descriptions of canisters and casks. Therefore, requirements listed represent engineering design features which will simplify the handling and storage operations; consequently, they are not to be viewed as absolute or non-negotiable. However, the core acquisition contract was negotiated with certain storage assumptions which effect costs of storage. Deviations from those assumptions which significantly effect costs would require approval by DOE-Idaho. If some stated requirements are too restrictive, modifications based on sound engineering principles may be negotiated with INEL. 11 figures

  8. Effect of helium pressure on the response of unirradiated UO2 subjected to thermal transients

    International Nuclear Information System (INIS)

    Fenske, G.R.; Chapello, P.M.; Emerson, J.E.; Poeppel, R.B.

    1983-01-01

    The effect of helium pressure on the transient response of unirradiated depleted UO 2 subjected to simulated hypothetical loss-of-flow accidents in a gas-cooled fast reactor was examined by use of the direct electrical heating technique. Transient tests were performed at pressures ranging from 7 to 10 X 10 5 Pa(7 to 10 atm) to 7 to 8 MPa (70 to 80 atm) on radially restrained and unrestrained fuel segments. The average heating rates ranged from about17 to 240 J/g x s. The results indicate that while the mechanical integrity of the fuel segment was independent of the test pressure, the rapid ejection of molten fuel from pellet interfaces of unrestrained fuel, observed at the lower pressures, was delayed or suppressed at the higher pressures

  9. Future requirements for petroleum fuels - an environmental perspective

    International Nuclear Information System (INIS)

    White, R.

    1998-01-01

    The environmental impacts of fuel emissions were discussed. Emissions from petroleum fuels are the largest contributor to a wide range of environmental problems including damage to the ozone layer and risks to human health. Forecasts indicate that future demand for fossil fuels for energy will continue to grow. The transportation sector is the largest single source of air emissions in Canada. The environmental requirements for all fuels will become progressively more stringent. The pollutants of primary concern include toxics, nitrogen oxides, volatile organic compounds, carbon monoxide, sulphur dioxide, and particulates. The U.S. auto-oil research program has conducted considerable research to understand the impact of fuel parameters of vehicle tailpipe emissions. In Canada, lead was removed from Canadian gas a decade ago. Since January 1998, low sulphur diesel (less than 500 ppm) is required for on-road use. Regulations have also been passed to reduce the level of benzene in gasoline to less than one per cent by mid-1999. It will be necessary to manage our fossil fuels to minimize the environmental impacts from combustion. In the longer term, it will be necessary to minimize fossil fuel use through conservation and shift to less polluting fuels

  10. Physical and transportation requirements for a FLIP fueled TRIGA

    International Nuclear Information System (INIS)

    Johnson, A.G.; Ringle, J.C.; Anderson, T.V.

    1977-01-01

    Several major changes to the OSTR Physical Security Plan were required by the NRC prior to the August 1976 receipt and installation of a new core consisting entirely of FLIP fuel. The general nature of these changes will be reviewed along with several decisions we faced during their implementation. At the previous TRIGA Owners' Conference in Salt Lake City, Utah, we reported on Oregon's regulatory program for research reactor emergency response planning and physical security. The latter program was of particular interest to us in light of the projected FLIP fuel shipments. The impact of the State's program for physical security of FLIP fuel during transportation will be presented. (author)

  11. Fusion fuel cycle: material requirements and potential effluents

    International Nuclear Information System (INIS)

    Teofilo, V.L.; Bickford, W.E.; Long, L.W.; Price, B.A.; Mellinger, P.J.; Willingham, C.E.; Young, J.K.

    1980-10-01

    Environmental effluents that may be associated with the fusion fuel cycle are identified. Existing standards for controlling their release are summarized and anticipated regulatory changes are identified. The ability of existing and planned environmental control technology to limit effluent releases to acceptable levels is evaluated. Reference tokamak fusion system concepts are described and the principal materials required of the associated fuel cycle are analyzed. These materials include the fusion fuels deuterium and tritium; helium, which is used as a coolant for both the blanket and superconducting magnets; lithium and beryllium used in the blanket; and niobium used in the magnets. The chemical and physical processes used to prepare these materials are also described

  12. Fusion fuel cycle: material requirements and potential effluents

    Energy Technology Data Exchange (ETDEWEB)

    Teofilo, V.L.; Bickford, W.E.; Long, L.W.; Price, B.A.; Mellinger, P.J.; Willingham, C.E.; Young, J.K.

    1980-10-01

    Environmental effluents that may be associated with the fusion fuel cycle are identified. Existing standards for controlling their release are summarized and anticipated regulatory changes are identified. The ability of existing and planned environmental control technology to limit effluent releases to acceptable levels is evaluated. Reference tokamak fusion system concepts are described and the principal materials required of the associated fuel cycle are analyzed. These materials include the fusion fuels deuterium and tritium; helium, which is used as a coolant for both the blanket and superconducting magnets; lithium and beryllium used in the blanket; and niobium used in the magnets. The chemical and physical processes used to prepare these materials are also described.

  13. Relation of fuel rod service parameters and design requirements to produced fuel rod and their components

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.

    1999-01-01

    Based on the presented material it is possible to state that there is a very close link between the fuel operational parameters and the requirements for its design and production process. The required performance and life-time of a fuel rod can be only assured by the correctly selected design and process solutions. The economical aspect of this problem is significantly depend on the commercial feasibility of a particular selected solution with the provision of an automated and comparative by inexpensive production of a fuel rod and its components. The operational conditions are also important for the life time of the fuel rods. If there are no special measures for the mitigation of the certain operation conditions the leakage of fuel elements can take place before the planned time. (authors)

  14. The spent fuel safety experiment

    International Nuclear Information System (INIS)

    Harmms, G.A.; Davis, F.J.; Ford, J.T.

    1995-01-01

    The Department of Energy is conducting an ongoing investigation of the consequences of taking fuel burnup into account in the design of spent fuel transportation packages. A series of experiments, collectively called the Spent Fuel Safety Experiment (SFSX), has been devised to provide integral benchmarks for testing computer-generated predictions of spent fuel behavior. A set of experiments is planned in which sections of unirradiated fuel rods are interchanged with similar sections of spent PWR fuel rods in a critical assembly. By determining the critical size of the arrays, one can obtain benchmark data for comparison with criticality safety calculations. The integral reactivity worth of the spent fuel can be assessed by comparing the measured delayed critical fuel loading with and without spent fuel. An analytical effort to model the experiments and anticipate the core loadings required to yield the delayed critical conditions runs in parallel with the experimental effort

  15. 40 CFR 80.581 - What are the batch testing and sample retention requirements for motor vehicle diesel fuel, NRLM...

    Science.gov (United States)

    2010-07-01

    ... retention requirements for motor vehicle diesel fuel, NRLM diesel fuel, and ECA marine fuel? 80.581 Section...) REGULATION OF FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel; Nonroad, Locomotive, and Marine Diesel Fuel... requirements for motor vehicle diesel fuel, NRLM diesel fuel, and ECA marine fuel? (a) Beginning on June 1...

  16. 40 CFR 80.608 - What requirements apply to diesel fuel and ECA marine fuel for use in the Territories?

    Science.gov (United States)

    2010-07-01

    ... Motor Vehicle Diesel Fuel; Nonroad, Locomotive, and Marine Diesel Fuel; and ECA Marine Fuel Exemptions... sulfur standards of § 80.520(a)(1) and (c) related to motor vehicle diesel fuel, of § 80.510(a), (b), and... 40 Protection of Environment 16 2010-07-01 2010-07-01 false What requirements apply to diesel fuel...

  17. World nuclear capacity and fuel cycle requirements 1992

    International Nuclear Information System (INIS)

    1992-12-01

    This analysis report presents the current status and projections of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. Long-term projections of US nuclear capacity, generation, fuel cycle requirements, and spent fuel discharges for three different scenarios through 2030 are provided in support of the Department of Energy's activities pertaining to the Nuclear Waste Policy Act of 1982 (as amended in 1987). The projections of uranium requirements also support the Energy Information Administration's annual report, Domestic Uranium Mining and Milling Industry: Viability Assessment for the Lower and Upper Reference case scenarios were obtained from the Office of Integrated Analysis and Forecasting, Energy Information Administration. Most of these projections were developed using the World Integrated Nuclear Evaluation System (WINES) model

  18. Fuels from microalgae: Technology status, potential, and research requirements

    Energy Technology Data Exchange (ETDEWEB)

    Neenan, B.; Feinberg, D.; Hill, A.; McIntosh, R.; Terry, K.

    1986-08-01

    Although numerous options for the production of fuels from microalgae have been proposed, our analysis indicates that only two qualify for extensive development - gasoline and ester fuel. In developing the comparisons that support this conclusion, we have identified the major areas of microalgae production and processing that require extensive development. Technology success requires developing and testing processes that fully utilize the polar and nonpolar lipids produced by microalgae. Process designs used in these analyses were derived from fragmented, preliminary laboratory data. These results must be substantiated and integrated processes proposed, tested, and refined to be able to evaluate the commercial feasibility from microalgae. The production of algal feedstocks for processing to gasoline or ester fuel requires algae of high productivity and high lipid content that efficiently utilize saline waters. Species screening and development suggest that algae can achieve required standards taken individually, but algae that can meet the integrated requirements still elude researchers. Effective development of fuels from microalgae technology requires that R and D be directed toward meeting the integrated standards set out in the analysis. As technology analysts, it is inappropriate for us to dictate how the R and D effort should proceed to meet these standards. We end our role by noting that alternative approaches to meeting the feasibility targets have been identified, and it is now the task of program managers and scientists to choose the appropriate approach to assure the greatest likelihood of realizing a commercially viable technology. 70 refs., 39 figs., 35 tabs.

  19. Thermal conditions and functional requirements for molten fuel containment

    International Nuclear Information System (INIS)

    Kang, C.S.; Torri, A.

    1980-05-01

    This paper discusses the configuration and functional requirements for the molten fuel containment system (MFCS) in the GCFR demonstration plant design. Meltdown conditions following a loss of shutdown cooling (LOSC) accident were studied to define the core debris volume for a realistic meltdown case. Materials and thicknesses of the molten fuel container were defined. Stainless steel was chosen as the sacrificial material and magnesium oxide was chosen as the crucible material. Thermal conditions for an expected quasi-steady state were analyzed. Highlights of the functional requirements which directly affect the MFCS design are discussed

  20. Fuel-cycle financing, capital requirements and sources of funds

    International Nuclear Information System (INIS)

    Manderbach, R.W.

    1977-01-01

    An issue of global importance today is the economic case fro nuclear power and the conservation of precious fossil resources. An important question is whether sufficient financial resources can be attracted to the nuclear industry in order to develop a complete fuel-cycle industry capable of meeting the requirements of a global nuclear power industry. Future growth of the nuclear power industry will depend largely on the timely development of a private competitive industry covering the total fuel cycle. The report of the Edison Electric Institute on Nuclear Fuels Supply estimates that by 1985 initial capital investmentor in the nuclear fuel cycle will total US$15x10 9 and by the year 2000, US$60x10 9 will be required. Although the amount of funding projected is manageable from a global availability standpoint, there is a hesitancy to commit financial resources to certain segments of the fuel cycle, because of the many unresolved problems in connection with the nuclear industry - uncertainty regarding local and international governmental regulations and legislation, environmental and alternative technological considerations coupled with the substantial long-term capital commitments needed in each of the several segments of the processes. Activities associated with the nuclear fuel cycle have unique investment requirements, which are needed in many diverse unrelated fields such as resource development and high technology process. This paper examines sources of capital on a national scale, such as net earnings, depreciation, capital market and public subsidies; and, in the broader context, capital investments in highly industrialized and developing countries. Possible areas of government guarantees and financing; and the situation on financing fuel-cycle projects in the USA and in other countries is also discussed. Comments are included on the money market and investment climate in developing countries, particularly regarding the development of uranium resources

  1. Shield requirement estimation for pin storage room in fuel fabrication plant

    International Nuclear Information System (INIS)

    Shanthi, M.M.; Keshavamurthy, R.S.; Sivashankaran, G.

    2012-01-01

    Fast Reactor Fuel Cycle Facility (FRFCF) is an upcoming project in Kalpakkam. It has the facility to recycle the fuel from PFBR. It is an integrated facility, consists of fuel reprocessing plant, fuel fabrication plant (FFP), core subassembly plant, Reprocessed Uranium plant (RUP) and waste management plant. The spent fuel from PFBR would be reprocessed in fuel reprocessing plant. The reprocessed fuel material would be sent to fuel fabrication plant. The main activity of fuel fabrication plant is the production of MOX fuel pins. The fuel fabrication plant has a fuel pin storage room. The shield requirement for the pin storage room has been estimated by Monte Carlo method. (author)

  2. Molten Salt Fuel Cycle Requirements for ADTT Applications

    International Nuclear Information System (INIS)

    Del Cul, G.D.; Toth, L.M.; Williams, D.F.

    1999-01-01

    The operation of an ADT system with the associated nuclear reactions has a profound effect upon the chemistry of the fuel - especially with regards to container compatibility and the chemical separations that may be required. The container can be protected by maintaining the redox chemistry within a relatively narrow, non-corrosive window. Neutron economy as well as other factors require a sophisticated regime of fission product separations. Neither of these control requirements has been demonstrated on the scale or degree of sophistication necessary to support an ADT device. We review the present situation with respect to fluoride salts, and focus on the critical issues in these areas which must be addressed. One requirement for advancement in this area - a supply of suitable materials - will soon be fulfilled by the remediation of ORNLs Molten Salt Reactor Experiment, and the removal of a total of 11,000 kg of enriched (Li-7 > 99.9%) coolant, flush, and fuel salts

  3. Test requirement for PIE of HANARO irradiated fuel rod

    International Nuclear Information System (INIS)

    Lim, I. C.; Cho, Y. G.

    2000-06-01

    Since the first criticality of HANARO reached in Feb. of 1995, the rod type U 3 Si-A1 fuel imported from AECL has been used. From the under-water fuel inspection which has been conducted since 1997, a ballooning-rupture type abnormality was observed in several fuel rods. In order to find the root cause of this abnormality and to find the resolution, the post irradiation examination(PIE) was proposed as the best way. In this document, the information from the under-water inspection as well as the PIE requirements are described. Based on the information in this document, a detail test plan will be developed by the project team who shall conduct the PIE

  4. Design requirements and performance requirements for reactor fuel recycle manipulator systems

    International Nuclear Information System (INIS)

    Grundmann, J.G.

    1975-01-01

    The development of a new generation of remote handling devices for remote production work in support of reactor fuel recycle systems is discussed. These devices require greater mobility, speed and visual capability than remote handling systems used in research activities. An upgraded manipulator system proposed for a High-Temperature Gas-Cooled Reactor fuel refabrication facility is described. Design and performance criteria for the manipulators, cranes, and TV cameras in the proposed system are enumerated

  5. Fuel cycle financing, capital requirements and sources of funds

    International Nuclear Information System (INIS)

    Manderbach, R.W.

    1977-01-01

    An issue of global importance today is the economic case for nuclear power and the conservation of precious fossil resources. A question important to all of us is can sufficient financial resources be attracted to the nuclear industry in order to develop a complete fuel cycle industry capable of meeting the requirements of a global nuclear power industry. Future growth of the nuclear power industry will depend to a large extent on the timely development of a private competitive industry covering the total fuel cycle. The report of the Edison Electric Institute on Nuclear Fuels Supply estimates that by 1985 initial capital investment in the nuclear fuel cycle will total $15 billion and by the year 2000, $60 billion will be required. Although undoubtedly the amount of funding projected is manageable from a global availability standpoint, there is a hesitancy to commit financial resources to certain segments of the fuel cycle. This is because of the many unresolved problems in connection with the nuclear industry such as uncertainty regarding local and international governmental regulations and legislation, environmental and alternative technological considerations coupled, of course, with the substantial capital long term commitments needed in each of the several segments of the processes. Activities associated with the nuclear fuel cycle have unique investment requirements. Investments are needed in many diverse unrelated fields such as resource development and high technology process some of which are not yet fully commercialized. Sources of capital will be examined on a national scale, such as net earnings, depreciation, capital market and public subsidies. The paper also examines, in the broader context, capital investments in highly industrialized and developing countries as well as discussing the possible areas of Government guarantees and financing. The intensive capital required in certain segments of the cycle, which are to be developed by private

  6. Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements

    International Nuclear Information System (INIS)

    KLEM, M.J.

    2000-01-01

    In 1998, a major change in the technical strategy for managing Multi Canister Overpacks (MCO) while stored within the Canister Storage Building (CSB) occurred. The technical strategy is documented in Baseline Change Request (BCR) No. SNF-98-006, Simplified SNF Project Baseline (MCO Sealing) (FDH 1998). This BCR deleted the hot conditioning process initially adopted for the Spent Nuclear Fuel Project (SNF Project) as documented in WHC-SD-SNF-SP-005, Integrated Process Strategy for K Basins Spent Nuclear Fuel (WHC 199.5). In summary, MCOs containing Spent Nuclear Fuel (SNF) from K Basins would be placed in interim storage following processing through the Cold Vacuum Drying (CVD) facility. With this change, the needs for the Hot Conditioning System (HCS) and inerting/pressure retaining capabilities of the CSB storage tubes and the MCO Handling Machine (MHM) were eliminated. Mechanical seals will be used on the MCOs prior to transport to the CSB. Covers will be welded on the MCOs for the final seal at the CSB. Approval of BCR No. SNF-98-006, imposed the need to review and update the CSB functions and requirements baseline documented herein including changing the document title to ''Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements.'' This revision aligns the functions and requirements baseline with the CSB Simplified SNF Project Baseline (MCO Sealing). This document represents the Canister Storage Building (CSB) Subproject technical baseline. It establishes the functions and requirements baseline for the implementation of the CSB Subproject. The document is organized in eight sections. Sections 1.0 Introduction and 2.0 Overview provide brief introductions to the document and the CSB Subproject. Sections 3.0 Functions, 4.0 Requirements, 5.0 Architecture, and 6.0 Interfaces provide the data described by their titles. Section 7.0 Glossary lists the acronyms and defines the terms used in this document. Section 8.0 References lists the

  7. Nuclear fuel requirements in Egypt up to the year 2000

    International Nuclear Information System (INIS)

    Hammad, F.H.; Ghoneim, M.M.; Abou-Zahra, A.A.

    1976-01-01

    The nuclear fuel demands (as U 3 O 8 ) in Egypt for the nuclear power and nuclear desalination programmes up to the year 2000 were calculated. The recent nuclear power forecasts of the International Atomic Energy Agency (IAEA) and the General Egyptian Electricity Coroporation (GEEC) were used as bases for calculation. The requirements were calculated for light and heavy water reactor (LWR and HWR) strategies. The cumulative uranium requirements for power and desalination programmes were found to be in the range of 19000-25000 tons in case of HWR strategy and 30000-400000 tons in case of LWR strategy. In the year 2000 alone the demands are expected to be 2000-2500 tons in case of HWR strategy and 3200-4000 tons in case of LWR strategy. Intensive uranium exploration coupled with a suitable reactor policy is essential to ensure the availability of these demands and for developing a local nuclear fuel technology

  8. 40 CFR 80.607 - What are the requirements for obtaining an exemption for diesel fuel or ECA marine fuel used for...

    Science.gov (United States)

    2010-07-01

    ... PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel; Nonroad, Locomotive... fuel will be segregated from motor vehicle diesel fuel, NRLM diesel fuel, or ECA marine fuel, as... documents associated with research and development motor vehicle diesel fuel must comply with requirements...

  9. Determination of fuel irradiation parameters. Required accuracies and available methods

    International Nuclear Information System (INIS)

    Mas, P.

    1977-01-01

    This paper reports on the present point of some main methods to determine the nuclear parameters of fuel irradiation in testing reactors (nuclear power, burn up, ...) The different methods (theoretical or experimental) are reviewed: neutron measurements and calculations, gamma scanning, heat balance, ... . The required accuracies are reviewed: they are of 3-5 % on flux, fluences, nuclear power, burn-up, conversion factor. These required accuracies are compared with the real accuracies available which are the present time of order of 5-20 % on these parameters

  10. New requirements for the WWER fuel and their consideration in designing the fuel assemblies

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Ananyev, Y.

    2003-01-01

    In 2001-2002 the base designs of the new generation fuel assemblies for the WWER-440 and WWER-1000 reactors were developed. The ways of their further modernisation were defined. The present report deals with the urgent requirements and how they have been implemented in these designs. The assessment of the efficiency of new designs is carried out on the basis of the existing data of the world market on the cost of: Uranium concentrate; dividing operations; fabrication. It is additionally possible also to take into account the cost of transportation, storage and processing of the irradiated fuel including burial of wastes

  11. Quality standards and requirements for solid recovered fuels: a review.

    Science.gov (United States)

    Flamme, Sabine; Geiping, Julia

    2012-04-01

    The utilization of solid recovered fuels (SRF) for energy recovery has been increasing steadily in recent years, and this development is set to continue. In order to use SRF efficiently, it is necessary to define quality standards and introduce targeted quality assurance measures. SRF can be used both in mono-incineration and in co-incineration systems, for instance in power generation and cement plants; but as quality requirements differ, it is necessary to unambiguously define the term 'solid recovered fuel'. The purpose of this article is to provide an overview of the origin, development and the current status of quality assurance for SRF. The basic principles of quality assurance for SRF are explained with reference to the development of the German RAL Quality Assurance System and in addition specifications that have emerged from European standardization work of CEN/TC 343 are analysed.

  12. 49 CFR Appendix D to Part 238 - Requirements for External Fuel Tanks on Tier I Locomotives

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Requirements for External Fuel Tanks on Tier I..., App. D Appendix D to Part 238—Requirements for External Fuel Tanks on Tier I Locomotives The... properties of the locomotive fuel tank to reduce the risk of fuel spillage to acceptable levels under...

  13. 40 CFR 85.1510 - Maintenance instructions, warranties, emission labeling and fuel economy requirements.

    Science.gov (United States)

    2010-07-01

    ..., emission labeling and fuel economy requirements. 85.1510 Section 85.1510 Protection of Environment..., warranties, emission labeling and fuel economy requirements. The provisions of this section are applicable to... for final admission. (d) Fuel economy labeling. (1) The certificate holder shall affix a fuel economy...

  14. FUEL HANDLING FACILITY BACKUP CENTRAL COMMUNICATIONS ROOM SPACE REQUIREMENTS CALCULATION

    International Nuclear Information System (INIS)

    SZALEWSKI, B.

    2005-01-01

    The purpose of the Fuel Handling Facility Backup Central Communications Room Space Requirements Calculation is to determine a preliminary estimate of the space required to house the backup central communications room in the Fuel Handling Facility (FHF). This room provides backup communications capability to the primary communication systems located in the Central Control Center Facility. This calculation will help guide FHF designers in allocating adequate space for communications system equipment in the FHF. This is a preliminary calculation determining preliminary estimates based on the assumptions listed in Section 4. As such, there are currently no limitations on the use of this preliminary calculation. The calculations contained in this document were developed by Design and Engineering and are intended solely for the use of Design and Engineering in its work regarding the FHF Backup Central Communications Room Space Requirements. Yucca Mountain Project personnel from Design and Engineering should be consulted before the use of the calculations for purposes other than those stated herein or use by individuals other than authorized personnel in Design and Engineering

  15. Bulk Fuel Storage and Delivery Systems Infrastructure Military Construction Requirements for Japan

    National Research Council Canada - National Science Library

    Padgett, Gary

    2000-01-01

    .... Specifically, this audit evaluated requirements for bulk fuel storage facilities at three locations in Japan. We also evaluated the management control program as it relates to the bulk fuel storage military construction requirements validation process.

  16. 40 CFR 80.511 - What are the per-gallon and marker requirements that apply to NRLM diesel fuel, ECA marine fuel...

    Science.gov (United States)

    2010-07-01

    ... PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel; Nonroad, Locomotive, and Marine Diesel Fuel; and ECA Marine Fuel General Information § 80.511 What are the per-gallon and... requirements that apply to NRLM diesel fuel, ECA marine fuel, and heating oil downstream of the refiner or...

  17. Rescue Effects: Irradiated Cells Helped by Unirradiated Bystander Cells

    Science.gov (United States)

    Lam, R. K. K.; Fung, Y. K.; Han, W.; Yu, K. N.

    2015-01-01

    The rescue effect describes the phenomenon where irradiated cells or organisms derive benefits from the feedback signals sent from the bystander unirradiated cells or organisms. An example of the benefit is the mitigation of radiation-induced DNA damages in the irradiated cells. The rescue effect can compromise the efficacy of radioimmunotherapy (RIT) (and actually all radiotherapy). In this paper, the discovery and subsequent confirmation studies on the rescue effect were reviewed. The mechanisms and the chemical messengers responsible for the rescue effect studied to date were summarized. The rescue effect between irradiated and bystander unirradiated zebrafish embryos in vivo sharing the same medium was also described. In the discussion section, the mechanism proposed for the rescue effect involving activation of the nuclear factor κB (NF-κB) pathway was scrutinized. This mechanism could explain the promotion of cellular survival and correct repair of DNA damage, dependence on cyclic adenosine monophosphate (cAMP) and modulation of intracellular reactive oxygen species (ROS) level in irradiated cells. Exploitation of the NF-κB pathway to improve the effectiveness of RIT was proposed. Finally, the possibility of using zebrafish embryos as the model to study the efficacy of RIT in treating solid tumors was also discussed. PMID:25625514

  18. 77 FR 18718 - Petroleum Reduction and Alternative Fuel Consumption Requirements for Federal Fleets

    Science.gov (United States)

    2012-03-28

    ... Petroleum Reduction and Alternative Fuel Consumption Requirements for Federal Fleets AGENCY: Office of...-required reduction in petroleum consumption and increase in alternative fuel consumption for Federal fleets... regulations for a statutorily-required reduction in petroleum consumption and increase in alternative fuel...

  19. 33 CFR 149.655 - What are the requirements for helicopter fueling facilities?

    Science.gov (United States)

    2010-07-01

    ... helicopter fueling facilities? 149.655 Section 149.655 Navigation and Navigable Waters COAST GUARD... EQUIPMENT Design and Equipment Helicopter Fueling Facilities § 149.655 What are the requirements for helicopter fueling facilities? Helicopter fueling facilities must comply with 46 CFR 108.489 or an equivalent...

  20. 40 CFR 80.521 - What are the standards and identification requirements for diesel fuel additives?

    Science.gov (United States)

    2010-07-01

    ... Motor Vehicle Diesel Fuel; Nonroad, Locomotive, and Marine Diesel Fuel; and ECA Marine Fuel Motor... consumer in diesel motor vehicles or nonroad diesel engines. [69 FR 39171, June 29, 2004] ... identification requirements for diesel fuel additives? 80.521 Section 80.521 Protection of Environment...

  1. 40 CFR 600.307-08 - Fuel economy label format requirements.

    Science.gov (United States)

    2010-07-01

    ... to battery electric vehicles, fuel cell vehicles, plug-in hybrid electric vehicles and vehicles... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Fuel economy label format requirements...) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy...

  2. Spent fuel disassembly hardware and other non-fuel bearing components: characterization, disposal cost estimates, and proposed repository acceptance requirements

    Energy Technology Data Exchange (ETDEWEB)

    Luksic, A.T.; McKee, R.W.; Daling, P.M.; Konzek, G.J.; Ludwick, J.D.; Purcell, W.L.

    1986-10-01

    There are two categories of waste considered in this report. The first is the spent fuel disassembly (SFD) hardware. This consists of the hardware remaining after the fuel pins have been removed from the fuel assembly. This includes end fittings, spacer grids, water rods (BWR) or guide tubes (PWR) as appropriate, and assorted springs, fasteners, etc. The second category is other non-fuel-bearing (NFB) components the DOE has agreed to accept for disposal, such as control rods, fuel channels, etc., under Appendix E of the standard utiltiy contract (10 CFR 961). It is estimated that there will be approximately 150 kg of SFD and NFB waste per average metric ton of uranium (MTU) of spent uranium. PWR fuel accounts for approximately two-thirds of the average spent-fuel mass but only 50 kg of the SFD and NFB waste, with most of that being spent fuel disassembly hardware. BWR fuel accounts for one-third of the average spent-fuel mass and the remaining 100 kg of the waste. The relatively large contribution of waste hardware in BWR fuel, will be non-fuel-bearing components, primarily consisting of the fuel channels. Chapters are devoted to a description of spent fuel disassembly hardware and non-fuel assembly components, characterization of activated components, disposal considerations (regulatory requirements, economic analysis, and projected annual waste quantities), and proposed acceptance requirements for spent fuel disassembly hardware and other non-fuel assembly components at a geologic repository. The economic analysis indicates that there is a large incentive for volume reduction.

  3. Reliability and availability requirements analysis for DEMO: fuel cycle system

    International Nuclear Information System (INIS)

    Pinna, T.; Borgognoni, F.

    2015-01-01

    The Demonstration Power Plant (DEMO) will be a fusion reactor prototype designed to demonstrate the capability to produce electrical power in a commercially acceptable way. Two of the key elements of the engineering development of the DEMO reactor are the definitions of reliability and availability requirements (or targets). The availability target for a hypothesized Fuel Cycle has been analysed as a test case. The analysis has been done on the basis of the experience gained in operating existing tokamak fusion reactors and developing the ITER design. Plant Breakdown Structure (PBS) and Functional Breakdown Structure (FBS) related to the DEMO Fuel Cycle and correlations between PBS and FBS have been identified. At first, a set of availability targets has been allocated to the various systems on the basis of their operating, protection and safety functions. 75% and 85% of availability has been allocated to the operating functions of fuelling system and tritium plant respectively. 99% of availability has been allocated to the overall systems in executing their safety functions. The chances of the systems to achieve the allocated targets have then been investigated through a Failure Mode and Effect Analysis and Reliability Block Diagram analysis. The following results have been obtained: 1) the target of 75% for the operations of the fuelling system looks reasonable, while the target of 85% for the operations of the whole tritium plant should be reduced to 80%, even though all the tritium plant systems can individually reach quite high availability targets, over 90% - 95%; 2) all the DEMO Fuel Cycle systems can reach the target of 99% in accomplishing their safety functions. (authors)

  4. Scintillation activity in an unirradiated single crystal of 3-hydroxyxanthine

    International Nuclear Information System (INIS)

    Cooke, D.W.; Jahan, M.S.; Alexander, C. Jr.

    1976-01-01

    A method of growing single crystals (approximately 4mm long) of 3-hydroxyxanthine is described. Observed scintillations occurring in an unirradiated single crystal of this potent oncogen as the temperature is lowered from 300 to 90 K are shown. It was found that these scintillations occur upon heating or cooling and do not diminish in activity as the number of heating and cooling cycles increase. It was found that a short duration u.v. exposure would terminate the scintillation activity and various attempts (such as annealing and pressure changes) to rejuvenate them were unsuccessful. With these observations in mind speculation is made concerning the mechanisms associated with the production of purine N-oxide derivatives. (U.K.)

  5. 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.

  6. 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

  7. 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...

  8. Foundations for the definition of MOX fuel quality requirements

    International Nuclear Information System (INIS)

    Bairiot, H.; Deramaix, P.; Vanderborck, Y.

    1991-01-01

    The quality of uranium-plutonium mixed oxide (MOX) fuel, as of any nuclear fuel, depends on the design optimization and on the fabrication process stability. The design optimization is essentially based on feed-back from irradiation experience through engineering assessment of the results; the stability of the process is necessary to justify minimal uncertainty margins in the fuel design. Since MOX fuel is quite similar to UO 2 fuel, the lessons learned from UO 2 fuels can complement the MOX experimental data base. MOX is however different from UO 2 fuel in some respects, among others: the industrial fabrication scale is a factor 10 lower than for UO 2 fuel, the fuel enrichment process takes place in the manufacturing plant, the radioactivity of Pu imposes handling constraints, Pu ages quite rapidly, altering its isotopic composition during storage, the incorporation of Pu alters the material physics and neutronic characteristics of the fuel. In this perspective, the paper outlines some quality attributes for which MOX fuel may or even must depart form UO 2 fuel. (orig.)

  9. 40 CFR 80.520 - What are the standards and dye requirements for motor vehicle diesel fuel?

    Science.gov (United States)

    2010-07-01

    ... requirements for motor vehicle diesel fuel? 80.520 Section 80.520 Protection of Environment ENVIRONMENTAL... Diesel Fuel; Nonroad, Locomotive, and Marine Diesel Fuel; and ECA Marine Fuel Motor Vehicle Diesel Fuel Standards and Requirements § 80.520 What are the standards and dye requirements for motor vehicle diesel...

  10. Chromosome dosimetry: the influence of culture media on the proliferation of irradiated and unirradiated human lymphocytes

    International Nuclear Information System (INIS)

    Purrott, R.J.; Lloyd, D.C.; Vulpis, N.

    1981-01-01

    The proliferation of phytohaemagglutinin stimulated human lymphocytes in four types of synthetic culture medium has been studied using the fluorescence plus Giemsa staining technique to determine cell cycle status. 48 hour cultures of unirradiated cells containing Ham's F10 or RPMI 1640 media yielded significant numbers of second cycle metaphases. Cultures containing Eagle's MEM or TC 199 media, however, required longer incubation times to produce appreciable numbers of second division cells. Intrinsic differences between donors in the rate of proliferation had little effect on the relative ranking of the media. Radiation induced mitotic delay of about 1 hour per Gray was observed for each medium. The relevance of these results to the accuracy of radiation dose estimation by chromosome aberration analysis is discussed. (author)

  11. Data requirements and maintenance of records for spent fuel management: A review

    International Nuclear Information System (INIS)

    2006-11-01

    Data collection and maintenance are an essential part of activities required in the lifetime management of spent fuel. Key data on spent fuel are required from the earliest phase of any project. To allow informed decisions for spent fuel management to be made, the data need to be maintained throughout the lifetime of spent fuel management including storage, transport, reprocessing or disposal. This publication is intended to provide a state-of-the-art review of spent fuel data management, including what data need to be gathered for the relevant activities in spent fuel management and how to maintain them by the responsible bodies at various stages of the nuclear fuel cycle. It provides some insights on a rational approach to spent fuel data management, considering the common requirements involved in spent fuel management for any Member State. In this regard, the information provided in these sections is mostly generic. After the introductory Section 1 and the Section 2 on data requirements for spent fuel management, Section 3 examines technical parameters that could specify spent fuel characteristics and associated conditions, followed by Section 4 on life cycle management of spent fuel data which includes the maintenance of records and other issues. Finally, some specific examples of the approaches already developed by a number of utilities and national organisations to characterise and track their spent-fuel data are presented in the Annex

  12. Meeting the special requirements of the fuel cycle industry

    International Nuclear Information System (INIS)

    McKenzie, N.C.

    1979-01-01

    The nuclear fuel cycle industry is generally thriving, despite the slow-down in nuclear power plant construction. Turnover is expected to grow from the present Pound4.7 billion to Pound12 billion in 1985 and nearly Pound16 billion in 1988. The emerging fuel cycle companies have large demands for investment and working capital but are not yet financially strong. Utilities also are finding the funding of fuel onerous. Special nuclear fuel finance companies have been formed to help both satisfy their financing needs. A particular problem is the security of loans and much creative thinking is being applied to the use of contracts as collaterals. (U.K.)

  13. PEM fuel cell bipolar plate material requirements for transportation applications

    Energy Technology Data Exchange (ETDEWEB)

    Borup, R.L.; Stroh, K.R.; Vanderborgh, N.E. [Los Alamos National Lab., NM (United States)] [and others

    1996-04-01

    Cost effective bipolar plates are currently under development to help make proton exchange membrane (PEM) fuel cells commercially viable. Bipolar plates separate individual cells of the fuel cell stack, and thus must supply strength, be electrically conductive, provide for thermal control of the fuel stack, be a non-porous materials separating hydrogen and oxygen feed streams, be corrosion resistant, provide gas distribution for the feed streams and meet fuel stack cost targets. Candidate materials include conductive polymers and metal plates with corrosion resistant coatings. Possible metals include aluminium, titanium, iron/stainless steel and nickel.

  14. Fuel cell systems for passenger cars - opportunities and requirements

    Energy Technology Data Exchange (ETDEWEB)

    Tachtler, J. [BMW, Munich (Germany); Bourne, C. [Rover Group, Coventry (United Kingdom)

    1996-12-31

    From the point of view of energy density, handling and economy, present-day motor fuels are superior to all known alternatives. The internal combustion engine powered by them satisfies the requirements of customers to an excellent degree. The search for alternatives can therefore only be justified if emissions can be avoided totally and non-fossil primary energy sources can be used or at least partially our dependence on mineral oil can be reduced. What was long suspected has been increasingly confirmed, not least by developments at BMW: electricity (stored in batteries) and hydrogen offer the best prerequisites for achieving these goals in the long term. These forms of energy can be produced in sufficient quantities and with relatively little effect on the environment. They promise to produce an absolute minimum of pollutants when used in vehicles. Natural gas, which is very similar to hydrogen, and hybrid systems, that would compensate for battery risks, could perform a valuable function in the transitional phase.

  15. The united kingdom's changing requirements for spent fuel storage

    International Nuclear Information System (INIS)

    Hodgson, Z.; Hambley, D.I.; Gregg, R.; Ross, D.N.

    2013-01-01

    The UK is adopting an open fuel cycle, and is necessarily moving to a regime of long term storage of spent fuel, followed by geological disposal once a geological disposal facility (GDF) is available. The earliest GDF receipt date for legacy spent fuel is assumed to be 2075. The UK is set to embark on a programme of new nuclear build to maintain a nuclear energy contribution of 16 GW. Additionally, the UK are considering a significant expansion of nuclear energy in order to meet carbon reduction targets and it is plausible to foresee a scenario where up to 75 GW from nuclear power production could be deployed in the UK by the mid 21. century. Such an expansion, could lead to spent fuel storage and its disposal being a dominant issue for the UK Government, the utilities and the public. If the UK were to transition a closed fuel cycle, then spent fuel storage should become less onerous depending on the timescales. The UK has demonstrated a preference for wet storage of spent fuel on an interim basis. The UK has adopted an approach of centralised storage, but a 16 GW new build programme and any significant expansion of this may push the UK towards distributed spent fuel storage at a number of reactors station sites across the UK

  16. Nuclear fuel licensing requirements: present status and future trends

    International Nuclear Information System (INIS)

    Gantchev, T.; Vitkova, M.; Gorinov, I.; Datcheva, D.; Rashkova, N.

    2001-01-01

    The nuclear fuel licensing process must be directed to establishing of criteria for licensing (fuel safety criteria) and relationship between safety limits, technical specifications and operational conditions. This paper discusses the fuel safety criteria as used by NRC and Russian vendor. A survey on the available fuel behavior, modeling and related computer codes is given with respect to help the licensing process including new safety features of general changes in fuel design and operational conditions. Several types of computer codes that are used in safety analysis are sensitive to fuel-related parameters. The need for further code development and verification has been stated on many occasions: new design elements, such as different cladding materials, higher burnup, different fuel microstructure and use of MOX fuel can affect the performance of these codes. Regulatory inspection practices during operation and refueling in different countries are also shown. Future trends are discussed in particular with regard to the coming high burnup and to new core management schemes

  17. 16 CFR Appendix A to Part 306 - Summary of Labeling Requirements for Biodiesel Fuels

    Science.gov (United States)

    2010-01-01

    ... Biodiesel Fuels A Appendix A to Part 306 Commercial Practices FEDERAL TRADE COMMISSION REGULATIONS UNDER... Part 306—Summary of Labeling Requirements for Biodiesel Fuels (Part 1 of 2) Fuel type Blends of 5 percent or less Blends of more than 5 but not more than 20 percent Header Text Color Biodiesel No label...

  18. 40 CFR 600.307-95 - Fuel economy label format requirements.

    Science.gov (United States)

    2010-07-01

    ... requirements. (a)(1) Fuel economy labels must be: (i) Rectangular in shape with a minimum height of 4.5 inches... equivalent fuel economy. To convert these values into units of miles per 100 cubic feet of natural gas... per 100 cubic feet of natural gas, multiply by 0.823.” may be replaced by the statement “The fuel...

  19. Bulk Fuel Storage Requirements for Maintenance, Repair, and Environmental Projects at Fort Hood, Texas

    National Research Council Canada - National Science Library

    Carros, Deborah

    2000-01-01

    This report is one in a series that addresses the accuracy and reliability of maintenance, repair, environmental, and construction requirements for bulk fuel storage and delivery systems infrastructure...

  20. Fuel handling and storage systems in nuclear power plants

    International Nuclear Information System (INIS)

    1984-01-01

    The scope of this Guide includes the design of handling and storage facilities for fuel assemblies from the receipt of fuel into the nuclear power plant until the fuel departs from that plant. The unirradiated fuel considered in this Guide is assumed not to exhibit any significant level of radiation so that it can be handled without shielding or cooling. This Guide also gives limited consideration to the handling and storage of certain core components. While the general design and safety principles are discussed in Section 2 of this Guide, more specific design requirements for the handling and storage of fuel are given in detailed sections which follow the general design and safety principles. Further useful information is to be found in the IAEA Technical Reports Series No. 189 ''Storage, Handling and Movement of Fuel and Related Components at Nuclear Power Plants'' and No. 198 ''Guide to the Safe Handling of Radioactive Wastes at Nuclear Power Plants''. However, the scope of the Guide does not include consideration of the following: (1) The various reactor physics questions associated with fuel and absorber loading and unloading into the core; (2) The design aspects of preparation of the reactor for fuel loading (such as the removal of the pressure vessel head for a light water reactor) and restoration after loading; (3) The design of shipping casks; (4) Fuel storage of a long-term nature exceeding the design lifetime of the nuclear power plant; (5) Unirradiated fuel containing plutonium

  1. High-Level Functional and Operational Requirements for the Advanced Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Charles Park

    2006-01-01

    This document describes the principal functional and operational requirements for the proposed Advanced Fuel Cycle Facility (AFCF). The AFCF is intended to be the world's foremost facility for nuclear fuel cycle research, technology development, and demonstration. The facility will also support the near-term mission to develop and demonstrate technology in support of fuel cycle needs identified by industry, and the long-term mission to retain and retain U.S. leadership in fuel cycle operations. The AFCF is essential to demonstrate a more proliferation-resistant fuel cycle and make long-term improvements in fuel cycle effectiveness, performance and economy

  2. 40 CFR 80.603 - What are the pre-compliance reporting requirements for NRLM diesel fuel?

    Science.gov (United States)

    2010-07-01

    ... requirements for NRLM diesel fuel? 80.603 Section 80.603 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel; Nonroad, Locomotive, and Marine Diesel Fuel; and ECA Marine Fuel Recordkeeping and Reporting...

  3. Unirradiated UO2 in irradiated zirconium alloy sheathing

    International Nuclear Information System (INIS)

    MacDonald, R.D.; Hardy, D.G.; Hunt, C.E.L.; Scoberg, J.A.

    1979-07-01

    Zircaloy-clad UO 2 fuel elements have defected in power reactors when element power outputs were raised significantly after a long irradiation at low power. We have irradiated fuel elements fabricated from fresh UO 2 pellets and zirconium alloy sheaths previously irradiated without fuel. This gave a fuel element with radiation-damaged low-ductility sheathing but with no fission products in the fuel. The elements were power boosted in-reactor to linear power outputs up to 84 kW/m for two five-day periods. No elements defected despite sheath strains of 0.82 percent at circumferential ridge postions. Half of these elements were subsequently soaked at low power to build up the fission product inventory in the fuel and then power boosted to 63 kW/m for a third time. Two elements defected on this final boost. We conclude that these defects were caused by fission product induced stress-corrosion cracking and that this mechanism plays an importent role in power reactor fuel defects. (auth)

  4. Fuel requirements for experimental devices in MTR reactors. A perturbation model for reactor core analysis

    International Nuclear Information System (INIS)

    Beeckmans de West-Meerbeeck, A.

    1991-01-01

    Irradiation in neutron absorbing devices, requiring high fast neutron fluxes in the core or high thermal fluxes in the reflector and flux traps, lead to higher density fuel and larger core dimensions. A perturbation model of the reactor core helps to estimate the fuel requirements. (orig.)

  5. Spent fuel storage requirements. An update of DOE/RL-85-2

    International Nuclear Information System (INIS)

    1986-10-01

    Utility projections of spent fuel storage capacities indicate that some commercial light water reactors (LWRs) have inadequate capacity to handle projected spent fuel discharges. This report presents estimates of potential near-term requirements for additional LWR spent fuel storage capacity, based on information supplied by utilities operating commercial nuclear power plants. These estimates provide information needed for planning the Department of Energy's (DOE) activities to be carried out under the DOE's Commercial Spent Fuel Management (CSFM) Program, in conjunction with the requirements of the Nuclear Waste Policy Act of 1982. This report is the latest in a series published by the DOE on LWR spent fuel storage requirements. The estimates in this report cover the period from the present through the year 2000. Although the DOE objective is to begin accepting spent fuel for final disposal in 1998, types of fuel and the receipt rates to be shipped are not yet known. Hence, this report makes no assumption regarding such fuel shipments. The report also assesses the possible impacts of increased fuel exposure and spent fuel transshipment on the requirements for additional storage capacity

  6. Analysis on the fuel cycle requirements of the FR systems

    International Nuclear Information System (INIS)

    Maki, Takashi; Horiuchi, Nobutake

    2002-01-01

    The functions of the nuclear fuel cycle amount analysis code, developed in 2001 were extended. This code is a program that calculates the change in characteristics with time of mass balance (for example, the amount of natural uranium demand, plutonium mass balance, environmental load reduction, etc.) in a nuclear fuel cycle, to examine the state of future reactor types or recycling facilities. In 2002, as for this code, calculation functions of reprocessing facilities on plutonium-thermal spent fuels, recovery uranium recycling, and multiple FR concepts were added, and the I/O function was improved according to it. Moreover, benchmark calculation to the extended amount analysis code was performed using the other tool, and it was confirmed that mass balance was calculated appropriately. Furthermore, the mass balance of a few typical FR cycle concepts was calculated in this analysis code, and the feature of each concept was clarified. (author)

  7. EDF requirements for hot cells examinations on irradiated fuel

    International Nuclear Information System (INIS)

    Segura, J.C.; Ducros, G.

