Research Reactor Preparations for the Air Shipment of Highly Enriched Uranium from Romania

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

K. J. Allen; I. Bolshinsky; L. L. Biro

2010-03-01

In June 2009 two air shipments transported both unirradiated (fresh) and irradiated (spent) Russian-origin highly enriched uranium (HEU) nuclear fuel from two research reactors in Romania to the Russian Federation for conversion to low enriched uranium. The Institute for Nuclear Research at Pitesti (SCN Pitesti) shipped 30.1 kg of HEU fresh fuel pellets to Dimitrovgrad, Russia and the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH) shipped 23.7 kilograms of HEU spent fuel assemblies from the VVR S research reactor at Magurele, Romania, to Chelyabinsk, Russia. Both HEU shipments were coordinated by the Russian Research Reactor Fuel Returnmore » Program (RRRFR) as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), were managed in Romania by the National Commission for Nuclear Activities Control (CNCAN), and were conducted in cooperation with the Russian Federation State Corporation Rosatom and the International Atomic Energy Agency. Both shipments were transported by truck to and from respective commercial airports in Romania and the Russian Federation and stored at secure nuclear facilities in Russia until the material is converted into low enriched uranium. These shipments resulted in Romania becoming the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the research reactor preparations and license approvals that were necessary to safely and securely complete these air shipments of nuclear fuel.« less

A pulse-controlled modified-burst test instrument for accident-tolerant fuel cladding

DOE PAGES

Cinbiz, M. Nedim; Brown, Nicholas R.; Terrani, Kurt A.; ...

2017-06-03

Pellet-cladding mechanical interaction due to thermal expansion of nuclear fuel pellets during a reactivity-initiated accident (RIA) is a potential mechanism for failure of nuclear fuel cladding. To investigate the mechanical behavior of cladding during an RIA, we developed a mechanical pulse-controlled modified burst test instrument that simulates transient events with a pulse width from 10 to