    2002-01-01

    The objectives of increasing French Nuclear Power Plants (NPP) availability while lengthening the fuel irradiation cycle and reaching higher burnups lead EDF to carry out on site and hot cell examinations. The data issued from such fuel behaviour monitoring programmes will be used to ascertain that the design criteria are met. Data are also needed for modelling, development and validation. The paper deals quickly with the logistics linked to the selection and transport of fuel rods from NPP to hot cell laboratory. Hot cell PIEs remain a valuable method to obtain data in such fields as PCI (Pellet-Cladding Interaction), internal pressure, FGR (Fission Gas Release), oxide thickness, metallurgical aspects. The paper introduces burnup determination methods, inner pressure evaluation, preparation of samples for further irradiation such as power ramps for PCI and RIA (Reactivity Initiated Accident) testing. The nuclear microprobe of Perre Suee laboratory is also presented. (author)

  8. Analysis of near-term spent fuel transportation hardware requirements and transportation costs

    International Nuclear Information System (INIS)

    Daling, P.M.; Engel, R.L.

    1983-01-01

    A computer model was developed to quantify the transportation hardware requirements and transportation costs associated with shipping spent fuel in the commercial nucler fuel cycle in the near future. Results from this study indicate that alternative spent fuel shipping systems (consolidated or disassembled fuel elements and new casks designed for older fuel) will significantly reduce the transportation hardware requirements and costs for shipping spent fuel in the commercial nuclear fuel cycle, if there is no significant change in their operating/handling characteristics. It was also found that a more modest cost reduction results from increasing the fraction of spent fuel shipped by truck from 25% to 50%. Larger transportation cost reductions could be realized with further increases in the truck shipping fraction. Using the given set of assumptions, it was found that the existing spent fuel cask fleet size is generally adequate to perform the needed transportation services until a fuel reprocessing plant (FRP) begins to receive fuel (assumed in 1987). Once the FRP opens, up to 7 additional truck systems and 16 additional rail systems are required at the reference truck shipping fraction of 25%. For the 50% truck shipping fraction, 17 additional truck systems and 9 additional rail systems are required. If consolidated fuel only is shipped (25% by truck), 5 additional rail casks are required and the current truck cask fleet is more than adequate until at least 1995. Changes in assumptions could affect the results. Transportation costs for a federal interim storage program could total about $25M if the FRP begins receiving fuel in 1987 or about $95M if the FRP is delayed until 1989. This is due to an increased utilization of federal interim storage facility from 350 MTU for the reference scenario to about 750 MTU if reprocessing is delayed by two years

  9. Characterization of un-irradiated and irradiated reactor graphite; Karakterizacija neozracenog i ozracenog reaktorskog grafita

    Energy Technology Data Exchange (ETDEWEB)

    Marinkovic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    This report contains three parts: characterization of Yugoslav nuclear graphite development of methods and obtained results, characterization of un-irradiated and irradiated domestic nuclear graphite; calculation of electrical conductivity changes due to vacancies in the graphite crystal lattice.

  10. Infrastructure Requirements for an Expanded Fuel Ethanol Industry

    Energy Technology Data Exchange (ETDEWEB)

    Reynolds, Robert E. [Downstream Alternatives, Inc., South Bend, IN (United States)

    2002-01-15

    This report provides technical information specifically related to ethanol transportation, distribution, and marketing issues. This report required analysis of the infrastructure requirements for an expanded ethanol industry.

  11. 40 CFR 80.593 - What are the reporting requirements for refiners and importers of motor vehicle diesel fuel...

    Science.gov (United States)

    2010-07-01

    ... for refiners and importers of motor vehicle diesel fuel subject to temporary refiner relief standards... PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel; Nonroad, Locomotive... the reporting requirements for refiners and importers of motor vehicle diesel fuel subject to...

  12. 40 CFR 80.604 - What are the annual reporting requirements for refiners and importers of NRLM diesel fuel?

    Science.gov (United States)

    2010-07-01

    ... Motor Vehicle Diesel Fuel; Nonroad, Locomotive, and Marine Diesel Fuel; and ECA Marine Fuel... requirements for refiners and importers of NRLM diesel fuel? 80.604 Section 80.604 Protection of Environment... importers of NRLM diesel fuel? Beginning with the annual compliance period that begins June 1, 2007, or the...

  13. KWIKPLAN: a computer program for projecting the annual requirements of nuclear fuel cycle operations

    International Nuclear Information System (INIS)

    Salmon, R.; Kee, C.W.

    1977-06-01

    The computer code KWIKPLAN was written to facilitate the calculation of projected nuclear fuel cycle activities. Using given projections of power generation, the code calculates annual requirements for fuel fabrication, fuel reprocessing, uranium mining, and plutonium use and production. The code uses installed capacity projections and mass flow data for six types of reactors to calculate projected fuel cycle activities and inventories. It calculates fissile uranium and plutonium flows and inventories after allowing for an economy with limited reprocessing capacity and a backlog of unreprocessed fuel. All calculations are made on a quarterly basis; printed and punched output of the projected fuel cycle activities are made on an annual basis. Since the punched information is used in another code to determine waste inventories, the code punches a table from which the effective average burnup can be calculated for the fuel being reprocessed

  14. Fueling Requirements for Steady State high butane current fraction discharges

    International Nuclear Information System (INIS)

    R.Raman

    2003-01-01

    The CT injector originally used for injecting CTs into 1T toroidal field discharges in the TdeV tokamak was shipped PPPL from the Affiliated Customs Brokers storage facility in Montreal during November 2002. All components were transported safely, without damage, and are currently in storage at PPPL, waiting for further funding in order to begin advanced fueling experiments on NSTX. The components are currently insured through the University of Washington. Several technical presentations were made to investigate the feasibility of the CT injector installation on NSTX. These technical presentations, attached to this document, were: (1) Motivation for Compact Toroida Injection in NSTX; (2) Assessment of the Engineering Feasibility of Installing CTF-II on NSTX; (3) Assessment of the Cost for CT Installation on NSTX--A Peer Review; and (4) CT Fueling for NSTX FY 04-08 steady-state operation needs

  15. Analysis on the fuel cycle requirements of the FR systems

    International Nuclear Information System (INIS)

    Maki, Takashi; Horiuchi, Nobutake

    2003-01-01

    The functions of the nuclear fuel cycle amount analysis code, developed in 2002 were extended. This code calculates the change in characteristics with time of mass balance (for example, the amount of natural uranium demand, plutonium mass balance, environmental load reduction, etc.) in nuclear fuel cycles, to examine the state of future reactor types or recycling facilities. In 2003, as for this code, calculation functions of automatic adjustment of FR capacity, LWR's recovery minor actinide (MA) recycling, were added, and the I/O function was improved according to it. Moreover, benchmark calculation to the extended amount analysis code was performed using the other tool, and it was confirmed that mass balance was calculated appropriately. Furthermore, the mass balance of a few typical FR cycle concepts was calculated with this analysis code, and the further of each concept was clarified. (author)

  16. Spent fuel storage requirements for nuclear utilities and OCRWM [Office of Civilian Radioactive Waste Management

    International Nuclear Information System (INIS)

    Wood, T.W.

    1990-03-01

    Projected spent fuel generation at US power reactors exceeds estimated aggregate pool storage capacity by approximately 30,000 metric tons of uranium (MTU). Based on the current repository schedule, little of the spent fuel inventory will be disposed of prior to shutdown of existing reactors, and a large additional capacity for surface storage of spent fuel will be required, either at reactors or at a centralized DOE storage site. Allocation of this storage requirement across the utility-DOE interface, and the resulting implications for reactor sites and the performance of the federal waste management system, were studied during the DOE MRS System Study and again subsequent to the reassessment of the repository schedule. Spent fuel logistics and cost results from these analyses will be used in definition of spent fuel storage capacity requirements for the federal system. 9 refs., 8 figs., 1 tab

  17. 10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

    Science.gov (United States)

    2010-01-01

    ... fuel in transit. 73.37 Section 73.37 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Physical Protection of Special Nuclear Material in Transit § 73.37 Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1...

  18. Radioactive materials and nuclear fuel transport requirements in Poland in the light of international regulations

    International Nuclear Information System (INIS)

    Musialowicz, T.

    1977-01-01

    National regulations for the transport of radioactive materials and nuclear fuel in Poland are discussed. Basic transport requirements and regulations, transport experience including transport accidents and emergency service are described. The comparison with international regulations is given

  19. Bulk Fuel Storage and Delivery Systems Infrastructure Military Construction Requirements for Japan

    National Research Council Canada - National Science Library

    Padgett, Gary

    2000-01-01

    This report is one in a series that addresses the accuracy and reliability of maintenance, repair, and environmental and construction requirements for bulk fuel storage and delivery systems infrastructure...

  20. 40 CFR 80.594 - What are the pre-compliance reporting requirements for motor vehicle diesel fuel?

    Science.gov (United States)

    2010-07-01

    ... requirements for motor vehicle diesel fuel? 80.594 Section 80.594 Protection of Environment ENVIRONMENTAL... Requirements § 80.594 What are the pre-compliance reporting requirements for motor vehicle diesel fuel? (a... June 1, 2005, all refiners and importers planning to produce or import motor vehicle diesel fuel...

  1. Spent fuel storage requirements. An update of DOE/RL-83-1

    International Nuclear Information System (INIS)

    1984-05-01

    Spent fuel storage capacities at some commercial light water reactors (LWRs) are inadequate to handle projected spent fuel discharges. This report presents estimates of potential near-term requirements for additional LWR spent fuel storage capacity, based on information voluntarily supplied by utilities operating commercial nuclear power plants. These estimates provide information needed for planning the Department of Energy's (DOE) Federal Interim Storage (FIS) Program and the spent fuel research, development, and demonstration (RD and D) activities to be carried out under the DOE's Commercial Spent Fuel Management (CSFM) Program, in conjunction with the requirements of the Nuclear Waste Policy Act of 1982. This report is the latest in a series published by the DOE on LWR spent fuel storage requirements. Since the planning needs of the CSFM program focus on the near-term management of spent fuel inventories from commercial nuclear power reactors, the estimates in this report cover the ten-year period from the present through 1983. The report also assesses the possible impacts of using various concepts to reduce the requirements for additional storage capacity

  2. Detection of irradiated spice in blend of irradiated and un-irradiated spices using thermoluminescence method

    International Nuclear Information System (INIS)

    Goto, Michiko; Yamazaki, Masao; Sekiguchi, Masayuki; Todoriki, Setsuko; Miyahara, Makoto

    2007-01-01

    Five blended spice sample were prepared by mixing irradiated and un-irradiated black pepper and paprika at different ratios. Blended black pepper containing 2%(w/w) of 5.4 kGy-irradiated black pepper showed no maximum at glow1. Irradiated black pepper samples, mixed to 5 or 10%(w/w), were identified as 'irradiated' or 'partially irradiated' or 'un-irradiated'. All samples with un-irradiated pepper up to 20%(w/w) were identified as irradiated'. In the case 5.0 kGy-irradiated paprika were mixed with un-irradiated paprika up to 5%(w/w), all samples were identified as irradiated'. The glow1 curves of samples, including irradiated paprika at 0.2%(w/w) or higher, exhibited a maximum between 150 and 250degC. The results suggest the existence of different critical mixing ratio for the detection of irradiation among each spices. Temperature range for integration of the TL glow intensity were compared between 70-400degC and approximate 150-250degC, and revealed that the latter temperature range was determined based on the measurement of TLD100. Although TL glow ratio in 150-250degC was lower than that of 70-400degC range, identification of irradiation was not affected. Treatment of un-irradiated black pepper and paprika with ultraviolet rays had no effect on the detection of irradiation. (author)

  3. Temperature dependence of magnetoresistance in neutron-irradiated and unirradiated high resistivity p-type silicon

    International Nuclear Information System (INIS)

    Yildirim, M.; Efeoglu, H.; Abay, B.; Yogurtcu, Y.K.

    1996-01-01

    The temperature dependence of the transverse magnetoresistance in irradiated and unirradiated p-type Si is studied in the range from 120 to 290 K. The magnetoresistance coefficients for the unirradiated left angle 001 right angle and left angle 1 anti 10 right angle samples increases with decreasing sample temperature in the range from 160 to 290 K, however, this behavior is reversed below 160 K. It is proposed that this reversal is due to the double injection effect. The magnetoresistance coefficient for the irradiated left angle 001 right angle sample increases with decreasing sample temperature in the range of 120 to 290 K and is greater than that for the unirradiated left angle 001 right angle sample. This result can be explained by increased scattering due to the increased number of defects produced by irradiation. On the other hand, the magnetoresistance coefficient for the unirradiated left angle 1 anti 10 right angle sample is found to be greater than that of the unirradiated left angle 001 right angle sample. (orig.)

  4. 75 FR 26121 - Regulation of Fuels and Fuel Additives: Alternative Affirmative Defense Requirements for Ultra...

    Science.gov (United States)

    2010-05-11

    ... standard at retail facilities. This alternative defense consists of a comprehensive program of quality... without change and may be made available online at http://www.regulations.gov , including any personal..., import, sell, store or transport ULSD highway diesel fuel to meet the standards specified in the diesel...

  5. 78 FR 23832 - Labeling Requirements for Alternative Fuels and Alternative Fueled Vehicles

    Science.gov (United States)

    2013-04-23

    ... will have the discretion to use the old FTC label if it chooses to omit such information from the EPA... inaccurate and misleading marketing claims. The Commission may consider these and other advertising issues as part of its ongoing review of its ``Guide Concerning Fuel Economy Advertising for New Automobiles,'' 16...

  6. 77 FR 36423 - Labeling Requirements for Alternative Fuels and Alternative Fueled Vehicles

    Science.gov (United States)

    2012-06-19

    ... delivered to the following address: Federal Trade Commission, Office of the Secretary, Room H-113 (Annex N... cell, advanced lean burn, and hybrid motor vehicles) that were added to the definition of ``alternative... legislation (i.e., lean burn, hybrid, and fuel cell vehicles). No comments opposed this approach. Edison...

  7. 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.

  8. Functional recovery in the irradiated kidney following removal of the contralateral unirradiated kidney

    International Nuclear Information System (INIS)

    Robbins, M.E.C.; Hopewell, J.W.; Golding, S.J.

    1986-01-01

    Radiation-induced damage to one kidney in the pig causes a fall in total renal function; this would be recognised and lead to a compensatory response in the unirradiated kidney. The presence of the unirradiated contralateral kidney may effectively prevent the irradiated kidney from expressing any potential for repair and/or recovery of function. If this were true then the question would obviously arise, does the irradiated kidney retain some capacity for recovery? In order to answer this question, the contralateral unirradiated kidney was removed from pigs 26 weeks after the irradiation of the other kidney. The subsequent response of the irradiated kidney to nephrectomy was assessed in terms of the changes in renal size and haemodynamics, i.e. GFR and effective renal plasma flow (ERPF). (Auth.)

  9. An analysis of facial nerve function in irradiated and unirradiated facial nerve grafts

    International Nuclear Information System (INIS)

    Brown, Paul D.; Eshleman, Jeffrey S.; Foote, Robert L.; Strome, Scott E.

    2000-01-01

    Purpose: The effect of high-dose radiation therapy on facial nerve grafts is controversial. Some authors believe radiotherapy is so detrimental to the outcome of facial nerve graft function that dynamic or static slings should be performed instead of facial nerve grafts in all patients who are to receive postoperative radiation therapy. Unfortunately, the facial function achieved with dynamic and static slings is almost always inferior to that after facial nerve grafts. In this retrospective study, we compared facial nerve function in irradiated and unirradiated nerve grafts. Methods and Materials: The medical records of 818 patients with neoplasms involving the parotid gland who received treatment between 1974 and 1997 were reviewed, of whom 66 underwent facial nerve grafting. Fourteen patients who died or had a recurrence less than a year after their facial nerve graft were excluded. The median follow-up for the remaining 52 patients was 10.6 years. Cable nerve grafts were performed in 50 patients and direct anastomoses of the facial nerve in two. Facial nerve function was scored by means of the House-Brackmann (H-B) facial grading system. Twenty-eight of the 52 patients received postoperative radiotherapy. The median time from nerve grafting to start of radiotherapy was 5.1 weeks. The median and mean doses of radiation were 6000 and 6033 cGy, respectively, for the irradiated grafts. One patient received preoperative radiotherapy to a total dose of 5000 cGy in 25 fractions and underwent surgery 1 month after the completion of radiotherapy. This patient was placed, by convention, in the irradiated facial nerve graft cohort. Results: Potential prognostic factors for facial nerve function such as age, gender, extent of surgery at the time of nerve grafting, preoperative facial nerve palsy, duration of preoperative palsy if present, or number of previous operations in the parotid bed were relatively well balanced between irradiated and unirradiated patients. However

  10. Identification and Analysis of Critical Gaps in Nuclear Fuel Cycle Codes Required by the SINEMA Program

    International Nuclear Information System (INIS)

    Miron, Adrian; Valentine, Joshua; Christenson, John; Hawwari, Majd; Bhatt, Santosh; Dunzik-Gougar, Mary Lou; Lineberry, Michael

    2009-01-01

    The current state of the art in nuclear fuel cycle (NFC) modeling is an eclectic mixture of codes with various levels of applicability, flexibility, and availability. In support of the advanced fuel cycle systems analyses, especially those by the Advanced Fuel Cycle Initiative (AFCI), University of Cincinnati in collaboration with Idaho State University carried out a detailed review of the existing codes describing various aspects of the nuclear fuel cycle and identified the research and development needs required for a comprehensive model of the global nuclear energy infrastructure and the associated nuclear fuel cycles. Relevant information obtained on the NFC codes was compiled into a relational database that allows easy access to various codes' properties. Additionally, the research analyzed the gaps in the NFC computer codes with respect to their potential integration into programs that perform comprehensive NFC analysis.

  11. Identification and Analysis of Critical Gaps in Nuclear Fuel Cycle Codes Required by the SINEMA Program

    Energy Technology Data Exchange (ETDEWEB)

    Adrian Miron; Joshua Valentine; John Christenson; Majd Hawwari; Santosh Bhatt; Mary Lou Dunzik-Gougar: Michael Lineberry

    2009-10-01

    The current state of the art in nuclear fuel cycle (NFC) modeling is an eclectic mixture of codes with various levels of applicability, flexibility, and availability. In support of the advanced fuel cycle systems analyses, especially those by the Advanced Fuel Cycle Initiative (AFCI), Unviery of Cincinnati in collaboration with Idaho State University carried out a detailed review of the existing codes describing various aspects of the nuclear fuel cycle and identified the research and development needs required for a comprehensive model of the global nuclear energy infrastructure and the associated nuclear fuel cycles. Relevant information obtained on the NFC codes was compiled into a relational database that allows easy access to various codes' properties. Additionally, the research analyzed the gaps in the NFC computer codes with respect to their potential integration into programs that perform comprehensive NFC analysis.

  12. A method for determining the spent-fuel contribution to transport cask containment requirements

    International Nuclear Information System (INIS)

    Sanders, T.L.; Seager, K.D.; Rashid, Y.R.; Barrett, P.R.; Malinauskas, A.P.; Einziger, R.E.; Jordan, H.; Reardon, P.C.

    1992-11-01

    This report examines containment requirements for spent-fuel transport containers that are transported under normal and hypothetical accident conditions. A methodology is described that estimates the probability of rod failure and the quantity of radioactive material released from breached rods. This methodology characterizes the dynamic environment of the cask and its contents and deterministically models the peak stresses that are induced in spent-fuel cladding by the mechanical and thermal dynamic environments. The peak stresses are evaluated in relation to probabilistic failure criteria for generated or preexisting ductile tearing and material fractures at cracks partially through the wall in fuel rods. Activity concentrations in the cask cavity are predicted from estimates of the fraction of gases, volatiles, and fuel fines that are released when the rod cladding is breached. Containment requirements based on the source term are calculated in terms of maximum permissible volumetric leak rates from the cask. Calculations are included for representative cask designs

  13. Fuel cells for transport: can the promise be fulfilled? Technical requirements and demands from customers

    Science.gov (United States)

    Klaiber, Thomas

    The paper discusses the technical requirements and the customer demands for vehicles that have an on-board methanol reformer and fuel cells. The research concentrates on the technical developmental risks which include minimizing volume, reducing weight and, at the same time, improving efficiency and system dynamics. Fuel cell powered vehicles with methanol reformers are not only suitable for a niche market but also these vehicles will compete with conventional vehicles. The greatest hindrance will be the price of the fuel cell. A possible progressive development of the number of fuel cell powered vehicles in conjunction with a reduction in costs will be discussed in the paper. When fuel cell vehicles come to the market it is necessary that an infrastructure for the fuel methanol or hydrogen is installed. Therefore, it will only be possible to introduce fuel cell vehicles into special markets, e.g. California. Such a process will need to be subsidized by additional incentives like tax concessions. Today there are many technical risks and unsolved problems relating to production technologies, infrastructure, and costs. Nevertheless, among the alternative power units, the fuel cell seems to be the only one that might be competitive to the conventional power unit, especially relating to emissions.

  14. 40 CFR 80.591 - What are the product transfer document requirements for additives to be used in diesel fuel?

    Science.gov (United States)

    2010-07-01

    ... Motor Vehicle Diesel Fuel; Nonroad, Locomotive, and Marine Diesel Fuel; and ECA Marine Fuel... content requirements for use in diesel motor vehicles and nonroad engines.”; or (2) For those additives... requirements for use in model year 2007 and newer diesel motor vehicles or model year 2011 and newer diesel...

  15. 40 CFR 80.510 - What are the standards and marker requirements for NRLM diesel fuel and ECA marine fuel?

    Science.gov (United States)

    2010-07-01

    ... Motor Vehicle Diesel Fuel; Nonroad, Locomotive, and Marine Diesel Fuel; and ECA Marine Fuel General... of marker solvent yellow 124. (2) All motor vehicle and NRLM diesel fuel shall be free of solvent... yellow 124 shall be considered motor vehicle diesel fuel or NRLM diesel fuel, as appropriate. (5) Any...

  16. New concepts, requirements and methods concerning the periodic inspection of the CANDU fuel channels

    International Nuclear Information System (INIS)

    Denis, J.R.

    1995-01-01

    Periodic inspection of fuel channels is essential for a proper assessment of the structural integrity of these vital components of the reactor. The development of wet channel technologies for non-destructive examination (NDE) of pressure tubes and the high technical performance and reliability of the CIGAR equipment have led, in less than 1 0 years, to the accumulation of a very significant volume of data concerning the flaw mechanisms and structural behaviour of the CANDU fuel channels. On this basis, a new form of the CAN/CSA-N285.4 Standard for Periodic Inspection of CANDU Nuclear Power Plant components was elaborated, introducing new concepts and requirements, in accord with the powerful NDE methods now available. This paper presents these concepts and requirements, and discusses the NDE methods, presently used or under development, to satisfy these requirements. Specific features regarding the fuel channel inspections of Cernavoda NGS Unit 1 are also discussed. (author)

  17. In vivo genetic toxicity studies in Chinese hamsters fed irradiated or unirradiated foodstuffs

    International Nuclear Information System (INIS)

    Altmann, H.

    1982-01-01

    Two in vivo genetic toxicity studies were performed in Chinese hamsters fed irradiated or unirradiated diets of chicken, fish or dates in order to detect possible mutagenic effects caused by irradiating these foodstuffs. The tests selected for study were: 1. Chromosomal analysis of bone narrow cells and 2. DNA metabolism in spleen cells. Chicken, fish and dates were irradiated with doses of 7, 2.5 and 1 kGy respectively. These investigations were subsequently extended to include the effects of irradiated dried onions, pulses and cocoa beans on DNA metabolism in Chinese hamster spleen cells only. Dried onions were irradiated with doses of 0.15, 9 and 15 kGy, pulses with 10 kGy and cocoa beans with 3.2 to 5 kGy. In addition, a fumigated cocoa bean group was included. No significant differences in chromosomal aberration rate were detected between groups fed irradiated or unirradiated diets. Dried dates, whether irradiated or not, showed some evidence of genetic toxicity in their effect on DNA metabolism in the spleen cells of Chinese hamsters. Both date diets caused more strand breaks DNA than are usual for Chinese hamster spleen cells, but DNA repair was not adversely affected. Chicken, both irradiated and unirradiated, was found to enhance replicative DNA synthesis but had no effect on the DNA repair process. Irradiated fish, however, caused enhanced DNA synthesis compared to unirradiated fish, but also had no adverse effect on DNA repair. Irradiated white beans also enhanced DNA synthesis compared to controls whereas unirradiated samples inhibited synthesis. (orig./MG)

  18. Uranium requirements for advanced fuel cycles in expanding nuclear power systems

    International Nuclear Information System (INIS)

    Banerjee, S.; Tamm, H.

    1978-01-01

    When considering advanced fuel cycle strategies in rapidly expanding nuclear power systems, equilibrium analyses do not apply. A computer simulation that accounts for system delay times and fissile inventories has been used to study the effects of different fuel cycles and different power growth rates on uranium consumption. The results show that for a given expansion rate of installed capacity, the main factors that affect resource requirements are the fissile inventory needed to introduce the advanced fuel cycle and the conversion (or breeding) ratio. In rapidly expanding systems, the effect of fissile inventory dominates, whereas in slowly expanding systems, conversion or breeding ratio dominates. Heavy-water-moderated and -cooled reactors, with their high conversion ratios, appear to be adaptable vehicles for accommodating fuel cycles covering a wide range of initial fissile inventories. They are therefore particularly suitable for conserving uranium over a wide range of nuclear power system expansion rates

  19. Recommended reactor coolant water chemistry requirements for WWER-1000 units with 235U higher enriched fuel

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.

    2011-01-01

    well as recommended reactor coolant water chemistry requirements for WWER-1000 Units with 235 U higher enriched fuel are presented

  20. Spent fuel storage requirements: the need for away-from-reactor storage

    International Nuclear Information System (INIS)

    1980-01-01

    The analyses of on-site storage capabilities of domestic utilities and estimates of timing and magnitude of away-from-reactor (AFR) storage requirements were presented in the report DOE/ET-0075 entitled Spent Fuel Storage Requirements: The Need For Away-From-Reactor Storage published in February 1979 by the US Department of Energy. Since utility plans and requirements continue to change with time, a need exists to update the AFR requirements estimates as appropriate. This short report updates the results presented in DOE/ET-0075 to reflect recent data on reactor operations and spent fuel storage. In addition to the updates of cases representing the range of AFR requirements in DOE/ET-0075, new cases of interest reflecting utility and regulatory trends are presented

  1. Determination of Light Water Reactor Fuel Burnup with the Isotope Ratio Method

    International Nuclear Information System (INIS)

    Gerlach, David C.; Mitchell, Mark R.; Reid, Bruce D.; Gesh, Christopher J.; Hurley, David E.

    2007-01-01

    For the current project to demonstrate that isotope ratio measurements can be extended to zirconium alloys used in LWR fuel assemblies we report new analyses on irradiated samples obtained from a reactor. Zirconium alloys are used for structural elements of fuel assemblies and for the fuel element cladding. This report covers new measurements done on irradiated and unirradiated zirconium alloys, Unirradiated zircaloy samples serve as reference samples and indicate starting values or natural values for the Ti isotope ratio measured. New measurements of irradiated samples include results for 3 samples provided by AREVA. New results indicate: 1. Titanium isotope ratios were measured again in unirradiated samples to obtain reference or starting values at the same time irradiated samples were analyzed. In particular, 49Ti/48Ti ratios were indistinguishably close to values determined several months earlier and to expected natural values. 2. 49Ti/48Ti ratios were measured in 3 irradiated samples thus far, and demonstrate marked departures from natural or initial ratios, well beyond analytical uncertainty, and the ratios vary with reported fluence values. The irradiated samples appear to have significant surface contamination or radiation damage which required more time for SIMS analyses. 3. Other activated impurity elements still limit the sample size for SIMS analysis of irradiated samples. The sub-samples chosen for SIMS analysis, although smaller than optimal, were still analyzed successfully without violating the conditions of the applicable Radiological Work Permit

  2. Laser pulse heating of nuclear fuels for simulation of reactor power

    Indian Academy of Sciences (India)

    Laser applications; nuclear fuel elements; nuclear safety. ... accident (LOCA) and reactivity initiated accident (RIA), a laser pulse heating system is under ... As a prelude to work on irradiated nuclear fuel specimens, pilot studies on unirradiated ...

  3. Use of fuel cells to meet military requirements for mobile power

    International Nuclear Information System (INIS)

    Andrukaitis, E.

    2004-01-01

    'Full text:' The use of fuel cell technology in military applications will depend on safe, high energy density systems being developed. An important part of using this technology is also the development of alternative hydrogen producing fuels with high energy densities and are easy to transport. Fuel cells are now a very large R and D effort for several military applications around the world. The major reason is because of the high power demands needed requires electrical energy sources that far exceed the capabilities of batteries currently being fielded for portable applications. Fuel cells are regarded as highly efficient, tactical energy converters that can be adapted for wide range of power requirements. They are potentially the lowest weight power source when coupled with batteries or capacitors to form hybrid systems. Generally electrical power is needed to support a number of applications from ultra-high power for electrical pulses (radios, sensors) to reliable, conditioned power for command and control systems. In the future, sustained power for electric drive systems, will also be required. Some of the promising applications in the military and the R and D challenges that remain to reach performance and reliability targets suitable for military requirements will be discussed. (author)

  4. 78 FR 77606 - Security Requirements for Facilities Storing Spent Nuclear Fuel

    Science.gov (United States)

    2013-12-24

    ... NUCLEAR REGULATORY COMMISSION 10 CFR Parts 72 and 73 [NRC-2009-0558] RIN 3150-AI78 Security... rulemaking that would revise the security requirements for storing spent nuclear fuel (SNF) in an independent... Nuclear Security and Incident Response, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001...

  5. Fuel characterization requirements for cofiring biomass in coal-fired boilers

    International Nuclear Information System (INIS)

    Prinzing, D.E.; Tillman, D.A.; Harding, N.S.

    1993-01-01

    The cofiring of biofuels with coal in existing boilers, or the cofiring of biofuels in combined cycle combustion turbine (CCCT) systems presents significant potential benefits to utilities, including reductions in SO 2 and NO x emissions as a function of reducing the mass flow of sulfur and nitrogen to the boiler, reducing CO 2 emissions from the combustion of fossil fuels; potentially reducing fuel costs both by the availability of wood residues and by the fact that biofuels are exempt from the proposed BTU tax; and providing support to industrial customers from the forest products industry. At the same time, cofiring requires careful attention to the characterization of the wood and coal, both singly and in combination. This paper reviews characterization requirements associated with cofiring biofuels and fossil fuels in boilers and CCCT installations with particular attention not only to such concerns as sulfur, nitrogen, moisture, and Btu content, but also to such issues as total ash content, base/acid ratio of the wood ash and the coal ash, alkali metal content in the wood ash and wood fuel (including converted fuels such as low Btu gas or pyrolytic oil), slagging and fouling indices, ash fusion temperature, and trace metal contents in the wood and coal. The importance of each parameter is reviewed, along with potential consequences of a failure to adequately characterize these parameters. The consequences of these parameters are reviewed with attention to firing biofuels with coal in pulverized coal (PC) and cyclone boilers, and firing biofuels with natural gas in CCCT installations

  6. Reassessment of the basis for NRC fuel damage criteria for reactivity transients

    International Nuclear Information System (INIS)

    McCardell, R.K.

    1994-01-01

    The present basis for NRC Fuel Damage Criteria was obtained from experiments performed in the Special Power Excursion Reactor Test (SPERT) IV Reactor Capsule Driver Core (CDC) at the Idaho National Engineering Laboratory (INEL) between 1967 and 1970. Most of the CDC test fuel rods were previously unirradiated and the failure threshold for these unirradiated fuel rods was measured to be about 200 calories per gram of UO 2 radially averaged fuel enthalpy at the axial peak

  7. Criticality safety requirements for transporting EBR-II fuel bottles stored at INTEC

    International Nuclear Information System (INIS)

    Lell, R. M.; Pope, C. L.

    2000-01-01

    Two carrier/shipping cask options are being developed to transport bottles of EBR-II fuel elements stored at INTEC. Some fuel bottles are intact, but some have developed leaks. Reactivity control requirements to maintain subcriticality during the hypothetical transport accident have been examined for both transport options for intact and leaking bottles. Poison rods, poison sleeves, and dummy filler bottles were considered; several possible poison materials and several possible dummy filler materials were studied. The minimum number of poison rods or dummy filler bottles has been determined for each carrier for transport of intact and leaking bottles

  8. Cladding properties under simulated fuel pin transients

    International Nuclear Information System (INIS)

    Hunter, C.W.; Johnson, G.D.

    1975-01-01

    A description is given of the HEDL fuel pin testing program utilizing a recently developed Fuel Cladding Transient Tester (FCTT) to generate the requisite mechanical property information on irradiated and unirradiated fast reactor fuel cladding under temperature ramp conditions. The test procedure is described, and data are presented

  9. Comparative economics for DUCRETE spent fuel storage cask handling, transportation, and capital requirements

    International Nuclear Information System (INIS)

    Powell, F.P.

    1995-04-01

    This report summarizes economic differences between a DUCRETE spent nuclear fuel storage cask and a conventional concrete storage cask in the areas of handling, transportation, and capital requirements. The DUCRETE cask is under evaluation as a new technology that could substantially reduce the overall costs of spent fuel and depleted U disposal. DUCRETE incorporates depleted U in a Portland cement mixture and functions as the cask's primary radiation barrier. The cask system design includes insertion of the US DOE Multi-Purpose Canister inside the DUCRETE cask. The economic comparison is from the time a cask is loaded in a spent fuel pool until it is placed in the repository and includes the utility and overall US system perspectives

  10. Crosscutting Requirements in the International Project on Innovative Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    Steur, Ronald; Lyubenov Yaven, Yanko; Gueorguiev, Boris; Mahadeva, Rao; Shen, Wenquan

    2002-01-01

    There are two categories of requirements: (i) user requirements that need to be met by the designers and manufacturers of innovative reactors and fuel cycles, and (ii) a wide spectrum of requirements that need to be met by countries, willing to successfully deploy innovative nuclear reactors for energy production. This part of the International Project on Innovative Reactors and Fuel Cycles will mainly deal with the second category of requirements. Both categories of requirements will vary depending on the institutional development, infrastructure availability and social attitude in any given country. Out of the need for sustainable development requirements will also more specific in the future. Over a 50-year time frame both categories of requirements will evolve with social and economic development as nuclear technology develops further. For example, the deployment of innovative reactors in countries with marginal or non-existing nuclear infrastructures would be possible only if the reactors are built, owned and operated by an international nuclear utility or if they are inherently safe and can be delivered as a 'black box - nuclear battery'. A number of issues will need to be addressed and conditions and requirements developed if this is going to become a reality. One general requirement for wider utilization of innovative nuclear power will be the public and environmental considerations, which will play a role in the decision making processes. Five main clusters of topics will be handled: - Infra-structural aspects, typology and consequences for nuclear development. - Industrial requirements for the different innovative concepts. - Institutional developments and requirements for future deployment of nuclear energy. (National as well as international) - Socio-political aspects, a.o. public acceptance and role of governments. - Sustainability: requirements following the need for sustainability Analysis will be made of the evolution of national and international

  11. High Temperature Tensile Properties of Unirradiated and Neutron Irradiated 20 Cr-35 Ni Austenitic Steel

    Energy Technology Data Exchange (ETDEWEB)

    Roy, R B; Solly, B

    1966-12-15

    The tensile properties of an unirradiated and neutron irradiated (at 40 deg C) 20 % Cr, 35 % Ni austenitic steel have been studied at 650 deg C, 750 deg C and 820 deg C. The tensile elongation and mode of fracture (transgranular) of unirradiated specimens tested at room temperature and 650 deg C are almost identical. At 750 deg C and 820 deg C the elongation decreases considerably and a large part of the total elongation is non-uniform. Furthermore, the mode of fracture at these temperatures is intergranular and microscopic evidence suggests that fracture is caused by formation and linkup of grain boundary cavities. YS and UTS decrease monotonically with temperature. Irradiated specimens show a further decrease in ductility and an increase in the tendency to grain boundary cracking. Irradiation has no significant effect on the YS, but the UTS are reduced. The embrittlement of the irradiated specimens is attributed to the presence of He and Li atoms produced during irradiation and the possible mechanisms are discussed. Prolonged annealing of irradiated and unirradiated specimens at 650 deg C appears to have no significant effect on tensile properties.

  12. Determining the minimum required uranium carbide content for HTGR UCO fuel kernels

    International Nuclear Information System (INIS)

    McMurray, Jacob W.; Lindemer, Terrence B.; Brown, Nicholas R.; Reif, Tyler J.; Morris, Robert N.; Hunn, John D.

    2017-01-01

    Highlights: • The minimum required uranium carbide content for HTGR UCO fuel kernels is calculated. • More nuclear and chemical factors have been included for more useful predictions. • The effect of transmutation products, like Pu and Np, on the oxygen distribution is included for the first time. - Abstract: Three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from O release when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. In the HTGR kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium apart from UO 2 in the form of a carbide, UC x and this fuel form is designated UCO. Here general oxygen balance formulas were developed for calculating the minimum UC x content to ensure negligible CO formation for 15.5% enriched UCO taken to 16.1% actinide burnup. Required input data were obtained from CALPHAD (CALculation of PHAse Diagrams) chemical thermodynamic models and the Serpent 2 reactor physics and depletion analysis tool. The results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmuted Pu and Np oxides on the oxygen distribution as the fuel kernel composition evolves with burnup.

  13. Fueling requirements of super-high-density plasmas towards innovative ignition regime

    International Nuclear Information System (INIS)

    Sakamoto, Ryuichi; Yamada, Hiroshi

    2014-01-01

    Highlights: • Self-burning scenario with internal diffusion barrier is investigated. • Peaked density profiles allow to sustain self-burning plasma at lower temperature. • Core fueling beyond internal diffusion barrier is essential to sustain peaked density. • Acceptable pellet size becomes small to prevent fusion out perturbation. • Very high velocity pellet injection beyond 10 km/s is inevitable for this scenario. - Abstract: Super-high-density plasma with an internal diffusion barrier which is observed in the Large Helical Device has been extrapolated to a fusion reactor grade plasma to explore an innovative ignition regime and to clarify essential requirements for pellet fueling. The peaked density profiles due to the internal diffusion barrier formation allow reduction in the required minimum temperature to sustain a self-burning plasma down to 10 keV. Direct core fueling beyond the internal diffusion barrier is essential to sustain the peaked density profile. Furthermore, the acceptable pellet size becomes small in terms of fusion output perturbation because the effective volume of the burning plasma becomes small with the peaked density profile. In order to sustain a self-burning plasma with an internal diffusion barrier, therefore, extremely high velocity pellet injection beyond 10 km/s is inevitable unless another solution to the core fueling is found

  14. Users' Requirements for Environmental Effects From Innovative Nuclear Energy Systems and Their Fuel Cycles

    International Nuclear Information System (INIS)

    Carreter, M.; Gray, M.; Falck, E.; Bonne, A.; Bell, M.

    2002-01-01

    The objective of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) is to support the safe, sustainable, economic and proliferation resistant use of nuclear technology to meet the needs of the 21. century. The first part of the project focusses on the development of an understanding of the requirements of possible users of innovative concepts for reactors and fuel cycle applications. This paper reports progress made on the identification of user requirements as they relate to the environment and environmental protection. The user requirements being formulated are intended to limit adverse environmental effects from the different facilities involved in the nuclear fuel cycles to be well below maximum acceptable levels. To determine if the user requirements are met, it is necessary to identify those factors that are relevant to assessment of the environmental performance of innovative nuclear systems. To this effect, Environmental Impact Assessment (EIA) and the Material Flow accounting (MFA) methodologies are being appraised for the suitability for application. This paper develops and provides the rationale for the 'users' requirements' as they are currently defined. Existing Environmental Impact Assessment and Materials Flow Accounting methodologies that can be applied to determine whether or not innovative technologies conform to the User Requirements are briefly described. It is concluded that after establishing fundamental principles, it is possible to formulate sets of general and specific users' requirements against which, the potential adverse environmental effects to be expected from innovative nuclear energy systems (INES) can be assessed. The application of these users' requirements should keep the adverse environmental effects from INES's within acceptable limits. (authors)

  15. An allowable cladding peak temperature for spent nuclear fuels in interim dry storage

    Science.gov (United States)

    Cha, Hyun-Jin; Jang, Ki-Nam; Kim, Kyu-Tae

    2018-01-01

    Allowable cladding peak temperatures for spent fuel cladding integrity in interim dry storage were investigated, considering hydride reorientation and mechanical property degradation behaviors of unirradiated and neutron irradiated Zr-Nb cladding tubes. Cladding tube specimens were heated up to various temperatures and then cooled down under tensile hoop stresses. Cool-down specimens indicate that higher heat-up temperature and larger tensile hoop stress generated larger radial hydride precipitation and smaller tensile strength and plastic hoop strain. Unirradiated specimens generated relatively larger radial hydride precipitation and plastic strain than did neutron irradiated specimens. Assuming a minimum plastic strain requirement of 5% for cladding integrity maintenance in interim dry storage, it is proposed that a cladding peak temperature during the interim dry storage is to keep below 250 °C if cladding tubes are cooled down to room temperature.

  16. System Design Description and Requirements for Modeling the Off-Gas Systems for Fuel Recycling Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Daryl R. Haefner; Jack D. Law; Troy J. Tranter

    2010-08-01

    This document provides descriptions of the off-gases evolved during spent nuclear fuel processing and the systems used to capture the gases of concern. Two reprocessing techniques are discussed, namely aqueous separations and electrochemical (pyrochemical) processing. The unit operations associated with each process are described in enough detail so that computer models to mimic their behavior can be developed. The document also lists the general requirements for the desired computer models.

  17. Spent-fuel-storage requirements: an update of DOE/RL-82-1

    International Nuclear Information System (INIS)

    1983-01-01

    Spent fuel storage requirements as projected through the year 2000 for US light water reactor (LWR) nuclear power plants were calculated using information supplied by the utilities reflecting plant status as of September 30, 1982. Projections through the year 2000 combined fuel discharge projections of the utilities with the assumed discharges of typical reactors required to meet the nuclear capacity of 132 gigawatts electrical (GWe) projected by the Energy Information Administration (EIA) for the year 2000. Three cases were developed and are summarized. A reference case, or maximum at-reactor (AR) capacity case, assumes that all reactor storage pools are increased to their maximum capacities, as estimated by the utilities, for spent fuel storage utilizing currently licensed technologies. Rod consolidation and dry storage technologies were not considered. The reference case assumes no transshipments between pools except as currently licensed by the Nuclear Regulatory Commission (NRC). This case identifies an initial requirement for 13 metric tons uranium (MTU) of additional storage in 1984, and a cumulative requirement for 13,090 MTU additional storage in the year 2000. The reference case is bounded by two alternative cases. One, a current capacity case, assumes that only those pool storage capacity increases currently planned by the operating utilities will occur. The second, or maximum capacity with transshipment case, assumes maximum development of pool storage capacity as described above and also assumes no constraints on transshipment of spent fuel among pools of reactors of like type (BWR) within a given utility. In all cases, a full core discharge capability is assumed to be maintained for each reactor, except that only one FCR is maintained when two reactors share a common pool. 1 figure, 12 tables

  18. Design impacts of safeguards and security requirements for a US MOX fuel fabrication facility

    International Nuclear Information System (INIS)

    Erkkila, B.H.; Rinard, P.M.; Thomas, K.E.; Zack, N.R.; Jaeger, C.D.

    1998-01-01

    The disposition of plutonium that is no longer required for the nation's defense is being structured to mitigate risks associated with the material's availability. In the 1997 Record of Decision, the US Government endorsed a dual-track approach that could employ domestic commercial reactors to effect the disposition of a portion of the plutonium in the form of mixed oxide (MOX) reactor fuels. To support this decision, the Office of Materials Disposition requested preparation of a document that would review US requirements for safeguards and security and describe their impact on the design of a MOX fuel fabrication facility. The intended users are potential bidders for the construction and operation of the facility. The document emphasizes the relevant DOE Orders but also considers the Nuclear Regulatory Commission (NRC) requirements. Where they are significantly different, the authors have highlighted this difference and provided guidance on the impact to the facility design. Finally, the impacts of International Atomic Energy Agency (IAEA) safeguards on facility design are discussed. Security and materials control and accountability issues that influence facility design are emphasized in each area of discussion. This paper will discuss the prepared report and the issues associated with facility design for implementing practical, modern safeguards and security systems into a new MOX fuel fabrication facility

  19. Spent nuclear fuel project multi-canister overpack, additional NRC requirements

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1998-01-01

    The US Department of Energy (DOE), established in the K Basin Spent Nuclear Fuel Project Regulatory Policy, dated August 4, 1995 (hereafter referred to as the Policy), the requirement for new Spent Nuclear Fuel (SNF) Project facilities to achieve nuclear safety equivalency to comparable US Nuclear Regulatory Commission (NRC)-licensed facilities. For activities other than during transport, when the Multi-Canister Overpack (MCO) is used and resides in the Canister Storage Building (CSB), Cold Vacuum Drying (CVD) facility or Hot Conditioning System, additional NRC requirements will also apply to the MCO based on the safety functions it performs and its interfaces with the SNF Project facilities. An evaluation was performed in consideration of the MCO safety functions to identify any additional NRC requirements needed, in combination with the existing and applicable DOE requirements, to establish nuclear safety equivalency for the MCO. The background, basic safety issues and general comparison of NRC and DOE requirements for the SNF Project are presented in WHC-SD-SNF-DB-002

  20. IAEA safety requirements for safety assessment of fuel cycle facilities and activities

    International Nuclear Information System (INIS)

    Jones, G.

    2013-01-01

    The IAEA's Statute authorises the Agency to establish standards of safety for protection of health and minimisation of danger to life and property. In that respect, the IAEA has established a Safety Fundamentals publication which contains ten safety principles for ensuring the protection of workers, the public and the environment from the harmful effects of ionising radiation. A number of these principles require safety assessments to be carried out as a means of evaluating compliance with safety requirements for all nuclear facilities and activities and to determine the measures that need to be taken to ensure safety. The safety assessments are required to be carried out and documented by the organisation responsible for operating the facility or conducting the activity, are to be independently verified and are to be submitted to the regulatory body as part of the licensing or authorisation process. In addition to the principles of the Safety Fundamentals, the IAEA establishes requirements that must be met to ensure the protection of people and the environment and which are governed by the principles in the Safety Fundamentals. The IAEA's Safety Requirements publication 'Safety Assessment for Facilities and Activities', establishes the safety requirements that need to be fulfilled in conducting and maintaining safety assessments for the lifetime of facilities and activities, with specific attention to defence in depth and the requirement for a graded approach to the application of these safety requirements across the wide range of fuel cycle facilities and activities. Requirements for independent verification of the safety assessment that needs to be carried out by the operating organisation, including the requirement for the safety assessment to be periodically reviewed and updated are also covered. For many fuel cycle facilities and activities, environmental impact assessments and non-radiological risk assessments will be required. The

  1. Radiation-resistant requirements analysis of device and control component for advanced spent fuel management process

    Energy Technology Data Exchange (ETDEWEB)

    Song, Tai Gil; Park, G. Y.; Kim, S. Y.; Lee, J. Y.; Kim, S. H.; Yoon, J. S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-02-01

    It is known that high levels of radiation can cause significant damage by altering the properties of materials. A practical understanding of the effects of radiation - how radiation affects various types of materials and components - is required to design equipment to operate reliably in a gamma radiation environment. When designing equipment to operate in a high gamma radiation environment, such as will be present in a nuclear spent fuel handling facility, several important steps should be followed. In order to active test of the advanced spent fuel management process, the radiation-resistant analysis of the device and control component for active test which is concerned about the radiation environment is conducted. Also the system design process is analysis and reviewed. In the foreign literature, 'threshold' values are generally reported. the threshold values are normally the dose required to begin degradation in a particular material property. The radiation effect analysis for the device of vol-oxidation and metalization, which are main device for the advanced spent fuel management process, is performed by the SCALE 4.4 code. 5 refs., 4 figs., 13 tabs. (Author)

  2. Legal aspects concerning the requirement for completion of the backend of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Strassburg, W. Dr.

    1987-01-01

    Most european states are forced by their political institutions, parliaments or governments to establish integrated programmes about the recycling and safe disposal of all kinds of radioactive materials resulting from the application of nuclear energy. A proper solution is not an option as on other continents, but an obligation, due to environmental and economical reasons. First, I will outline the ''Fuel Management Concept'' in the Federal Republic of Germany, than, I will say something about the institutional responsibilities, the package deal clause and the actual practice of management and disposal. We will look into the legal aspects of the atomic energy act, the hazard protection and precaution, the administrative requirements, the discretionary rules of law and some undefined legal conceptions, the practice of public licencing, the advanced public participation and the immediate effectiveness as the substance of my reflections. Finally there shall be some closing remarks on assessing the German fuel management concept

  3. The effect of compressive stress on the Young's modulus of unirradiated and irradiated nuclear graphites

    International Nuclear Information System (INIS)

    Oku, T.; Usui, T.; Ero, M.; Fukuda, Y.

    1977-01-01

    The Young's moduli of unirradiated and high temperature (800 to 1000 0 C) irradiated graphites for HTGR were measured by the ultrasonic method in the direction of applied compressive stress during and after stressing. The Young's moduli of all the tested graphites decreased with increasing compressive stress both during and after stressing. In order to investigate the reason for the decrease in Young's modulus by applying compressive stress, the mercury pore diameter distributions of a part of the unirradiated and irradiated specimens were measured. The change in pore distribution is believed to be associated with structural changes produced by irradiation and compressive stressing. The residual strain, after removing the compressive stress, showed a good correlation with the decrease in Young's modulus caused by the compressive stress. The decrease in Young's modulus by applying compressive stress was considered to be due to the increase in the mobile dislocation density and the growth or formation of cracks. The results suggest, however, that the mechanism giving the larger contribution depends on the brand of graphite, and in anisotropic graphite it depends on the direction of applied stress and the irradiation conditions. (author)

  4. Molecular dynamics study on interfacial thermal conductance of unirradiated and irradiated SiC/C

    International Nuclear Information System (INIS)

    Wang, Qingyu; Wang, Chenglong; Zhang, Yue; Li, Taosheng

    2014-01-01

    SiC f /SiC composite materials have been considered as candidate structural materials for several types of advanced nuclear reactors. Both experimental and computer simulations studies have revealed the degradation of thermal conductivity for this material after irradiation. The objective of this study is to investigate the effect of SiC/graphite interface structure and irradiation on the interfacial thermal conductance by using molecular dynamics simulation. Five SiC/graphite composite models were created with different interface structures, and irradiation was introduced near the interfaces. Thermal conductance was calculated by means of reverse-NEMD method. Results show that there is a positive correlation between the interfacial energy and interfacial C–Si bond quantity, and irradiated models showed higher interfacial energy compared with their unirradiated counterparts. Except the model with graphite atom plane parallel to the interface, the interfacial thermal conductance of unirradiated and irradiated (1000 eV) models, increases as the increase of interfacial energy, respectively. For all irradiated models, lattice defects are of importance in impacting the interfacial thermal conductance depending on the interface structure. For the model with graphite layer parallel to the interface, the interfacial thermal conductance increased after irradiation, for the other models the interfacial thermal conductance decreased. The vibrational density of states of atoms in the interfacial region was calculated to analyze the phonon mismatch at the interface

  5. Thermogravimetric studies of the thermo-oxidative stability of irradiated and unirradiated polyethylene

    International Nuclear Information System (INIS)

    Novakovic, L.; Gal, O.; Markovic, V.; Stannett, V.T.

    1985-01-01

    In part one of this series the effects of a phenolic, an amine and a thioester antioxidant on the thermo-oxidative stability of irradiated and unirradiated low-density polyethylene was reported. In this paper the effects of combined phenolic and thioester stabilizers are described. Isothermal thermogravimetric analysis was used to study the systems. Pronounced synergism was observed with the induction periods, the time when the initial weight loss begins and the 5% weight loss. At about 50% of each stabilizer increases greater than twofold were observed both with the unirradiated and irradiated polymers. The rate constants for oxygen uptake were decreased. However, the rates of degradation at 5% weight loss fell between the values of the two pure stabilizers with no pronounced synergism in either case. In the absence of oxygen little effect of either antioxidant or their mixtures was observed. The corresponding activation energies were somewhat higher, however, with the irradiated samples containing antioxidants. Dynamic thermogravimetry was used for this study. A kinetic analysis indicated that there were somewhat different modes of degradation at lower- and higher-temperature ranges. (author)

  6. Thermogravimetric studies of the thermo-oxidative stability of irradiated and unirradiated polyethylene. 2. Combined antioxidants

    Energy Technology Data Exchange (ETDEWEB)

    Novakovic, L; Gal, O; Markovic, V; Stannett, V T

    1985-01-01

    In part one of this series the effects of a phenolic, an amine and a thioester antioxidant on the thermo-oxidative stability of irradiated and unirradiated low-density polyethylene was reported. In this paper the effects of combined phenolic and thioester stabilizers are described. Isothermal thermogravimetric analysis was used to study the systems. Pronounced synergism was observed with the induction periods, the time when the initial weight loss begins and the 5% weight loss. At about 50% of each stabilizer increases greater than twofold were observed both with the unirradiated and irradiated polymers. The rate constants for oxygen uptake were decreased. However, the rates of degradation at 5% weight loss fell between the values of the two pure stabilizers with no pronounced synergism in either case. In the absence of oxygen little effect of either antioxidant or their mixtures was observed. The corresponding activation energies were somewhat higher, however, with the irradiated samples containing antioxidants. Dynamic thermogravimetry was used for this study. A kinetic analysis indicated that there were somewhat different modes of degradation at lower- and higher-temperature ranges. (author).

  7. Histochemical differentiation between unirradiated and gamma-irradiated tissue in commercial use of some irradiated vegetables

    International Nuclear Information System (INIS)

    Foa, E.

    1978-01-01

    The use of gamma irradiation as a commercial method for the preservation of fruits and vegetables calls for methods of differentiation between unirradiated and irradiated products. A new approach to studying the influence of gamma irradiation on vegetable tissue at the cellular level by histochemical techniques has been developed by the authors and already applied to a number of fruits and vegetables. The possibility of evidencing radiation effects in the polysaccharide components of the cell wall suggested that these detected differences could be applied to differentiate irradiated from unirradiated tomatoes and potatoes. Some work done to determine changes in the cell wall polysaccharides of gamma-irradiated potatoes and tomatoes and to relate these changes to some other factors, such as storage time and vegetable variety, is reported here. While significant differences have been found in the optical densities of the total polysaccharides of the cell wall as a function of irradiation and of the other variables mentioned, it is not yet possible to use these values as a means of reliable differentiation. (author)

  8. Hydrogen in CANDU fuel elements

    International Nuclear Information System (INIS)

    Sejnoha, R.; Manzer, A.M.; Surette, B.A.

    1995-01-01

    Unirradiated and irradiated CANDU fuel cladding was tested to compare the role of stress-corrosion cracking and of hydrogen in the development of fuel defects. The results of the tests are compared with information on fuel performance in-reactor. The role of hydriding (deuteriding) from the coolant and from the fuel element inside is discussed, and the control of 'hydrogen gas' content in the element is confirmed as essential for defect-free fuel performance. Finally, implications for fuel element design are discussed. (author)

  9. Program requirements to determine and relate fuel damage and failure thresholds to anticipated conditions in pressurized water reactors

    International Nuclear Information System (INIS)

    Loyd, R.F.; Croucher, D.W.

    1980-03-01

    Anticipated transients, licensing criteria, and damage mechanisms for PWR fuel rods are reviewed. Potential mechanistic fuel rod damage limits for PWRs are discussed. An expermental program to be conducted out-of-pile and in the Engineering Test Reactor (ETR) to generate a safety data base to define mechanistic fuel damage and failure thresholds and to relate these thresholds to the thermal-hydraulic and power conditions in a PWR is proposed. The requirements for performing the tests are outlined. Analytical support requirements are defined

  10. User requirements in the area of safety of innovative nuclear reactors and fuel cycle installations

    International Nuclear Information System (INIS)

    Kuczera, B.; Juhn, P.E.; Fukuda, K.; )

    2002-01-01

    Full text: Against the background of already existing IAEA and INSAC publications in the area of safety, in the framework of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) a set of user requirements for the safety of future nuclear installations has been established. Five top-level requirements are expected to apply to any type of innovative design. They should foster an increased level of safety that is transparent to and fully accepted by the general public. The approach to future reactor safety includes two complementary strategies: increased emphasis on inherent safety characteristics and enhancement of defense in depth. As compared to existing plants, the effectiveness of preventing measures should be highly enhanced, resulting in fewer mitigation measures. The targets and possible approaches of each of the five levels of defense developed for innovative reactor designs are outlined in the paper

  11. 40 CFR 600.006-87 - Data and information requirements for fuel economy vehicles.

    Science.gov (United States)

    2010-07-01

    ..., motor controller, battery configuration, or other components performed within 2,000 miles prior to fuel... fuel economy vehicles. 600.006-87 Section 600.006-87 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel...

  12. 40 CFR 600.006-08 - Data and information requirements for fuel economy vehicles.

    Science.gov (United States)

    2010-07-01

    ... controller, battery configuration, or other components performed within 2,000 miles prior to fuel economy... fuel economy vehicles. 600.006-08 Section 600.006-08 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel...

  13. In-core power sharing and fuel requirement study for a decommissioning Boiling Water Reactor using the linear reactivity model

    International Nuclear Information System (INIS)

    Chen, Chung-Yuan; Tung, Wu-Hsiung; Yaur, Shung-Jung; Kuo, Weng-Sheng

    2014-01-01

    Highlights: • Linear reactivity model (LRM) was modified and applied to Boiling Water Reactor. • The power sharing and fuel requirement study of the last cycle and two cycles before decommissioning was implemented. • The loading pattern design concept for the cycles before decommissioning is carried out. - Abstract: A study of in-core power sharing and fuel requirement for a decommissioning BWR (Boiling Water Reactor) was carried out using the linear reactivity model (LRM). The power sharing of each fuel batch was taken as an independent variable, and the related parameters were set and modified to simulate actual cases. Optimizations of the last cycle and two cycles before decommissioning were both implemented; in the last-one-cycle optimization, a single cycle optimization was carried out with different upper limits of fuel batch power, whereas, in the two-cycle optimization, two cycles were optimized with different cycle lengths, along with two different optimization approaches which are the simultaneous optimization of two cycles (MO) and two successive single-cycle optimizations (SO). The results of the last-one-cycle optimization show that it is better to increase the fresh fuel power and decrease the thrice-burnt fuel power as much as possible. It also shows that relaxing the power limit is good to the fresh fuel requirement which will be reduced under lower power limit. On the other hand, the results of the last-two-cycle (cycle N-1 and N) optimization show that the MO is better than SO, and the power of fresh fuel batch should be decreased in cycle N-1 to save its energy for the next cycle. The results of the single-cycle optimization are found to be the same as that in cycle N of the multi-cycle optimization. Besides that, under the same total energy requirement of two cycles, a long-short distribution of cycle length design can save more fresh fuel

  14. Thermal studies on unirradiated and γ-irradiated polymer of allyl diglycol carbonate

    International Nuclear Information System (INIS)

    Kalsi, P.C.; Pandey, A.K.; Iyer, R.H.; Singh Mudher, K.D.

    1995-01-01

    The thermal decomposition of unirradiated and γ-irradiated (5.93-15.5 MRad dose range) allyl diglycol carbonate polymer (trade name, CR-39) was studied by thermogravimetry (TG) and differential thermal analysis (DTA). These studies indicate four main decomposition steps in CR-39 polymer in air. Assessment of the influence of radiation dose on the above range shows that while the 5.93 MRad γ-irradiated polymer CR-39 degrades in three steps, the 15.5 MRad γ-irradiated polymer degrades in only two steps. The kinetics of the different stages of degradation were also evaluated from the TG curves. Irradiation enhances the decomposition rate and the effect increases further with increasing radiation dose. The activation energy values calculated for all the decomposition stages decrease on irradiation

  15. Reactor Materials Program electrochemical potential measurements by ORNL with unirradiated and irradiated stainless steel specimens

    International Nuclear Information System (INIS)

    Baumann, E.W.; Caskey, G.R. Jr.

    1993-07-01

    Effect of irradiation of stainless steel on electrochemical potential (ECP) was investigated by measurements in dilute HNO 3 and H 2 O 2 solutions, conditions simulating reactor moderator. The electrodes were made from unirradiated/irradiated, unsensitized/sensitized specimens from R-reactor piping. Results were inconclusive because of budgetary restrictions. The dose rate may have been too small to produce a significant radiolytic effect. Neither the earlier CERT corrosion susceptibility tests nor the present ECP measurements showed a pronounced effect of irradiation on susceptibility of the stainless steel to IGSCC; this is confirmed by the absence in the stainless steel of the SRS reactor tanks (except for the C Reactor tank knuckle area)

  16. Bakery products from irradiated and unirradiated eggs - detection of irradiation in a processed food

    International Nuclear Information System (INIS)

    Pfordt, J.; Grabowski, H.U. von

    1995-01-01

    The detection of radiation-specific degradation products in fat has become an established method which has successfully been applied to egg products. This study is making evident the detectability of irradiated eggs as an ingredient of specified processed foods. Tart layers were produced from both irradiated and non-irradiated liquid whole egg. When the fat components were isolated from the tart layers and investigated by GC/MS, the presence of irradiated eggs could clearly be shown. While the radiation-induced hydrocarbons and 2-alkylcyclobutanones could not be found in unirradiated samples, tart layers from irradiated eggs contained these substances. Especially for the hydrocarbons a satisfying correlation between radiation dose and concentration could be observed. The concentrations of radiation-induced compounds were generally lower in the tart layers than in the liquid egg samples they had been produced from. (orig.) [de

  17. Evaluation of staphylococcus aureus growth in unirradiated and irradiated cured meats using the Gompertz equation

    International Nuclear Information System (INIS)

    Szczawinski, J.; Szczawinska, M.

    1993-01-01

    1. Sodium nitrite, in concentration of 156 mg/kg, exerted little (from practical point of view) inhibitory effect on S. aureus in cured meat. Its addition caused lag time extension, increase of exponential growth rate, decrease of generation time and maximum population density of staphylococci in samples of cured meat stored at 20 C. 2. Growth curves of Staphylococcus aureus in unirradiated and irradiated (50 kGy) meats cured without sodium nitrite were similar. 3. Irradiation of meat cured with 156 mg/kg sodium nitrite with increasing doses of ionizing radiation (0, 10, 30, 50 kGy) progressively decreased growth rates and lag time and increased generation time and maximum population densities of Staphylococcus aureus in samples inoculated after radiation treatment and incubated at 20 C. (orig.)

  18. Reactor Materials Program electrochemical potential measurements by ORNL with unirradiated and irradiated stainless steel specimens

    Energy Technology Data Exchange (ETDEWEB)

    Baumann, E.W.; Caskey, G.R. Jr.

    1993-07-01

    Effect of irradiation of stainless steel on electrochemical potential (ECP) was investigated by measurements in dilute HNO{sub 3} and H{sub 2}O{sub 2} solutions, conditions simulating reactor moderator. The electrodes were made from unirradiated/irradiated, unsensitized/sensitized specimens from R-reactor piping. Results were inconclusive because of budgetary restrictions. The dose rate may have been too small to produce a significant radiolytic effect. Neither the earlier CERT corrosion susceptibility tests nor the present ECP measurements showed a pronounced effect of irradiation on susceptibility of the stainless steel to IGSCC; this is confirmed by the absence in the stainless steel of the SRS reactor tanks (except for the C Reactor tank knuckle area).

  19. Erythrokinetics in mice bearing tumours in either preirradiated or unirradiated tissue

    International Nuclear Information System (INIS)

    Jirtle, R.L.; Clifton, K.H.

    1978-01-01

    Experiments were designed to clarify the causes of anaemia in hosts bearing tumours in either unirradiated or preirradiated tissue. Isotopic methods are described which enable the estimation of erythrocyte destruction and production rates, and the potential red cell life spans in tumour-bearing animals. In this experimental system, anaemia (a) is in large part due to accelerated random erythrocyte loss, (b) is exacerbated as tumours grow by a progressive reduction in the potential erythrocyte life span due to intrinsic erythrocyte defects. (c) is accompanied by an increase in erythrocyte production of six- to ten-fold and (d) is postponed in onset and decreased in magnitude by preirradiation of the tumour transplant site. (author)

  20. Kinetics of thermal decomposition of γ-irradiated and unirradiated complexes of mandelhydroxamic acid

    International Nuclear Information System (INIS)

    Hassan, R.M.; Farid, T.; El-Bellihi, A.A.

    1992-01-01

    The thermal decomposition of γ-irradiated and unirradiated complexes of mandelhydroxamic acid (HMA), Co (HMA) 2 .1/2H 2 O, Mn (HMA) 2 .2H 2 O, Ba (HMA) 2 .2H 2 O and Cd (HMA) 2 .2H 2 O have been studied thermogravimetrically (under isothermal conditions). The thermal dehydration of each complex occured in one step, while the decomposition of dehydrated complexes occured in two steps. The kinetic parameters for dehydration were computed by different models. The thermal dehydration is regulated by random nucleation A 3 for Co-, Mn-, and Cd-complexes and by phase-boundary (R 3 ) for Ba-complex. The effect of γ-irradiation on the kinetic parameters of thermal decomposition is discussed. Radiation did not modify the mechanism of the reaction but accelarated the dehydration steps in the case of Mn- and Co-complexes. (author) 7 refs.; 1 fig.; 4 tabs

  1. Overview and Critical Assessment of the Tensile Properties of unirradiated and irradiated EUROFER97

    Energy Technology Data Exchange (ETDEWEB)

    Lucon, E.; Vandermeulen, W.

    2007-10-15

    Material research represents a crucial issue for the assessment of fusion as a future viable source of energy. Structural materials, in particular, need to show a superior mechanical and chemical behaviour to guarantee the safe operation of the reactor during its whole lifetime, while retaining low activation characteristics to minimise the environmental impact of the produced waste. In this context, specific efforts have been focused for the last twenty years in Europe, Japan and the US, on developing suitable Reduced Activation Ferritic Martensitic (RAFM) steels as candidate structural materials. EUROFER97 has recently emerged in Europe as the reference material for the DEMO design. In the framework of the Long-Term Programme of EFDA (European Fusion Development Agreement), a coordinated effort has been launched aimed at providing a critical assessment of the mechanical and microstructural properties of EUROFER97 in the unirradiated and irradiated conditions, based on the results accumulated since the late 90ies within numerous EFDA tasks.

  2. Erythrokinetics in mice bearing tumours in either preirradiated or unirradiated tissue

    Energy Technology Data Exchange (ETDEWEB)

    Jirtle, R L; Clifton, K H [Wisconsin Univ., Madison (USA). Div. of Clinical Oncology

    1978-11-01

    Experiments were designed to clarify the causes of anaemia in hosts bearing tumours in either unirradiated or preirradiated tissue. Isotopic methods are described which enable the estimation of erythrocyte destruction and production rates, and the potential red cell life spans in tumour-bearing animals. In this experimental system, anaemia (a) is in large part due to accelerated random erythrocyte loss, (b) is exacerbated as tumours grow by a progressive reduction in the potential erythrocyte life span due to intrinsic erythrocyte defects. (c) is accompanied by an increase in erythrocyte production of six- to ten-fold and (d) is postponed in onset and decreased in magnitude by preirradiation of the tumour transplant site.

  3. Industry requirements for introduction of alternative energies with emphasis on hydrogen fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Delabbio, F. [Rio Tinto, Canadian Exploration Ltd., Toronto, ON (Canada); Starbuck, D. [Newmont Mining Corp., Denver, CO (United States); Akerman, A. [CVRD-Inco, Toronto, ON (Canada); Betournay, M.C. [Natural Resources Canada, Ottawa, ON (Canada). CANMET Mining and Mineral Sciences Laboratories

    2007-07-01

    This paper discussed issues related to the use of alternate sources of energy in underground mining applications. Hydrogen power systems were examined in relation to operational drivers, available commercial supplies, site supplies, health and safety issues, capital and operating costs, mine production, and the role of government. Hydrogen power systems are being considered for mining applications in an effort to reduce greenhouse gas (GHG) emissions and reduce cooling and ventilation requirements. This article examined a range of issues that must be addressed before alternate energy systems such as hydrogen fuel cell technology can be used in larger-scale underground mining applications. The mining industry supports the development of new technologies. However, the introduction of alternate energy technologies must proceed in steps which include proof of concept testing, the development of generic infrastructure, power systems and regulations, and whole operating system studies. 13 refs., 1 fig.

  4. Prediction of requirements on labor force in the fuel and power generation sector

    International Nuclear Information System (INIS)

    Kaveckova, R.

    1990-01-01

    One of the aspects of socio-economic assessment of development is quantification of the expected requirements on the number of personnel. Predictions are discussed for the period before the year 2005 for solid fuel mining and treatment, gas production and bitumen mining, power and heat generation and also for the production of electricity and heat by nuclear power plants. They are based on an analysis of past development and the present state, on presumed implementation of various concept variants, on the type structure of nuclear power plants, on the rules of the electric power supply system, and also on foreign materials. It is expected that in 2005, nuclear power will employ 15,654 personnel. (M.D.). 4 tabs., 16 refs

  5. Sweden's radiation protection regulations for spent fuel and nuclear waste: Requirements and compliance

    International Nuclear Information System (INIS)

    Norden, M.; Jensen, M.; Larsson, C.M.; Avila, R.; Bergman, S.S.; Wiebert, A.; Wiklund, A.

    2000-01-01

    The Swedish regulations on radiation protection in connection with spent fuel and nuclear waste disposal concern protection of human health and the environment. The reasoning behind the regulations is in observance with the Rio declaration, in the sense that they take into consideration sustainable development also in continued presence of multiple sources of radioactive effluents. Optimisation and best available technique are used as methods for risk reduction. For human health, a risk concept is used, whereas for environmental protection, focus is set on protection of biological resources and diversity. Compliance with the health and environmental goals is discussed using generic definition of the environment. The hypothetical outflow from a repository takes place in the different compartments and the resulting spread in doses are discussed and compared to the requirements of the individual dose standard, and other environmental effects are assessed. (author)

  6. Requirements for near-real-time accounting of strategic nuclear materials in nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Hakkila, E.A.; Cobb, D.D.; Dietz, R.J.; Shipley, J.P.; Smith, D.B.

    1978-01-01

    A Purex-based nuclear fuel reprocessing plant has been studied for possible incorporation of near-real-time accounting to supplement conventional accounting procedures. Near-real-time accounting of special nuclear materials relies on in-line or at-line flow measurements and plutonium assay of product and waste streams, complemented by conventional analytical chemistry for daily instrument calibrations. In-line alpha monitors could be used for waste stream measurements of plutonium, even in the presence of high beta-gamma fluxes from fission products. X-ray absorption edge densitometry using either K- or L-absorption edges could be used for plutonium concentration measurements in main product streams. Some problem areas identified in waste stream measurements include measurements of leached hulls and of centrifuge sludge. Conventional analytical chemical methods for measuring plutonium in weapons grade material can be modified for reprocessed plutonium. Analytical techniques requiring special precautions will be reviewed

  7. Development of a fresh MOX fuel transport package for disposition of weapons plutonium

    International Nuclear Information System (INIS)

    Ludwig, S.B.; Pope, R.B.; Shappert, L.B.; Michelhaugh, R.D.; Chae, S.M.

    1998-01-01

    The US Department of Energy announced its Record of Decision on January 14, 1997, to embark on a dual-track approach for disposition of surplus weapons-usable plutonium using immobilization in glass or ceramics and burning plutonium as mixed-oxide (MOX) fuel in reactors. In support of the MOX fuel alternative, Oak Ridge National Laboratory initiated development of conceptual designs for a new package for transporting fresh (unirradiated) MOX fuel assemblies between the MOX fabrication facility and existing commercial light-water reactors in the US. This paper summarizes progress made in development of new MOX transport package conceptual designs. The development effort has included documentation of programmatic and technical requirements for the new package and development and analysis of conceptual designs that satisfy these requirements

  8. Fission gas induced fuel swelling in low and medium burnup fuel during high temperature transients

    International Nuclear Information System (INIS)

    Vinjamuri, K.

    1980-01-01

    The behavior of light water reactor fuel elements under postulated accident conditions is being studied by the EG and G Idaho, Inc., Thermal Fuels Behavior Program for the Nuclear Regulatory Commission. As a part of this program, unirradiated and previously irradiated, pressurized-water-reactor type fuel rods were tested under power-cooling-mismatch (PCM) conditions in the Power Burst Facility (PBF). During these integral in-reactor experiments, film boiling was produced on the fuel rods which created high fuel and cladding temperatures. Fuel rod diameters increased in the film boiling region to a greater extent for irradiated rods than for unirradiated rods. The purpose of the study was to investigate and assess the fuel swelling which caused the fuel rod diameter increases and to evaluate the ability of an analytical code, the Gas Release and Swelling Subroutine - Steady-State and Transient (GRASS-SST), to predict the results

  9. Fission gas induced fuel swelling in low and medium burnup fuel during high temperature transients. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Vinjamuri, K.

    1980-01-01

    The behavior of light water reactor fuel elements under postulated accident conditions is being studied by the EG and G Idaho, Inc., Thermal Fuels Behavior Program for the Nuclear Regulatory Commission. As a part of this program, unirradiated and previously irradiated, pressurized-water-reactor type fuel rods were tested under power-cooling-mismatch (PCM) conditions in the Power Burst Facility (PBF). During these integral in-reactor experiments, film boiling was produced on the fuel rods which created high fuel and cladding temperatures. Fuel rod diameters increased in the film boiling region to a greater extent for irradiated rods than for unirradiated rods. The purpose of the study was to investigate and assess the fuel swelling which caused the fuel rod diameter increases and to evaluate the ability of an analytical code, the Gas Release and Swelling Subroutine - Steady-State and Transient (GRASS-SST), to predict the results.

  10. Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

    Science.gov (United States)

    Vislov, I. S.; Pischulin, V. P.; Kladiev, S. N.; Slobodyan, S. M.

    2016-08-01

    The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium-plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel

  11. Environmental Assessment: Relocation and storage of TRIGA reg-sign reactor fuel, Hanford Site, Richland, Washington

    International Nuclear Information System (INIS)

    1995-08-01

    In order to allow the shutdown of the Hanford 308 Building in the 300 Area, it is proposed to relocate fuel assemblies (101 irradiated, three unirradiated) from the Mark I TRIGA Reactor storage pool. The irradiated fuel assemblies would be stored in casks in the Interim Storage Area in the Hanford 400 Area; the three unirradiated ones would be transferred to another TRIGA reactor. The relocation is not expected to change the offsite exposure from all Hanford Site 300 and 400 Area operations

  12. 40 CFR Table 6 to Subpart Ddddd of... - Fuel Analysis Requirements

    Science.gov (United States)

    2010-07-01

    ... Institutional Boilers and Process Heaters Pt. 63, Subpt. DDDDD, Table 6 Table 6 to Subpart DDDDD of Part 63—Fuel... of pounds of pollutant per MMBtu of heat content. 3. Hydrogen Chloride * * * a. Collect fuel samples...

  13. 77 FR 14482 - Petroleum Reduction and Alternative Fuel Consumption Requirements for Federal Fleets

    Science.gov (United States)

    2012-03-12

    ...-road or non-road vehicle. \\2\\ The definition of the term ``low-speed electric vehicle,'' as used... alternative fuel used in Federal fleet motor vehicles. This report also would include the alternative fuel...-- (1) a motor vehicle that operates solely on alternative fuel; or (2) a low-speed electric vehicle. (g...

  14. 14 CFR Special Federal Aviation... - Fuel Tank System Fault Tolerance Evaluation Requirements

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel Tank System Fault Tolerance Evaluation..., SFAR No. 88 Special Federal Aviation Regulation No. 88—Fuel Tank System Fault Tolerance Evaluation... certificates that may affect the airplane fuel tank system, for turbine-powered transport category airplanes...

  15. 40 CFR 600.006-86 - Data and information requirements for fuel economy vehicles.

    Science.gov (United States)

    2010-07-01

    ... fuel economy vehicles. 600.006-86 Section 600.006-86 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy Regulations for 1977 and Later Model Year Automobiles-General Provisions § 600.006-86 Data and...

  16. 40 CFR 600.006-89 - Data and information requirements for fuel economy vehicles.

    Science.gov (United States)

    2010-07-01

    ... fuel economy vehicles. 600.006-89 Section 600.006-89 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy Regulations for 1977 and Later Model Year Automobiles-General Provisions § 600.006-89 Data and...

  17. Sperm quantity and size variation in un-irradiated and irradiated males of the malaria mosquito Anopheles arabiensis Patton

    NARCIS (Netherlands)

    Helinski, M.; Knols, B.G.J.

    2009-01-01

    Anopheles mosquitoes are important candidates for genetic control strategies. However, little is known about sperm quality and quantity as determinants of male reproductive success. In this study, sperm quantity and length variation were assessed in testes of un-irradiated and irradiated Anopheles

  18. Facilitation of syngeneic stem cell engraftment by anti-class I monoclonal antibody pretreatment of unirradiated recipients

    International Nuclear Information System (INIS)

    Voralia, M.; Semeluk, A.; Wegmann, T.G.

    1987-01-01

    We have established a murine model of syngeneic bone marrow transplantation based on the use of monoclonal antibody as the sole conditioning regimen in unirradiated recipients. Administration of a single injection of monoclonal antibody directed against major histocompatibility complex-encoded class I determinants facilitated permanent hemopoietic stem cell engraftment without any apparent side-effects. Whereas untreated hosts exhibited a maximal chimerism of 15% at donor cell doses of up to 12 X 10(7) bone marrow cells, pretreatment by 2 mg of anti-class I antibody one week prior to transplantation of 3 X 10(7) syngeneic bone marrow cells resulted in a mean donor representation of about 80%. The antibody can be given up to four weeks prior to transplantation, and the degree of donor engraftment observed is a function of the dose of antibody administered. The fact that specific antibody enhanced engraftment in two strain combinations indicates that antibody is the active agent in facilitating engraftment and that facilitation is not strain-restricted. Anti-class I antibodies of the IgG2a, but not IgG1, isotype are effective in promoting engraftment. Although the isotype requirement suggests a role for antibody-mediated cytotoxicity in promoting stem cell engraftment, the extensive time-frame of facilitation suggests that other effects of the antibody may also be involved. The model of syngeneic bone marrow transplantation we describe here will be useful in studying the mechanisms regulating stem cell engraftment and may have potential clinical application as an approach to autologous marrow transplantation

  19. Designing consideration for a HLW / Spent Fuel DGR in Germany with retrievability requirements

    International Nuclear Information System (INIS)

    Thomauske, Bruno

    2014-01-01

    Since 2012 retrievability is part of the German waste disposal concept. In the preliminary safety studies of waste disposal in the Gorleben salt dome, retrievability had been included. The waste disposal concept on this new basis seems to be feasible. The new requirement to include retrievability for spent fuel and high level waste in the waste disposal concept led to a few but manageable consequences: waste containers must fulfill the requirement not to release aerosols in the first 500 years after closure of the repository; there are no consequences for the horizontal disposal of the waste containers in galleries; for the vertical disposal of the unshielded waste containers in boreholes the boreholes have to be stabilized by cylindrical liners; after transport of the waste containers above surface they have to be stored in interim storage facilities: these interim storage facilities, the waste handling facilities and the waste containers needed for long term storage have to be available in case waste has to be retrieved

  20. Integration of the AVLIS (atomic vapor laser isotopic separation) process into the nuclear fuel cycle. [Effect of AVLIS feed requirements on overall fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Hargrove, R.S.; Knighton, J.B.; Eby, R.S.; Pashley, J.H.; Norman, R.E.

    1986-08-01

    AVLIS RD and D efforts are currently proceeding toward full-scale integrated enrichment demonstrations in the late 1980's and potential plant deployment in the mid 1990's. Since AVLIS requires a uranium metal feed and produces an enriched uranium metal product, some change in current uranium processing practices are necessitated. AVLIS could operate with a UF/sub 6/-in UF/sub 6/-out interface with little effect to the remainder of the fuel cycle. This path, however, does not allow electric utility customers to realize the full potential of low cost AVLIS enrichment. Several alternative processing methods have been identified and evaluated which appear to provide opportunities to make substantial cost savings in the overall fuel cycle. These alternatives involve varying levels of RD and D resources, calendar time, and technical risk to implement and provide these cost reduction opportunities. Both feed conversion contracts and fuel fabricator contracts are long-term entities. Because of these factors, it is not too early to start planning and making decisions on the most advantageous options so that AVLIS can be integrated cost effectively into the fuel cycle. This should offer economic opportunity to all parties involved including DOE, utilities, feed converters, and fuel fabricators. 10 refs., 11 figs., 2 tabs.

  1. Concerning major items in government ordinance requiring modification of part of enforcement regulation for law relating to control of nuclear material, nuclear fuel and nuclear reactor

    International Nuclear Information System (INIS)

    1989-01-01

    The report describes major items planned to be incorporated into the enforcement regulations for laws relating to control of nuclear material, nuclear fuel and nuclear reactor. The modifications have become necessary for the nation to conclude a nuclear material protection treaty with other countries. The modification include the definitions of 'special nuclear fuel substances' and 'special nuclear fuel substances' and 'special nuclear fuel substances subject to protection'. The modifications require that protective measures be taken when handling and transporting special nuclear fuel substances subject to protection. Transport of special nuclear fuel substances requires approval from the Prime Minister or Transport Minister. Transport of special nuclear fuel substances subject to protection should be conducted after notifying the prefectural Public Safety Commission. Transport of special nuclear fuel substances subject to protection requires the conclusion of arrangements among responsible persons and approval of them from the Prime Minister. (N.K.)

  2. AREVA's fuel assemblies addressing high performance requirements of the worldwide PWR fleet

    International Nuclear Information System (INIS)

    Anniel, Marc; Bordy, Michel-Aristide

    2009-01-01

    Taking advantage of its presence in the fuel activities since the start of commercial nuclear worldwide operation, AREVA is continuing to support the customers with the priority on reliability, to: >participate in plant operational performance for the in core fuel reliability, the Zero Tolerance for Failure ZTF as a continuous improvement target and the minimisation of manufacturing/quality troubles, >guarantee the supply chain a proven product stability and continuous availability, >support performance improvements with proven design and technology for fuel management updating and cycle cost optimization, >support licensing assessments for fuel assembly and reloads, data/methodologies/services, >meet regulatory challenges regarding new phenomena, addressing emergent performance issues and emerging industry challenges for changing operating regimes. This capacity is based on supplies by AREVA accumulating very large experience both in manufacturing and in plant operation, which is demonstrated by: >manufacturing location in 4 countries including 9 fuel factories in USA, Germany, Belgium and France. Up to now about 120,000 fuel assemblies and 8,000 RCCA have been released to PWR nuclear countries, from AREVA European factories, >irradiation performed or in progress in about half of PWR world wide nuclear plants. Our optimum performances cover rod burn ups of to 82GWD/tU and fuel assemblies successfully operated under various world wide fuel management types. AREVA's experience, which is the largest in the world, has the extensive support of the well known fuel components such as the M5'TM'cladding, the MONOBLOC'TM'guide tube, the HTP'TM' and HMP'TM' structure components and the comprehensive services brought in engineering, irradiation and post irradiation fields. All of AREVA's fuel knowledge is devoted to extend the definition of fuel reliability to cover the whole scope of fuel vendor support. Our Top Reliability and Quality provide customers with continuous

  3. A Criticality Safety Study on Storing Unirradiated Cintichem-Type Targets at Sandia National Laboratories

    International Nuclear Information System (INIS)

    Romero, D.J.; Parma, E.J.; Busch, R.D.

    1999-01-01

    This criticality safety analysis is performed to determine the effective multiplication factor (k eff ) for a storage cabinet filled with unirradiated Cintichem-type targets. These targets will be used to produce 99 Mo at Sandia National Laboratories and will be stored on-site prior to irradiation in the Annular Core Research Reactor. The analysis consisted of using the Monte Carlo code MCNP (Version 4A) to model and predict the k eff for the proposed dry storage configuration under credible loss of geometry and moderator control. Effects of target pitch, non-uniform loading, and target internal/external flooding are evaluated. Further studies were done with deterministic methods to verify the results obtained from MCNP and to obtain a clearer understanding of the parameters affecting system criticality. The diffusion accelerated neutral particle transport code ONEDANT was used to model the target in a one-dimensional, infinite half-slab geometry and determine the critical slab thickness. Hand calculations were also completed to determine the critical slab thickness with modified one-group, and one-group, two region approximations. Results obtained from ONEDANT and the hand calculations were compared to applicable cases in a commonly used criticality safety analysis handbook. Overall, the critical slab thicknesses obtained in the deterministic analysis were much larger than the dimensions of the cabinet and further support the predictions by MCNP that a critical system cannot be attained for the base case or in conditions where loss of geometry and moderation control occur

  4. Fusarium growth on culture media made of tissue juice from irradiated and unirradiated potato tubers

    International Nuclear Information System (INIS)

    Taczanowski, M.

    1994-01-01

    Fusarium Sulphureum Schlecht is one of the tuber pathogens causing potato storage disease knowing as dry rot. Because irradiation can disturb the tissue defence mechanism against the pathogen, it was decided to carry out experiments on influence of the treatment on subsequent tuber tissue reaction to a maceration process. The maceration as a physical stress was a substitute for the pathogen activity. Tubers of two potato varieties were tested: Mila -a resistant variety to Fusarium and Atol - susceptible one. Tubers of both varieties were irradiated with a dose of 105 kGy. Unirradiated tubers were taken as a control. A day after irradiation the cortex tissue was macerated using an ordinary rasper and the resulted tissue pulp was strained through medical gauze to obtain crude juice. The juice was clarified by centrifugation and then added to dissolved PDA. The volume ratio of juice to PDA was 1:1. The prepared media were dispensed into Petri dishes. Small pieces of the Fusarium culture were put on the surface of the medium at the centre of each Petri dish. Subsequent growth of the fungus was assessed by measurement of culture diameters every 24 hours. Linear functions of the Fusarium growth were obtained for Mila control and Atol control. In the case of Mila, the Fusarium found more favourable conditions for its growth in the presence of juice from irradiated tubers than from the control ones. Making the same comparison for Atol, no difference was detected. (author)

  5. The reaction of unirradiated and irradiated nuclear graphites with water vapor in helium

    International Nuclear Information System (INIS)

    Imai, Hisashi; Nomura, Shinzo; Kurosawa, Takeshi; Fujii, Kimio; Sasaki, Yasuichi

    1980-10-01

    Nuclear graphites more than 10 brands were oxidized with water vapor in helium and then some selected graphites were irradiated with fast neutron in the Japan Materials Testing Reactor to clarify the effect of radiation damage of graphite on their reaction behaviors. The reaction was carried out under a well defined condition in the temperature range 800 -- 1000 0 C at concentrations of water vapor 0.38 -- 1.30 volume percent in helium flow of total pressure of 1 atm. The chemical reactivity of graphite irradiated at 1000 +- 50 0 C increased linearly with neutron fluence until irradiation of 3.2 x 10 21 n/cm 2 . The activation energy for the reaction was found to decrease with neutron fluence for almost all the graphites, except for a few ones. The order of reaction increased from 0.5 for the unirradiated graphite to 1.0 for the graphite irradiated up to 6.0 x 10 20 n/cm 2 . Experiment was also performed to study a superposed effect between the influence of radiation damage of graphite and the catalytic action of barium on the reaction rate, as well as the effect of catalyser of barium. It was shown that these effects were not superposed upon each other, although barium had a strong catalytic action on the reaction. (author)

  6. Spent fuel storage and transport cask decontamination and modification. An overview of management requirements and applications based on practical experience

    International Nuclear Information System (INIS)

    1999-04-01

    A large increase in the number of casks required for transport and/or storage of spent fuel is forecast into the next century. The principal requirement will be for increased number of storage and dual purpose (transport/storage) casks for interim storage of spent fuel prior to reprocessing or permanent disposal in both on-site and off-site storage facilities. Through contact with radioactive materials spent fuel casks will be contaminated on both internal and external surfaces. In broad terms, cask contamination management can be defined by three components: minimisation, prevention and decontamination. This publication is a compilation of international experience with cask contamination problems and decontamination practices. The objective is to present current knowledge and experience as well as developments, trends and potential for new applications in this field. Furthermore, the report may assist in new design or modification of existing casks, cask handling systems and decontamination equipment

  7. Dissolution of Used Nuclear Fuel Using a TBP/N-Paraffin Solvent

    Energy Technology Data Exchange (ETDEWEB)

    Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Shehee, T. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jones, D. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); DelCul, G. D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-10-02

    The dissolution of unirradiated used nuclear fuel (UNF) pellets pretreated for tritium removal was demonstrated using a tributly phosphate (TBP) solvent. Dissolution of pretreated fuel in TBP could potentially combine dissolution with two cycle of solvent extraction required for separating the actinides and lanthanides from other fission products. Dissolutions were performed using UNF surrogates prepared from both uranyl nitrate and uranium trioxide produced from the pretreatment process by adding selected actinide and stable fission product elements. In laboratory-scale experiments, the U dissolution efficiency ranged from 80-99+% for both the nitrate and oxide surrogate fuels. On average, 80% of the Pu and 50% of the Np and Am in the nitrate surrogate dissolved; however, little of the transuranic elements dissolved in the oxide form. The majority of the 3+ lanthanide elements dissolved. Only small amounts of Sr (0-1.6%) and Mo (0.1-1.7%) and essentially no Cs, Ru, Zr, or Pd dissolved.

  8. 40 CFR 80.1430 - Requirements for exporters of renewable fuels.

    Science.gov (United States)

    2010-07-01

    ... of renewable fuel in the blend. (2) Determination of the renewable portion of the blend using Method... method as approved by the EPA. (3) Assuming the maximum concentration of the renewable fuel in the blend... its neat form or blended with gasoline or diesel, that is exported from any of the regions described...

  9. Closed Nuclear Fuel Cycle Technologies to Meet Near-Term and Transition Period Requirements

    International Nuclear Information System (INIS)

    Collins, E.D.; Felker, L.K.; Benker, D.E.; Campbell, D.O.

    2008-01-01

    A scenario that very likely fits conditions in the U.S. nuclear power industry and can meet the goals of cost minimization, waste minimization, and provisions of engineered safeguards for proliferation resistance, including no separated plutonium, to close the fuel cycle with full actinide recycle is evaluated. Processing aged fuels, removed from the reactor for 30 years or more, can provide significant advantages in cost reduction and waste minimization. The UREX+3 separations process is being developed to separate used fuel components for reuse, thus minimizing waste generation and storage in geologic repositories. Near-term use of existing and new thermal spectrum reactors can be used initially for recycle actinide transmutation. A transition period will eventually occur, when economic conditions will allow commercial deployment of fast reactors; during this time, recycled plutonium can be diverted into fast reactor fuel and conversion of depleted uranium into additional fuel material can be considered. (authors)

  10. Closed Nuclear Fuel Cycle Technologies to Meet Near-Term and Transition Period Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Collins, E.D.; Felker, L.K.; Benker, D.E.; Campbell, D.O. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, Tennessee, 37831-6152 (United States)

    2008-07-01

    A scenario that very likely fits conditions in the U.S. nuclear power industry and can meet the goals of cost minimization, waste minimization, and provisions of engineered safeguards for proliferation resistance, including no separated plutonium, to close the fuel cycle with full actinide recycle is evaluated. Processing aged fuels, removed from the reactor for 30 years or more, can provide significant advantages in cost reduction and waste minimization. The UREX+3 separations process is being developed to separate used fuel components for reuse, thus minimizing waste generation and storage in geologic repositories. Near-term use of existing and new thermal spectrum reactors can be used initially for recycle actinide transmutation. A transition period will eventually occur, when economic conditions will allow commercial deployment of fast reactors; during this time, recycled plutonium can be diverted into fast reactor fuel and conversion of depleted uranium into additional fuel material can be considered. (authors)

  11. LOCA scenario tests of irradiated fuel rod specimens

    International Nuclear Information System (INIS)

    Scott, Harold

    2004-01-01

    Full text: The NRC's cladding performance program at Argonne National Laboratory (ANL) is testing fueled high-burnup segments subjected to LOCA integral phenomena. The data are provided to NRC and the nuclear industry for their independent assessment of the adequacy of licensing criteria for LOCA events. The tests are being conducted with high-burnup 30 cm segments from Limerick (9x9 Zry-2) and H.B. Robinson (15x15 Zry-4) reactors. Prior to testing, sibling samples are characterized with respect to fuel morphology, fuel-cladding bond, cladding oxide layer thickness, hydrogen content and high-temperature steam oxidation kinetics. Specimens that survive quench are subjected to four-point bend tests, followed by local diametral compression tests. The retention of post-quench ductility is a more limiting requirement than surviving thermal stresses during quench. Companion tests are conducted with unirradiated cladding to generate baseline data for comparison with the high-burnup fuel results. LOCA integral tests have the following sequential steps: stabilization of temperature, internal pressure and steam flow at 300 C, ramping of temperature (∼5C/s) through ballooning and burst to 1204 C, hold at 1204 C for 1-5 minutes, slow-cooling (∼3C/s) to 800 C, and water quenching at ∼800C. Two high-burnup tests were completed in 2002 with Limerick BWR rod segments: ramp to burst in argon followed by slow cooling; and the LOCA test with 5-minute hold time at 1204 C, followed by slow cooling. With the exception of burst-opening shape, results for burst temperature, burst pressure, burst length, and ballooning strain profile are more similar to, than different from, results for unirradiated Zry-2 cladding exposed to the same time-temperature history. The 3rd Limerick test with quench was performed in December 2003, and a 4th Limerick test was performed in March 2004. Tests on high-burnup Robinson PWR fuel segments are scheduled to begin in June 2004. The presentation points

  12. Fuel for ADS: State-of-the-art, requirements, current and future programmes

    International Nuclear Information System (INIS)

    Gabrielli, F.

    2015-01-01

    Fuels are the cornerstone of research/development programs of accelerator-driven systems for transmuting minor actinides (MA). Compared with the fuels for critical reactors, fuels for ADS are generally U-free to improve the transmutation performance and contain high volumetric concentrations (about 50%) of MA and Pu compounds. Their specific fabrication, reprocessing, design and safety issues are being currently investigated. This paper provides an overview of the current state-of-the-art of the assessment of these innovative fuels. Emphasis has been placed on the main outcomes of the work performed within the Sixth FP EU EUROTRANS Programme, which provided a decisive step forward in improving knowledge on fabrication, properties, and behaviour under irradiation of these challenging fuels. In EUROTRANS, for the ADS application, the focus was on Ceramic-Ceramic (Pu,MA)O 2 + MgO and Ceramic-Metallic (Pu,MA)O 2 + 92 Mo composite fuels, which were recommended for the European Facility for Industrial Transmutation (EFIT). The fuels consist of particles of (Pu,MA)O 2 phases dispersed in a magnesia or molybdenum matrix. Solid nitride (Pu,MA,Zr)O 2 fuels have also been considered as a back-up solution. This paper presents the main experimental results from out-of-pile and in-pile experiments of the Ceramic-Ceramic and Ceramic-Metallic composite fuels as well as related safety assessments. Further, the on-going research and development activities on fuels loaded with large amount of MA will be presented. (author)

  13. The data requirements for the verification and validation of a fuel performance code - the transuranus perspective

    International Nuclear Information System (INIS)

    Schubert, A.; Di Marcello, V.; Rondinella, V.; Van De Laar, J.; Van Uffelen, P.

    2013-01-01

    In general, the verification and validation (V and V) of a fuel performance code like TRANSURANUS consists of three basic steps: a) verifying the correctness and numerical stability of the sub-models; b) comparing the sub-models with experimental data; c) comparing the results of the integral fuel performance code with experimental data Only the second and third steps of the V and V rely on experimental information. This scheme can be further detailed according to the physical origin of the data: on one hand, in-reactor ('in-pile') experimental data are generated in the course of the irradiation; on the other hand ex-reactor ('out-of-pile') experimental data are obtained for instance from various postirradiation examinations (PIE) or dedicated experiments with fresh samples. For both categories, we will first discuss the V and V of sub-models of TRANSURANUS related to separate aspects of the fuel behaviour: this includes the radial variation of the composition and fissile isotopes, the thermal properties of the fuel (e.g. thermal conductivity, melting temperature, etc.), the mechanical properties of fuel and cladding (e.g. elastic constants, creep properties), as well as the models for the fission product behaviour. Secondly, the integral code verification will be addressed as it treats various aspects of the fuel behaviour, including the geometrical changes in the fuel and the gas pressure and composition of the free volume in the rod. (authors)

  14. Classification of the railway accident in accordance with the requirement of the safety analysis of transporting spent fuel

    International Nuclear Information System (INIS)

    Wu Tao

    1993-01-01

    Based on the analysis of the difference between the accident severity categorization used in the Ministry of Railway and that used in the safety analysis of the transporting spent fuel, a method used for the classification of the railway accident in accordance with the requirement of the safety analysis of transporting spent fuel is suggested. The method classifies the railway accidents into 10 scenarios and make it possible to scale the accident through directly using the data documented by the Ministry of Railway without any additional effort

  15. On the evaluation of feed requirements and costs analysis of preirradiated fuel cycle operations for uranium fuelled light water reactors

    International Nuclear Information System (INIS)

    El Osery, I.A.; Yasso, K.A.; Abdel Salam, A.S.

    1982-01-01

    This work is a part of an integrated scheme for nuclear power cost evaluation. The paper gives a brief description of the different operations included to get enriched UO 2 in its final form. A Material Balance Sheet is developed to estimate quantitatively the input material reguirements to each fuel operation. An improved approach for fuel cost analysis is developed. The paper includes a complete FORTRAN listing for the computer program carried out for this purpose together with description of the program input data requirements and output facilities. Illustrative numerical results are provided

  16. System comparison of hydrogen with other alternative fuels in terms of EPACT requirements

    Energy Technology Data Exchange (ETDEWEB)

    Barbir, F.; Oezay, K.; Veziroglu, T.N. [Univ. of Miami, Coral Gables, FL (United States)

    1996-10-01

    The feasibility of several alternative fuels, namely natural gas, methanol, ethanol, hydrogen and electricity, to replace 10% of gasoline by the year 2000 has been investigated. The analysis was divided in two parts: (i) analysis of vehicle technologies, and (ii) analysis of fuel production storage and distribution, from the primary energy sources to the refueling station. Only technologies that are developed to at least demonstration level were considered. The amount and type of the primary energy sources have been determined for each of the fuels being analyzed. A need for a common denominator for different types of energy has been identified.

  17. The French program on the spent nuclear fuel long term evolution: Major results, uncertainties and new requirements

    International Nuclear Information System (INIS)

    Ferry, Cecile; Poinssot, Christophe; Gras, Jean-Marie

    2006-01-01

    conditions; oxidation kinetics of UO 2 and spent fuel are assessed by coupling a microscopic approach with macroscopic classical methods; - in water, this boundary condition corresponds to the nominal scenario after the breaching of the canister during geological disposal; in this case, the effect of water a radiolysis on the spent fuel matrix dissolution is investigated. An overview of the state of knowledge on the long-term behaviour of spent fuel under these various conditions has been provided by Ferry et al. (2005). This report constitutes the scientific synthesis due at the term of the law. This state of the art was derived from the results obtained under the PRECCI project as well as from a review of the literature and some of the data issued from the 5. PRDC European project 'Spent Fuel Stability under Repository Conditions' (Poinssot et al, 2005). Regarding these results, this paper presents for each initial operational question and concept, the main scientific issues, the major results obtained since 1999, and the remaining uncertainties. The new requirements in the frame of a broader context which includes transport and in-pool storage issues are also identified. The contents of this paper is as follows: 1 Introduction; 2 The retrievability of spent fuel assemblies after storage; 2.1 Main scientific issues; 2.2 Major results; 2.2.1 Mechanical evolution of the cladding and structural materials; 2.2.2 Specific case of the defected fuel rods - Evolution of the spent fuel pellets; 2.3 Remaining uncertainties and new requirements for each operational context; 3 The treatment of spent fuel after a long period of storage; 3.1 Main scientific issues; 3.2 Major results; 3.3 Remaining uncertainties and new requirements; 4 The definition of the radionuclide source terms; 4.1 Main scientific issues; 4.2 Major results; 4.3 Remaining uncertainties and new requirements for each operational context; 5 The compatibility between dry storage and subsequent disposal; 5.1 Main

  18. chemical determination of burnup ratio in nuclear fuels

    International Nuclear Information System (INIS)

    Guereli, L.

    1997-01-01

    Measurements of the extent of fission are important to determine the irradiation performance of a nuclear fuel. The energy released per unit mass of uranium (burnup) can be determined from measurement of the percent of heavy atoms that have fissioned during irradiation.The preferred method for this determination is choosing a suitable fission monitor (usually ''1''4''8Nd) and its determination after separation from the fuel matrix. In thermal reactor fuels where the only heavy element in the starting material is uranium, uranium depletion can be used for burnup determination. ''2''3''5U depletion method requires measurement of uranium isotopic ratios of both irradiated and unirradiated fuel. Isotopic ratios can be determined by thermal ionization mass spectrometer following separation of uranium from the fuel matrix. Separation procedures include solvent extraction, ion exchange and anion exchange chromatography. Another fission monitor used is ''1''3''9La determination by HPLC. Because La is monoisotopic (''1''3''9La) in the fuel, it can be determined by chemical analysis techniques

  19. Radioactive decay properties of CANDU fuel. Volume 1: the natural uranium fuel cycle

    International Nuclear Information System (INIS)

    Clegg, L.J.; Coady, J.R.

    1977-01-01

    The computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 1 presents these data for unirradiated fuel, uranium ore and uranium mill tailings. In Part 2 they have been computed for fuel irradiated to levels of burnup ranging from 140 GJ/kg U to 1150 GJ/kg U. (author)

  20. Induction of lethal mutations in the x-chromosome of unirradiated Drosophila oocytes after fertilization by irradiated spermatozoa

    International Nuclear Information System (INIS)

    Shaposhnikov, M.V.; Zainullin, V.G.

    2003-01-01

    Full text: In primary study on Drosophila it was found that irradiated male X-chromosomes induce recessive lethals in unirradiated female homologues (Abeleva et al., 1961, Radiobiologya. 1:123-126). The same effects were obtained in Drosophila in some recent investigations. The mechanisms of these effects is unknown. However it may be responsible for low-dose radiation effects as it induce mutations in unirradiated DNA. We assume that this effect may be a result of activation of error prone repair in response to preliminary DNA lesions in irradiated chromosome. In this research we analyse the frequencies of the recessive lethal mutations in the X-chromosome of Drosophila females mated with irradiated Basc males. We used acute irradiation with a dose rate of 10 Gy. For testing our hypothesis we use the mus209 and mei-41 mutant females. Mus209 is a PCNA gene homologue and mei-41 is a homologue of ATM gene. These genes are involved in post-replication DNA repair which may be error prone repair in Drosophila. It was obtained the tendency to decreasing the mutation rate at the mei-41[D5] background and decreasing mutation rate in mus209[B1] background in comparison with wild type strains CS (p<0.05). The obtained results demonstrate the possible role of mus209[B1] and mei-41[D5] genes in the inducing of mutations in the unirradiated X-chromosome in the presence of irradiated homologue

  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. Design verification of the CANFLEX fuel bundle - quality assurance requirements for mechanical flow testing

    International Nuclear Information System (INIS)

    Alavi, P.; Oldaker, I.E.; Chung, C.H.; Suk, H.C.

    1997-01-01

    As part of the design verification program for the new fuel bundle, a series of out-reactor tests was conducted on the CANFLEX 43-element fuel bundle design. These tests simulated current CANDU 6 reactor normal operating conditions of flow, temperature and pressure. This paper describes the Quality Assurance (QA) Program implemented for the tests that were run at the testing laboratories of Atomic Energy of Canada Limited (AECL) and Korea Atomic energy Research Institute (KAERI). (author)

  3. Assessment of technical risks and R and D requirements for a magnetic confinement fusion fuel system. Final report

    International Nuclear Information System (INIS)

    DeFreece, D.A.

    1983-11-01

    This report documents a specific use and results of a novel technique for assessing the technical risks associated with the hardware development of a possible future commercial fusion power plant fuel system. Technical risk is defined as the risk that a particular technology or component which is currently under development will not achieve a set of required technical specifications. A technical risk assessment is the quantification of this risk. This Technical Risk Assessment (TRA) methodology was applied to a deuterium-tritium fuel system for a magnetic-confinement fusion power plant. The fuel system is defined to support a generic commercial reactor with at least two viable options for each critical subsystem. Each subsystem option is defined in detail including nominal performance requirements and subsystem interfaces. Subsystem experts were canvassed to obtain values for past, present and future technical performance parameters for each of the subsystem options. These forecasts are presented as probabilities of achieving given levels of performance in specific time periods for assumed funding scenarios. Several funding scenarios were examined to discern whether performance limitations are caused by funding or technology. A computerized Fuel System simulation is described which uses these subsystem performance parameter forecasts as inputs

  4. Modern requirements to quality assurance and control in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Weidinger, H.G.

    1999-01-01

    This lecture have shown a new type of quality assurance management has already successfully introduced in various industries and now starts to be used increasingly in the nuclear fuel industry. Static authority regulations and a tendency to bureaucratic understanding and handling of these regulations lead to a delayed start and a relatively slow progress of these quality strategies in the nuclear fuel technology. However, the economic pressure of strong competition and increasing demands of the utilities as the user of nuclear fuel result in a more determined introduction also to this area. The different use of statistical methods of two different fuel vendors are shown. Vendor A uses old fashioned methods. The focus is on the expensive final product control and few emphasis is on design of experiments and process control. Consequently, this vendor will have high costs, not only for QC and rejection but also for repair and replace actions after delivery. To the contrary, vendor B invests primarily in the design of experiments and process control. This vendor will profit only from lower direct costs but also from being at the front line of technical development and from enjoying a satisfied and happy customer. Many well examined quality management tools are available today which help not only to improve the quality but also decrease the costs. Still, the progress in using these techniques in nuclear fuel technology is limited and not comparable to the progress in other industries like automobile production or the electronic industry. (author)

  5. Advanced Characterization Techniques for Silicon Carbide and Pyrocarbon Coatings on Fuel Particles for High Temperature Reactors (HTR)

    Energy Technology Data Exchange (ETDEWEB)

    Basini, V.; Charollais, F. [CEA Cadarache, DEN/DEC/SPUA, BP 1, 13108 St Paul Lez Durance (France); Dugne, O. [CEA Marcoule, DEN/DTEC/SCGS BP 17171 30207 Bagnols sur Ceze (France); Garcia, C. [Laboratoire des Composites Thermostructuraux (LCTS), UMR CNRS 5801, 3 allee de La Boetie, 33600 Pessac (France); Perez, M. [CEA Grenoble DRT/DTH/LTH, 17 rue des Martyrs, 38054 Grenoble cedex 9 (France)

    2008-07-01

    Cea and AREVA NP have engaged an extensive research and development program on HTR (high temperature reactor) fuel. The improving of safety of (very) high temperature reactors (V/HTR) is based on the quality of the fuel particles. This requires a good knowledge of the properties of the four-layers TRISO particles designed to retain the uranium and fission products during irradiation or accident conditions. The aim of this work is to characterize exhaustively the structure and the thermomechanical properties of each unirradiated layer (silicon carbide and pyrocarbon coatings) by electron microscopy (SEM, TEM), selected area electronic diffraction (SEAD), thermo reflectance microscopy and nano-indentation. The long term objective of this study is to define pertinent parameters for fuel performance codes used to better understand the thermomechanical behaviour of the coated particles. (authors)

  6. Evaluation of research reactor fuel reliability in support of regulatory requirements

    International Nuclear Information System (INIS)

    Sokolov, Eugene N.

    2005-01-01

    This standards, codes and practices survey is devoted to the problem of reliability of R and D especially research reactor fuel (RRF) performance-related processes. Regulatory R and D evaluations were based on one standard and just few of them provide correlation to other relative standards whereas synthetic process approach reflects actual status of particular R and D practices. Fuel performance regulatory parameters are based on quality standards. A reliability process-based method similar to PSA/FMEA is proposed to evaluate RRF performance- related parameters in terms of reactor safety. (author)

  7. Evaluation of research reactor fuel reliability in support of regulatory requirements

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Eugene N [Chalk River Laboratories, AECL, Chalk River, ON, K0J 1J0 (Canada)

    2005-07-01

    This standards, codes and practices survey is devoted to the problem of reliability of R and D especially research reactor fuel (RRF) performance-related processes. Regulatory R and D evaluations were based on one standard and just few of them provide correlation to other relative standards whereas synthetic process approach reflects actual status of particular R and D practices. Fuel performance regulatory parameters are based on quality standards. A reliability process-based method similar to PSA/FMEA is proposed to evaluate RRF performance- related parameters in terms of reactor safety. (author)

  8. 14 CFR 91.151 - Fuel requirements for flight in VFR conditions.

    Science.gov (United States)

    2010-01-01

    ... begin a flight in an airplane under VFR conditions unless (considering wind and forecast weather conditions) there is enough fuel to fly to the first point of intended landing and, assuming normal cruising speed— (1) During the day, to fly after that for at least 30 minutes; or (2) At night, to fly after that...

  9. 40 CFR 600.307-86 - Fuel economy label format requirements.

    Science.gov (United States)

    2010-07-01

    ... metering system, including number of carburetor barrels, if applicable; (7) Transmission class; (8... Regulations for 1977 and Later Model Year Automobiles-Labeling § 600.307-86 Fuel economy label format... the city and highway estimates by 0.85, then rounding to the next lower integer value. (2) The upper...

  10. Fuel requirements for isotope production and reasearch reactors: Possible alternative ways of meeting non-proliferation objectives

    International Nuclear Information System (INIS)

    There is a continuing need for access to medium-to-high flux research reactors of intermediate power level (5-50 MW) for the production of industrial and medical radioisotopes, for the provision of neutron beams and for materials research. The construction of further reactors of this type is likely. To obtain the required flux levels in adequate volumes and at the lowest capital cost, past practice has been to design a small-core reactor around a fuel element concept using fully enriched uranium, that is, uranium enriched to 80% U-235 or greater. In recent years, however, it has been recognised that the use of fully enriched uranium in research reactors could give rise to significant risks of nuclear weapons proliferation. Accordingly, there would be advantage if research reactors could be operated on low enriched fuel, that is, enrichment levels of 20% or less. It is the purpose of this paper to explore the implications for proliferation of the enrichment level of research reactor fuel and to draw attention to possible options for reducing proliferation concerns which warrant further study. It does not, however, consider research reactors using very low enriched or natural uranium fuel. The paper is offered to stimulate discussion of the issues and the views expressed do not necessarily represent any formal Australian position

  11. 40 CFR 80.570 - What labeling requirements apply to retailers and wholesale purchaser-consumers of diesel fuel...

    Science.gov (United States)

    2010-07-01

    ... FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel; Nonroad, Locomotive, and Marine Diesel Fuel; and ECA..., motor vehicle diesel fuel subject to the 15 ppm sulfur standard of § 80.520(a)(1), must affix the... dispensing, motor vehicle diesel fuel subject to the 500 ppm sulfur standard of § 80.520(c), must prominently...

  12. Requirements and testing methods for surfaces of metallic bipolar plates for low-temperature PEM fuel cells

    Science.gov (United States)

    Jendras, P.; Lötsch, K.; von Unwerth, T.

    2017-03-01

    To reduce emissions and to substitute combustion engines automotive manufacturers, legislature and first users aspire hydrogen fuel cell vehicles. Up to now the focus of research was set on ensuring functionality and increasing durability of fuel cell components. Therefore, expensive materials were used. Contemporary research and development try to substitute these substances by more cost-effective material combinations. The bipolar plate is a key component with the greatest influence on volume and mass of a fuel cell stack and they have to meet complex requirements. They support bending sensitive components of stack, spread reactants over active cell area and form the electrical contact to another cell. Furthermore, bipolar plates dissipate heat of reaction and separate one cell gastight from the other. Consequently, they need a low interfacial contact resistance (ICR) to the gas diffusion layer, high flexural strength, good thermal conductivity and a high durability. To reduce costs stainless steel is a favoured material for bipolar plates in automotive applications. Steel is characterized by good electrical and thermal conductivity but the acid environment requires a high chemical durability against corrosion as well. On the one hand formation of a passivating oxide layer increasing ICR should be inhibited. On the other hand pitting corrosion leading to increased permeation rate may not occur. Therefore, a suitable substrate lamination combination is wanted. In this study material testing methods for bipolar plates are considered.

  13. GRIST-2 preliminary test plan and requirements for fuel fabrication and preirradiation

    International Nuclear Information System (INIS)

    Tang, I.M.; Harmon, D.P.; Torri, A.

    1978-12-01

    The preliminary version of the GRIST-2 test plan has been developed for the planned initial 5 years (1984 to 1989) of TREAT-Upgrade in-pile tests. These tests will be employed to study the phenomenology and integral behavior of GCFR core disruptive accidents (CDAs) and to support the Final Safety Analysis Report (FSAR) CDA analyses for the demonstration plant licensing. The preliminary test plan is outlined. Test Phases I and II are for the fresh fuel (preconditioned or not) CDA behavior at the beginning-of-life (BOL) reactor state. Phase III is for the reactor state that contains irradiated fuel with a saturated content of helium and fission gas. Phase IV is for larger bundle tests and scaling effects

  14. The climate responsibilities of the fossil fuel industry: why the Paris Agreement goals require an end to growth

    Science.gov (United States)

    Trout, K.; Muttitt, G.; Kretzmann, S.; Stockman, L.; Doukas, A.

    2017-12-01

    In December 2015, governments agreed in Paris to limit global average temperature rise to well below 2°C, and to aim to limit it to 1.5°C, compared to pre-industrial levels. Achieving these goals would require greenhouse gas emissions to reach net zero early in the second half of this century, and consequently most fossil fuel use to be phased out. This has clear implications for fossil fuel industry, and shines light on what should be expected of the industry in its business decisions - complementing the discussion of the industry's role in the scientific debate. This presentation shares the results to date of ongoing research into the committed emissions from oilfields, gasfields and coal mines, compared to carbon budgets. Building on prior work on fossil fuel reserves (notably Meinshausen 2009), our research focuses just on the developed reserves, from already-producing fields and mines. We estimate developed reserves of oil and gas using industry databases, and of coal using analysis by the International Energy Agency, and compare with carbon budgets published in the IPCC's 5th Assessment Report. The key findings are that: Developed reserves of oil, gas, and coal are more than we can afford to burn while keeping likely warming below 2°C. Developed reserves of oil and gas alone would take the world beyond 1.5°C. The implications are that development of any new fields or mines will either push the world beyond agreed climate limits, or cause some existing extraction assets to become stranded. This suggests that fossil fuel companies should stop developing new infrastructure, and governments should oversee a managed decline of the industry over the coming decades, combined with an upscaling of clean energy, as existing fossil fuel reserves are depleted.

  15. Interaction of spermine with DNA, vitamin C and bovine serum albumin in the unirradiated and gamma irradiated states

    International Nuclear Information System (INIS)

    Upadhyay, S.N.; Lal, C.; Bhardwaj, R.; Chaturvedi, S.; Chaudhury, N.K.

    2006-01-01

    Structural deformability of spermine with radiation dose (maximum 10 Gy) has been proved. Complex formation of spermine with DNA, vitamin C and BSA took place. Calibration and radiation-induced absorption changes in spermine by ninhydrin reagent has been followed quantitatively. Interaction of vitamin C with DNA and their radiation-induced changes have been reported. Interaction of spermine with DNA in the unirradiated and gamma irradiated states in 10 -3 M phosphate buffer and water have been compared. Addition of spermine and vitamin C to DNA makes DNA structure more condensed. Bovine serum albumin also binds with spermine and protects it from radiation-induced degradation. (author)

  16. Requirements on fuel management for a safe and optimum operation of the German research reactor FRJ-2

    International Nuclear Information System (INIS)

    Nabbi, R.; Bormann, H.J.; Wolters, J.

    1997-01-01

    In case of a coarse control arm (CCAs) at FRJ-2, reactivity is added to the reactor. The amount of this reactivity depends on the efficiency of the individual CCAs which has been measured as 180% of the average reactivity of the six arms for the central arm. For this design basis accident, it is required that only 4 out of 5 residual arms must be capable of shutting down the reactor. This minimum shutdown reactivity is provided by an optimum fuel management including an experimental reactivity determination. Calculation of fuel burnup and material densities is performed by the depletion code SUSAN, which has been verified by separate calculations using ORIGEN. The difference in the reactivity values (between calculation and measurement) is mainly a consequence of the limitation of the inverse kinetic method, which is not capable of covering the effects of the flux deformation and interaction of the CCAs and core in the process of reactor scram. (author)

  17. Fuel Behaviour at High During RIA and LOCA Accidents

    International Nuclear Information System (INIS)

    Barrio del Juanes, M. T.; Garcia Cuesta, J. C.; Vallejo Diaz, I.; Herranz Puebla

    2001-01-01

    Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs

  18. Transient survivability of LMR oxide fuel pins

    International Nuclear Information System (INIS)

    Weber, E.T.; Pitner, A.L.; Bard, F.E.; Culley, G.E.; Hunter, C.W.

    1986-01-01

    Fuel pin integrity during transient events must be assessed for both the core design and safety analysis phases of a reactor project. A significant increase in the experience related to limits of integrity for oxide fuel pins in transient overpower events has been realized from testing of fuel pins irradiated in FFTF and PFR. Fourteen FFTF irradiated fuel pins were tested in TREAT, representing a range of burnups, overpower ramp rates and maximum overpower conditions. Results of these tests along with similar testing in the PFR/TREAT program, provide a demonstration of significant safety margins for oxide fuel pins. Useful information applied in analytical extrapolation of fuel pin test data have been developed from laboratory transient tests on irradiated fuel cladding (FCTT) and on unirradiated fuel pellet deformation. These refinements in oxide fuel transient performance are being applied in assessment of transient capabilities of long lifetime fuel designs using ferritic cladding

  19. Standard Compliance: Guidelines to Help State and Alternative Fuel Provider Fleets Meet Their Energy Policy Act Requirements, 10 CFR Part 490 (Book)

    Energy Technology Data Exchange (ETDEWEB)

    2014-03-01

    This guidebook addresses the primary requirements of the Alternative Fuel Transportation Program to help state and alternative fuel provider fleets comply with the Energy Policy Act via the Standard Compliance option. It also addresses the topics that covered fleets ask about most frequently.

  20. Characterization of SiC–SiC composites for accident tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Deck, C.P., E-mail: Christian.Deck@ga.com; Jacobsen, G.M.; Sheeder, J.; Gutierrez, O.; Zhang, J.; Stone, J.; Khalifa, H.E.; Back, C.A.

    2015-11-15

    Silicon carbide (SiC) is being investigated for accident tolerant fuel cladding applications due to its high temperature strength, exceptional stability under irradiation, and reduced oxidation compared to Zircaloy under accident conditions. An engineered cladding design combining monolithic SiC and SiC–SiC composite layers could offer a tough, hermetic structure to provide improved performance and safety, with a failure rate comparable to current Zircaloy cladding. Modeling and design efforts require a thorough understanding of the properties and structure of SiC-based cladding. Furthermore, both fabrication and characterization of long, thin-walled SiC–SiC tubes to meet application requirements are challenging. In this work, mechanical and thermal properties of unirradiated, as-fabricated SiC-based cladding structures were measured, and permeability and dimensional control were assessed. In order to account for the tubular geometry of the cladding designs, development and modification of several characterization methods were required.

  1. Nuclear fuel cycle requirements and supply considerations, through the long-term

    International Nuclear Information System (INIS)

    1978-02-01

    The OECD Nuclear Energy Agency and the International Atomic Energy Agency have for many years published a joint report entitled ''Uranium Resources, Production and Demand'', and a revised edition of this work, dated December 1977, is now available. This report, on the other hand, is the result of a separate study of the supply and demand outlook for all fuel cycle services, as well as for uranium, through the long-term. The work was undertaken by the Nuclear Energy Agency's Working Party on Uranium Demand, whose members are listed in Appendix III. The intent here has been to contribute to the orderly development of nuclear power, by: 1. identifying potential problems in the supply of uranium and fuel cycle services, and possible areas for international co-operation in the resolution of such problems; 2. examining several long-range scenarios to determine the comparative needs of advanced reactors for uranium and for supporting services, thereby establishing the basis for the further development of uranium resources and specific reactor systems; and 3. assisting those having responsibilities in planning, forecasting, and programme management. This report is the work of a group of technical experts and does not necessarily reflect official policy or endorsement of the report's projections and conclusions by the Member Governments of the Nuclear Energy Agency

  2. Nondestructive examination techniques on Candu fuel elements

    International Nuclear Information System (INIS)

    Gheorghe, G.; Man, I.

    2013-01-01

    During irradiation in nuclear reactor, fuel elements undergo dimensional and structural changes, and changes of surface conditions sheath as well, which can lead to damages and even loss of integrity. Visual examination and photography of Candu fuel elements are among the non-destructive examination techniques, next to dimensional measurements that include profiling (diameter, bending, camber) and length, sheath integrity control with eddy currents, measurement of the oxide layer thickness by eddy current techniques. Unirradiated Zircaloy-4 tubes were used for calibration purposes, whereas irradiated Zircaloy-4 tubes were actually subjected to visual inspection and dimensional measurements. We present results of measurements done by eddy current techniques on Zircaloy- 4 tubes, unirradiated, but oxidized in an autoclave prior to examinations. The purpose of these nondestructive examination techniques is to determine those parameters that characterize the behavior and performance of nuclear fuel operation. (authors)

  3. Proposed Rule for Control of Air Pollution From New Motor Vehicles: Proposed Heavy-Duty Engine and Vehicle Standards and Highway Diesel Fuel Sulfur Control Requirements

    Science.gov (United States)

    Rule summary, CFR citations and additional resources concerning proposed new emission standards that will begin to take effect in 2007 and corresponding diesel fuel requirements that take effect in 2006.

  4. Quality assurance requirements for control of procurement items and services for nuclear fuel reprocessing facilities

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    Requirements and guidelines are provided for the control of activities to be exercised during procurement of items and services which affect the quality of nuclear facilities. These requirements and guidelines apply to procurement activities for items and services such as designing, purchasing, fabricating, handling, shipping, storing, cleaning, constructing, erecting, installing, inspecting, texting, maintaining and modifying

  5. Proof tests of irradiated and unirradiated EBR-II subassembly ducts

    International Nuclear Information System (INIS)

    Ruther, W.E.; Chopra, P.S.; Lambert, J.D.B.

    1977-01-01

    A series of dynamic pressure tests have been conducted within EBR-II subassembly ducts. The tests were designed to simulate bursting of a driver-fuel element in a cluster of such elements at their burnup limit during off-normal conditions in EBR-II. The major objective of the tests was to assure that such failure, which might cause rapid release of stored fission gas, would not deform or otherwise damage subassembly ducts in a way that would hinder movement of a control rod. The test results are described

  6. Review of research on non-destructive analysis of reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Ruby, S L; Gardner, D G [Westinghouse Electric Corp. (United States). Atomic Power Dept.

    1959-01-01

    This report contains a selected bibliography on the nondestructive assay of irradiated and unirradiated reactor fuel assemblies. The report also includes a discussion and evaluation of these papers in the light of the problems presented by variations in enrichment, alloying, cladding, geometrical arrangement, etc., among the many kinds of fuel assemblies. Suggestions for further development work are made. (author).

  7. Reviewing reactor engineering and fuel handling

    International Nuclear Information System (INIS)

    1991-12-01

    Experience has shown that the better operating nuclear power plants have well defined and effectively administered policies and procedures for governing reactor engineering and fuel handling (RE and FH) activities. This document provides supplementary guidance to OSART experts for evaluating the RE and FH programmes and activities at a nuclear power plant and assessing their effectiveness and adequacy. It is in no way intended to conflict with existing regulations and rules, but rather to exemplify those characteristics and features that are desirable for an effective, well structured RE and FH programme. This supplementary guidance addresses those aspects of RE and FH activities that are required in order to ensure optimum core operation for a nuclear reactor without compromising the limits imposed by the design, safety considerations of the nuclear fuel. In the context of this document, reactor engineering refers to those activities associated with in-core fuel and reactivity management, whereas fuel handling refers to the movement, storage, control and accountability of unirradiated and irradiated fuel. The document comprises five main sections and several appendices. In Section 2 of this guide, the essential aspects of an effective RE and FH programme are discussed. In Section 3, the various types of documents and reference materials needed for the preparatory work and investigation are listed. In Section 4, specific guidelines for investigation of RE and FH programmes are presented. In Section 5, the essential attributes of an excellent RE and FH programme are listed. The supplementary guidance is concluded with a series of appendices exemplifying the various qualities and attributes of a sound, well defined RE and FH programme

  8. Estimating indigenous resources for fuel-wood and poles and plantation requirements in the tribal trust lands of Zimbabwe Rhodesia

    Energy Technology Data Exchange (ETDEWEB)

    Furness, C K

    1981-01-01

    The difficulties encountered in planning for the conservation of indigenous timber resources and in estimating the timber consumption in tribal trust land are outlined in this paper. An estimate of these resources and of the consumption of timber, together with an estimate of exotic plantations required to make up any shortfall of timber, is given. Some 66,000 ha of eucalypts are currently required in the tribal trust lands, where planting has thus far provided only 3800 ha. The types of plantations established and the species used are mentioned. The rural population has, generally speaking, shown only limited enthusiasm for growing exotics, one of the reasons being the traditional use of indigenous timber which is still available in most areas without cost, and the preference for indigenous timber compared to eucalypts. The need for more reliable data for future planning is emphasized. Substitutes for fuel-wood are discussed and the need to reserve areas of indigenous timber in tribal trust land for the protection of soil and water and for fuel-wood are proposed. (Refs. 1).

  9. Fuel processing

    International Nuclear Information System (INIS)

    Allardice, R.H.

    1990-01-01

    The technical and economic viability of the fast breeder reactor as an electricity generating system depends not only upon the reactor performance but also on a capability to recycle plutonium efficiently, reliably and economically through the reactor and fuel cycle facilities. Thus the fuel cycle is an integral and essential part of the system. Fuel cycle research and development has focused on demonstrating that the challenging technical requirements of processing plutonium fuel could be met and that the sometimes conflicting requirements of the fuel developer, fuel fabricator and fuel reprocessor could be reconciled. Pilot plant operation and development and design studies have established both the technical and economic feasibility of the fuel cycle but scope for further improvement exists through process intensification and flowsheet optimization. These objectives and the increasing processing demands made by the continuing improvement to fuel design and irradiation performance provide an incentive for continuing fuel cycle development work. (author)

  10. Special Analysis for the Disposal of the INL Waste Associated with the Unirradiated Light Water Breeder Reactor (LWBR) Waste Stream at the Area 5 Radioactive Waste Management Site

    Energy Technology Data Exchange (ETDEWEB)

    Shott, Gregory [National Security Technologies, LLC, Las Vegas, NV (United States)

    2017-03-21

    This special analysis (SA) evaluates whether the Idaho National Laboratory (INL) Waste Associated with the Unirradiated Light Water Breeder Reactor (LWBR) waste stream (INEL167203QR1, Revision 0) is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the INL Waste Associated with the Unirradiated LWBR waste meets all U.S. Department of Energy (DOE) Manual DOE M 435.1-1, “Radioactive Waste Management Manual,” Chapter IV, Section P performance objectives (DOE 1999). The INL Waste Associated with the Unirradiated LWBR waste stream is recommended for acceptance with the condition that the total uranium-233 (233U) inventory be limited to 2.7E13 Bq (7.2E2 Ci).

  11. U.S. petroleum refining: Meeting requirements for cleaner fuels and refineries

    International Nuclear Information System (INIS)

    Warden, R.B.

    1993-01-01

    A review is presented of a study carried out by the National Petroleum Council that assessed the ability of the U.S. oil industry to manufacture and supply the quantity and quality of products required in the 1990s and beyond. The competitiveness of domestic supply vs product imports was analyzed and the investment requirements and other costs associated with meeting new environmental legislation and regulations on petroleum products and refineries were addressed. In particular, the requirements of the 1990 Clean Air Act Amendments and other environmental, health, and safety initiatives, both current and prospective, were evaluated. Refineries and the logistics system were studied but not crude oil supply or service stations. The costs of stationary source health, safety and environmental regulations and requirements were evaluated for the years 1995, 2000 and 2010, and sources of U.S. light products and U.S. refinery utilization were modelled for these years. Three demand scenarios were considered: growth, no-growth, and decline. Annual expenditures for health, safety and environment programs inside refineries are expected to double in the 1990s. Expenditures of $106 billion are projected over the period 1990-2010 for new facilities and programs necessary for current and anticipated stationary source regulations. Refining and logistics costs will increase substantially. Other conclusions related to capital expenditures, refining capability, product compatibility, oxygenates and foreign product supply cost are drawn. 26 figs

  12. 40 CFR 1060.102 - What permeation emission control requirements apply for fuel lines?

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false What permeation emission control... AGENCY (CONTINUED) AIR POLLUTION CONTROLS CONTROL OF EVAPORATIVE EMISSIONS FROM NEW AND IN-USE NONROAD AND STATIONARY EQUIPMENT Emission Standards and Related Requirements § 1060.102 What permeation...

  13. 40 CFR 1060.103 - What permeation emission control requirements apply for fuel tanks?

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false What permeation emission control... AGENCY (CONTINUED) AIR POLLUTION CONTROLS CONTROL OF EVAPORATIVE EMISSIONS FROM NEW AND IN-USE NONROAD AND STATIONARY EQUIPMENT Emission Standards and Related Requirements § 1060.103 What permeation...

  14. An evaluation model for the definition of regulatory requirements on spent fuel pool cooling systems

    International Nuclear Information System (INIS)

    Izquierdo, J.M.

    1979-01-01

    A calculation model is presented for establishing regulatory requirements in the SFPCS System. The major design factors, regulatory and design limits and key parameters are discussed. A regulatory position for internal use is proposed. Finally, associated problems and experience are presented. (author)

  15. Application profile and requirements on fuel cell heat-and-power stations; Einsatzprofil und Anforderungen an Brennstoffzellen-HKW

    Energy Technology Data Exchange (ETDEWEB)

    Kruhl, J

    1999-12-31

    In centralized power supply systems, maximum electrical efficiency is required in order to achieve competitiveness with high-efficiency combined-cycle systems. Combined systems of high-temperature fuel cells (MCFC/SOFC) with integrated gas or steam turbine processes are a promising option. In decentralized power supply, low-temperature and medium-temperature fuel cells (PEMFC/PAFC) may be feasible as well provided that the temperature characteristic of the demand side permits the use of the heat generated. In view of the decreasing heat demand of classic cogeneration structures, costly district heating will in the long run be replaced by building-specific cogeneration solutions. In this field, high-efficiency fuel cell systems have advantages over small-scale cogeneration units. If power-specific investments are low, the new technology may be more favourable than separate supply systems with networks and heating boilers and, in consequence, may gain wider acceptance. (orig.) [Deutsch] In der zentralen Stromerzeugung erfordert der Wettbewerb mit hocheffizienten GuD-Anlagen die Maximierung des elektrischen Wirkungsgrades. Hierzu bieten sich Kombianlagen aus Hochtemperatur-Brennstoffzellen (MCFC/SOFC) mit integrierten Dampf- bzw. Gasturbinen-Prozessen an. Im dezentralen Versorgungsbereich bieten zusaetzlich Nieder- und Mitteltemperatur-Brennstoffzellen (PEMFC/PAFC) ein hohes Einsatzpotential, sofern die Temperaturcharakteristik des Bedarfs eine Nutzbarkeit der entstehenden Waerme zulaesst. Unter dem Aspekt des zukuenftig sinkenden Waermebedarfs klassischer KWK-Strukturen wird die kostenintensive Nahwaermeversorgung zunehmend durch gebaeudespezifische KWK-Loesungen verdraengt werden. In diesem Bereich bieten hocheffiziente Brennstoffzellen-Systeme deutliche Vorteile gegenueber motorischen Klein-BHKW. Werden hier geringe leistungsspezifische Investitionen erreicht, kann eine Wirtschaftlichkeit gegenueber der getrennten Versorgung durch Netz und Heizkessel gegeben sein

  16. Comparison of the growth of Listeria monocytogenes in unirradiated and irradiated cook-chill roast beef and gravy at refrigeration temperatures

    International Nuclear Information System (INIS)

    Grant, I.R.; Nixon, C.R.; Patterson, M.F.

    1993-01-01

    Specific growth rates of two strains of Listeria monocytogenes in unirradiated and irradiated (2 kGy) roast beef and gravy stored at 5° and 10°C were found to be similar. However, exponential growth of L. monocytogenes after irradiation was preceded by an extended lag period of 6–9 d at 5°C and 3–4 d at 10°C, compared with lag periods of 1–2 d and <0.1 d in unirradiated beef and gravy stored similarly

  17. Challenges for Incorporation of additional safeguards requirements in a fuel fabrication facility in Japan

    International Nuclear Information System (INIS)

    Ishikawa, Tadatsugu; Suzuki, Katsuyuki

    2004-01-01

    Recent introduction of strengthened SG measure (i.e.: SNRI (Short Notice Random Inspection)) necessitated semi-real-time reporting including non-prescribed data in the FA for vital processes. These requirements have been bet by endeavors of both Inspectorates and operators. Using the integrated databases it will also be possible to produce most of reports required for safeguards purposes on a semi-real time basis, albeit with additional investment for softwares. However, in order to fully enjoy the advancements of information technology it is necessary to streamline the legal and procedural platform in addition to technical matters, in particular what are 'must' for safeguards and abolish if they are not really needed anymore. (author)

  18. Fuel corrosion processes under waste disposal conditions

    International Nuclear Information System (INIS)

    Shoesmith, D.W.

    2000-01-01

    The release of the majority of radionuclides from spent nuclear fuel under permanent disposal conditions will be controlled by the rate of dissolution of the UO 2 fuel matrix. In this manuscript the mechanism of the coupled anodic (fuel dissolution) and cathodic (oxidant reduction) reactions which constitute the overall fuel corrosion process is reviewed, and the many published observations on fuel corrosion under disposal conditions discussed. The primary emphasis is on summarizing the overall mechanistic behaviour and establishing the primary factors likely to control fuel corrosion. Included are discussions on the influence of various oxidants including radiolytic ones, pH, temperature, groundwater composition, and the formation of corrosion product deposits. The relevance of the data recorded on unirradiated UO 2 to the interpretation of spent fuel behaviour is included. Based on the review, the data used to develop fuel corrosion models under the conditions anticipated in Yucca Mountain (NV, USA) are evaluated

  19. Theoretical requirements to tolerances to be imposed on fuel rod design parameters for RBEC-M lead-bismuth fast reactor

    International Nuclear Information System (INIS)

    Vasiliev, A.; Alekseev, P.; Mikityuk, K.; Fomichenko, P.; Shestopalov, A.

    2002-01-01

    Development of advanced reactors with innovative materials requires comprehensive analysis of fuel rod design parameters as well as tolerances to be imposed on these parameters. Currently, it is considered traditional to estimate uncertainties in core neutronics parameters on the basis of known tolerances imposed on fuel rod design parameters. However, requirements to some core neutronics parameters of advanced reactors can be first formulated and then taken into account, while developing the technologies for innovative fuel rod manufacturing, i.e. an 'inverse' problem can be solved. The aim of this problem is to find combinations of fuel rod design tolerances which provide that selected core neutronics parameters remain within specified deviations during base irradiation. (authors)

  20. Comparative requirements for electric energy for production of hydrogen fuel and/or recharging of battery electric automobile fleets in New Zealand and the United States

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Paul [Stanford University, Stanford, CA 94305 (United States); Leaver, Jonathan D. [Department of Civil Engineering, Unitec NZ, Auckland 1142 (New Zealand)

    2010-10-15

    Within the current outlook for sustainable electric energy supply with concomitant reduction in emission of greenhouse gases, accelerated attention is focusing on the long-term development of hydrogen fuel cell and all-electric battery vehicles to provide alternative fuels to replace petroleum-derived fuels for automotive national fleets. The potential varies significantly between large industrially developed nations and smaller industrially developing nations. The requirement for additional electric energy supply from low-specific energy renewable resources and high-specific energy nuclear resources depends strongly on individual national economic, environmental, and political factors. Analysis of the additional electric energy supply required for the two potential large-scale technologies for fueling future national transportation sectors is compared for a large Organization for Economic Co-operation and Development (OECD) nation (USA) with a small OECD nation (New Zealand), normalized on a per-capita basis. (author)

  1. Toward a risk assessment of the spent fuel and high-level nuclear waste disposal system. Risk assessment requirements, literature review, methods evaluation: an interim report

    International Nuclear Information System (INIS)

    Hamilton, L.D.; Hill, D.; Rowe, M.D.; Stern, E.

    1986-04-01

    This report provides background information for a risk assessment of the disposal system for spent nuclear fuel and high-level radioactive waste (HLW). It contains a literature review, a survey of the statutory requirements for risk assessment, and a preliminary evaluation of methods. The literature review outlines the state of knowledge of risk assessment and accident consequence analysis in the nuclear fuel cycle and its applicability to spent fuel and HLW disposal. The survey of statutory requirements determines the extent to which risk assessment may be needed in development of the waste-disposal system. The evaluation of methods reviews and evaluates merits and applicabilities of alternative methods for assessing risks and relates them to the problems of spent fuel and HLW disposal. 99 refs

  2. Toward a risk assessment of the spent fuel and high-level nuclear waste disposal system. Risk assessment requirements, literature review, methods evaluation: an interim report

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, L.D.; Hill, D.; Rowe, M.D.; Stern, E.

    1986-04-01

    This report provides background information for a risk assessment of the disposal system for spent nuclear fuel and high-level radioactive waste (HLW). It contains a literature review, a survey of the statutory requirements for risk assessment, and a preliminary evaluation of methods. The literature review outlines the state of knowledge of risk assessment and accident consequence analysis in the nuclear fuel cycle and its applicability to spent fuel and HLW disposal. The survey of statutory requirements determines the extent to which risk assessment may be needed in development of the waste-disposal system. The evaluation of methods reviews and evaluates merits and applicabilities of alternative methods for assessing risks and relates them to the problems of spent fuel and HLW disposal. 99 refs.

  3. Assessment of the requirements for placing and maintaining Savannah River Site spent fuel storage basins under International Atomic Energy Agency safeguards

    International Nuclear Information System (INIS)

    Amacker, O.P. Jr.; Curtis, M.M.; Delegard, C.H.; Hsue, S.T.; Whitesel, R.N.

    1997-03-01

    The United States is considering the offer of irradiated research reactor spent fuel (RRSF) for international safeguards applied by the International Atomic Energy Agency (IAEA). The offer would be to add one or more spent fuel storage basins to the list of facilities eligible for IAEA safeguards. The fuel to be safeguarded would be stored in basins on the Savannah River Site (SRS). This RRSF potentially can include returns of Material Test Reactor (MTR) VAX fuel from Argentina, Brazil, and Chile (ABC); returns from other foreign research reactors; and fuel from domestic research reactors. Basins on the SRS being considered for this fuel storage are the Receiving Basin for Offsite Fuel (RBOF) and the L-Area Disassembly Basin (L-Basin). A working group of SRS, U.S. Department of Energy International Safeguards Division (NN-44), and National Laboratory personnel with experience in IAEA safeguards was convened to consider the requirements for applying the safeguards to this material. The working group projected the safeguards requirements and described alternatives

  4. A study on items necessary to develop the requirements for the management of serious accidents postulated in fuel fabrication, enrichment and reprocessing facilities

    International Nuclear Information System (INIS)

    Takanashi, Mitsuhiro; Yamate, Kazuki; Asada, Kazuo; Yamada, Takashi; Endo, Shigeki

    2013-05-01

    The purpose of this study is to supply the points to discuss on new rules of fuel fabrication, enrichment and reprocessing facilities (hereinafter referred to as 'fuel cycle facilities') conducted by Nuclear Regulation Authority. Requirements for management of serious accidents in the fuel cycle facilities were summarized in this study. Taking into account the lessons learned from the accident of TEPCO Fukushima Daiichi Nuclear Power Plant in Mar. 2011, Act for the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors was amended in June 2012. The main items of the amendment were as follows: Preparation for the management of serious accidents, Introduction of evaluation system for safety improvement, Application of new standards to existing nuclear facilities (back-fitting). Japan Nuclear Energy Safety organization (JNES) conducted a fundamental study on serious accidents and their management in the fuel cycle facilities and made a report. In the report, the concept of Defense in Depth and the definition of serious accidents for the fuel cycle facilities were discussed. Those discussions were conducted by reference to new regulation rules (draft) for power reactors and from the view of features of the fuel cycle facilities. However, further detailed studies are necessary in order to clarify some issues in it. It was also reflected opinions from experts in JNES technical meetings on accident management of the fuel cycle facilities to brush up this report. (author)

  5. Comparison of Commercial Aircraft Fuel Requirements in Regards to FAR, Flight Profile Simulation, and Flight Operational Techniques

    Science.gov (United States)

    Heitzman, Nicholas

    There are significant fuel consumption consequences for non-optimal flight operations. This study is intended to analyze and highlight areas of interest that affect fuel consumption in typical flight operations. By gathering information from actual flight operators (pilots, dispatch, performance engineers, and air traffic controllers), real performance issues can be addressed and analyzed. A series of interviews were performed with various individuals in the industry and organizations. The wide range of insight directed this study to focus on FAA regulations, airline policy, the ATC system, weather, and flight planning. The goal is to highlight where operational performance differs from design intent in order to better connect optimization with actual flight operations. After further investigation and consensus from the experienced participants, the FAA regulations do not need any serious attention until newer technologies and capabilities are implemented. The ATC system is severely out of date and is one of the largest limiting factors in current flight operations. Although participants are pessimistic about its timely implementation, the FAA's NextGen program for a future National Airspace System should help improve the efficiency of flight operations. This includes situational awareness, weather monitoring, communication, information management, optimized routing, and cleaner flight profiles like Required Navigation Performance (RNP) and Continuous Descent Approach (CDA). Working off the interview results, trade-studies were performed using an in-house flight profile simulation of a Boeing 737-300, integrating NASA legacy codes EDET and NPSS with a custom written mission performance and point-performance "Skymap" calculator. From these trade-studies, it was found that certain flight conditions affect flight operations more than others. With weather, traffic, and unforeseeable risks, flight planning is still limited by its high level of precaution. From this

  6. 40 CFR 80.590 - What are the product transfer document requirements for motor vehicle diesel fuel, NRLM diesel...

    Science.gov (United States)

    2010-07-01

    ... distillates? 80.590 Section 80.590 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR... oil, ECA marine fuel, and other distillates? (a) This paragraph (a) applies on each occasion that any... distillates used or intended to be used as MVNRLM diesel fuel, heating oil, or ECA marine fuel) except when...

  7. Seismic procurement requirements at the FPR (Fuel Processing Restoration) facility at INEL (Idaho National Engineering Laboratory)

    International Nuclear Information System (INIS)

    Bingham, G.E.; Hardy, G.S.; Griffin, M.J.

    1989-01-01

    Traditional methods used to seismically qualify equipment for new facilities has been either by testing or analysis. Testing programs are generally expensive and their input loadings are conservative. It is also generally recognized that standard seismic analysis techniques produce overly conservative results. Seismic loads and response levels for equipment are typically calculated that far exceed the values actually experienced in earthquakes. A more efficient method for demonstrating the seismic adequacy of equipment has been developed which is based on conclusions derived from studying the performance of equipment that has been subjected to actual earthquake excitations. The earthquake experience data concludes that damage or malfunction to most types of equipment subjected to earthquakes is far less than that predicted by traditional testing and analysis techniques. The use of conclusions derived from experience data provides a more realistic approach in assessing the seismic ruggedness of equipment. By recognizing this inherently higher capacity that exists in specific classes of equipment, vendors can often supply off the shelf equipment without the need to perform expensive modifications to meet requirements imposed by conservative qualification analyses. This paper will describe the development of the experienced based method for equipment seismic qualification and its application at the FPR facility

  8. Radiation quality-dependence of bystander effect in unirradiated fibroblasts is associated with TGF-β1-Smad2 pathway and miR-21 in irradiated keratinocytes

    Science.gov (United States)

    Yin, Xiaoming; Tian, Wenqian; Wang, Longxiao; Wang, Jingdong; Zhang, Shuyu; Cao, Jianping; Yang, Hongying

    2015-01-01

    Traditional radiation biology states that radiation causes damage only in cells traversed by ionizing radiation. But radiation-induced bystander effect (RIBE), which refers to the biological responses in unirradiated cells when the neighboring cells are exposed to radiation, challenged this old dogma and has become a new paradigm of this field. By nature, RIBEs are the consequences of intercellular communication between irradiated and unirradiated cells. However, there are still some important questions remain unanswered such as whether RIBE is dependent on radiation quality, what are the determining factors if so, etc. Using a transwell co-culture system, we found that HaCaT keratinocytes irradiated with α-particles but not X-rays could induce bystander micronucleus formation in unirradiated WS1 fibroblasts after co-culture. More importantly, the activation of TGF-β1-Smad2 pathway and the consistent decrease of miR-21 level in α-irradiated HaCaT cells were essential to the micronucleus induction in bystander WS1 cells. On the other hand, X-irradiation did not induce bystander effect in unirradiated WS1 cells, accompanied by lack of Smad2 activation and consistent decrease of miR-21 in X-irradiated HaCaT cells. Taken together, these results suggest that the radiation quality-dependence of bystander effect may be associated with the TGF-β1-Smad2 pathway and miR-21 in irradiated cells. PMID:26080011

  9. 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.

  10. Investigation of Nuclide Importance to Functional Requirements Related to Transport and Long-Term Storage of LWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Broadhead, B.L.

    1995-01-01

    The radionuclide characteristics of light-water-reactor (LWR) spent fuel play key roles in the design and licensing activities for radioactive waste transportation systems, interim storage facilities, and the final repository site. Several areas of analysis require detailed information concerning the time-dependent behavior of radioactive nuclides including (1) neutron/gamma-ray sources for shielding studies, (2) fissile/absorber concentrations for criticality safety determinations, (3) residual decay heat predictions for thermal considerations, and (4) curie and/or radiological toxicity levels for materials assumed to be released into the ground/environment after long periods of time. The crucial nature of the radionuclide predictions over both short and long periods of time has resulted in an increased emphasis on thorough validation for radionuclide generation/depletion codes. Current radionuclide generation/depletion codes have the capability to follow the evolution of some 1600 isotopes during both irradiation and decay time periods. Of these, typically only 10 to 20 nuclides dominate contributions to each analysis area. Thus a quantitative ranking of nuclides over various time periods is desired for each of the analysis areas of shielding, criticality, heat transfer, and environmental dose (radiological toxicity). These rankings should allow for validation and data improvement efforts to be focused only on the most important nuclides. This study investigates the relative importances of the various actinide, fission-product, and light-element isotopes associated with LWR spent fuel with respect to five analysis areas: criticality safety (absorption fractions), shielding (dose rate fractions), curies (fractional curies levels), decay heat (fraction of total watts), and radiological toxicity (fraction of potential committed effective dose equivalent). These rankings are presented for up to six different burnup/enrichment scenarios and at decay times from 2 to

  11. Requirements for a Dynamic Solvent Extraction Module to Support Development of Advanced Technologies for the Recycle of Used Nuclear Fuel

    International Nuclear Information System (INIS)

    Law, Jack; Rutledge, Veronica; Pereira, Candido; Copple, Jackie; Frey, Kurt; Krebs, John; Maggos, Laura; Nichols, Kevin; Wardle, Kent; Sadasivan, Pratap; DeAlmieda, Valmor; Depaoli, David

    2011-01-01

    The Department of Energy's Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program has been established to create and deploy next generation, verified and validated nuclear energy modeling and simulation capabilities for the design, implementation, and operation of future nuclear energy systems to improve the U.S. energy security. As part of the NEAMS program, Integrated Performance and Safety Codes (IPSC's) are being produced to significantly advance the status of modeling and simulation of energy systems beyond what is currently available to the extent that the new codes be readily functional in the short term and extensible in the longer term. The four IPSC areas include Safeguards and Separations, Reactors, Fuels, and Waste Forms. As part of the Safeguards and Separations (SafeSeps) IPSC effort, interoperable process models are being developed that enable dynamic simulation of an advanced separations plant. A SafeSepss IPSC 'toolkit' is in development to enable the integration of separation process modules and safeguards tools into the design process by providing an environment to compose, verify and validate a simulation application to be used for analysis of various plant configurations and operating conditions. The modules of this toolkit will be implemented on a modern, expandable architecture with the flexibility to explore and evaluate a wide range of process options while preserving their stand-alone usability. Modules implemented at the plant-level will initially incorporate relatively simple representations for each process through a reduced modeling approach. Final versions will incorporate the capability to bridge to subscale models to provide required fidelity in chemical and physical processes. A dynamic solvent extraction model and its module implementation are needed to support the development of this integrated plant model. As a stand-alone application, it will also support solvent development of extraction flowsheets and integrated

  12. Requirements for a Dynamic Solvent Extraction Module to Support Development of Advanced Technologies for the Recycle of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jack Law; Veronica Rutledge; Candido Pereira; Jackie Copple; Kurt Frey; John Krebs; Laura Maggos; Kevin Nichols; Kent Wardle; Pratap Sadasivan; Valmor DeAlmieda; David Depaoli

    2011-06-01

    The Department of Energy's Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program has been established to create and deploy next generation, verified and validated nuclear energy modeling and simulation capabilities for the design, implementation, and operation of future nuclear energy systems to improve the U.S. energy security. As part of the NEAMS program, Integrated Performance and Safety Codes (IPSC's) are being produced to significantly advance the status of modeling and simulation of energy systems beyond what is currently available to the extent that the new codes be readily functional in the short term and extensible in the longer term. The four IPSC areas include Safeguards and Separations, Reactors, Fuels, and Waste Forms. As part of the Safeguards and Separations (SafeSeps) IPSC effort, interoperable process models are being developed that enable dynamic simulation of an advanced separations plant. A SafeSepss IPSC 'toolkit' is in development to enable the integration of separation process modules and safeguards tools into the design process by providing an environment to compose, verify and validate a simulation application to be used for analysis of various plant configurations and operating conditions. The modules of this toolkit will be implemented on a modern, expandable architecture with the flexibility to explore and evaluate a wide range of process options while preserving their stand-alone usability. Modules implemented at the plant-level will initially incorporate relatively simple representations for each process through a reduced modeling approach. Final versions will incorporate the capability to bridge to subscale models to provide required fidelity in chemical and physical processes. A dynamic solvent extraction model and its module implementation are needed to support the development of this integrated plant model. As a stand-alone application, it will also support solvent development of extraction flowsheets

  13. The Analysis of the General Performance and Mechanical Behavior of Unirradiated FeCrAl Alloys Before and After Welding

    Energy Technology Data Exchange (ETDEWEB)

    Gussev, Maxim N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-06-03

    The present report summarizes and discusses the preliminary results for the in-depth characterization of the modern, nuclear-grade FeCrAl alloys currently under development. The alloys were designed for enhanced radiation tolerance and weldability, and the research is currently being pursued by the Department of Energy (DOE) Nuclear Energy Enabling Technologies (NEET) program. Last year, seven candidate FeCrAl alloys with well-controlled chemistry and microstructures were designed and produced; welding was performed under well-controlled conditions. The structure and general performance of unirradiated alloys were assessed using standardized and advanced microstructural characterization techniques and mechanical testing. The primary objective is to identify the best candidate alloy, or at a minimum to identify the contributing factors that increase the weldability and radiation tolerance of FeCrAl alloys, therefore enabling future generations of FeCrAl alloys to deliver better performance parameters. This report is structured so as to describe these critical assessments of the weldability; radiation tolerance will be reported on in later reports from this program.

  14. Semi-conservative deoxyribonucleic acid synthesis in unirradiated and ultraviolet-irradiated xeroderma pigmentosum and normal human skin fibroblasts

    International Nuclear Information System (INIS)

    Rude, J.M.; Friedberg, E.C.

    1977-01-01

    Rates of semiconservative DNA synthesis have been investigated in asynchronous xeroderma pigmentosum (XP), XP variant, and normal human skin fibroblasts using the technique of cellular autoradiography. In unirradiated cells, no differences in DNA synthesis rates were detected among the three cell strains. Exposure to UV radiation caused the rate of DNA synthesis to decrease for at least three hours in all three cell strains. In the normal cell strain, recovery of the DNA synthetic rate occurred at later times following a UV fluence of 5 J/m 2 . At this same UV fluence, recovery was absent in classical XP cells during a 24 h post-irradiation period while it was slower than normal in XP variant cells. When the UV fluence to classical XP and XP variant cells was reduced so that survival in all three cell strains was approximately the same (25%), recovery of the DNA synthetic rate was similar in all three cell strains. These results are discussed in terms of current models of DNA replication in UV-irradiated cells and indicate: (1) that pyrimidine dimers are very effective blocks to DNA synthesis and (2) that there is no inherent defect in semiconservative DNA synthesis in either classical XP or XP variant cells which is independent of a defect in DNA repair capacity

  15. Semi-conservative deoxyribonucleic acid synthesis in unirradiated and ultraviolet-irradiated xeroderma pigmentosum and normal human skin fibroblasts

    Energy Technology Data Exchange (ETDEWEB)

    Rude' e, J.M.; Friedberg, E.C.

    1977-03-01

    Rates of semiconservative DNA synthesis have been investigated in asynchronous xeroderma pigmentosum (XP), XP variant, and normal human skin fibroblasts using the technique of cellular autoradiography. In unirradiated cells, no differences in DNA synthesis rates were detected among the three cell strains. Exposure to uv radiation caused the rate of DNA synthesis to decrease for at least three hours in all three cell strains. In the normal cell strain, recovery of the DNA synthetic rate occurred at later times following a uv fluence of 5 J/m2. At this same uv fluence, recovery was absent in classical XP cells during a 24 h post-irradiation period while it was slower than normal in XP variant cells. When the uv fluence to classical XP and XP variant cells was reduced so that survival in all three cell strains was approximately the same (25%), recovery of the DNA synthetic rate was similar in all three cell strains. These results are discussed in terms of current models of DNA replication in uv-irradiated cells and indicate: (1) that pyrimidine dimers are very effective blocks to DNA synthesis and (2) that there is no inherent defect in semi-conservative DNA synthesis in either classical XP or XP variant cells which is independent of a defect in DNA repair capacity.

  16. Tensile and electrical properties of unirradiated and irradiated Hycon 3HP{trademark} CuNiBe

    Energy Technology Data Exchange (ETDEWEB)

    Zinkle, S.J.; Eatherly, W.S. [Oak Ridge National Lab., TN (United States)

    1996-10-01

    The unirradiated tensile properties of two different heats of Hycon 3HP{trademark} CuNiBe (HT Temper) have been measured over the temperature range of 20-500{degrees}C for longitudinal and long transverse orientations. The room temperature electrical conductivity has also been measured for both heats. Both heats exhibited a very good combination of strength and conductivity at room temperature. The strength remained relatively high at all test temperatures, with a yield strength of 420-520 MPa at 500{degrees}C. However, low levels of ductility (<5% uniform elongation) were observed at test temperatures above 200-250{degrees}C, due to flow localization adjacent to grain boundaries. Fission neutron irradiation to a dose of {approximately}0.7 dpa at temperatures between 100 and 240{degrees}C produced a slight increase in strength and a significant decrease in ductility. The measured tensile elongation increased with increasing irradiation temperature, with a uniform elongation of {approximately}3.3% observed at 240{degrees}C. The electrical conductivity decreased slightly following irradiation, due to the presence of defect clusters and Ni, Zn, Co transmutation products. The data indicate that CuNiBe alloys have irradiated tensile and electrical properties comparable or superior to CuCrZr and oxide dispersion strengthened copper at temperatures <250{degrees}C, and may be suitable for certain fusion energy structural applications.

  17. Comparison of toxin production by clostridium botulinum type E in irradiated and unirradiated vacuum-packed trout (Salmo gairdneri)

    International Nuclear Information System (INIS)

    Hussain, M.; Ehlermann, D.; Diehl, J.F.

    1977-01-01

    Trouts obtained from a nearby Fish farm were slaughtered, gutted, cut into 100g samples and inoculated with 10 1 , 10 3 and 10 5 spores per g of Clostridium botulinum type E. The vacuum-packed samples were stored under melting ice (0 0 C) and at temperatures of 5 0 and 10 0 C for periods of up to 8 weeks. At weekly intervals, occurrence of spoilage and toxin production were determined. Only at 10 0 C storage, the irradiated samples showed toxin production before spoilage was observed. When the fishes were stored at 5 0 C, no toxicity occured before spoilage was observed even in samples treated with doses as high as 200 krad. Samples stored under melting ice, irradiated or unirradiated, never showed toxin production. It is concluded that the radurization of fish at doses of about 100 or 200 krad and at storage temperatures of melting ice or up to 5 0 C is safe with regard to a possible botulism risk. (orig.) [de

  18. Comparison of toxin production by clostridium botulinum type E in irradiated and unirradiated vacuum-packed trout (Salmo gairdneri)

    International Nuclear Information System (INIS)

    Hussain, A.M.; Ehlermann, D.; Diehl, J.F.

    1977-01-01

    Trouts obtained from a nearby fish farm were slaughtered, gutted, cut into 100 g samples and inoculated with 10 1 , 10 3 and 10 5 spores per g of Clostridium botulinum type E. The vacuum-packed samples were stored under melting ice (0 0 C) and at temperatures of 5 0 and 10 0 C for periods of up to 8 weeks. At weekly intervals, occurrence of spoilage and toxin production were determined. Only at 10 0 C storage, the irradiated samples showed toxin production before spoilage was observed. When the fishes were stored at 5 0 C, no toxicity occurred before spoilage was observed even in samples treated with doses as high as 200 krad. Samples stored under melting ice, irradiated or unirradiated, never showed toxin production. It is concluded that the radurization of fish at doses of about 100 or 200 krad and at storage temperatures of melting ice or up to 5 0 C is safe with regard to a possible botulism risk. (orig.) [de

  19. Nuclear fuels

    International Nuclear Information System (INIS)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

    2009-01-01

    , Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO 2 ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO 2 and MOX ceramics (Chromium oxide-doped UO 2 fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The nature of spent nuclear

  20. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    irradiation, Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO{sub 2} ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO{sub 2} and MOX ceramics (Chromium oxide-doped UO{sub 2} fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The

  1. Key natural analogue input required to build a safety case for direct disposal of spent nuclear fuel in Japan

    Energy Technology Data Exchange (ETDEWEB)

    McKinley, I.G.; Hardie, S.M.L.; Klein, E. [MCM Consulting, Baden-Dättwil (Switzerland); Kawamura, H. [Obayashi Corporation, Nuclear Facilities Division, Tokyo (Japan); Beattie, T.M. [MCM Consulting, Bristol (United Kingdom)

    2015-06-15

    Natural analogues have been previously used to support the safety case for direct disposal of spent nuclear fuel, but the focus of such work was very dependent on the key barriers of specific national disposal concepts. Investigations of the feasibility of such disposal in Japan are at an early stage but, nevertheless, it is clear that building a robust safety case will be very challenging and would benefit from focused support from natural analogue studies—both in terms of developing/testing required models and, as importantly, presenting safety arguments to a wide range of stakeholders. This paper identifies key analogues that support both longevity and spread of failure times of massive steel overpacks, the effectiveness of buffering of radiolytic oxidants and the chemical and physical mechanisms retarding release of radionuclides from the engineered barriers. It is concluded that, for countries like Japan where performance needs to be assessed as realistically as possible, natural analogues can complement the existing laboratory and theoretical knowledge base and contribute towards development of a robust safety case. (authors)

  2. 40 CFR 63.6604 - What fuel requirements must I meet if I own or operate an existing stationary CI RICE?

    Science.gov (United States)

    2010-07-01

    ... I own or operate an existing stationary CI RICE? 63.6604 Section 63.6604 Protection of Environment....6604 What fuel requirements must I meet if I own or operate an existing stationary CI RICE? If you own or operate an existing non-emergency CI stationary RICE with a site rating of more than 300 brake HP...

  3. 40 CFR 80.620 - What are the additional requirements for diesel fuel or distillates produced by foreign...

    Science.gov (United States)

    2010-07-01

    ... audits of the foreign refinery. (i) Inspections and audits may be either announced in advance by EPA, or... diesel fuel or distillate was produced, assurance that the diesel fuel or distillate remained segregated...: (i) Be approved in advance by EPA, based on a demonstration of ability to perform the procedures...

  4. Fuel rod analysis to respond to high burnup and demanding loading requirements. Probabilistic methodology recovers design margins narrowed by degrading fuel thermal conductivity and progressing FGR

    Energy Technology Data Exchange (ETDEWEB)

    Eberle, R; Heins, L; Sontheimer, F [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

    1997-08-01

    The proof that fuel rods will safely withstand all loads arising from inpile service conditions is generally achieved through the assessment of a number of design criteria by using a conservative analysis methodology in conjunction with design limits ``on the safe side``. The classical approach is the application of a fuel rod code to the Worst Case which is defined by the combination of most unfavorable conditions and assumptions with respect to the criterion under consideration. As it is evident that the deterministic construction of such Worst Cases imply an (unknown but) intuitively very high degree of conservatism, it is not surprising that this will develop to cause problems the more demanding fuel insertion conditions have to be anticipated (increased burnup, high efficiency loading schemes, etc.). A certain relief can be gained form cautious revisions of single design limits based on grown performance experience. But this increase of knowledge allows as well to change the established deterministic ``go/no-go`` conception into a better differentiating assessment methodology by which the quantification of the implied conservatism and the remaining design margins is possible: the Probabilistic Design Methodology (PDM). Principles and elements of the PDM are described. An essential prerequisite is a best-estimate fuel rod code which incorporates the latest state of knowledge about potential performance limiting phenomena (e.g. burnup degradation of fuel oxide thermal conductivity) as Siemens/KWU`s CARO-E does. An example is given how input distributions for rod data and model parameters transfer into a frequency distribution of maximum rod internal pressure, and indications are given how this is to be interpreted in view of a probabilistically re-formulated design criterion. The PDM provides a realistic conservative assessment of design criteria and will thus recover design margins for increasingly aggravated loading conditions. (author). 9 refs, 9 figs, 2 tabs.

  5. Fuel rod analysis to respond to high burnup and demanding loading requirements. Probabilistic methodology recovers design margins narrowed by degrading fuel thermal conductivity and progressing FGR

    International Nuclear Information System (INIS)

    Eberle, R.; Heins, L.; Sontheimer, F.

    1997-01-01

    The proof that fuel rods will safely withstand all loads arising from inpile service conditions is generally achieved through the assessment of a number of design criteria by using a conservative analysis methodology in conjunction with design limits ''on the safe side''. The classical approach is the application of a fuel rod code to the Worst Case which is defined by the combination of most unfavorable conditions and assumptions with respect to the criterion under consideration. As it is evident that the deterministic construction of such Worst Cases imply an (unknown but) intuitively very high degree of conservatism, it is not surprising that this will develop to cause problems the more demanding fuel insertion conditions have to be anticipated (increased burnup, high efficiency loading schemes, etc.). A certain relief can be gained form cautious revisions of single design limits based on grown performance experience. But this increase of knowledge allows as well to change the established deterministic ''go/no-go'' conception into a better differentiating assessment methodology by which the quantification of the implied conservatism and the remaining design margins is possible: the Probabilistic Design Methodology (PDM). Principles and elements of the PDM are described. An essential prerequisite is a best-estimate fuel rod code which incorporates the latest state of knowledge about potential performance limiting phenomena (e.g. burnup degradation of fuel oxide thermal conductivity) as Siemens/KWU's CARO-E does. An example is given how input distributions for rod data and model parameters transfer into a frequency distribution of maximum rod internal pressure, and indications are given how this is to be interpreted in view of a probabilistically re-formulated design criterion. The PDM provides a realistic conservative assessment of design criteria and will thus recover design margins for increasingly aggravated loading conditions. (author). 9 refs, 9 figs, 2 tabs

  6. Reconciling the expectations of community participants with the requirements of non-fossil fuel obligation: the experience of Harlock Hill windfarm

    International Nuclear Information System (INIS)

    Harrop, J.

    1998-01-01

    Is it possible to reconcile the aspirations of community participants in a wind energy project with the requirements imposed by the Non-Fossil Fuel Obligation legislation and procedure? This paper considers the practical experience of the framework that was adopted at Harlock Hill wind farm for community participation and the legal structure that were required to ensure that the project retained the full benefit of the premium price arrangements with the Non-Fossil Purchasing Agency Limited. (Author)

  7. Fuel gases

    International Nuclear Information System (INIS)

    Anon.

    1996-01-01

    This paper gives a brief presentation of the context, perspectives of production, specificities, and the conditions required for the development of NGV (Natural Gas for Vehicle) and LPG-f (Liquefied Petroleum Gas fuel) alternative fuels. After an historical presentation of 80 years of LPG evolution in vehicle fuels, a first part describes the economical and environmental advantages of gaseous alternative fuels (cleaner combustion, longer engines life, reduced noise pollution, greater natural gas reserves, lower political-economical petroleum dependence..). The second part gives a comparative cost and environmental evaluation between the available alternative fuels: bio-fuels, electric power and fuel gases, taking into account the processes and constraints involved in the production of these fuels. (J.S.)

  8. Effects of fabrication requirements on fuel performance in relation to operating conditions. The views of Electricite de France (EDF)

    International Nuclear Information System (INIS)

    Ponticq, M.; Richer, P.; Scribe, G.

    1979-01-01

    Because of the operational constraints relating to fuel behaviour, imperfect knowledge of the behaviour of a defective fuel assembly and, in the near future, the need to adapt reactor power to grid following (load following and remote control), EDF is aiming to reduce the present rate of fuel failure. While the phenomena affecting fuel behaviour have now been listed and analysed, the efforts at reducing their consequences have yet to be completed. This can be achieved, firstly, by reducing or eliminating fabrication defects, which are responsible for failure at the beginning of fuel life, through establishment of a good quality assurance organization and the search for still higher efficiency of quality control and fabrication equipment, and secondly, by developing fabrication techniques minimizing in particular cladding-pellet interactions and the stress corrosion of the cladding, which are responsible for fuel failure as from mid-life. However, the reactor operating conditions likely to apply in the near future may lead, for a given fuel configuration, to a re-evaluation of the fabrication parameters of cladding and UO 2 pellets. (author)

  9. Revised ANL-reported tensile data for unirradiated and irradiated (FFTF, HFIR) V-Ti and V-Cr-Ti alloys

    International Nuclear Information System (INIS)

    Billone, M.C.

    1998-01-01

    The tensile data for all unirradiated and irradiated vanadium alloys samples tested at Argonne National Laboratory (ANL) have been critically reviewed and, when necessary, revised. The review and revision are based on reanalyzing the original load-displacement strip chart recordings by a methodology consistent with current ASTM standards. For unirradiated alloys (162 samples), the revised values differ from the previous values as follows: -11±19 MPa (-4±6%) for yield strength (YS), -3±15 MPa (-1±3%) for ultimate tensile strength (UTS), -5±2% strain for uniform elongation (UE), and -4±2% strain for total elongation (TE). Of these changes, the decrease in -1±6 MPa (0±1%) for UTS, -5±2% for UE, and -4±2% for TE. Of these changes, the decrease in UE values for alloys irradiated and tested at 400--435 C is the most significant. This decrease results from the proper subtraction of nongauge-length deformation from measured crosshead deformation. In previous analysis of the tensile curves, the nongauge-length deformation was not correctly determined and subtracted from the crosshead displacement. The previously reported and revised tensile values for unirradiated alloys (20--700 C) are tabulated in Appendix A. The revised tensile values for the FFTF-irradiated (400--600 C) and HFIR-irradiated (400 C) alloys are tabulated in Appendix B, along with the neutron damage and helium levels. Appendix C compares the revised values to the previously reported values for irradiated alloys. Appendix D contains previous and revised values for the tensile properties of unirradiated V-5Cr-5Ti (BL-63) alloy exposed to oxygen

  10. The influence of late-stage pupal irradiation and increased irradiated: un-irradiated male ratio on mating competitiveness of the malaria mosquito Anopheles arabiensis Patton.

    Science.gov (United States)

    Helinski, M E H; Knols, B G J

    2009-06-01

    Competitiveness of released males in genetic control programmes is of critical importance. In this paper, we explored two scenarios to compensate for the loss of mating competitiveness after pupal stage irradiation in males of the malaria mosquito Anopheles arabiensis. First, competition experiments with a higher ratio of irradiated versus un-irradiated males were performed. Second, pupae were irradiated just prior to emergence and male mating competitiveness was determined. Males were irradiated in the pupal stage with a partially or fully-sterilizing dose of 70 or 120 Gy, respectively. Pupae were irradiated aged 20-26 h (young) as routinely performed, or the pupal stage was artificially prolonged by cooling and pupae were irradiated aged 42-48 h (old). Irradiated males competed at a ratio of 3:1:1 to un-irradiated males for mates in a large cage design. At the 3:1 ratio, the number of females inseminated by males irradiated with 70 Gy as young pupae was similar to the number inseminated by un-irradiated males for the majority of the replicates. At 120 Gy, significantly fewer females were inseminated by irradiated than by un-irradiated males. The irradiation of older pupae did not result in a significantly improved male mating competitiveness compared to the irradiation of young pupae. Our findings indicate that the loss of competitiveness after pupal stage irradiation can be compensated for by a threefold increase of irradiated males, but only for the partially-sterilizing dose. In addition, cooling might be a useful tool to facilitate handling processes of large numbers of mosquitoes in genetic control programmes.

  11. Tensile and fracture toughness properties of copper alloys and their HIP joints with austenitic stainless steel in unirradiated and neutron irradiated condition

    International Nuclear Information System (INIS)

    Taehtinen, S.; Pyykkoenen, M.; Singh, B.N.; Toft, P.

    1998-03-01

    The tensile strength and ductility of unirradiated CuAl25 IG0 and CuCrZr alloys decreased continuously with increasing temperature up to 350 deg C. Fracture toughness of unirradiated CuAl25 IG0 alloy decreased continuously with increasing temperature from 20 deg C to 350 deg C whereas the fracture toughness of unirradiated CuCrZr alloy remained almost constant at temperatures up to 100 deg C, was decreased significantly at 200 deg C and slightly increased at 350 deg C. Fracture toughness of HIP joints were lower than that of corresponding copper alloy and fracture path in HIP joint specimen was always within copper alloy side of the joint. Neutron irradiation to a dose level of 0.3 dpa resulted in hardening and reduction in uniform elongation to about 2-4% at 200 deg C in both copper alloys. At higher temperatures softening was observed and uniform elongation increased to about 5% and 16% for CuAl25 IG0 and CuCrZr alloys, respectively. Fracture toughness of CuAl25 IG0 alloy reduced markedly due to neutron irradiation in the temperature range from 20 deg C to 350 deg C. The fracture toughness of the irradiated CuCrZr alloy also decreased in the range from 20 deg C to 350 deg C, although it remained almost unaffected at temperatures below 200 deg C and decreased significantly at 350 deg C when compared with that of unirradiated CuCrZr alloy. (orig.)

  12. Reprocessing fuel from the Southwest Experimental Fast Oxide Reactor at the Savannah River Plant

    International Nuclear Information System (INIS)

    Gray, L.W.; Campbell, T.G.

    1985-11-01

    The irradiated fuel, reject fuel tubes, and fuel fabrication scrap from the Southwest Experimental Fast Oxide Reactor (SEFOR) were transferred to the Savannah River Plant (SRP) for uranium and plutonium recovery. The unirradiated material was declad and dissolved at SRP; dissolution was accomplished in concentrated nitric acid without the addition of fluoride. The irradiated fuel was declad at Atomics International and repacked in aluminum. The fuel and aluminum cans were dissolved at SRP using nitric acid catalyzed by mercuric nitrate. As this fuel was dissolved in nongeometrically favorable tanks, boron was used as a soluble neutron poison

  13. Neutron radiography of fuel pins

    International Nuclear Information System (INIS)

    Jackson, C.N. Jr.; Powers, H.G.; Burgess, C.A.

    1975-01-01

    Neutron radiography performed with a reactor source has been shown to be a superior radiographic method for the examination of unirradiated mixed oxide fuel pins at the Hanford Engineering Development Laboratory. Approximately 1,700 fuel pins were contained in a sample that demonstrated the capability of the method for detecting laminations, structural flaws, fissile density variation, hydrogenous inclusions and voids in assembled fuel pins. The nature, extent, and importance of the detected conditions are substantiated by gamma autoradiography and by destructive analysis employing alpha autoradiography, electron microprobe and visual inspection. Also, a series of radiographs illustrate the response of neutron radiography as compared to low voltage and high voltage x-ray and gamma source Iridium 192 radiography. (U.S.)

  14. Comparison of wall/divertor deuterium retention and plasma fueling requirements on the DIII-D, TdeV, and ASDEX-upgrade tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Maingi, R. [Oak Ridge Associated Universities, TN (United States); Terreault, B. [Inst. National de la Recherche Scientifique, Varennes, Quebec (Canada); Haas, G. [Max Planck Inst. fuer Plasmaphysik, Garching (Germany)] [and others

    1996-06-01

    The authors present a comparison of the wall deuterium retention and plasma fueling requirements of three diverted tokamaks, DIII-D, TdeV, and ASDEX-Upgrade, with different fractions of graphite coverage of stainless steel or Inconel outer walls and different heating modes. Data from particle balance experiments on each tokamak demonstrate well-defined differences in wall retention of deuterium gas, even though all three tokamaks have complete graphite coverage of divertor components and all three are routinely boronized. This paper compares the evolution of the change in wall loading and net fueling efficiency for gas during dedicated experiments without Helium Glow Discharge Cleaning on the DIII-D and TdeV tokamaks. On the DIII-D tokamak, it was demonstrated that the wall loading could be increased by > 1,250 Torr-1 (equivalent to 150 {times} plasma particle content) plasma inventories resulting in an increase in fueling efficiency from 0.08 to 0.25, whereas the wall loading on the TdeV tokamak could only be increased by < 35 Torr-{ell} (equivalent to 50{times} plasma particle content) plasma inventories at a maximum fueling efficiency {approximately} 1. Data from the ASDEX-Upgrade tokamak suggests qualitative behavior of wall retention and fueling efficiency similar to DIII-D.

  15. Fuel Behaviour at High During RIA and LOCA Accidents; Comportamiento del Combustible de Alto Quemado en Accidents RIA y LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Barrio del Juanes, M T; Garcia Cuesta, J C; Vallejo Diaz, I; Puebla, Herranz

    2001-07-01

    Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs.

  16. Fuel Behaviour at High During RIA and LOCA Accidents; Comportamiento del Combustible de Alto Quemado en Accidents RIA y LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Barrio del Juanes, M.T.; Garcia Cuesta, J.C.; Vallejo Diaz, I.; Herranz Puebla

    2001-07-01

    Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs.

  17. Irradiation effects on thermal properties of LWR hydride fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-01

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

  18. Integral nuclear fuel element assembly

    International Nuclear Information System (INIS)

    Schluderberg, D. C.

    1985-01-01

    An integral nuclear fuel element assembly utilizes longitudinally finned fuel pins. The continuous or interrupted fins of the fuel pins are brazed to fins of juxtaposed fuel pins or directly to the juxtaposed fuel pins or both. The integrally brazed fuel assembly is designed to satisfy the thermal and hydraulic requirements of a fuel assembly lattice having moderator to fuel atom ratios required to achieve high conversion and breeding ratios

  19. Reclamation and reuse of LEU silicide fuel from manufacturing scrap

    International Nuclear Information System (INIS)

    Gale, G.R.; Pace, B.W.; Evans, R.S.

    2004-01-01

    In order to provide an understanding of the organization which is the sole supplier of United States plate type research and test reactor fuel and LEU core conversions, a brief description of the structure and history is presented. Babcock and Wilcox (B and W) is a part of McDermott International, Inc. which is a large diversified corporation employing over 20,000 people primarily in engineering and construction for the off-shore oil and power generation industries throughout the world. B and W provides many energy related products requiring precision machining and high quality systems. This is accomplished by using state-of-the-art equipment, technology and highly skilled people. The RTRFE group within B and W has the ability to produce various complexly shaped fuel elements with a wide variety of fuels and enrichments. B and W RTRFE has fabricated over 200,000 plates since 1981 and gained the diversified experience necessary to satisfy many customer requirements. This accomplishment was possible with the support of McDermott International and all of its resources. B and W has always had a commitment to high quality and integrity. This is apparent by the success and longevity (125 years) of the company. A lower cost to convert cores to LEU provides direct support to RERTR and demonstrates Babcock and Wilcox's commitment to the program. As a supporter of RERTR reactor conversion from HEU to LEU, B and W has contributed a significant amount of R and D money to improve the silicide fuel process which ultimately lowers the LEU core costs. In the most recent R and D project, B and W is constructing a LEU silicide reclamation facility to re-use the unirradiated fuel scrap generated from the production process. Remanufacturing use of this fuel completes the fuel cycle and provides a contribution to LEU cores by reducing scrap inventory and handling costs, lowering initial purchase of fuel due to increasing the process yields, and lowering the replacement costs. This

  20. WWER spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

    Bower, C C; Lettington, C [GEC Alsthom Engineering Systems Ltd., Whetstone (United Kingdom)

    1994-12-31

    Selection criteria for PAKS NPP dry storage system are outlined. They include the following: fuel temperature in storage; sub-criticality assurance (avoidance of criticality for fuel in the unirradiated condition without having to take credit for burn-up); assurance of decay heat removal; dose uptake to the operators and public; protection of environment; volume of waste produced during operation and decommissioning; physical protection of stored irradiated fuel assemblies; IAEA safeguards assurance; storage system versus final disposal route; cost of construction and extent of technology transfer to Hungarian industry. Several available systems are evaluated against these criteria, and as a result the GEC ALSTHOM Modular Vault Dry Store (MVDS) system has been selected. The MVDS is a passively cooled dry storage facility. Its most important technical, safety, licensing and technology transfer characteristics are outlined. On the basis of the experience gained some key questions and considerations related to the East European perspective in the field of spent fuel storage are discussed. 8 figs.

  1. WWER spent fuel storage

    International Nuclear Information System (INIS)

    Bower, C.C.; Lettington, C.

    1994-01-01

    Selection criteria for PAKS NPP dry storage system are outlined. They include the following: fuel temperature in storage; sub-criticality assurance (avoidance of criticality for fuel in the unirradiated condition without having to take credit for burn-up); assurance of decay heat removal; dose uptake to the operators and public; protection of environment; volume of waste produced during operation and decommissioning; physical protection of stored irradiated fuel assemblies; IAEA safeguards assurance; storage system versus final disposal route; cost of construction and extent of technology transfer to Hungarian industry. Several available systems are evaluated against these criteria, and as a result the GEC ALSTHOM Modular Vault Dry Store (MVDS) system has been selected. The MVDS is a passively cooled dry storage facility. Its most important technical, safety, licensing and technology transfer characteristics are outlined. On the basis of the experience gained some key questions and considerations related to the East European perspective in the field of spent fuel storage are discussed. 8 figs

  2. Fuels processing for transportation fuel cell systems

    Science.gov (United States)

    Kumar, R.; Ahmed, S.

    Fuel cells primarily use hydrogen as the fuel. This hydrogen must be produced from other fuels such as natural gas or methanol. The fuel processor requirements are affected by the fuel to be converted, the type of fuel cell to be supplied, and the fuel cell application. The conventional fuel processing technology has been reexamined to determine how it must be adapted for use in demanding applications such as transportation. The two major fuel conversion processes are steam reforming and partial oxidation reforming. The former is established practice for stationary applications; the latter offers certain advantages for mobile systems and is presently in various stages of development. This paper discusses these fuel processing technologies and the more recent developments for fuel cell systems used in transportation. The need for new materials in fuels processing, particularly in the area of reforming catalysis and hydrogen purification, is discussed.

  3. Failed fuel monitoring at nuclear power plants with RBMK reactors: operating parameters, requirements and decision making criteria

    International Nuclear Information System (INIS)

    Zhukov, I.V.

    1993-01-01

    The procedure for estimating the number of failed fuel rods in the core and the prediction of their discharge efficiency during operation is presented. The procedure is based on the FFM data base and the I-131 and Xe-133 coolant activity. (author)

  4. Current state of knowledge of water radiolysis effects on spent nuclear fuel corrosion

    International Nuclear Information System (INIS)

    Christensen, H.; Sunder, S.

    2000-07-01

    Literature data on the effect of water radiolysis products on spent-fuel oxidation and dissolution are reviewed. Effects of gamma radiolysis, alpha radiolysis, and dissolved O 2 or H 2 O 2 in unirradiated solutions are discussed separately. Also, the effect of carbonate in gamma-irradiated solutions and radiolysis effects on leaching of spent fuel are reviewed. In addition, a kinetic model for calculating the corrosion rates of UO 2 in solutions undergoing radiolysis is discussed. The model gives good agreement between calculated and measured corrosion rates in the case of gamma radiolysis and in unirradiated solutions containing dissolved oxygen or hydrogen peroxide. However, the model fails to predict the results of alpha radiolysis. In a recent study , it was shown that the model gave good agreement with measured corrosion rates of spent fuel exposed in deionized water. The applications of radiolysis studies for geologic disposal of used nuclear fuel are discussed. (author)

  5. R and D status and requirements for PIE in the fields of the HTGR fuel and the innovative basic research on High-Temperature Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Sawa, Kazuhiro; Tobita, Tsutomu; Sumita, Junya [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Ishihara, Masahiro; Hayashi, Kimio; Hoshiya, Taiji; Sekino, Hajime; Ooeda, Etsurou

    1999-09-01

    The High Temperature Engineering Test Reactor (HTTR), which is the first high temperature gas-cooled reactor (HTGR) in Japan, achieved its first criticality in November 1998 at the Oarai Research Establishment of the Japan Atomic Energy Research Institute (JAERI). In the field of HTGR fuel development, JAERI will proceed research and development (R and D) works by the following steps: (STEP-1) confirmation of irradiation performance of the first-loading fuel of the HTTR, (STEP-2) study on irradiation performance of high burnup SiC-coated fuel particle and (STEP-3) development of ZrC-coated fuel particle. Requirements for post-irradiation examination (PIE) are different for each R and D step. In STEP-1, firstly, hot cells will be prepared in the HTTR reactor building to handle spent fuels. In parallel, general equipments such as those for deconsolidation of fuel compacts and for handling coated fuel particles will be installed in the Hot Laboratory at Oarai. In STEP-2, precise PIE techniques, for example, Raman spectroscopy for measurement of stress on irradiated SiC layer, will be investigated. In STEP-3, new PIE techniques should be developed to investigate irradiation behavior of ZrC-coated particle. In the field of the innovative basic research on high-temperature engineering, some preliminary tests have been made on the research areas of (1) new materials development, (2) fusion technology, (3) radiation chemistry and (4) high-temperature in-core instrumentation. Requirements for PIE are under investigation, in particular in the field of the new materials development. Besides more general apparatuses including transmission electron microscopy (TEM), some special apparatuses such as an electron spin resonance (ESR) spectrometer, a specific resistance/Hall coefficient measuring system and a differential scanning calorimeter (DSC) are planned to install in the Hot Laboratory at Oarai. Acquisition of advanced knowledge on the irradiation behavior is expected in

  6. Specification of requirements to get a license for an Independent Spent Fuel Dry Storage Installation (ISFSI) at the site of the NPP-LV

    International Nuclear Information System (INIS)

    Serrano R, M. L.

    2015-09-01

    This article describes some of the work done in the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS) to define specifically the requirements that the Federal Electricity Commission (CFE) shall meet to submit for consideration of CNSNS an operation request of an Independent Spent Fuel Dry Storage Installation (ISFSI). The project of a facility of this type arose from the need to provide storage capacity for spent nuclear fuel in the nuclear power plant of Laguna Verde (NPP-LV) and to continue the operation at the same facility in a safe manner. The licensing of these facilities in the United States of America has two modes: specific license or general license. The characteristics of these licenses are described in this article. However, in Mexico the existing national legislation is not designed for such license types, in fact there is a lack of standards or regulations in this regard. The regulatory law of Article 27 of the Constitution in the nuclear matter, only generally establishes that this type of facility requires an authorization from the Ministry of Energy. For this reason and because there is not a national legislation, was necessary to use the legislation that provides the Nuclear Regulatory Commission of USA, the US NRC. However, it cannot be applied as is established, so was necessary that the CNSNS analyze one by one the requirements of both types of license and determine what would be required to NPP-LV to submit its operating license of ISFSI. The American regulatory applicable to an ISFSI, the 10-Cfr-72 of the US NRC, establishes the requirements for both types of licenses. Chapter 10-Cfr was analyzed in all its clauses and coupled to the laws, regulations and standards as well as to the requirements established by CNSNS, all associated with a store spent fuel on site; the respective certification of containers for spent fuel dry storage was not included in this article, even though the CNSNS also performed that activity under the

  7. Behaviour of rock-like oxide fuels under reactivity-initiated accident conditions

    International Nuclear Information System (INIS)

    Kazuyuki, Kusagaya; Takehiko, Nakamura; Makio, Yoshinaga; Hiroshi, Akie; Toshiyuki, Yamashita; Hiroshi, Uetsuka

    2002-01-01

    Pulse irradiation tests of three types of un-irradiated rock-like oxide (ROX) fuel - yttria-stabilised zirconia (YSZ) single phase, YSZ and spinel (MgAl 2 O 4 ) homogeneous mixture and particle-dispersed YSZ/spinel - were conducted in the Nuclear Safety Research Reactor to investigate the fuel behaviour under reactivity-initiated accident conditions. The ROX fuels failed at fuel volumetric enthalpies above 10 GJ/m 3 , which was comparable to that of un-irradiated UO 2 fuel. The failure mode of the ROX fuels, however, was quite different from that of the UO 2 fuel. The ROX fuels failed with fuel pellet melting and a part of the molten fuel was released out to the surrounding coolant water. In spite of the release, no significant mechanical energy generation due to fuel/coolant thermal interaction was observed in the tested enthalpy range below∼12 GJ/m 3 . The YSZ type and homogenous YSZ/spinel type ROX fuels failed by cladding burst when their temperatures peaked, while the particle-dispersed YSZ/spinel type ROX fuel seemed to have failed by cladding local melting. (author)

  8. Thermal conductivity of beginning-of-life uranium-plutonium mixed oxide fuel for fast reactor (Interim report)

    International Nuclear Information System (INIS)

    Inoue, Masaki; Mizuno, Tomoyasu; Asaga, Takeo

    1997-11-01

    Thermal conductivity of uranium-plutonium mixed oxide fuel for fast reactor at beginning-of-life was correlated based on the recent results in order to apply to the fuel design and the fuel performance analysis. A number of experimental results of unirradiated fuel specimens were corrected from open literatures and PNC internal reports and examined for the database. In this work two porosity correction factors were needed for high density fuel and low density fuel (around the current Monju specification). The universal porosity correction factor was not determined in this work. In the next step, theoretical and analytical considerations should be taken into account. (J.P.N.)

  9. Dynamic Thermal Loads and Cooling Requirements Calculations for V ACs System in Nuclear Fuel Processing Facilities Using Computer Aided Energy Conservation Models

    International Nuclear Information System (INIS)

    EL Fawal, M.M.; Gadalla, A.A.; Taher, B.M.

    2010-01-01

    In terms of nuclear safety, the most important function of ventilation air conditioning (VAC) systems is to maintain safe ambient conditions for components and structures important to safety inside the nuclear facility and to maintain appropriate working conditions for the plant's operating and maintenance staff. As a part of a study aimed to evaluate the performance of VAC system of the nuclear fuel cycle facility (NFCF) a computer model was developed and verified to evaluate the thermal loads and cooling requirements for different zones of fuel processing facility. The program is based on transfer function method (TFM) and it is used to calculate the dynamic heat gain by various multilayer walls constructions and windows hour by hour at any orientation of the building. The developed model was verified by comparing the obtained calculated results of the solar heat gain by a given building with the corresponding calculated values using finite difference method (FDM) and total equivalent temperature different method (TETD). As an example the developed program is used to calculate the cooling loads of the different zones of a typical nuclear fuel facility the results showed that the cooling capacities of the different cooling units of each zone of the facility meet the design requirements according to safety regulations in nuclear facilities.

  10. Electrochemical corrosion studies on copper-base waste package container materials in unirradiated 0.1 N NaNO3 at 95 degrees C

    International Nuclear Information System (INIS)

    Akkaya, M.; Verink, E.D. Jr.; Van Konynenburg, R.A.

    1988-05-01

    Three candidate materials were investigated in this study in terms of their electrochemical corrosion behavior in unirradiated 0.1 N NaNO 3 solutions at 95 degrees C. Anodic polarization experiments were conducted to determine the passive current densities, pitting potentials, and other parameters, together with Cyclic Current Reversal Voltammetry tests to evaluate the stability and protectiveness of the passive oxides formed. X-ray diffraction and Auger Electron Spectroscopy were used for identification of the corrosion products as well as Scanning Electron Microscopy for the surface morphology studies. 2 refs., 22 figs., 2 tabs

  11. Nuclear Fuel Safety Criteria Technical Review - Second edition

    International Nuclear Information System (INIS)

    Beck, Winfried; Blanpain, Patrick; Fuketa, Toyoshi; Gorzel, Andreas; Hozer, Zoltan; Kamimura, Katsuichiro; Koo, Yang-Hyun; Maertens, Dietmar; Nechaeva, Olga; Petit, Marc; Rehacek, Radomir; Rey-Gayo, Jose Maria; Sairanen, Risto; Sonnenburg, Heinz-Guenther; Valach, Mojmir; Waeckel, Nicolas; Yueh, Ken; Zhang, Jinzhao; Voglewede, John

    2012-01-01

    Most of the current nuclear fuel safety criteria were established during the 1960's and early 1970's. Although these criteria were validated against experiments with fuel designs available at that time, a number of tests were based on unirradiated fuels. Additional verification was performed as these designs evolved, but mostly with the aim of showing that the new designs adequately complied with existing criteria, and not to establish new limits. In 1996, the OECD Nuclear Energy Agency (NEA) reviewed existing fuel safety criteria, focusing on new fuel and core designs, new cladding materials and industry manufacturing processes. The results were published in the Nuclear Fuel Safety Criteria Technical Review of 2001. The NEA has since re-examined the criteria. A brief description of each criterion and its rationale are presented in this second edition, which will be of interest to both regulators and industry (fuel vendors, utilities)

  12. Thermomechanical behavior and modeling of zircaloy cladding tubes from an unirradiated state to high burn-up

    International Nuclear Information System (INIS)

    Schaeffler-Le Pichon, I.; Geyer, P.; Bouffioux, P.

    1997-01-01

    Creep laws are nowadays commonly used to simulate the fuel rod response to the solicitations it faces during its life. These laws are sufficient for describing the base operating conditions (where only creep appears), but they have to be improved for power ramp conditions (where hardening and relaxation appear). The modification due to a neutronic irradiation of the thermomechanical behavior of stress-relieved Zircaloy 4 fuel tubes that have been analysed for five different fluences ranging from a non-irradiated material to a material for which the combustion rate was very high is presented. In the second part, a viscoplastic model able to simulate, for different isotherms, out-of-flux anisotropic mechanical behavior of the cladding tubes irradiated until high burn-up is proposed. Finally, results of numerical simulations show the ability of the model to reproduce the totality of the thermomechanical experiments. (author)

  13. 77 FR 72746 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard and Diesel...

    Science.gov (United States)

    2012-12-06

    ... Fuels and Fuel Additives: Modifications to Renewable Fuel Standard and Diesel Sulfur Programs AGENCY... Fuel Standard (``RFS'') program under section 211(o) of the Clean Air Act. The direct final rule also... marine diesel fuel produced by transmix processors, and the fuel marker requirements for 500 ppm sulfur...

  14. Investigation of nuclide importance to functional requirements related to transport and long-term storage of LWR spent fuel

    International Nuclear Information System (INIS)

    Broadhead, B.L.; DeHart, M.D.; Ryman, J.C.; Tang, J.S.; Parks, C.V.

    1995-06-01

    This study investigates the relative importances of the various actinide, fission-product, and light-element isotopes associated with LWR spent fuel with respect to five analysis areas: criticality safety (absorption fractions), shielding (dose rate fractions), curies (fractional curies levels), decay heat (fraction of total watts), and radiological toxicity (fraction of potential committed effective dose equivalent). These rankings are presented for up to six different burnup/enrichment scenarios and at decay times from 2 to 100,000 years. Ranking plots for each of these analysis areas are given in an Appendix for completeness, as well as summary tables in the main body of the report. Summary rankings are presented in terms of high (greater than 10% contribution to the total), medium (between 1% and 10% contribution), and low (less than 1% contribution) for both short- and long-term cooling. When compared with the expected measurement accuracies, these rankings show that most of the important isotopes can be characterized sufficiently for the purpose of radionuclide generation/depletion code validation in each of the analysis areas. Because the main focus of this work is on the relative importances of isotopes associated with L at sign spent fuel, some conclusions may not be applicable to similar areas such as high-level waste (HLW) and nonfuel-bearing components (NFBC)

  15. 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

  16. Vapor pressures of oxide reactor fuels above 3000 K: Review and perspective

    International Nuclear Information System (INIS)

    Breitung, W.

    1982-03-01

    Vapor pressures of liquid oxide reactor fuels are among the most important material data required for theoretical analyses of Hypothetical Core Disruptive Accidents in Fast Breeder Reactors. This report is an attempt to completely summarize and critically review the numerous theoretical and experimental results published for the pressure-temperature and pressure-energy relation of unirradiated UO 2 and (U,Pu)O 2 . First - to define the research goal - the precision in the saturation vapor pressure is quantified which is required for the purpose of HCDA calculations. Then the various theoretical and experimental methods used for the determination of p-T and p-U data are reviewed with respect to their principles, results and uncertainties. The achievements of the individual methods are discussed in the light of the research goal and - in view of the widely scattered data - recommendations are made concerning the p-T and p-U relation of UO 2 . Finally, the most important future research areas are identified, including some specific research proposals which aim at reducing the still large uncertainties in fuel vapor pressures down to the desired level. (orig.) [de

  17. 75 FR 37733 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard Program

    Science.gov (United States)

    2010-06-30

    ... Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard Program AGENCY... direct final rule to amend the Renewable Fuel Standard program requirements on May 10, 2010. Because EPA... Fuel Standard program requirements, published on May 10, 2010. We stated in that direct final rule that...

  18. Fuel manufacturing and utilization

    International Nuclear Information System (INIS)

    2005-01-01

    The efficient utilisation of nuclear fuel requires manufacturing facilities capable of making advanced fuel types, with appropriate quality control. Once made, the use of such fuels requires a proper understanding of their behaviour in the reactor environment, so that safe operation for the design life can be achieved. The International Atomic Energy Agency supports Member States to improve in-pile fuel performance and management of materials; and to develop advanced fuel technologies for ensuring reliability and economic efficiency of the nuclear fuel cycle. It provides assistance to Member States to support fuel-manufacturing capability, including quality assurance techniques, optimization of manufacturing parameters and radiation protection. The IAEA supports the development fuel modelling expertise in Member States, covering both normal operation and postulated and severe accident conditions. It provides information and support for the operation of Nuclear Power Plant to ensure that the environment and water chemistry is appropriate for fuel operation. The IAEA supports fuel failure investigations, including equipment for failed fuel detection and for post-irradiation examination and inspection, as well as fuel repair, it provides information and support research into the basic properties of fuel materials, including UO 2 , MOX and zirconium alloys. It further offers guidance on the relationship with back-end requirement (interim storage, transport, reprocessing, disposal), fuel utilization and management, MOX fuels, alternative fuels and advanced fuel technology

  19. Current state of knowledge in radiolysis effects on spent fuel corrosion

    International Nuclear Information System (INIS)

    Christensen, H.; Sunder, S.

    1998-09-01

    Literature data on the effect of water radiolysis products on spent fuel oxidation and dissolution have been reviewed. Effects of γ-radiolysis, α-radiolysis and dissolved O 2 or H 2 O 2 in unirradiated solutions have been discussed separately. Also the effect of carbonate in γ-irradiated solutions and radiolysis effects on leaching of spent fuels have been reviewed. In addition a radiolysis model for calculation of corrosion rates of UO 2 , presented previously, has been discussed. The model has been shown to give a good agreement between calculated and measured corrosion rates in the case of γ-radiolysis and in unirradiated solutions of dissolved oxygen or hydrogen peroxide. The model has failed to predict the results of α-radiolysis. In a recent study it was shown that the model gave a good agreement with measured corrosion rates of spent fuel exposed in deionized water

  20. Fuel assemblies

    International Nuclear Information System (INIS)

    Echigoya, Hironori; Nomata, Terumitsu.

    1983-01-01

    Purpose: To render the axial distribution relatively flat. Constitution: First nuclear element comprises a fuel can made of zircalloy i.e., the metal with less neutron absorption, which is filled with a plurality of UO 2 pellets and sealed by using a lower end plug, a plenum spring and an upper end plug by means of welding. Second fuel element is formed by substituting a part of the UO 2 pellets with a water tube which is sealed with water and has a space for allowing the heat expansion. The nuclear fuel assembly is constituted by using the first and second fuel elements together. In such a structure, since water reflects neutrons and decrease their leakage to increase the temperature, reactivity is added at the upper portion of the fuel assembly to thereby flatten the axial power distribution. Accordingly, stable operation is possible only by means of deep control rods while requiring no shallow control rods. (Sekiya, K.)

  1. Customer satisfaction techniques applied to the transport of nuclear fuels to meet ISO 9001:2000 requirements

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, G.K. [BNFL (United Kingdom)

    2004-07-01

    Understanding the needs of potential and existing customers is seen as essential to the continuing success of the nuclear fuel transport business. A recognised process utilised for this is the recording and measurement of customer satisfaction. This paper looks at what customer satisfaction means to BNFL International Transport in terms of, understanding what the customers needs are, supplying those needs or proposing appropriate alternatives, delivering on contractual obligations, in terms of the product or service being supplied to cost and time, listening to the changing needs of customers should this occur and monitoring the perception of customers in terms of performance. Within BNFL as a whole customer satisfaction is managed through the application of a Management Framework. Within this framework are set a number of values, these are focused on acknowledging social and environmental responsibilities while satisfying the needs of customers. Customer satisfaction is a key aspect of this philosophy.

  2. Study of advanced professional educational requirements relative to nuclear fuel cycle engineering in industry and government. Final report

    International Nuclear Information System (INIS)

    Jur, T.A.; Huhns, M.N.; Keating, D.A.; Orloff, D.I.; Rhodes, C.A.; Stanford, T.G.; Stephens, L.M.; Tatterson, G.B.; Van Brunt, V.

    1978-12-01

    An assessment was conducted of educational needs among engineers working in nuclear fuel cycle-related areas, focusing on the nuclear industry in the Southeast. Educational needs addressed were those at the post-baccalaureate professional level. As a result of the study, a list of subject areas has been compiled as best representing the current content of an educational program. In addition to identifying subject areas, a set of course descriptions and reference materials has been developed around each subject. Each course description contains information regarding objectives, anticipated audience, and prerequisites and offers a suggested course outline. An initial modest program of implementation is recommended which would continue to concentrate on the Southeast as a target area

  3. Study of advanced professional educational requirements relative to nuclear fuel cycle engineering in industry and government. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Jur, T.A.; Huhns, M.N.; Keating, D.A.; Orloff, D.I.; Rhodes, C.A.; Stanford, T.G.; Stephens, L.M.; Tatterson, G.B.; Van Brunt, V.

    1978-12-01

    Under contract with the U.S. Department of Energy, the College of Engineering at the University of South Carolina has conducted an assessment of educational needs among engineers working in nuclear fuel cycle related areas. The study was initiated as a regional effort focusing on the concentration of nuclear industry in the Southeast. Educational needs addressed were those at the post-baccalaureate professional level. The project was envisioned as providing base line information for the eventual implementation of a program in line with the needs of the Southeast's nuclear community. Specific objectives were to establish the content of such a program and to determine those specialized features which would make the program most attractive and useful.

  4. Study of advanced professional educational requirements relative to nuclear fuel cycle engineering in industry and government. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Jur, T.A.; Huhns, M.N.; Keating, D.A.; Orloff, D.I.; Rhodes, C.A.; Stanford, T.G.; Stephens, L.M.; Tatterson, G.B.; Van Brunt, V.

    1978-12-01

    An assessment was conducted of educational needs among engineers working in nuclear fuel cycle-related areas, focusing on the nuclear industry in the Southeast. Educational needs addressed were those at the post-baccalaureate professional level. As a result of the study, a list of subject areas has been compiled as best representing the current content of an educational program. In addition to identifying subject areas, a set of course descriptions and reference materials has been developed around each subject. Each course description contains information regarding objectives, anticipated audience, and prerequisites and offers a suggested course outline. An initial modest program of implementation is recommended which would continue to concentrate on the Southeast as a target area.

  5. Customer satisfaction techniques applied to the transport of nuclear fuels to meet ISO 9001:2000 requirements

    International Nuclear Information System (INIS)

    Fisher, G.K.

    2004-01-01

    Understanding the needs of potential and existing customers is seen as essential to the continuing success of the nuclear fuel transport business. A recognised process utilised for this is the recording and measurement of customer satisfaction. This paper looks at what customer satisfaction means to BNFL International Transport in terms of, understanding what the customers needs are, supplying those needs or proposing appropriate alternatives, delivering on contractual obligations, in terms of the product or service being supplied to cost and time, listening to the changing needs of customers should this occur and monitoring the perception of customers in terms of performance. Within BNFL as a whole customer satisfaction is managed through the application of a Management Framework. Within this framework are set a number of values, these are focused on acknowledging social and environmental responsibilities while satisfying the needs of customers. Customer satisfaction is a key aspect of this philosophy

  6. Study of advanced professional educational requirements relative to nuclear fuel cycle engineering in industry and government. Final report

    International Nuclear Information System (INIS)

    Jur, T.A.; Huhns, M.N.; Keating, D.A.; Orloff, D.I.; Rhodes, C.A.; Stanford, T.G.; Stephens, L.M.; Tatterson, G.B.; Van Brunt, V.

    1978-12-01

    Under contract with the U.S. Department of Energy, the College of Engineering at the University of South Carolina has conducted an assessment of educational needs among engineers working in nuclear fuel cycle related areas. The study was initiated as a regional effort focusing on the concentration of nuclear industry in the Southeast. Educational needs addressed were those at the post-baccalaureate professional level. The project was envisioned as providing base line information for the eventual implementation of a program in line with the needs of the Southeast's nuclear community. Specific objectives were to establish the content of such a program and to determine those specialized features which would make the program most attractive and useful

  7. Researches of WWER fuel rods behaviour under RIA accident conditions

    International Nuclear Information System (INIS)

    Nechaeva, O.; Medvedev, A.; Novikov, V.; Salatov, A.

    2003-01-01

    Unirradiated fuel rod and refabricated fuel rod tests in the BIGR as well as acceptance criteria proving absence of fragmentation and the settlement modeling of refabricated fuel rods thermomechanical behavior in the BIGR-tests using RAPTA-5 code are discussed in this paper. The behaviour of WWER type simulators with E110 and E635 cladding was researched at the BIGR reactor under power pulse conditions simulating reactivity initiated accident. The results of the tests in four variants of experimental conditions are submitted. The behaviour of 12 WWER type refabricated fuel rods was researched in the BIGR reactor under power pulse conditions simulating reactivity initiated accident: burnup 48 and 60 MWd/kgU, pulse width 3 ms, peak fuel enthalpy 115-190 cal/g. The program of future tests in the research reactor MIR with high burnup fuel rod (up to 70 MWd/kgU) under conditions simulating design RIA in WWER-1000 is presented

  8. Comparison of the intermediate storage periods and areas required for final storage of high-level radioactive waste and spent fuel in various types of host rock

    International Nuclear Information System (INIS)

    Mueller-Hoeppe, N.; Lerch, C.; Jobmann, M.; Filbert, W.

    2005-01-01

    The present new version of the German concept for radioactive waste and spent fuel management is based on the assumption that a repository for high-level waste and spent fuel will not be required until 2030. One reason frequently given for this date is the intermediate storage period of at least forty years to allow the very high initial heat generation to decay. However, calculations performed by the authors have shown that the minimum intermediate storage period for a repository in rock salt is only between four and nineteen years, depending on the final storage concept and the load of the waste package. In clay as a host rock, the minimum intermediate storage times were calculated to be between 31 and 142 years; the same time spans are expected to apply to final storage in magmatic rock, such as granite. The maximum permissible loads of a container holding spent fuel in salt are many times those in clay and granite, respectively. It was also seen that the area requirement for final storage of the same waste structures is roughly a factor of ten higher in clay than in salt. The differences between granite and salt are similar. The reasons for these grave differences, on the one hand, are the better thermal conductivity of salt and, on the other hand, the better heat tolerance of the crushed salt used as backfill material compared to that of bentonite used in the clay and granite concepts. While salt will allow temperatures of up to 200 C, the maximum temperature in bentonite is limited to 100 C. (orig.)

  9. The oxidative dissolution of unirradiated UO2 by hydrogen peroxide as a function of pH

    International Nuclear Information System (INIS)

    Clarens, F.; Pablo, J. de; Casas, I.; Gimenez, J.; Rovira, M.; Merino, J.; Cera, E.; Bruno, J.; Quinones, J.; Martinez-Esparza, A.

    2005-01-01

    The dissolution of non-irradiated UO 2 was studied as a function of both pH and hydrogen peroxide concentration (simulating radiolytic generated product). At acidic pH and a relatively low hydrogen peroxide concentration (10 -5 mol dm -3 ), the UO 2 dissolution rate decreases linearly with pH while at alkaline pH the dissolution rate increases linearly with pH. At higher H 2 O 2 concentrations (10 -3 mol dm -3 ) the dissolution rates are lower than the ones at 10 -5 mol dm -3 H 2 O 2 , which has been attributed to the precipitation at these conditions of studtite (UO 4 . 4H 2 O, which was identified by X-ray diffraction), together with the possibility of hydrogen peroxide decomposition. In the literature, spent fuel dissolution rates determined in the absence of carbonate fall in the H 2 O 2 concentration range 5 x 10 -7 - 5 x 10 -5 mol dm -3 according to our results, which is in agreement with H 2 O 2 concentrations determined in spent fuel leaching experiments

  10. Calculation of axial hydrogen redistribution on the spent fuels during interim dry storage

    International Nuclear Information System (INIS)

    Sasahara, Akihiro; Matsumura, Tetsuo

    2006-01-01

    One of the phenomena that will affect fuel integrity during a spent fuel dry storage is a hydrogen axial migration in cladding. If there is a hydrogen pickup in cladding in reactor operation, hydrogen will move from hotter to colder cladding region in the axial direction under fuel temperature gradient during dry storage. Then hydrogen beyond solubility limit in colder region will be precipitated as hydride, and consequently hydride embrittlement may take place in the cladding. In this study, hydrogen redistribution experiments were carried out to obtain the data related to hydrogen axial migration by using actually twenty years dry (air) stored spent PWR-UO 2 fuel rods of which burn-ups were 31 and 58 MWd/kg HM. From the hydrogen redistribution experiments, the heat of transport of hydrogen of zircaloy-4 cladding from twenty years dry stored spent PWR-UO 2 fuel rods were from 10.1 to 18.6 kcal/mol and they were significantly larger than that of unirradiated zircaloy-4 cladding. This means that hydrogen in irradiated cladding can move easier than that in unirradiated cladding. In the hydrogen redistribution experiments, hydrogen diffusion coefficients and solubility limit were also obtained. There are few differences in the diffusion coefficients and solubility limits between the irradiated cladding and unirradiated cladding. The hydrogen redistribution in the cladding after dry storage for forty years was evaluated by one-dimensional diffusion calculation using the measured values. The maximum values as the heat of transports, diffusion coefficients and solubility limits of the irradiated cladding and various spent fuel temperature profiles reported were used in the calculation. The axial hydrogen migration was not significant after dry storage for forty years in helium atmosphere and the maximum values as the heat of transports, diffusion coefficients and solubility limits of the unirradiated cladding gave conservative evaluation for hydrogen redistribution

  11. Dearomatization of jet fuel on irradiated platinum-supported catalyst

    International Nuclear Information System (INIS)

    Mucka, V.; Ostrihonova, A.; Kopernicky, I.; Mikula, O.

    1983-01-01

    The effect of ionizing radiation ( 60 Co #betta#-rays) on Pt-supported catalyst used for the dearomatization of jet fuel with distillation in the range 395 to 534 K has been studied. Pre-irradiation of the catalyst with doses in the range 10 2 to 5 x 10 4 Gy leads to the partial catalyst activation. Irradiation of the catalyst enhances its resistance to catalyst poisons, particularly to sulphur-compounds, and this is probably the reason for its catalytic activity being approx. 60 to 100% greater than that of un-irradiated catalyst. Optimum conditions for dearomatization on the irradiated catalyst were found and, by means of a rotary three-factorial experiment, it was shown that these lie at lower temperatures and lower pressures than those for un-irradiated catalyst. (author)

  12. Fluidized-bed calcination of LWR fuel-reprocessing HLLW: requirements and potential for off-gas cleanup

    International Nuclear Information System (INIS)

    Schindler, R.E.

    1979-01-01

    Fluidized-bed solidification (calcination) was developed on a pilot scale for a variety of simulated LWR high-level liquid-waste (HLLW) and blended high-level and intermediate-level liquid-waste (ILLW) compositions. It has also been demonstrated with ICPP fuel-reprocessing waste since 1963 in the Waste Calcining Facility (WCF) at gross feed rates of 5 to 12 m 3 /day. A fluidized-bed calciner produces a relatively large volume of off-gas. A calciner solidifying 6 m 3 /day of liquid waste would generate about 13 standard m 3 /min of off-gas containing 10 to 20 g of entrained solids per standard m 3 of off-gas. Use of an off-gas system similar to that of the WCF could provide an overall process decontamination factor for particulates of about 2 x 10 10 . A potential advantage of fluidized-bed calcination over other solidification methods is the ability to control ruthenium volatilization from the calciner at less than 0.01% by calcining at 500 0 C or above. Use of an off-gas system similar to that of the WCF would provide an overall process decontamination factor for volatile ruthenium of greater than 1.6 x 10 7

  13. Final disposal of spent fuel in the Finnish bedrock. Scope and requirements for site-specific safety analysis

    International Nuclear Information System (INIS)

    1996-12-01

    The report is a summary of the research conducted in the period 1993 to 1996 into safety of spent fuel final disposal. The principal goal of the research in this period, as set in 1993, was to develop a strategy for site-specific safety analysis. At the same time efforts were to be continued to gather data and validate the technical approach for the analysis. The work aimed at having the data needed for the analysis available at the end of year 1998. A safety assessment update, TILA-96, prepared by VTT Energy, is published as a separate report. The assessment is based on the TVO-92 safety analysis, but takes into account the knowledge acquired after 1992 on safety aspects of the disposal system and the data gathered from the site investigations made by TVO and from the beginning of 1996, by Posiva. Since the site investigations are still ongoing and much of the data gathered still pending interpretation, only limited amount of new site-specific information has been available for the present assessment. (172 refs.)

  14. Nuclear fuel safety studies by laser pulse heating

    International Nuclear Information System (INIS)

    Viswanadham, C.S.; Kumar, Santosh; Dey, G.K.; Kutty, T.R.G.; Khan, K.B.; Kumar, Arun; Jathar, V.P.; Sahoo, K.C.

    2009-01-01

    The behaviour of nuclear fuels under transient heating conditions is vital to nuclear safety. A laser pulse based heating system to simulate the transient heating conditions experienced by the fuel during reactor accidents like LOCA and RIA is under development at BARC, Mumbai. Some of the concepts used in this system are under testing in pilot studies. This paper describes the results of some pilot studies carried out on unirradiated UO 2 specimens by laser pulse heating, followed by metallography and X-ray diffraction measurements. (author)

  15. Spent fuel dissolution studies FY 1991 to 1994

    International Nuclear Information System (INIS)

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

    1995-12-01

    Dissolution and transport as a result of groundwater flow are generally accepted as the primary mechanisms by which radionuclides from spent fuel placed in a geologic repository could be released to the biosphere. To help provide a source term for performance assessment calculations, dissolution studies on spent fuel and unirradiated uranium oxides have been conducted over the past few years at Pacific Northwest National Laboratory (PNNL) in support of the Yucca Mountain Site Characterization Project. This report describes work for fiscal years 1991 through 1994. The objectives of these studies and the associated conclusions, which were based on the limited number of tests conducted so far, are described in the following subsections

  16. Spent Nuclear Fuel project stage and store K basin SNF in canister storage building functions and requirements. Revision 1

    International Nuclear Information System (INIS)

    Womack, J.C.

    1995-01-01

    This document establishes the functions and requirements baseline for the implementation of the Canister Storage Building Subproject. The mission allocated to the Canister Storage Building Subproject is to provide safe, environmentally sound staging and storage of K Basin SNF until a decision on the final disposition is reached and implemented

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  18. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

    Lotts, A.L.; Coobs, J.H.

    1976-08-01

    The status of fuel and fuel cycle technology for high-temperature gas-cooled reactors (HTGRs) is reviewed. The all-ceramic core of the HTGRs permits high temperatures compared with other reactors. Core outlet temperatures of 740 0 C are now available for the steam cycle. For advanced HTGRs such as are required for direct-cycle power generation and for high-temperature process heat, coolant temperatures as high as 1000 0 C may be expected. The paper discusses the variations of HTGR fuel designs that meet the performance requirements and the requirements of the isotopes to be used in the fuel cycle. Also discussed are the fuel cycle possibilities, which include the low-enrichment cycle, the Th- 233 U cycle, and plutonium utilization in either cycle. The status of fuel and fuel cycle development is summarized

  19. Dissolution of unirradiated UO{sub 2} and UO{sub 2} doped with {sup 233}U under reducing conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ollila, K. [VTT Processes (Finland); Oversby, V.M. [VMO Konsult (Sweden)

    2005-01-01

    Experiments have been conducted to determine an upper limit to the dissolution rate of UO{sub 2} under reducing conditions appropriate to those in a geologic repository for spent fuel disposal in Finland and Sweden. Test duration ranged from 52 to 140 days. The total amount of U recovered in each test was converted into a dissolution rate per year for the sample. The dissolution rate was then used to calculate an expected lifetime for the samples under the test conditions. The dissolution rate did not depend on the length of the testing period. Rather, the dissolution rate appeared to decrease as the samples were exposed to sequential testing periods. This indicates that the results are still influenced by transient effects such as high-energy surface sites, which implies that the dissolution rates measured are upper limits. The sample lifetimes calculated from the last two testing periods, which had a total of 269 days, ranged from 7 to 10 million years. There was no indication of an effect of alpha radiolysis on the dissolution rate results for samples with doping levels of 0, 5, and 10% {sup 233}U.

  20. On the requirement for remodelling the spent nuclear fuel transportation casks for research reactors. A review of the drop impact analyses of JRC-80Y-20T

    International Nuclear Information System (INIS)

    2005-07-01

    The Japan Atomic Energy Research Institute (JAERI) constructed two stainless steel transportation casks, JRC-80Y-20T, for spent nuclear fuels of research reactors and had utilized them for transportation since 1981. A modification of the design was applied to the United States of America (USA) for transportation of silicide fuels. Additional analyses employing the impact analysis code LS-DYNA that was often used for safety analysis were submitted by the JAERI to the USA in 2003 to show integrity of the packages; the casks were still not approved, because inelastic deformation was occurred on the surface of the lid touching to the body. To resolve this problem on design approval of transportation casks, a review group was formed in June 2004. The group examined the impact analyses by reviewing the input data and performing the sensitivity analyses. As the drop impact analyses were found to be practically reasonable, it was concluded that the approval of the USA for the transportation casks could not be obtained just by revising the analyses; therefore, remodelling the casks is required. (author)

  1. Fuels Registration, Reporting, and Compliance Help

    Science.gov (United States)

    Information about the requirements for registration and health effects testing of new fuels or fuel additives and mandatory registration for fuels reporting and about mandatory reporting forms for parties regulated under EPA fuel programs.

  2. United States Domestic Research Reactor Infrastructure TRIGA Reactor Fuel Support

    International Nuclear Information System (INIS)

    Morrell, Douglas

    2011-01-01

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  3. Spent nuclear fuel project recommended reaction rate constants for corrosion of N-Reactor fuel

    International Nuclear Information System (INIS)

    Cooper, T.D.; Pajunen, A.L.

    1998-01-01

    The US Department of Energy (DOE) established the Spent Nuclear Fuel Project (SNF Project) to address safety and environmental concerns associated with deteriorating spent nuclear fuel presently stored in the Hanford Site's K Basins. The SNF Project has been tasked by the DOE with moving the spent N-Reactor fuel from wet storage to contained dry storage in order to reduce operating costs and environmental hazards. The chemical reactivity of the fuel must be understood at each process step and during long-term dry storage. Normally, the first step would be to measure the N-fuel reactivity before attempting thermal-hydraulic transfer calculations; however, because of the accelerated project schedule, the initial modeling was performed using literature values for uranium reactivity. These literature values were typically found for unirradiated, uncorroded metal. It was fully recognized from the beginning that irradiation and corrosion effects could cause N-fuel to exhibit quite different reactivities than those commonly found in the literature. Even for unirradiated, uncorroded uranium metal, many independent variables affect uranium metal reactivity resulting in a wide scatter of data. Despite this wide reactivity range, it is necessary to choose a defensible model and estimate the reactivity range of the N-fuel until actual reactivity can be established by characterization activities. McGillivray, Ritchie, and Condon developed data and/or models that apply for certain samples over limited temperature ranges and/or reaction conditions (McGillivray 1994, Ritchie 1981 and 1986, and Condon 1983). These models are based upon small data sets and have relatively large correlation coefficients

  4. NWTS program criteria for mined geologic disposal of nuclear waste: functional requirements and performance criteria for waste packages for solidified high-level waste and spent fuel

    International Nuclear Information System (INIS)

    1982-07-01

    The Department of Energy (DOE) has primary federal responsibility for the development and implementation of safe and environmentally acceptable nuclear waste disposal methods. Currently, the principal emphasis in the program is on emplacement of nuclear wastes in mined geologic repositories well beneath the earth's surface. A brief description of the mined geologic disposal system is provided. The National Waste Terminal Storage (NWTS) program was established under DOE's predecessor, the Energy Research and Development Administration, to provide facilities for the mined geologic disposal of radioactive wastes. The NWTS program includes both the development and the implementation of the technology necessary for designing, constructing, licensing, and operating repositories. The program does not include the management of processing radioactive wastes or of transporting the wastes to repositories. The NWTS-33 series, of which this document is a part, provides guidance for the NWTS program in the development and implementation of licensed mined geologic disposal systems for solidified high-level and transuranic (TRU) wastes. This document presents the functional requirements and performance criteria for waste packages for solidified high-level waste and spent fuel. A separate document to be developed, NWTS-33(4b), will present the requirements and criteria for waste packages for TRU wastes. The hierarchy and application of these requirements and criteria are discussed in Section 2.2

  5. Failed fuel rod detector

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, Katsuya; Matsuda, Yasuhiko

    1984-05-02

    The purpose of the project is to enable failed fuel rod detection simply with no requirement for dismantling the fuel assembly. A gamma-ray detection section is arranged so as to attend on the optional fuel rods in the fuel assembly. The fuel assembly is adapted such that a gamma-ray shielding plate is detachably inserted into optional gaps of the fuel rods or, alternatively, the fuel assembly can detachably be inserted to the gamma-ray shielding plate. In this way, amount of gaseous fission products accumulated in all of the plenum portions in the fuel rods as the object of the measurement can be determined without dismantling the fuel assembly. Accordingly, by comparing the amounts of the gaseous fission products, the failed fuel rod can be detected.

  6. In-pile experiments on fuel rod behaviour during a LOCA

    International Nuclear Information System (INIS)

    Sepold, E.H.; Karb, E.H.; Pruessmann, M.

    1981-07-01

    This report describes the results of the Test Series G2/3 within the in-pile experimental program for the investigation of LWR fuel rod behavior. The results were obtained with single rods of a PWR design in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechanisms of fuel rod failure were being performed with irradiated and unirradiated rods. The main parameter of the test program ist the burnup, ranging from 2500 to 35000 MWd/t. The results of test series G2/3 (35000 MWd/t) with respect to the burst data, i.e. burst temperature, burst pressure, and burst strain, do not indicate major differences from the in-pile tests with unirradiated test specimens. (orig.) [de

  7. In-pile experiemts on fuel rod behavior during a LOCA

    International Nuclear Information System (INIS)

    Pruessmann, M.; Karb, E.H.; Sepold, L.

    1981-02-01

    This report describes the results of the Test Series G1 within the in-pile experimental program for the investigation of LWR fuel rod behavior. The results were obtained with single rods of a PWR design in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechansims of fuel rod failure were being performed with irradiated and unirradiated rods. The main parameter of the test program is the burnup ranging from 2500 to 35 000 MWd/t. The results of test series G1 (35 000 MWd/t) with respect to the burst data, i.e. burst temperature, burst pressure, and burst strain, do not indicate major differences from the in-pile tests with unirradiated test specimens. (orig.) [de

  8. Prompt neutron decay constant for the Oak Ridge Research Reactor with 20 wt % 235U enriched fuel

    International Nuclear Information System (INIS)

    Ragan, G.E.; Mihalczo, J.T.

    1986-01-01

    This paper describes measurements of the prompt neutron decay constant at delayed criticality for the Oak Ridge Research Reactor (ORR) using 20 wt % 235 U enriched fuel and compares these measurements with similar measurements using 93.2 wt % 235 U enriched fuel. This reactor parameter is of interest because it affects the transient behavior of the reactor in prompt criticality accident situations. This experiment is part of a program to investigate the differences in the performance of research reactors fueled with highly enriched and low enriched uranium. The prompt neutron decay constants were obtained using noise analysis measurement techniques for a core with newly fabricated, unirradiated fuel elements

  9. Macroscopic behavior of fast reactor fuel subjected to simulated thermal transients

    International Nuclear Information System (INIS)

    Fenske, G.R.; Emerson, J.E.; Savoie, F.E.

    1983-06-01

    High-speed cinematography has been used to characterize the macroscopic behavior of irradiated and unirradiated fuel subjected to thermal transients prototypical of fast reactor transients. The results demonstrate that as the cladding melts, the fuel can disperse via spallation if the fuel contains in excess of approx. 16 μmoles/gm of fission gas. Once the cladding has melted, the macroscopic behavior (time to failure and dispersive nature) was strongly influenced by the presence of volatile fission products and the heating rate

  10. Integrated fuel processor development

    International Nuclear Information System (INIS)

    Ahmed, S.; Pereira, C.; Lee, S. H. D.; Krumpelt, M.

    2001-01-01

    The Department of Energy's Office of Advanced Automotive Technologies has been supporting the development of fuel-flexible fuel processors at Argonne National Laboratory. These fuel processors will enable fuel cell vehicles to operate on fuels available through the existing infrastructure. The constraints of on-board space and weight require that these fuel processors be designed to be compact and lightweight, while meeting the performance targets for efficiency and gas quality needed for the fuel cell. This paper discusses the performance of a prototype fuel processor that has been designed and fabricated to operate with liquid fuels, such as gasoline, ethanol, methanol, etc. Rated for a capacity of 10 kWe (one-fifth of that needed for a car), the prototype fuel processor integrates the unit operations (vaporization, heat exchange, etc.) and processes (reforming, water-gas shift, preferential oxidation reactions, etc.) necessary to produce the hydrogen-rich gas (reformate) that will fuel the polymer electrolyte fuel cell stacks. The fuel processor work is being complemented by analytical and fundamental research. With the ultimate objective of meeting on-board fuel processor goals, these studies include: modeling fuel cell systems to identify design and operating features; evaluating alternative fuel processing options; and developing appropriate catalysts and materials. Issues and outstanding challenges that need to be overcome in order to develop practical, on-board devices are discussed

  11. Experiences in certification of packages for transportation of fresh nuclear fuel in the context of new safety requirements established by IAEA regulations (IAEA-96 regulations, ST-1) for air transportation of nuclear materials (requirements to C-type packages)

    Energy Technology Data Exchange (ETDEWEB)

    Dudai, V.I.; Kovtun, A.D.; Matveev, V.Z.; Morenko, A.I.; Nilulin, V.M.; Shapovalov, V.I.; Yakushev, V.A.; Bobrovsky, V.S.; Rozhkov, V.V.; Agapov, A.M.; Kolesnikov, A.S. [Russian Federal Nuclear Centre - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation)]|[JSC ' ' MSZ' ' , Electrostal (Russian Federation)]|[JSC ' ' NPCC' ' , Novosibirsk (Russian Federation)]|[Minatom of Russia, Moscow (Russian Federation)]|[Gosatomnadzor of Russia, Moscow (Russian Federation)

    2004-07-01

    Every year in Russia, a large amount of domestic and international transportation of fresh nuclear fuel (FNF) used in Russian and foreign energy and research atomic reactors and referred to fissile materials based on IAEA Regulations is performed. Here, bulk transportation is performed by air, and it concerns international transportation in particular. According to national ''Main Regulations for Safe Transport and physical Protection of Nuclear Materials (OPBZ- 83)'' and ''Regulations for the Safe Transport of Radioactive Materials'' of the International Atomic Energy Agency (IAEA Regulations), nuclear and radiation security under normal (accident free) and accident conditions of transport must be completely provided by the package design. In this context, high requirements to fissile packages exposed to heat and mechanical loads in transport accidents are imposed. A long-standing experience in accident free transportation of FM has shown that such approach to provide nuclear and radiation security pays for itself completely. Nevertheless, once in 10 years the International Atomic Energy Agency on every revision of the ''Regulations for the Safe Transport of Radioactive Materials'' places more stringent requirements upon the FM and transportation thereof, resulting from the objectively increasing risk associated with constant rise in volume and density of transportation, and also strained social and economical situation in a number of regions in the world. In the new edition of the IAEA Regulations (ST-1), published in 1996 and brought into force in 2001 (IAEA-96 Regulations), the requirements to FM packages conveyed by aircraft were radically changed. These requirements are completely presented in new Russian ''Regulations for the Safe Transport of Radioactive Materials'' (PBTRM- 2004) which will be brought into force in the time ahead.

  12. 40 CFR 80.602 - What records must be kept by entities in the NRLM diesel fuel, ECA marine fuel, and diesel fuel...

    Science.gov (United States)

    2010-07-01

    ... (CONTINUED) AIR PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel; Nonroad, Locomotive, and Marine Diesel Fuel; and ECA Marine Fuel Recordkeeping and Reporting Requirements... in the NRLM diesel fuel, ECA marine fuel, and diesel fuel additive production, importation, and...

  13. Instrumentation of fuel elements and fuel plates

    International Nuclear Information System (INIS)

    Durand, J.P.; Fanjas, Y.

    1993-01-01

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

  14. Instrumentation of fuel elements and fuel plates

    International Nuclear Information System (INIS)

    Durand, J.P.; Fanjas, Y.

    1994-01-01

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

  15. 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

  16. Characterization of spent fuel hulls and dissolution residues

    International Nuclear Information System (INIS)

    Gue, J.P.; Andriessen, H.

    1985-04-01

    The main results obtained within the framework of CEC programmes, by KFK, UKAEA and CEA, are reviewed concerning the characterization of dissolution wastes. The contents were determined of the main radioactive emitters contained in the hulls originating in a whole fuel assembly sampled at the La Hague plant, or from Dounreay PFR fuels. Radiochemical characterizations were carried out by different methods including neutron emission measurement, alpha and beta-gamma spectrometry, and mass spectrometry. Decontamination of the hulls by using rinsings and supplementary treatment were also dealt with. The ignition and explosion risks associated with the zircaloy fines formed during the shearing of LWR fuels were examined, and the ignition properties of irradiated and unirradiated zircaloy powders were determined and compared. The physical properties and compositions of the dissolution residues of PFR fuels were defined, in order to conduct tests on the immobilization of these wastes in cement

  17. Proposed high throughput electrorefining treatment for spent N- Reactor fuel

    International Nuclear Information System (INIS)

    Gay, E.C.; Miller, W.E.; Laidler, J.J.

    1996-01-01

    A high-throughput electrorefining process is being adapted to treat spent N-Reactor fuel for ultimate disposal in a geologic repository. Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the type of fragmentation necessary to provide fuel segments suitable for this process. Based on these tests, a conceptual design was produced of a plant-scale electrorefiner. In this design, the diameter of an electrode assembly is about 1.07 m (42 in.). Three of these assemblies in an electrorefiner would accommodate a 3-metric-ton batch of N-Reactor fuel that would be processed at a rate of 42 kg of uranium per hour

  18. Fuel Exhaling Fuel Cell.

    Science.gov (United States)

    Manzoor Bhat, Zahid; Thimmappa, Ravikumar; Devendrachari, Mruthyunjayachari Chattanahalli; Kottaichamy, Alagar Raja; Shafi, Shahid Pottachola; Varhade, Swapnil; Gautam, Manu; Thotiyl, Musthafa Ottakam

    2018-01-18

    State-of-the-art proton exchange membrane fuel cells (PEMFCs) anodically inhale H 2 fuel and cathodically expel water molecules. We show an unprecedented fuel cell concept exhibiting cathodic fuel exhalation capability of anodically inhaled fuel, driven by the neutralization energy on decoupling the direct acid-base chemistry. The fuel exhaling fuel cell delivered a peak power density of 70 mW/cm 2 at a peak current density of 160 mA/cm 2 with a cathodic H 2 output of ∼80 mL in 1 h. We illustrate that the energy benefits from the same fuel stream can at least be doubled by directing it through proposed neutralization electrochemical cell prior to PEMFC in a tandem configuration.

  19. Dissolution of mixed oxide fuel as a function of fabrication variables

    International Nuclear Information System (INIS)

    Lerch, R.E.

    1979-08-01

    Dissolution properties of mechanically blended mixed oxide fuel were very dependent on the six fuel fabrication variables studied. Fuel sintering temperature, source of PuO 2 and PuO 2 content of the fuel had major effects: (1) as the sintering temperature was increased from 1400 to 1700 0 C, pellet dissolution was more complete; (2) pellets made from burned metal derived PuO 2 were more completely dissolved than pellets made from calcined nitrate derived PuO 2 which in turn were more completely dissolved than pellets made from calcined nitrate derived PuO 2 ; (3) as the PuO 2 content decreased from 25 to 15 wt % PuO 2 , pellet dissolution was more complete. Preferential dissolution of uranium occurred in all the mechanically blended mixed oxide. Unirradiated mixed oxide fuel pellets made by the Sol Gel process were generally quite soluble in nitric acid. Unirradiated mixed oxide fuel pellets made by the coprecipitation process dissolved completely and rapidly in nitric acid. Fuel made by the coprecipitation process was more completely dissolved than fuel made by the Sol Gel process which, in turn, was more completely dissolved than fuel made by mechanically blending UO 2 and PuO 2 as shown below. Addition of uncomplexed fluoride to nitric acid during fuel dissolution generally rendered all fuel samples completely dissolvable. In boiling 12M nitric acid, 95 to 99% of the plutonium which was going to dissolve did so in the first hour. Irradiated mechanically blended mixed oxide fuel with known fuel fabrication conditions was also subjected to fuel dissolution tests. While irradiation was shown to increase completeness of plutonium dissolution, poor dissolubility due to adverse fabrication conditions (e.g., low sintering temperature) remained after irradiation

  20. 33 CFR 183.570 - Fuel filters and strainers: Installation.

    Science.gov (United States)

    2010-07-01

    ... SECURITY (CONTINUED) BOATING SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Manufacturer Requirements... the engine or boat structure independent from its fuel line connections, unless the fuel filter or...

  1. 14 CFR 25.991 - Fuel pumps.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel pumps. 25.991 Section 25.991... STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System Components § 25.991 Fuel pumps. (a) Main pumps. Each fuel pump required for proper engine operation, or required to meet the fuel system...

  2. General requirements applicable to the production, inspection, processing, packaging and storage of various types of waste resulting from the reprocessing of fuels irradiated in pressurized light water reactors

    International Nuclear Information System (INIS)

    1982-09-01

    The Fundamental Safety Rules applicable to certain types of nuclear installation are intended to clarify the conditions of which observance, for the type of installation concerned and for the subject that they deal with, is considered as equivalent to compliance with regulatory French technical practice. These Rules should facilitate safety analysises and the clear understanding between persons interested in matters related to nuclear safety. They in no way reduce the operator's liability and pose no obstacle to statutory provisions in force. For any installation to which a Fundamental Safety Rule applies according to the foregoing paragraph, the operator may be relieved from application of the Rule if he shows proof that the safety objectives set by the Rule are attained by other means that he proposes within the framework of statutory procedures. Furthermore, the Central Service for the Safety of Nuclear Installations reserves the right at all times to alter any Fundamental Safety Rule, as required, should it deem this necessary, while specifying the applicability conditions. This rule is intended to define the general provisions applicable to the production, inspection, processing, packaging and storage of the different types of wastes resulting from the reprocessing of fuels irradiated in a PWR

  3. Design Package for Fuel Retrieval System Fuel Handling Tool Modification

    International Nuclear Information System (INIS)

    TEDESCHI, D.J.

    2000-01-01

    This is a design package that contains the details for a modification to a tool used for moving fuel elements during loading of MCO Fuel Baskets for the Fuel Retrieval System. The tool is called the fuel handling tool (or stinger). This document contains requirements, development design information, tests, and test reports

  4. Fuel Assembly Damping Summary

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  5. DUPIC fuel compatibility assessment

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Rho, G. H.; Park, J. W. [and others

    2000-03-01

    The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition.

  6. DUPIC fuel compatibility assessment

    International Nuclear Information System (INIS)

    Choi, Hang Bok; Rho, G. H.; Park, J. W. and others

    2000-03-01

    The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition

  7. Transport of MOX fuel

    International Nuclear Information System (INIS)

    Porter, I.R.; Carr, M.

    1997-01-01

    The regulatory framework which governs the transport of MOX fuel is set out, including packages, transport modes and security requirements. Technical requirements for the packages are reviewed and BNFL's experience in plutonium and MOX fuel transport is described. The safety of such operations and the public perception of safety are described and the question of gaining public acceptance for MOX fuel transport is addressed. The paper concludes by emphasising the need for proactive programmes to improve the public acceptance of these operations. (Author)

  8. Fissile fuel doubling time characteristics for reactor lifetime fuel logistics

    International Nuclear Information System (INIS)

    Heindler, M.; Harms, A.A.

    1978-01-01

    The establishment of nuclear fuel requirements and their efficient utilization requires a detailed knowledge of some aspects of fuel dynamics and processing during the reactor lifetime. It is shown here that the use of the fuel stockpile inventory concept can serve effectively for this fuel management purpose. The temporal variation of the fissile fuel doubling time as well as nonequilibrium core conditions are among the characteristics which thus become more evident. These characteristics - rather than a single figure-of-merit - clearly provide an improved description of the expansion capacity and/or fuel requirements of a nuclear reactor energy system

  9. Spent fuel storage facility, Kalpakkam

    International Nuclear Information System (INIS)

    Shreekumar, B.; Anthony, S.

    2017-01-01

    Spent Fuel Storage Facility (SFSF), Kalpakkam is designed to store spent fuel arising from PHWRs. Spent fuel is transported in AERB qualified/authorized shipping cask by NPCIL to SFSF by road or rail route. The spent fuel storage facility at Kalpakkam was hot commissioned in December 2006. All systems, structures and components (SSCs) related to safety are designed to meet the operational requirements

  10. The Effect of E-Selen Antioxidant on the Fatty Acids Content of the Homogenate of Unirradiated and Irradiated Pupae of Ceratitis Capitata

    International Nuclear Information System (INIS)

    Zaghloul, Y.S.; Abbassy, S.A.; Elakhdar, E.A.H.; Elakhdar, E.A.H.

    2011-01-01

    As antioxidant E-selen was added to the larval artificial diets of the Mediterranean fruit fly, Ceratitis capitata. The produced full grown pupae were exposed to gamma rays at dose rate of 90 Gy. The fatty acid contents of the normal and irradiated insects were analyzed to test to what extent, the uptake of the antioxidant will ameliorate the physiological damage induced to the medfly pupae, as a result of their exposure to irradiation. The results obtained by using Gas-liquid chromatography of fatty acid methyl esters, showed the palmitic and oleic acids to be the most predominant fatty acids in all pupal homogenates in both normal and irradiated pupae, either by the uptake of doses of the E-selen or without. The uptake of E-Selen with a dose of 0.3 mg led to a complete absence of the lauric and myristic acids in the homogenates of the unirradiated pupae of the medfly. This absence of lauric and myristic acids was accompanied by a decrease in the concentrations of the palmitic and linoleic acids reaching 25.17% and 32.98%, respectively as compared to pupae without the uptake of antioxidant. The concentrations of both the stearic and oleic acids showed an obvious increment reaching 199.22% and 58.94%, respectively, relative to those reared on the defined media and no added E.Selen. Exposure of the pupae to a dose of 90 Gy, resulted in an increment in the concentrations of the saturated lauric and myristic acid reaching 96.77% and 34.07%, respectively, relative to the untreated ones. While, at the same dose level, there were decrements in the concentrations of the palmitic, stearic, oleic and linoleic acids with percentages reaching 6.4, 9.22, 3.65 and 1.15%, respectively as compared to unirradiated controls. Irradiation of the pupae with sterilizing dose (90 Gy) after up taking the E-Selen in their larval diets, led to the increase of the concentrations of lauric, oleic and linoleic acids by ratios of 16.82, 8.84 and 29.26%, respectively as compared to their

  11. Thermal conductivity of fully dense unirradiated UO2: A new formulation from experimental results between 100 deg. C and 2500 deg. C, and associated fundamental properties

    International Nuclear Information System (INIS)

    Delette, G.; Charles, M.

    1997-01-01

    The various contributions to the thermal conductivity of UO 2 are first reviewed: contribution from phonons is preponderant up to 1600 deg. C; radiative contribution is negligible in the case of a polycrystalline sample, and is unable to account for the increase in conductivity observed above 1600 deg. C; electronic contribution, which seems patently to be responsible for this increase, is efficient from 1400 deg. C (electron-hole pairs treated as small polarons). Given the lack of decisive arguments on their actual efficiency, it was decided that, for temperatures above 2400 deg. C, neither a possible decrease in the electronic contribution due to vacancies, nor an additional possible contribution from Frenkel pairs would be described. We do not go therefore beyond the above-mentioned electronic contribution. In the light of these considerations, the law established by CEA for the thermal conductivity of unirradiated UO 2 , on the basis of a homogeneous set of measurements between 100 deg. C and 2500 deg. C, has been revised. A least-square method applied to the above measurements was used to derive simultaneously the four adjustable constants of the law which is henceforth proposed: λ = 1/A + BT+C/T 2 exp(-W/kT). The values of A, B, C, and W obtained, have the advantage of giving a better account of the high temperature results compared to the Martin recommendations (which has, by the way, the same physical form as above). Furthermore, this new law allows an extrapolation towards the melting point which is better founded than the earlier CEA law, while still respecting the value of the integrated thermal conductivity up to melting. Finally, various burnup effects (impurities, stoichiometry, etc.) can be formulated in physical terms. Moreover, the values of various fundamental properties connected with the derived constants have been determined: Debye temperature, Grueneisen parameter, Mott-Hubbard energy, electrical conductivity. These values are discussed and

  12. Fuel transporting device

    International Nuclear Information System (INIS)

    Shiratori, Hirozo.

    1979-01-01

    Purpose: In a liquid-metal cooled reactor, to reduce the waiting time of fuel handling apparatuses and shorten the fuel exchange time. Constitution: A fuel transporting machine is arranged between a reactor vessel and an out-pile storage tank, thereby dividing the transportation line of the pot for contracting fuel and transporting the same. By assuming such a construction, the flow of fuel transportation which has heretofore been carried out through fuel transportation pipes is not limited to one direction but the take-out of fuels from the reactor and the take-in thereof from the storage tank can be carried out constantly, and much time is not required for fuel exchange. (Kamimura, M.)

  13. Fuel cell opportunities

    Energy Technology Data Exchange (ETDEWEB)

    Harris, K. [Hydrogenics Corporation, Mississauga, ON (Canada)

    2002-07-01

    The opportunities for fuel cell development are discussed. Fuel cells are highly efficient, reliable and require little maintenance. They also produce virtually zero emissions. The author stated that there are some complicated issues to resolve before fuel cells can be widely used. These include hydrogen availability and infrastructure. While the cost of fuel cells is currently very high, these costs are constantly coming down. The industry is still in the early stages of development. The driving forces for the development of fuel cells are: deregulation of energy markets, growing expectations for distributed power generation, discontinuity between energy supply and demand, and environmental concerns. 12 figs.

  14. Hinkley Point 'C' power station public inquiry: proof of evidence on the need for Hinkley Point 'C' to help meet capacity requirement and the non-fossil-fuel proportion economically

    International Nuclear Information System (INIS)

    Jenkin, F.P.

    1988-09-01

    A public inquiry has been set up to examine the planning application made by the Central Electricity Generating Board (CEGB) for the construction of a 1200 MW Pressurized Water Reactor power station at Hinkley Point (Hinkley Point ''C'') in the United Kingdom. The purpose of this evidence to the Inquiry is to show why there is a need now to go ahead with the construction of Hinkley Point ''C'' generating station to help meet the non-fossil-fuel proportion of generation economically and also to help meet future generating capacity requirement. The CEGB submits that it is appropriate to compare Hinkley Point ''C'' with other non-fossil-fuel alternatives under various bases. Those dealt with by this proof of evidence are as follows: i) ability to contribute to capacity need and in assisting the distribution companies to meet their duty to supply electricity; ii) ability to contribute to the non-fossil-fuel proportion; iii) relative economic merit. (author)

  15. Safety analysis of MOX fuels by fuel performance code

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-12-01

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

  16. Fuel-cycle cost comparisons with oxide and silicide fuels

    International Nuclear Information System (INIS)

    Matos, J.E.; Freese, K.E.

    1982-01-01

    This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data are presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires that they are successfully demonstrated and licensed

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

    International Nuclear Information System (INIS)

    Domanus, J. C.

    1976-03-01

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

  18. U-Mo fuels handbook. Version 1.0

    International Nuclear Information System (INIS)

    Rest, Jeffrey; Kim, Yeon Soo; Hofman, Gerard L.; Meyer, Mitchell K.; Hayes, Steven L.

    2006-01-01

    This handbook provides an overview of property data and fuel performance topics with an emphasis on data available for U-Mo alloys. These data often exist only in report format and have not been widely disseminated in the journal literature. For some topics there is more than one source of data, which are sometimes inconsistent. In this situation, the authors have attempted to select the best dataset to provide a standard for fuel designers and reactor operators. Following the section on unirradiated and irradiated materials properties for the monolithic U-Mo alloy, property data for cladding and matrix aluminum are presented. Property data for cladding aluminum are more widely available, and are not presented in great depth. Finally, some properties of (U-Mo)/Al dispersions are also included in this document. Where no data are available, best estimate correlations are provided. Best fits to the data are presented in order to facilitate use by fuel designers and reactor operators.

  19. Demonstration of pyrometallurgical processing for metal fuel and HLW

    International Nuclear Information System (INIS)

    Tadafumi, Koyama; Kensuke, Kinoshita; Takatoshi, Hizikata; Tadashi, Inoue; Ougier, M.; Rikard, Malmbeck; Glatz, J.P.; Lothar, Koch

    2001-01-01

    CRIEPI and JRC-ITU have started a joint study on pyrometallurgical processing to demonstrate the capability of this type of process for separating actinide elements from spent fuel and HLW. The equipment dedicated for this experiments has been developed and installed in JRC-ITU. The stainless steel box equipped with tele-manipulators is operated under pure Ar atmosphere, and prepared for later installation in a hot cell. Experiments on pyro-processing of un-irradiated U-Pu-Zr metal alloy fuel by molten salt electrorefining has been carried out. Recovery of U and Pu from this type alloy fuel was first demonstrated with using solid iron cathode and liquid Cd cathode, respectively. (author)

  20. Fuel assemblies

    International Nuclear Information System (INIS)

    Mukai, Hideyuki

    1987-01-01

    Purpose: To prevent bending of fuel rods caused by the difference of irradiation growth between coupling fuel rods and standards fuel rods thereby maintain the fuel rod integrity. Constitution: The f value for a fuel can (the ratio of pole of zirconium crystals in the entire crystals along the axial direction of the fuel can) of a coupling fuel rod secured by upper and lower tie plates is made smaller than the f value for the fuel can of a standard fuel rod not secured by the upper and the lower tie plates. This can make the irradiation growth of the fuel can of the coupling fuel rod greater than the irradiation growth of the fuel can of the standard fuel rod and, accordingly, since the elongation of the standard fuel rod can always by made greater, bending of the standard fuel rod can be prevented. (Yoshihara, M.)

  1. Thermal conductivity of fully dense unirradiated UO{sub 2}: A new formulation from experimental results between 100 deg. C and 2500 deg. C, and associated fundamental properties

    Energy Technology Data Exchange (ETDEWEB)

    Delette, G; Charles, M [Commissariat a l` Energie Atomique, Grenoble (France)

    1997-08-01

    The various contributions to the thermal conductivity of UO{sub 2} are first reviewed: contribution from phonons is preponderant up to 1600 deg. C; radiative contribution is negligible in the case of a polycrystalline sample, and is unable to account for the increase in conductivity observed above 1600 deg. C; electronic contribution, which seems patently to be responsible for this increase, is efficient from 1400 deg. C (electron-hole pairs treated as small polarons). Given the lack of decisive arguments on their actual efficiency, it was decided that, for temperatures above 2400 deg. C, neither a possible decrease in the electronic contribution due to vacancies, nor an additional possible contribution from Frenkel pairs would be described. We do not go therefore beyond the above-mentioned electronic contribution. In the light of these considerations, the law established by CEA for the thermal conductivity of unirradiated UO{sub 2}, on the basis of a homogeneous set of measurements between 100 deg. C and 2500 deg. C, has been revised. A least-square method applied to the above measurements was used to derive simultaneously the four adjustable constants of the law which is henceforth proposed: {lambda} = 1/A + BT+C/T{sup 2}exp(-W/kT). The values of A, B, C, and W obtained, have the advantage of giving a better account of the high temperature results compared to the Martin recommendations (which has, by the way, the same physical form as above). Furthermore, this new law allows an extrapolation towards the melting point which is better founded than the earlier CEA law, while still respecting the value of the integrated thermal conductivity up to melting. Finally, various burnup effects (impurities, stoichiometry, etc.) can be formulated in physical terms. Moreover, the values of various fundamental properties connected with the derived constants have been determined: Debye temperature, Grueneisen parameter, Mott-Hubbard energy, electrical conductivity. These values

  2. 40 CFR 79.55 - Base fuel specifications.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 16 2010-07-01 2010-07-01 false Base fuel specifications. 79.55... (CONTINUED) REGISTRATION OF FUELS AND FUEL ADDITIVES Testing Requirements for Registration § 79.55 Base fuel specifications. (a) General Characteristics. (1) The base fuel(s) in each fuel family shall serve as the group...

  3. Fuel cells for commercial energy

    Science.gov (United States)

    Huppmann, Gerhard; Weisse, Eckart; Bischoff, Manfred

    1990-04-01

    The development of various types of fuel cells is described. Advantges and drawbacks are considered for alkaline fuel cells, phosphoric acid fuel cells, and molten carbonate fuel cells. It is shown that their modular construction is particularly adapted to power heat systems. A comparison which is largely in favor of fuel cells, is made between coal, oil, natural gas power stations, and fuel cells. Safety risks in operation are also compared with those of conventional power stations. Fuel cells are particularly suited for dwellings, shopping centers, swimming pools, other sporting installations, and research facilities, whose high current and heat requirements can be covered by power heat coupling.

  4. Fuels: market, quality, emissions in France

    International Nuclear Information System (INIS)

    Philippon, A.

    1997-01-01

    Here is a study about the automobile fuels market. From the market trends, we find the evolution of fuels quality; but in front of the concurrence and with the imbalance between diesel fuels and gasoline fuels, the improvement in fuels quality that requires investments does not increase as well as the air quality should necessitate. (N.C.)

  5. A literature survey on the dissolution mechanism of spent fuel under disposal conditions

    International Nuclear Information System (INIS)

    Ollila, Kaija

    1989-06-01

    In Finland spent nuclear fuel is planned to be disposed of at large depths in crystalline bedrock. As part of the YJT (Nuclear Waste Commission of Finnish Power Companies) - program, the solubiliy and dissolution mechanisms of unirradiated UO 2 are experimentally investigated as a function of groundwater conditions. This study is a literature survey on the leaching and dissolution studies carried out with spent fuel. It consists first a review on characterization studies of spent fuel. Then the solubilities and release mechanisms of the radionuclides from spent fuel in granitic or related groundwaters are discussed, including the dissolution of UO 2 matrix, and the leaching of fission products and actinides. Lastly approaches to modelling the dissolution of spent fuel are shortly discussed

  6. Plans and status for accepting spent nuclear fuel from foreign research reactors

    International Nuclear Information System (INIS)

    Zeitoun, Abe; Williams, John; Brown, Keith; Chacey, Kenneth

    1996-01-01

    In May 1996, the Department of Energy, acting with the cooperation of the Department of State, announced adoption of a policy that will have a significant influence on international efforts to prevent the spread of nuclear weapons. This policy is concerned with the management of spent nuclear fuel from foreign research reactors. Spent nuclear fuel, unirradiated fuel, and target material accepted under the policy must contain uranium enriched in the U.S. Although such spent fuel will comprise a relatively small part of the Department of Energy's (the Department's) overall inventory of spent nuclear fuel, the policy invokes actions that provide a cornerstone of U.S. nonproliferation activities. Implementation of this policy is now underway. This paper describes the Department's implementation strategy with the emphasis on those actions that will affect foreign research reactor operators. (author)

  7. Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

    International Nuclear Information System (INIS)

    Amaya, Masaki; Nakamura, Jinichi; Nagase, Fumihisa; Fuketa, Toyoshi

    2011-01-01

    The thermal conductivity formula of fuel pellet which contains the effects of burnup and plutonium (Pu) addition was proposed based on the Klemens' theory and reported thermal conductivities of unirradiated (U, Pu) O 2 and irradiated UO 2 pellets. The thermal conductivity of high burnup MOX pellet was formulated by applying a summation rule between phonon scattering parameters which show the effects of plutonium addition and burnup. Temperature of high burnup MOX fuel was evaluated based on the thermal conductivity integral which was calculated from the above-mentioned thermal conductivity formula. Calculated fuel temperatures were plotted against the linear heat rates of the fuel rods, and were compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it was confirmed that the proposed thermal conductivity formula of MOX pellets is adequate.

  8. Fuel assembly storage pool

    International Nuclear Information System (INIS)

    Hiranuma, Hiroshi.

    1976-01-01

    Object: To remove limitation of the number of storage of fuel assemblies to increase the number of storage thereof so as to relatively reduce the water depth required for shielding radioactive rays. Structure: Fuel assembly storage rack containers for receiving a plurality of spent fuel assembly racks are stacked in multi-layer fashion within a storage pool filled with water for shielding radioactive rays and removing heat. (Furukawa, Y.)

  9. German nuclear fuel exports and imports 1991

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    The statistics compiled by the German Federal Office for Trade and Industry (Bundesamt fuer Wirtschaft) for the Federal Ministry for the Environment, Conservation of Nature, and Reactor Safety of imports and exports of nuclear fuels and source materials in 1991 show a major drop by 33.8% in imports and a pronounced rise by 191.5% in exports, compared to the levels in the previous year. Source material for the purposes of these statistics refers only to uranium concentrate. Quantitatively, the biggest import items are source materials, depleted uranium, and uranium enriched up to 3%. Exports of unirradiated material quantitatively comprise mainly depleted uranium, source material, and uranium enriched up to 10%. (orig.) [de

  10. Financing the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Stephany, M.

    1975-01-01

    While conventional power stations usually have fossil fuel reserves for only a few weeks, nuclear power stations, because of the relatively long time required for uranium processing from ore extraction to the delivery of the fuel elements and their prolonged in-pile time, require fuel reserves for a period of several years. Although the specific fuel costs of nuclear power stations are much lower than those of conventional power stations, this results in consistently higher financial requirements. But the problems involved in financing the nuclear fuel do not only include the aspect of financing the requirements of reactor operators, but also of financing the facilities of the nuclear fuel cycle. As far as the fuel supply is concerned, the true financial requirements greatly exceed the mere purchasing costs because the costs of financing are rather high as a consequence of the long lead times. (orig./UA) [de

  11. Nuclear fuel storage

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1981-01-01

    A nuclear fuel storage apparatus for use in a water-filled pool is fabricated of a material such as stainless steel in the form of an egg crate structure having vertically extending openings. Fuel may be stored in this basic structure in a checkerboard pattern with high enrichment fuel, or in all openings when the fuel is of low effective enrichment. Inserts of a material such as stainless steel are adapted to fit within these openings so that a water gap and, therefore, a flux trap is formed between adjacent fuel storage locations. These inserts may be added at a later time and fuel of a higher enrichment may be stored in each opening. When it is desired to store fuel of still greater enrichment, poison plates may be added to the water gap formed by the installed insert plates, or substituted for the insert plates. Alternately, or in addition, fuel may be installed in high neutron absorption poison boxes which surround the fuel assembly. The stainless steel inserts and the poison plates are each not required until the capacity of the basic egg crate structure is approached. Purchase of these items can, therefore, be deferred for many years. Should the fuel to be stored be of higher enrichment than initially forecast, the deferred decision on the poison plates makes it possible to obtain increased poison in the plates to satisfy the newly discovered requirement

  12. Toward sustainable fuel cells

    DEFF Research Database (Denmark)

    Stephens, Ifan; Rossmeisl, Jan; Chorkendorff, Ib

    2016-01-01

    to a regular gasoline car. However, current fuel cells require 0.25 g of platinum (Pt) per kilowatt of power (2) as catalysts to drive the electrode reactions. If the entire global annual production of Pt were devoted to fuel cell vehicles, fewer than 10 million vehicles could be produced each year, a mere 10...

  13. Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities, Sections 15-19

    International Nuclear Information System (INIS)

    Schneider, K.J.

    1982-09-01

    Information is presented under the following section headings: fuel reprocessing; spent fuel and high-level and transuranic waste storage; spent fuel and high-level and transuranic waste disposal; low-level and intermediate-level waste disposal; and, transportation of radioactive materials in the nuclear fuel cycle. In each of the first three sections a description is given on the mainline process, effluent processing and waste management systems, plant layout, and alternative process schemes. Safety information and a summary are also included in each. The section on transport of radioactive materials includes information on the transportation of uranium ore, uranium ore concentrate, UF 6 , PuO 2 powder, unirradiated uranium and mixed-oxide fuel assemblies, spent fuel, solidified high-level waste, contact-handled transuranic waste, remote-handled transuranic waste, and low and intermediate level nontransuranic waste. A glossary is included

  14. Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities, Sections 15-19

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.

    1982-09-01

    Information is presented under the following section headings: fuel reprocessing; spent fuel and high-level and transuranic waste storage; spent fuel and high-level and transuranic waste disposal; low-level and intermediate-level waste disposal; and, transportation of radioactive materials in the nuclear fuel cycle. In each of the first three sections a description is given on the mainline process, effluent processing and waste management systems, plant layout, and alternative process schemes. Safety information and a summary are also included in each. The section on transport of radioactive materials includes information on the transportation of uranium ore, uranium ore concentrate, UF/sub 6/, PuO/sub 2/ powder, unirradiated uranium and mixed-oxide fuel assemblies, spent fuel, solidified high-level waste, contact-handled transuranic waste, remote-handled transuranic waste, and low and intermediate level nontransuranic waste. A glossary is included. (JGB)

  15. Thermal and irradiation effects on high-temperature mechanical properties of materials for SCWR fuel cladding

    International Nuclear Information System (INIS)

    Kano, F.; Tsuchiya, Y.; Oka, K.

    2009-01-01

    The thermal and irradiation effects on high-temperature mechanical properties are examined for candidate alloys for fuel cladding of supercritical water-cooled reactors (SCRWs). JMTR (Japan Materials Testing Reactor) and Experimental Fast Reactor JOYO were utilized for neutron irradiation tests, considering their fluence and temperature. Irradiation was performed with JMTR at 600degC up to 4x10 24 n/m 2 and with JOYO at 600degC and 700degC up to 6x10 25 n/m 2 . Tensile test, creep test and hardness measurement were carried out for high-temperature mechanical properties. Based on the uniaxial creep test, the extrapolation curves were drawn with time-temperature relationships utilizing the Larson and Miller Parameter. Several candidate alloys are expected to satisfy the design requirement from the estimation of the creep rupture stress for 50000 hours. Comparing the creep strengths under irradiated and unirradiated conditions, it was inferred that creep deformation was dominated by the thermal effect rather than the irradiation at SCWR core condition. The microstructure was examined using transmission electron microscope (TEM) analysis, focusing on void swelling and helium (He) bubble formation. Void formation was observed in the materials irradiated with JOYO at 600degC but not at 700degC. However, its effect on the deformation of components was estimated to be tolerable since their size and density were negligibly small. The manufacturability of the thin-wall, small-diameter tube was confirmed for the potential candidate alloys through the trial tests in the factory where the fuel cladding tube is manufactured. (author)

  16. Alternative Fuels

    Science.gov (United States)

    Alternative fuels include gaseous fuels such as hydrogen, natural gas, and propane; alcohols such as ethanol, methanol, and butanol; vegetable and waste-derived oils; and electricity. Overview of alternative fuels is here.

  17. Radiological implications of plutonium recycle and the use of thorium fuels in power reactor operations

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, H. F.

    1976-01-15

    As economically attractive sources of natural uranium are gradually depleted attention will turn to recycling plutonium or to the use of thorium fuels. The radiological implications of these fuel cycles in terms of fuel handling and radioactive waste disposal are investigated in relation to a conventional /sup 235/U enriched oxide fuel. It is suggested that a comparative study of this nature may be an important aspect of the overall optimization of future fuel cycle strategies. It is shown that the use of thorium based fuels has distinct advantages in terms of neutron dose rates from irradiated fuels and long term proportional to decay heating commitment compared with conventional uranium/plutonium fuels. However, this introduces a ..gamma.. dose rate problem in the fabrication and handling of unirradiated /sup 233/U fuels. For both plutonium and thorium fuels these radiological problems increase during storage of the fuel prior to reactor irradiation. The novel health physics problems which arise in the handling and processing of thorium fuels are reviewed in an appendix.

  18. Review of oxidation rates of DOE spent nuclear fuel : Part 1 : nuclear fuel

    International Nuclear Information System (INIS)

    Hilton, B.A.

    2000-01-01

    reaction rates of irradiated and unirradiated fuel were determined to be similar. The apparent reaction rate of irradiated metallic fuel increases as a function of swelling due to the increased surface area

  19. Equipment for the conditioning of core components in the fuel element storage pool with particular respect to the design required by the conditions for nuclear facilities in operation and the surveillance in accordance with atomic rules and regulations

    International Nuclear Information System (INIS)

    Dumpe, J.; Schwiertz, V.; Geiser, C.; Prucker, E.

    2001-01-01

    In nuclear power plants worn out and activated parts from the reactor core (core components) which are placed into the fuel element storage pool arise on a regular basis during the technical maintenance and the review. The disposal of these core components due to radiation protection aspects is only feasible within the fuel element storage pool during the operation of the NPP using techniques of the under water conditioning. Therefore, special GNS equipment is used for the conditioning, using under water conditioning equipment, such as UWS, BZ, and ZVA, a number of lifting and auxiliary equipment for mounting and dismantling purposes and the handling of the core components and the waste casks within the fuel element storage pool. These components must meet particular safety requirements with regard to their integrity and reliability. They are designed according to the requirements on nuclear components (KTA). The manipulating equipment must be partly redundant and the protection goals for nuclear accidents must be met. The Bavarian Ministry for Development and Environment tasked the TUeV Sueddeutschland with the surveillance and control. The conditioning equipment of GNS is therefore designed in co-ordination with the examiner of the Governmental Regulating Agency, in particular respect to all safety aspects. Furthermore the examiners perform reviews of the construction and the documentation during the design and construction phase. (orig.)

  20. Demonstration of an instrumental technique in the measurement of solution weight in the accountability vessels of a fuel reprocessing plant

    International Nuclear Information System (INIS)

    Nakajima, K.

    1977-04-01

    Load cells were installed on the input accountability vessel of a commercial reactor fuel reprocessing facility to determine if this proven principle of mass measurement is in fact applicable in such a severe radiation environment over a long period of time. Two other locations selected were the plutonium product nitrate solution accountability vessel and the plutonium product nitrate solution storage vessel. The latter two environments, while not severely radio-active, require a high degree of contamination control. All three vessels are of different geometrical configuration and capacity. Each vessel was carefully calibrated for volume measurements by adding controlled pre-measured increments of water. Measurements were made using the conventional dip-tube manometer system and the load cell - digital voltmeter. Standard deviation of the measurements on the input vessel and the plutonium storage vessel were in both cases 0.3%; for the plutonium accountability vessel 1.9%. Measurements taken of the input vessel during the ''cold run'' over a six-month period using solutions of unirradiated uranium showed a standard deviation of 0.4% and a bias of 0.8% in the summer months and 0.7% and 0.6% respectively in the winter months FINAL STOP CODE

  1. Fuel assembly

    International Nuclear Information System (INIS)

    Chaki, Masao; Nishida, Koji; Karasawa, Hidetoshi; Kanazawa, Toru; Orii, Akihito; Nagayoshi, Takuji; Kashiwai, Shin-ichi; Masuhara, Yasuhiro

    1998-01-01

    The present invention concerns a fuel assembly, for a BWR type nuclear reactor, comprising fuel rods in 9 x 9 matrix. The inner width of the channel box is about 132mm and the length of the fuel rods which are not short fuel rods is about 4m. Two water rods having a circular cross section are arranged on a diagonal line in a portion of 3 x 3 matrix at the center of the fuel assembly, and two fuel rods are disposed at vacant spaces, and the number of fuel rods is 74. Eight fuel rods are determined as short fuel rods among 74 fuel rods. Assuming the fuel inventory in the short fuel rod as X(kg), and the fuel inventory in the fuel rods other than the short fuel rods as Y(kg), X and Y satisfy the relation: X + Y ≥ 173m, Y ≤ - 9.7X + 292, Y ≤ - 0.3X + 203 and X > 0. Then, even when the short fuel rods are used, the fuel inventory is increased and fuel economy can be improved. (I.N.)

  2. Fuel assembly

    International Nuclear Information System (INIS)

    Yamazaki, Hajime.

    1995-01-01

    In a fuel assembly having fuel rods of different length, fuel pellets of mixed oxides of uranium and plutonium are loaded to a short fuel rod. The volume ratio of a pellet-loaded portion to a plenum portion of the short fuel rod is made greater than the volume ratio of a fuel rod to which uranium fuel pellets are loaded. In addition, the volume of the plenum portion of the short fuel rod is set greater depending on the plutonium content in the loaded fuel pellets. MOX fuel pellets are loaded on the short fuel rods having a greater degree of freedom relevant to the setting for the volume of the plenum portion compared with that of a long rod fuel, and the volume of the plenum portion is ensured greater depending on the plutonium content. Even if a large amount of FP gas and He gas are discharged from the MOX fuels compared with that from the uranium fuels, the internal pressure of the MOX fuel rod during operation is maintained substantially identical with that of the uranium fuel rod, so that a risk of generating excess stresses applied to the fuel cladding tubes and rupture of fuels are greatly reduced. (N.H.)

  3. Fuel loads and fuel type mapping

    Science.gov (United States)

    Chuvieco, Emilio; Riaño, David; Van Wagtendonk, Jan W.; Morsdof, Felix; Chuvieco, Emilio

    2003-01-01

    Correct description of fuel properties is critical to improve fire danger assessment and fire behaviour modeling, since they guide both fire ignition and fire propagation. This chapter deals with properties of fuel that can be considered static in short periods of time: biomass loads, plant geometry, compactness, etc. Mapping these properties require a detail knowledge of vegetation vertical and horizontal structure. Several systems to classify the great diversity of vegetation characteristics in few fuel types are described, as well as methods for mapping them with special emphasis on those based on remote sensing images.

  4. Materials for high-temperature fuel cells

    CERN Document Server

    Jiang, San Ping; Lu, Max

    2013-01-01

    There are a large number of books available on fuel cells; however, the majority are on specific types of fuel cells such as solid oxide fuel cells, proton exchange membrane fuel cells, or on specific technical aspects of fuel cells, e.g., the system or stack engineering. Thus, there is a need for a book focused on materials requirements in fuel cells. Key Materials in High-Temperature Fuel Cells is a concise source of the most important and key materials and catalysts in high-temperature fuel cells with emphasis on the most important solid oxide fuel cells. A related book will cover key mater

  5. Materials for low-temperature fuel cells

    CERN Document Server

    Ladewig, Bradley; Yan, Yushan; Lu, Max

    2014-01-01

    There are a large number of books available on fuel cells; however, the majority are on specific types of fuel cells such as solid oxide fuel cells, proton exchange membrane fuel cells, or on specific technical aspects of fuel cells, e.g., the system or stack engineering. Thus, there is a need for a book focused on materials requirements in fuel cells. Key Materials in Low-Temperature Fuel Cells is a concise source of the most important and key materials and catalysts in low-temperature fuel cells. A related book will cover key materials in high-temperature fuel cells. The two books form part

  6. Development of PEM fuel cell technology at international fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, D.J.

    1996-04-01

    The PEM technology has not developed to the level of phosphoric acid fuel cells. Several factors have held the technology development back such as high membrane cost, sensitivity of PEM fuel cells to low level of carbon monoxide impurities, the requirement to maintain full humidification of the cell, and the need to pressurize the fuel cell in order to achieve the performance targets. International Fuel Cells has identified a hydrogen fueled PEM fuel cell concept that leverages recent research advances to overcome major economic and technical obstacles.

  7. ITER fuel cycle

    International Nuclear Information System (INIS)

    Leger, D.; Dinner, P.; Yoshida, H.

    1991-01-01

    Resulting from the Conceptual Design Activities (1988-1990) by the parties involved in the International Thermonuclear Experimental Reactor (ITER) project, this document summarizes the design requirements and the Conceptual Design Descriptions for each of the principal subsystems and design options of the ITER Fuel Cycle conceptual design. The ITER Fuel Cycle system provides for the handling of all tritiated water and gas mixtures on ITER. The system is subdivided into subsystems for fuelling, primary (torus) vacuum pumping, fuel processing, blanket tritium recovery, and common processes (including isotopic separation, fuel management and storage, and processes for detritiation of solid, liquid, and gaseous wastes). After an introduction describing system function and conceptual design procedure, a summary of the design is presented including a discussion of scope and main parameters, and the fuel design options for fuelling, plasma chamber vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary and common processes. Design requirements are defined and design descriptions are given for the various subsystems (fuelling, plasma vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary/common processes). The document ends with sections on fuel cycle design integration, fuel cycle building layout, safety considerations, a summary of the research and development programme, costing, and conclusions. Refs, figs and tabs

  8. Fuels and Lubricants Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Modern naval aircraft and turbine-powered craft require reliable and high-quality fuels and lubricants to satisfy the demands imposed upon them for top performance...

  9. Requirements for low-cost electricity and hydrogen fuel production from multiunit inertial fusion energy plants with a shared driver and target factory

    International Nuclear Information System (INIS)

    Logan, G.B.; Moir, R.W.; Hoffmman, M.A.

    1995-01-01

    The economy of scale for multiunit inertial fusion energy (IFE) power plants is explored based on the molten salt HYLIFE-II fusion chamber concept, for the purpose of producing lower cost electricity and hydrogen fuel. The cost of electricity (CoE) is minimized with a new IFE systems code IFEFUEL5 for a matrix of plant cases with one to eight fusion chambers of 250 to 2000-MW (electric) net output each, sharing a common heavy-ion driver and target factory. Improvements to previous HYLIFE-II models include a recirculating induction linac driver optimized as a function of driver energy and rep-rate (average driver power), inclusion of beam switchyard costs, a fusion chamber cost scaling dependence on both thermal power and fusion yield, and a more accurate bypass pump power scaling with chamber rep-rate. A CoE less than 3 cents/kW(electric)-h is found for plant outputs greater than 2 GW(electric), allowing hydrogen fuel production by wafer electrolysis to provide lower fuel cost per mile for higher efficiency hydrogen engines compared with gasoline engines. These multiunit, multi-GW(electric) IFE plants allow staged utility plant deployment, lower optimum chamber rep-rates, less sensitivity to driver and target fabrication costs, and a CoE possibly lower than future fission, fossil, and solar competitors. 37 refs., 12 figs., 4 tabs

  10. Study on the FBR cycle introduction scenario. 4. Evaluation of the FBR cycle introduction scenario from the viewpoints of the fuel cycle requirements

    International Nuclear Information System (INIS)

    Ono, Kiyoshi; Shiotani, Hiroki; Hirao, Kazunori

    2003-07-01

    This report is intended to explain the outline of the scenario studies on FBR (Fast Breeder Reactor) cycle introduction. Recently, people value the reduction of environmental impact in addition to the recycle of energy resources and the energy security in these scenario studies. This report summarizes the analysis about the necessity of plutonium recycling in LWR (Light water Reactor) from short-term view and about the necessity of FBR cycle introduction from a long-term view in Japan, by comparing 'FBR scenario' with 'LWR once-through scenario' and 'Pu recycle in LWR scenario', from the viewpoints of cumulative uranium demand, spent fuel storage, radioactive waste arising, etc. It becomes clear that the plutonium recycling in LWR has a good effect on the reduction of spent fuel storage and the cumulative natural uranium demand before FBR cycle introduction, from short-term view (20-30 years). On the other hand, this analysis also shows that there is much effect of FBR deployment not only on saving amount of uranium use and energy security but also on reduction of high-level radioactive waste (spent fuels and vitrified waste) and minor actinide arising, from long-term view (100-200 years). (author)

  11. Diesel fuel filtration system

    International Nuclear Information System (INIS)

    Schneider, D.

    1996-01-01

    The American nuclear utility industry is subject to tight regulations on the quality of diesel fuel that is stored at nuclear generating stations. This fuel is required to supply safety-related emergency diesel generators--the backup power systems associated with the safe shutdown of reactors. One important parameter being regulated is the level of particulate contamination in the diesel fuel. Carbon particulate is a natural byproduct of aging diesel fuel. Carbon particulate precipitates from the fuel's hydrocarbons, then remains suspended or settles to the bottom of fuel oil storage tanks. If the carbon particulate is not removed, unacceptable levels of particulate contamination will eventually occur. The oil must be discarded or filtered. Having an outside contractor come to the plant to filter the diesel fuel can be costly and time consuming. Time is an even more critical factor if a nuclear plant is in a Limiting Condition of Operation (LCO) situation. A most effective way to reduce both cost and risk is for a utility to build and install its own diesel fuel filtration system. The cost savings associated with designing, fabricating and operating the system inhouse can be significant, and the value of reducing the risk of reactor shutdown because of uncertified diesel fuel may be even higher. This article describes such a fuel filtering system

  12. MTR fuel inspection at CERCA

    International Nuclear Information System (INIS)

    Fanjas, Y.

    1992-01-01

    The stringent specifications for MTR fuel plates and fuel elements require various sophisticated inspection techniques. In particular, the development of low enriched silicide fuels made it necessary to adapt these techniques to high density plates. This paper presents the status of inspection technology at CERCA. (author)

  13. Nondestructive verification and assay systems for spent fuels. Technical appendixes

    International Nuclear Information System (INIS)

    Cobb, D.D.; Phillips, J.R.; Baker, M.P.

    1982-04-01

    Six technical appendixes are presented that provide important supporting technical information for the study of the application of nondestructive measurements to spent-fuel storage. Each appendix addresses a particular technical subject in a reasonably self-contained fashion. Appendix A is a comparison of spent-fuel data predicted by reactor operators with measured data from reprocessors. This comparison indicates a rather high level of uncertainty in previous burnup calculations. Appendix B describes a series of nondestructive measurements at the GE-Morris Operation Spent-Fuel Storage Facility. This series of experiments successfully demonstrated a technique for reproducible positioning of fuel assemblies for nondestructive measurement. The experimental results indicate the importance of measuring the axial and angular burnup profiles of irradiated fuel assemblies for quantitative determination of spent-fuel parameters. Appendix C is a reasonably comprehensive bibliography of reports and symposia papers on spent-fuel nondestructive measurements to April 1981. Appendix D is a compendium of spent-fuel calculations that includes isotope production and depletion calculations using the EPRI-CINDER code, calculations of neutron and gamma-ray source terms, and correlations of these sources with burnup and plutonium content. Appendix E describes the pulsed-neutron technique and its potential application to spent-fuel measurements. Although not yet developed, the technique holds the promise of providing separate measurements of the uranium and plutonium fissile isotopes. Appendix F describes the experimental program and facilities at Los Alamos for the development of spent-fuel nondestructive measurement systems. Measurements are reported showing that the active neutron method is sensitive to the replacement of a single fuel rod with a dummy rod in an unirradiated uranium fuel assembly

  14. Nuclear fuels

    International Nuclear Information System (INIS)

    Gangwani, Saloni; Chakrabortty, Sumita

    2011-01-01

    Nuclear fuel is a material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned for energy. Nuclear fuels are the most dense sources of energy available. Nuclear fuel in a nuclear fuel cycle can refer to the fuel itself, or to physical objects (for example bundles composed of fuel rods) composed of the fuel material, mixed with structural, neutron moderating, or neutron reflecting materials. Long-lived radioactive waste from the back end of the fuel cycle is especially relevant when designing a complete waste management plan for SNF. When looking at long-term radioactive decay, the actinides in the SNF have a significant influence due to their characteristically long half-lives. Depending on what a nuclear reactor is fueled with, the actinide composition in the SNF will be different. The following paper will also include the uses. advancements, advantages, disadvantages, various processes and behavior of nuclear fuels

  15. Extended fuel cycle length

    International Nuclear Information System (INIS)

    Bruyere, M.; Vallee, A.; Collette, C.

    1986-09-01

    Extended fuel cycle length and burnup are currently offered by Framatome and Fragema in order to satisfy the needs of the utilities in terms of fuel cycle cost and of overall systems cost optimization. We intend to point out the consequences of an increased fuel cycle length and burnup on reactor safety, in order to determine whether the bounding safety analyses presented in the Safety Analysis Report are applicable and to evaluate the effect on plant licensing. This paper presents the results of this examination. The first part indicates the consequences of increased fuel cycle length and burnup on the nuclear data used in the bounding accident analyses. In the second part of this paper, the required safety reanalyses are presented and the impact on the safety margins of different fuel management strategies is examined. In addition, systems modifications which can be required are indicated

  16. Fuel and nuclear fuel cycle

    International Nuclear Information System (INIS)

    Prunier, C.

    1998-01-01

    The nuclear fuel is studied in detail, the best choice and why in relation with the type of reactor, the properties of the fuel cans, the choice of fuel materials. An important part is granted to the fuel assembly of PWR type reactor and the performances of nuclear fuels are tackled. The different subjects for research and development are discussed and this article ends with the particular situation of mixed oxide fuels ( materials, behavior, efficiency). (N.C.)

  17. FAILED FUEL DISPOSITION STUDY

    Energy Technology Data Exchange (ETDEWEB)

    THIELGES, J.R.

    2004-12-20

    In May 2004 alpha contamination was found on the lid of the pre-filter housing in the Sodium Removal Ion Exchange System during routine filter change. Subsequent investigation determined that the alpha contamination likely came from a fuel pin(s) contained in an Ident-69 (ID-69) type pin storage container serial number 9 (ID-69-9) that was washed in the Sodium Removal System (SRS) in January 2004. Because all evidence indicated that the wash water interacted with the fuel, this ID49 is designated as containing a failed fuel pin with gross cladding defect and was set aside in the Interim Examination and Maintenance (IEM) Cell until it could be determined how to proceed for long term dry storage of the fuel pin container. This ID49 contained fuel pins from the driver fuel assembly (DFA) 16392, which was identified as a Delayed Neutron Monitor (DNM) leaker assembly. However, this DFA was disassembled and the fuel pin that was thought to be the failed pin was encapsulated and was not located in this ID49 container. This failed fuel disposition study discusses two alternatives that could be used to address long term storage for the contents of ID-69-9. The first alternative evaluated utilizes the current method of identifying and storing DNM leaker fuel pin(s) in tubes and thus, verifying that the alpha contamination found in the SRS came from a failed pin in this pin container. This approach will require unloading selected fuel pins from the ID-69, visually examining and possibly weighing suspect fuel pins to identify the failed pin(s), inserting the failed pin(s) in storage tubes, and reloading the fuel pins into ID49 containers. Safety analysis must be performed to revise the 200 Area Interim Storage Area (ISA) Final Safety Analysis Report (FSAR) (Reference 1) for this fuel configuration. The second alternative considered is to store the failed fuel as-is in the ID-69. This was evaluated to determine if this approach would comply with storage requirements. This

  18. FAILED FUEL DISPOSITION STUDY

    International Nuclear Information System (INIS)

    THIELGES, J.R.

    2004-01-01

    In May 2004 alpha contamination was found on the lid of the pre-filter housing in the Sodium Removal Ion Exchange System during routine filter change. Subsequent investigation determined that the alpha contamination likely came from a fuel pin(s) contained in an Ident-69 (ID-69) type pin storage container serial number 9 (ID-69-9) that was washed in the Sodium Removal System (SRS) in January 2004. Because all evidence indicated that the wash water interacted with the fuel, this ID49 is designated as containing a failed fuel pin with gross cladding defect and was set aside in the Interim Examination and Maintenance (IEM) Cell until it could be determined how to proceed for long term dry storage of the fuel pin container. This ID49 contained fuel pins from the driver fuel assembly (DFA) 16392, which was identified as a Delayed Neutron Monitor (DNM) leaker assembly. However, this DFA was disassembled and the fuel pin that was thought to be the failed pin was encapsulated and was not located in this ID49 container. This failed fuel disposition study discusses two alternatives that could be used to address long term storage for the contents of ID-69-9. The first alternative evaluated utilizes the current method of identifying and storing DNM leaker fuel pin(s) in tubes and thus, verifying that the alpha contamination found in the SRS came from a failed pin in this pin container. This approach will require unloading selected fuel pins from the ID-69, visually examining and possibly weighing suspect fuel pins to identify the failed pin(s), inserting the failed pin(s) in storage tubes, and reloading the fuel pins into ID49 containers. Safety analysis must be performed to revise the 200 Area Interim Storage Area (ISA) Final Safety Analysis Report (FSAR) (Reference 1) for this fuel configuration. The second alternative considered is to store the failed fuel as-is in the ID-69. This was evaluated to determine if this approach would comply with storage requirements. This

  19. Nuclear fuel management via fuel quality factor averaging

    International Nuclear Information System (INIS)

    Mingle, J.O.

    1978-01-01

    The numerical procedure of prime number averaging is applied to the fuel quality factor distribution of once and twice-burned fuel in order to evolve a fuel management scheme. The resulting fuel shuffling arrangement produces a near optimal flat power profile both under beginning-of-life and end-of-life conditions. The procedure is easily applied requiring only the solution of linear algebraic equations. (author)

  20. The burnup dependence of light water reactor spent fuel oxidation

    International Nuclear Information System (INIS)

    Hanson, B.D.

    1998-07-01

    Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO 2 is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO 2 to higher oxides. The oxidation of UO 2 has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO 2 oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO 2 to UO 2.4 was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO 2.4 to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO 2 oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO 2 and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies associated with spent fuel oxidation (Section 5)

  1. Sophisticated fuel handling system evolved

    International Nuclear Information System (INIS)

    Ross, D.A.

    1988-01-01

    The control systems at Sellafield fuel handling plant are described. The requirements called for built-in diagnostic features as well as the ability to handle a large sequencing application. Speed was also important; responses better than 50ms were required. The control systems are used to automate operations within each of the three main process caves - two Magnox fuel decanners and an advanced gas-cooled reactor fuel dismantler. The fuel route within the fuel handling plant is illustrated and described. ASPIC (Automated Sequence Package for Industrial Control) which was developed as a controller for the plant processes is described. (U.K.)

  2. Development of spent fuel remote handling technology

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Sup; Park, B S; Park, Y S; Oh, S C; Kim, S H; Cho, M W; Hong, D H

    1997-12-01

    Since the nation`s policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs

  3. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

    Yoon, Ji Sup; Park, B. S.; Park, Y. S.; Oh, S. C.; Kim, S. H.; Cho, M. W.; Hong, D. H.

    1997-12-01

    Since the nation's policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs

  4. News from the fuel elements industry

    International Nuclear Information System (INIS)

    Racine, R.; Delannay, M.; Dehon, C.; Jouan, J.; Beuneche, M.

    1981-01-01

    This article deals successively with: the re-structuring of the PWR fuel industry in France, with the setting up of Fragema and Cogema Framatome Combustible; Fragema products, from standard fuel assembly to the development of a new advanced fuel assembly; Framatome's experience with PWR fuel; fuel performances in the light of requirements imposed by network needs follow-up; devices developed by Fragema for on-site analysis of irradiated fuel [fr

  5. Progress on matrix SiC processing and properties for fully ceramic microencapsulated fuel form

    International Nuclear Information System (INIS)

    Terrani, K.A.; Kiggans, J.O.; Silva, C.M.; Shih, C.; Katoh, Y.; Snead, L.L.

    2015-01-01

    The consolidation mechanism and resulting properties of the silicon carbide (SiC) matrix of fully ceramic microencapsulated (FCM) fuel form are discussed. The matrix is produced via the nano-infiltration transient eutectic-forming (NITE) process. Coefficient of thermal expansion, thermal conductivity, and strength characteristics of this SiC matrix have been characterized in the unirradiated state. An ad hoc methodology for estimation of thermal conductivity of the neutron-irradiated NITE–SiC matrix is also provided to aid fuel performance modeling efforts specific to this concept. Finally, specific processing methods developed for production of an optimal and reliable fuel form using this process are summarized. These various sections collectively report the progress made to date on production of optimal FCM fuel form to enable its application in light water and advanced reactors

  6. Fuel assembly

    International Nuclear Information System (INIS)

    Sakuyama, Tadashi; Mukai, Hideyuki.

    1988-01-01

    Purpose: To prevent the bending of a fuel rod caused by the difference in the elongation between a joined fuel rod and a standard fuel rod thereby maintain the fuel rod integrity. Constitution: A joined fuel rod is in a thread engagement at its lower end plug thereof with a lower plate, while passed through at its upper end plug into an upper tie plate and secured with a nut. Further, a standard fuel rod is engaged at its upper end plug and lower end plug with the upper tie plate and the lower tie plate respectively. Expansion springs are mounted to the upper end plugs of these bonded fuel rods and the standard fuel rods for preventing this lifting. Each of the fuel rods comprises a plurality of sintered pellets of nuclear fuel materials laminated in a zircaloy fuel can. The content of the alloy ingredient in the fuel can of the bonded fuel rod is made greater than that of the alloy ingredient of the standard fuel rod. this can increase the elongation for the bonded fuel rod, and the spring of the standard fuel rod is tightly bonded to prevent the bending. (Yoshino, Y.)

  7. Spent fuel: prediction model development

    International Nuclear Information System (INIS)

    Almassy, M.Y.; Bosi, D.M.; Cantley, D.A.

    1979-07-01

    The need for spent fuel disposal performance modeling stems from a requirement to assess the risks involved with deep geologic disposal of spent fuel, and to support licensing and public acceptance of spent fuel repositories. Through the balanced program of analysis, diagnostic testing, and disposal demonstration tests, highlighted in this presentation, the goal of defining risks and of quantifying fuel performance during long-term disposal can be attained

  8. Radiation induced synthesis of In{sub 2}O{sub 3} nanoparticles - Part II: Synthesis of In{sub 2}O{sub 3} nanoparticles by thermal decomposition of un-irradiated and γ-irradiated indium acetylacetonate

    Energy Technology Data Exchange (ETDEWEB)

    Al-Resheedi, Ajayb Saud; Alhokbany, Norah Saad [Department of Chemistry, College of Science, King Saud University, KSU, (Saudi Arabia); Mahfouz, Refaat Mohammed, E-mail: rmhfouz@science.au.edu.eg [Chemistry Department, Faculty of Science, Assiut University, AUN, (Egypt)

    2015-09-15

    Pure cubic phase, In{sub 2}O{sub 3} nanoparticles with porous structure were synthesized by solid state thermal oxidation of un-irradiated and γ-irradiated indium acetyl acetonate in presence and absence of sodium dodecyl sulphate as surfactant. The as- synthesized In{sub 2}O{sub 3} nanoparticles were characterized by X-ray diffraction (XRD), fourier transformation infrared spectroscopy (FT-IR), scanning electron microscopy (SEM), transition electron microscopy (TEM) and thermogravimetry (TG). The shapes and morphologies of as- synthesized In{sub 2}O{sub 3} nanoparticles were highly affected by γ-irradiation of indium acetyl acetonate precursor and by addition of sodium dodecyl sulphate as surfactant. Calcination of un-irradiated indium acetyl acetonate precursor to 4 hours of 600 °C leads to the formation of spherical- shaped accumulative and merged In{sub 2}O{sub 3} nanoparticles with porous structure, whereas irregular porous architectures composed of pure In{sub 2}O{sub 3} nanoparticles were obtained by using γ-irradiated indium acetylacetonate precursor. The as- prepared In{sub 2}O{sub 3} nano products exhibit photoluminescence emission (PL) property and display thermal stability in a wide range of temperature (25-800 °C) which suggest possible applications in nanoscale optoelectronic devices. (author)

  9. The influence of Boron on creep-rupture behaviour of austenitic unstabilized and Nb-stabilized stainless steel X8CrNi 1613 in unirradiated and irradiated condition

    International Nuclear Information System (INIS)

    Sen, Susant Kumar.

    1976-10-01

    The present study deals with influence of boron on creep-rupture behaviour in unirradiated condition at 650 0 C along with precipitation behaviour, heat-treatment and recrystallization of unstabilized and stabilized steel. The results of creep-rupture tests on unirradiated specimens show that boron exerts a beneficial effect on the rupture life and ductility. Boron losses its beneficial effect on creep properties in unstabilized steel by prolong creeping. The magnitude of beneficial effect of Boron on creep properties depends upon the initial boron distribution which influences the number, size and distribution of the precipitates. Boron promotes the precipitation of type M 23 C 6 Carbides in the grain as well as at the grain boundary. Boron segregates in atomic form during slow cooling from austenitizing temperature. The recrystallization will be delayed by the presence of boron. The results of creep tests at 650 0 C shows that boron exerts a beneficial effect on creep life of irradiated steels. (orig./GSC) [de

  10. Overview of the CANDU fuel handling system for advanced fuel cycles

    International Nuclear Information System (INIS)

    Koivisto, D.J.; Brown, D.R.

    1997-01-01

    Because of its neutron economies and on-power re-fuelling capabilities the CANDU system is ideally suited for implementing advanced fuel cycles because it can be adapted to burn these alternative fuels without major changes to the reactor. The fuel handling system is adaptable to implement advanced fuel cycles with some minor changes. Each individual advanced fuel cycle imposes some new set of special requirements on the fuel handling system that is different from the requirements usually encountered in handling the traditional natural uranium fuel. These changes are minor from an overall plant point of view but will require some interesting design and operating changes to the fuel handling system. Some preliminary conceptual design has been done on the fuel handling system in support of these fuel cycles. Some fuel handling details were studies in depth for some of the advanced fuel cycles. This paper provides an overview of the concepts and design challenges. (author)

  11. User requirements for innovative nuclear reactors and fuel cycle technologies in the area of economics, environment, safety, waste management, proliferation resistance and cross cutting issues, and methodology for innovative technologies assessment

    International Nuclear Information System (INIS)

    Kupitz, Juergen; Depisch, Frank; Allan, Colin

    2003-01-01

    The IAEA General Conference in 2000 has invited ''all interested Member States to combine their efforts under the aegis of the Agency in considering the issues of the nuclear fuel cycle, in particular by examining innovative and proliferation-resistant nuclear technology''. In response to this invitation, the IAEA initiated an ''International Project on Innovative Nuclear Reactors and Fuel Cycles'', INPRO. The overall objectives of INPRO is to help to ensure that nuclear energy is available to contribute in fulfilling in a sustainable manner energy needs in the 21st century, and to bring together all interested Member States, both technology holders and technology users, to consider jointly the international and national actions required to achieve desired innovations in nuclear reactors and fuel cycles that use sound and economically competitive technology. Phase-I of INPRO was initiated in May 2001. During Phase-I, work was subdivided in two sub phase: Phase 1A (finished in June 2003) and Phase 1B (started in June 2003). Phase 1A dealt with the definition of Basic Principles, User Requirements and Criteria, and the development of a methodology for the evaluation of innovative nuclear technologies. In Phase 1A, task groups for several areas were established: (a) Prospects and Potentials of Nuclear Power, (b) Economics; (c) Sustainability and Environment, (d) Safety of Nuclear Installations, (e) Waste Management, (f) Proliferation Resistance, (g) Crosscutting issues and (h) for the Methodology for Assessment. In Phase-IB evaluations of innovative nuclear energy technologies will be performed by Member States against the INPRO Basic Principles, User Requirements and Criteria. This paper summarizes the results achieved in the Phase 1A of INPRO and is a cooperative effort of the INPRO team, consisting of all INPRO cost free experts and task managers. (author)

  12. Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

    International Nuclear Information System (INIS)

    Choi, Hang Bok; Ryu, Ho Jin; Roh, Gyu Hong; Jeong, Chang Joon; Park, Chang Je; Song, Kee Chan; Lee, Jung Won

    2005-10-01

    This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor

  13. Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Ryu, Ho Jin; Roh, Gyu Hong; Jeong, Chang Joon; Park, Chang Je; Song, Kee Chan; Lee, Jung Won

    2005-10-15

    This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor.

  14. Reactor fuel element and fuel assembly

    International Nuclear Information System (INIS)

    Okada, Seiji; Ishida, Tsuyoshi; Ikeda, Atsuko.

    1997-01-01

    A mixture of fission products and burnable poisons is disposed at least to a portion between MOX pellets to form a burnable poison-incorporated fuel element without mixing burnable poisons to the MOX pellets. Alternatively, a mixture of materials other than the fission products and burnable poisons is formed into disks, a fuel lamination portion is divided into at least to two regions, and the ratio of number of the disks of the mixture relative to the volume of the region is increased toward the lower portion of the fuel lamination portion. With such a constitution, the axial power distribution of fuels can be made flat easily. Alternatively, the thickness of the disk of the mixture is increased toward the lower region of the fuel lamination portion to flatten the axial power distribution of the fuels in the same manner easily. The time and the cost required for the manufacture are reduced, and MOX fuels filled with burnable poisons with easy maintenance and control can be realized. (N.H.)

  15. Reprocessing of nuclear fuels

    International Nuclear Information System (INIS)

    Hatfield, G.W.

    1960-11-01

    One of the persistent ideas concerning nuclear power is that the fuel costs are negligible. This, of course, is incorrect and, in fact, one of the major problems in the development of economic nuclear power is to get the cost of the fuel cycles down to an acceptable level. The irradiated fuel removed from the nuclear power reactors must be returned as fresh fuel into the system. Aside from the problems of handling and shipping involved in the reprocessing cycles, the two major steps are the chemical separation and the refabrication. The chemical separation covers the processing of the spent fuel to separate and recover the unburned fuel as well as the new fuel produced in the reactor. This includes the decontamination of these materials from other radioactive fission products formed in the reactor. Refabrication involves the working and sheathing of recycled fuel into the shapes and forms required by reactor design and the economics of the fabrication problem determines to a large extent the quality of the material required from the chemical treatment. At present there appear to be enough separating facilities in the United States and the United Kingdom to handle the recycling of fuel from power reactors for the next few years. However, we understand the costs of recycling fuel in these facilities will be high or low depend ing on whether or not the capital costs of the plant are included in the processing cost. Also, the present plants may not be well adapted to carry out the chemical processing of the very wide variety of power reactor fuel elements which are being considered and will continue to be considered over the years to come. (author)

  16. 16 CFR 309.10 - Alternative vehicle fuel rating.

    Science.gov (United States)

    2010-01-01

    ... 16 Commercial Practices 1 2010-01-01 2010-01-01 false Alternative vehicle fuel rating. 309.10... LABELING REQUIREMENTS FOR ALTERNATIVE FUELS AND ALTERNATIVE FUELED VEHICLES Requirements for Alternative Fuels Duties of Importers, Producers, and Refiners of Non-Liquid Alternative Vehicle Fuels (other Than...

  17. Automotive Fuel Processor Development and Demonstration with Fuel Cell Systems

    Energy Technology Data Exchange (ETDEWEB)

    Nuvera Fuel Cells

    2005-04-15

    The potential for fuel cell systems to improve energy efficiency and reduce emissions over conventional power systems has generated significant interest in fuel cell technologies. While fuel cells are being investigated for use in many applications such as stationary power generation and small portable devices, transportation applications present some unique challenges for fuel cell technology. Due to their lower operating temperature and non-brittle materials, most transportation work is focusing on fuel cells using proton exchange membrane (PEM) technology. Since PEM fuel cells are fueled by hydrogen, major obstacles to their widespread use are the lack of an available hydrogen fueling infrastructure and hydrogen's relatively low energy storage density, which leads to a much lower driving range than conventional vehicles. One potential solution to the hydrogen infrastructure and storage density issues is to convert a conventional fuel such as gasoline into hydrogen onboard the vehicle using a fuel processor. Figure 2 shows that gasoline stores roughly 7 times more energy per volume than pressurized hydrogen gas at 700 bar and 4 times more than liquid hydrogen. If integrated properly, the fuel processor/fuel cell system would also be more efficient than traditional engines and would give a fuel economy benefit while hydrogen storage and distribution issues are being investigated. Widespread implementation of fuel processor/fuel cell systems requires improvements in several aspects of the technology, including size, startup time, transient response time, and cost. In addition, the ability to operate on a number of hydrocarbon fuels that are available through the existing infrastructure is a key enabler for commercializing these systems. In this program, Nuvera Fuel Cells collaborated with the Department of Energy (DOE) to develop efficient, low-emission, multi-fuel processors for transportation applications. Nuvera's focus was on (1) developing fuel

  18. Fuel cells : a viable fossil fuel alternative

    Energy Technology Data Exchange (ETDEWEB)

    Paduada, M.

    2007-02-15

    This article presented a program initiated by Natural Resources Canada (NRCan) to develop proof-of-concept of underground mining vehicles powered by fuel cells in order to eliminate emissions. Recent studies on American and Canadian underground mines provided the basis for estimating the operational cost savings of switching from diesel to fuel cells. For the Canadian mines evaluated, the estimated ventilation system operating cost reductions ranged from 29 per cent to 75 per cent. In order to demonstrate the viability of a fuel cell-powered vehicle, NRCan has designed a modified Caterpillar R1300 loader with a 160 kW hybrid power plant in which 3 stacks of fuel cells deliver up to 90 kW continuously, and a nickel-metal hydride battery provides up to 70 kW. The battery subsystem transiently boosts output to meet peak power requirements and also accommodates regenerative braking. Traction for the loader is provided by a brushless permanent magnet traction motor. The hydraulic pump motor is capable of a 55 kW load continuously. The loader's hydraulic and traction systems are operated independently. Future fuel cell-powered vehicles designed by the program may include a locomotive and a utility vehicle. Future mines running their operations with hydrogen-fueled equipment may also gain advantages by employing fuel cells in the operation of handheld equipment such as radios, flashlights, and headlamps. However, the proton exchange membrane (PEM) fuel cells used in the project are prohibitively expensive. The catalytic content of a fuel cell can add hundreds of dollars per kW of electric output. Production of catalytic precious metals will be strongly connected to the scale of use and acceptance of fuel cells in vehicles. In addition, the efficiency of hydrogen production and delivery is significantly lower than the well-to-tank efficiency of many conventional fuels. It was concluded that an adequate hydrogen infrastructure will be required for the mining industry

  19. Distillate Fuel Trends: International Supply Variations and Alternate Fuel Properties

    Science.gov (United States)

    2013-01-31

    fuel in NATO countries will have some amount of FAME present. There is some work being done on hydrocarbon alternatives but the regulatory structure ... synthesis or hydrotreatment – Requirements and test methods.” According to the specification, paraffinic diesel fuel does not meet the current requirements...or international specification for triglyceride based fuel oils (straight vegetable oil / raw vegetable oil). The same holds true for alcohol-based

  20. Probabilistic assessment of light water reactor fuel performance

    International Nuclear Information System (INIS)

    Misfeldt, I.

    1978-10-01

    A computer system for the statistical evaluation of LWR fuel performance has been developed. The computer code FRP, Fuel Reliability Predictor, calculates the distributions for parameters characterizing the fuel performance and failure probability. The statistical methods employed are either Monte Carlo simulations or low order Taylor approximation. Included in the computer system is a deterministic fuel performance code, which has been verified by comparison with data from irradiation experiments. The distributions for all material data utilized in the fuel simulations are estimations from the best available information in the literature. For the failure prediction, a stress corrosion failure criterion has been derived. The failure criterion is based on data from out-of-reactor stress corrosion experiments performed on unirradiated and irradiated zircaloy with iodine present. By means of an example the typical distributions of the variables characterizing the fuel performance and the accuracy of the methods themselves have been investigated. The application of the computer system is illustrated by a number of examples, these include the evaluation of irradiation experiments, design comparisons, and analyses of minor accidents. (author)

  1. Alternate aircraft fuels prospects and operational implications

    Science.gov (United States)

    Witcofski, R. D.

    1977-01-01

    The paper discusses NASA studies of the potentials of coal-derived aviation fuels, specifically synthetic aviation kerosene, liquid methane, and liquid hydrogen. Topics include areas of fuel production, air terminal requirements for aircraft fueling (for liquid hydrogen only), and the performance characteristics of aircraft designed to utilize alternate fuels. Energy requirements associated with the production of each of the three selected fuels are determined, and fuel prices are estimated. Subsonic commercial air transports using liquid hydrogen fuel have been analyzed, and their performance and the performance of aircraft which use commercial aviation kerosene are compared. Environmental and safety issues are considered.

  2. Catalytic processes for cleaner fuels

    International Nuclear Information System (INIS)

    Catani, R.; Marchionna, M.; Rossini, S.

    1999-01-01

    More stringent limitations on vehicle emissions require different measurement: fuel reformulation is one of the most important and is calling for a noticeable impact on refinery assets. Composition rangers of the future fuels have been defined on a time scale. In this scenario the evolution of catalytic technologies becomes a fundamental tool for allowing refinery to reach the fixed-by-law targets. In this paper, the refinery process options to meet each specific requirements of reformulated fuels are surveyed [it

  3. Non-destructive test for VHTR fuel using 160kV X-ray system in Hotcell

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jun; Yoo, Boung Ok; Choo, Yong sun; Baik Sang youl; Kim, Hee Moon; Ahn, Sang Bok [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The research for VHTR which is one of the next generation reactor has been actively carried out. As a part of the research for VHTR, an irradiation examination for the VHTR fuel was performed to confirm an in-pile behavior in HANARO. The non-destructive test for the irradiated fuel is very important to understand the in-pile behavior of the fuel. Especially, the X-ray system is useful to observe the fuel shape without destruction. A dimensional change and defect of the fuel can be confirmed thorough the Xray system. Also, using the 3-D software and CT technology, the fuel shape can be intuitionally observed. The 450kV and 160kV X-ray system were installed and operated in IMEF hotcell. The 160kV X-ray system relatively using a low voltage is suitable to a small scale sample. And high resolution images can be obtained. In this study, the non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. Through these test, the possibility for the X-ray inspection of irradiated fuel was confirmed. The non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. The clear images of the irradiated coated particle were produced without the radiation damage during the Xray inspection. The X-ray images of the VHTR fuel will be utilized as the in-pile performance validation data.

  4. Non-destructive test for VHTR fuel using 160kV X-ray system in Hotcell

    International Nuclear Information System (INIS)

    Kim, Young Jun; Yoo, Boung Ok; Choo, Yong sun; Baik Sang youl; Kim, Hee Moon; Ahn, Sang Bok

    2016-01-01

    The research for VHTR which is one of the next generation reactor has been actively carried out. As a part of the research for VHTR, an irradiation examination for the VHTR fuel was performed to confirm an in-pile behavior in HANARO. The non-destructive test for the irradiated fuel is very important to understand the in-pile behavior of the fuel. Especially, the X-ray system is useful to observe the fuel shape without destruction. A dimensional change and defect of the fuel can be confirmed thorough the Xray system. Also, using the 3-D software and CT technology, the fuel shape can be intuitionally observed. The 450kV and 160kV X-ray system were installed and operated in IMEF hotcell. The 160kV X-ray system relatively using a low voltage is suitable to a small scale sample. And high resolution images can be obtained. In this study, the non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. Through these test, the possibility for the X-ray inspection of irradiated fuel was confirmed. The non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. The clear images of the irradiated coated particle were produced without the radiation damage during the Xray inspection. The X-ray images of the VHTR fuel will be utilized as the in-pile performance validation data.

  5. Test plan for reactions between spent fuel and J-13 well water under unsaturated conditions

    International Nuclear Information System (INIS)

    Finn, P.A.; Wronkiewicz, D.J.; Hoh, J.C.; Emery, J.W.; Hafenrichter, L.D.; Bates, J.K.

    1993-01-01

    The Yucca Mountain Site Characterization Project is evaluating the long-term performance of a high-level nuclear waste form, spent fuel from commercial reactors. Permanent disposal of the spent fuel is possible in a potential repository to be located in the volcanic tuff beds near Yucca Mountain, Nevada. During the post-containment period the spent fuel could be exposed to water condensation since of the cladding is assumed to fail during this time. Spent fuel leach (SFL) tests are designed to simulate and monitor the release of radionuclides from the spent fuel under this condition. This Test Plan addresses the anticipated conditions whereby spent fuel is contacted by small amounts of water that trickle through the spent fuel container. Two complentary test plans are presented, one to examine the reaction of spent fuel and J-13 well water under unsaturated conditions and the second to examine the reaction of unirradiated UO 2 pellets and J-13 well water under unsaturated conditions. The former test plan examines the importance of the water content, the oxygen content as affected by radiolysis, the fuel burnup, fuel surface area, and temperature. The latter test plant examines the effect of the non-presence of Teflon in the test vessel

  6. 10 CFR 503.14 - Fuels search.

    Science.gov (United States)

    2010-01-01

    ... ENERGY (CONTINUED) ALTERNATE FUELS NEW FACILITIES General Requirements for Exemptions § 503.14 Fuels search. Prior to submitting a petition for a permanent exemption for lack of alternate fuel supply, site... where OFE and the petitioner can reach accord, it may evaluate use of a different alternate fuel in lieu...

  7. 49 CFR 229.217 - Fuel tank.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Fuel tank. 229.217 Section 229.217 Transportation... tank. (a) External fuel tanks. Locomotives equipped with external fuel tanks shall, at a minimum... to the fuel tank safety requirements of § 238.223 or § 238.423 of this chapter. The Director of the...

  8. Fuel assembly

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1970-01-01

    Herein disclosed is a fuel assembly in which a fuel rod bundle is easily detachable by rotating a fuel rod fastener rotatably mounted to the upper surface of an upper tie-plate supporting a fuel bundle therebelow. A locking portion at the leading end of each fuel rod protrudes through the upper tie-plate and is engaged with or separated from the tie-plate by the rotation of the fastener. The removal of a desired fuel rod can therefore be remotely accomplished without the necessity of handling pawls, locking washers and nuts. (Owens, K.J.)

  9. Nuclear fuel

    International Nuclear Information System (INIS)

    D Hondt, P.

    1998-01-01

    The research and development programme on nuclear fuel at the Belgian Nuclear Research Centre SCK/CEN is described. The objective of this programme is to enhance the quantitative prediction of the operational limits of nuclear fuel and to assess the behaviour of fuel under incidental and accidental conditions. Progress is described in different domains including the modelling of fission gas release in LWR fuel, thermal conductivity, basic physical phenomena, post-irradiation examination for fuel performance assessment, and conceptual studies of incidental and accidental fuel experiments

  10. Fuels and Combustion

    KAUST Repository

    Johansson, Bengt

    2016-08-17

    This chapter discusses the combustion processes and the link to the fuel properties that are suitable for them. It describes the basic three concepts, including spark ignition (SI) and compression ignition (CI), and homogeneous charge compression ignition (HCCI). The fuel used in a CI engine is vastly different from that in an SI engine. In an SI engine, the fuel should sustain high pressure and temperature without autoignition. Apart from the dominating SI and CI engines, it is also possible to operate with a type of combustion: autoignition. With HCCI, the fuel and air are fully premixed before combustion as in the SI engine, but combustion is started by the increased pressure and temperature during the compression stroke. Apart from the three combustion processes, there are also a few combined or intermediate concepts, such as Spark-Assisted Compression Ignition (SACI). Those concepts are discussed in terms of the requirements of fuel properties.

  11. Fuels and Combustion

    KAUST Repository

    Johansson, Bengt

    2016-01-01

    This chapter discusses the combustion processes and the link to the fuel properties that are suitable for them. It describes the basic three concepts, including spark ignition (SI) and compression ignition (CI), and homogeneous charge compression ignition (HCCI). The fuel used in a CI engine is vastly different from that in an SI engine. In an SI engine, the fuel should sustain high pressure and temperature without autoignition. Apart from the dominating SI and CI engines, it is also possible to operate with a type of combustion: autoignition. With HCCI, the fuel and air are fully premixed before combustion as in the SI engine, but combustion is started by the increased pressure and temperature during the compression stroke. Apart from the three combustion processes, there are also a few combined or intermediate concepts, such as Spark-Assisted Compression Ignition (SACI). Those concepts are discussed in terms of the requirements of fuel properties.

  12. Uranium plutonium oxide fuels

    International Nuclear Information System (INIS)

    Cox, C.M.; Leggett, R.D.; Weber, E.T.

    1981-01-01

    Uranium plutonium oxide is the principal fuel material for liquid metal fast breeder reactors (LMFBR's) throughout the world. Development of this material has been a reasonably straightforward evolution from the UO 2 used routinely in the light water reactor (LWR's); but, because of the lower neutron capture cross sections and much lower coolant pressures in the sodium cooled LMFBR's, the fuel is operated to much higher discharge exposures than that of a LWR. A typical LMFBR fuel assembly is shown. Depending on the required power output and the configuration of the reactor, some 70 to 400 such fuel assemblies are clustered to form the core. There is a wide variation in cross section and length of the assemblies where the increasing size reflects a chronological increase in plant size and power output as well as considerations of decreasing the net fuel cycle cost. Design and performance characteristics are described

  13. Castor oil polyurethane as a coating option for spent nuclear fuel disposal containment

    Energy Technology Data Exchange (ETDEWEB)

    Mortley, A.; Bonin, H.W.; Bui, V.T. [Dept. of Chemistry and Chemical Engineering, Royal Military College of Canada, Kingston, Ontario (Canada)

    2009-07-01

    Castor oil polyurethane (COPU) coatings are being proposed as an additional barrier in the design of the copper containers to store spent nuclear fuel in Canada. The present work investigates the variation in the physicomechanical properties of two COPUs, based on an aliphatic and aromatic diisocyanate, as a function of ionizing radiation dose and dose rate. The changes in physicomechanical properties have shown that radiation, regardless of dose rate and isocyanate structure, increases the values of the modulus and the ultimate tensile strength when compared with those of the unirradiated samples, with aromatic based polyurethanes being more susceptible to variation than aliphatic based ones. (author)

  14. Castor oil polyurethane as a coating option for spent nuclear fuel disposal containment

    International Nuclear Information System (INIS)

    Mortley, A.; Bonin, H.W.; Bui, V.T.

    2009-01-01

    Castor oil polyurethane (COPU) coatings are being proposed as an additional barrier in the design of the copper containers to store spent nuclear fuel in Canada. The present work investigates the variation in the physicomechanical properties of two COPUs, based on an aliphatic and aromatic diisocyanate, as a function of ionizing radiation dose and dose rate. The changes in physicomechanical properties have shown that radiation, regardless of dose rate and isocyanate structure, increases the values of the modulus and the ultimate tensile strength when compared with those of the unirradiated samples, with aromatic based polyurethanes being more susceptible to variation than aliphatic based ones. (author)

  15. Fuel assembly

    International Nuclear Information System (INIS)

    Wataumi, Kazutoshi; Tajiri, Hiroshi.

    1992-01-01

    In a fuel assembly of a BWR type reactor, a pellet to be loaded comprises an external layer of fissile materials containing burnable poisons and an internal layer of fissile materials not containing burnable poison. For example, there is provided a dual type pellet comprising an external layer made of UO 2 incorporated with Gd 2 O 3 at a predetermined concentration as the burnable poisons and an internal layer made of UO 2 not containing Gd 2 O 3 . The amount of the burnable poisons required for predetermined places is controlled by the thickness of the ring of the external layer. This can dissipate an unnecessary poisoning effect at the final stage of the combustion cycle. Further, since only one or a few kinds of powder mixture of the burnable poisons and the fissile materials is necessary, production and product control can be facilitated. (I.N.)

  16. Quality management of nuclear fuel

    International Nuclear Information System (INIS)

    2006-01-01

    The Guide presents the quality management requirements to be complied with in the procurement, design, manufacture, transport, receipt, storage, handling and operation of nuclear fuel. The Guide also applies to control rods and shield elements to be placed in the reactor. The Guide is mainly aimed for the licensee responsible for the procurement and operation of fuel, for the fuel designer and manufacturer and for other organisations, whose activities affect fuel quality and the safety of fuel transport, storage and operation. General requirements for nuclear fuel are presented in Section 114 of the Finnish Nuclear Energy Decree and in Section 15 of the Government Decision (395/1991). Regulatory control of the safety of fuel is described in Guides YVL6.1, YVL6.2 and YVL6.3. An overview of the regulatory control of nuclear power plants carried out by STUK (Radiation and Nuclear Safety Authority, Finland) is clarified in Guide YVL1.1

  17. Measurement of soluble nuclide dissolution rates from spent fuel

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  18. Fuel storage

    International Nuclear Information System (INIS)

    Palacios, C.; Alvarez-Miranda, A.

    2009-01-01

    ENSA is a well known manufacturer of multi-system primary components for the nuclear industry and is totally prepared to satisfy future market requirements in this industry. At the same time that ENSA has been gaining a reputation world wider for the supply of primary components, has been strengthening its commitment and experience in supplying spent fuel components, either pool racks or storage and transportation casks, and offers not only fabrication but also design capabilities for its products. ENSA has supplied Spent Fuel Pool Racks, in spain, Finland, Taiwan, Korea, China, and currently it is in the process of licensing its own rack design in the United States of America for the ESBWR along with Ge-Hitachi. ENSA has supplied racks for 20 pools and 22 different reactors and it has also manufactured racks under all available technologies and developed a design known as Interlock Cell Matrix whose main features are outlined in this article. Another ENSA achievement in rack technology is the use of remote control for re-racking activities instead of using divers, which improves the ALARA requirements. Regarding casks for storage and transportation, ENSA also has al leading worldwide position, with exports prevailing over the Spanish market where ENSA has supplied 16 storage and transportation casks to the Spanish nuclear power Trillo. In some cases, ENSA acts as subcontractor for other clients. Foreign markets are still a major challenge for ENSA. ENSA-is well known for its manufacturing capabilities in the nuclear industry, but has been always involved in design activities through its engineering division, which carries out different tasks: components Design; Tooling Design; Engineering and Documentation; Project Engineering; Calculations, Design and Development Engineering. (Author)

  19. Spent fuel management

    International Nuclear Information System (INIS)

    2005-01-01

    The production of nuclear electricity results in the generation of spent fuel that requires safe, secure and efficient management. Appropriate management of the resulting spent fuel is a key issue for the steady and sustainable growth of nuclear energy. Currently about 10,000 tonnes heavy metal (HM) of spent fuel are unloaded every year from nuclear power reactors worldwide, of which 8,500 t HM need to be stored (after accounting for reprocessed fuel). This is the largest continuous source of civil radioactive material generated, and needs to be managed appropriately. Member States have referred to storage periods of 100 years and even beyond, and as storage quantities and durations extend, new challenges arise in the institutional as well as in the technical area. The IAEA gives high priority to safe and effective spent fuel management. As an example of continuing efforts, the 2003 International Conference on Storage of Spent Fuel from Power Reactors gathered 125 participants from 35 member states to exchange information on this important subject. With its large number of Member States, the IAEA is well-positioned to gather and share information useful in addressing Member State priorities. IAEA activities on this topic include plans to produce technical documents as resources for a range of priority topics: spent fuel performance assessment and research, burnup credit applications, cask maintenance, cask loading optimization, long term storage requirements including records maintenance, economics, spent fuel treatment, remote technology, and influence of fuel design on spent fuel storage. In addition to broader topics, the IAEA supports coordinated research projects and technical cooperation projects focused on specific needs

  20. The effect of fuel chemistry on UO{sub 2} dissolution

    Energy Technology Data Exchange (ETDEWEB)

    Casella, Amanda, E-mail: amanda.casella@pnnl.gov [Pacific Northwest National Laboratory, PO Box 999, MSIN P7-25, Richland, WA 99352 (United States); Hanson, Brady, E-mail: brady.hanson@pnnl.gov [Pacific Northwest National Laboratory, PO Box 999, MSIN P7-27, Richland, WA 99352 (United States); Miller, William [University of Missouri Research Reactor, 1513 Research Park Drive, Columbia, MO 65211 (United States)

    2016-08-01

    The dissolution rate of both unirradiated UO{sub 2} and used nuclear fuel has been studied by numerous countries as part of the performance assessment of proposed geologic repositories. In the scenario of waste package failure and groundwater contact with the fuel, the effects of variables such as temperature, dissolved oxygen, and water and fuel chemistry on the dissolution rates of the fuel are necessary to provide a quantitative estimate of the potential release over geologic time frames. The primary objective of this research was to determine the influence these parameters, with primary focus on the fuel chemistry, have on the dissolution rate of unirradiated UO{sub 2} under oxidizing repository conditions and compare them to the rates predicted by current dissolution models. Both unirradiated UO{sub 2} and UO{sub 2} doped with varying concentrations of Gd{sub 2}O{sub 3}, to simulate used fuel composition after long time periods when radiolysis has minor contributions to dissolution, were examined. In general, a rise in temperature increased the dissolution rate of UO{sub 2} and had a larger effect on pure UO{sub 2} than on those doped with Gd{sub 2}O{sub 3}. Oxygen dependence was observed in the UO{sub 2} samples with no dopant and increased as the temperature rose; in the doped fuels less dependence was observed. The addition of gadolinia into the UO{sub 2} matrix resulted in a significant decrease in the dissolution rate. The matrix stabilization effect resulting from the dopant proved even more beneficial in lowering the dissolution rate at higher temperatures and dissolved O{sub 2} concentrations in the leachate where the rates would typically be elevated. - Highlights: • UO{sub 2} dissolution rates were measured for a matrix of repository relevant conditions. • Dopants in the UO{sub 2} matrix lowered the dissolution rate. • Reduction in rates by dopants were increased at elevated temperature and O{sub 2} levels. • UO{sub 2} may be overly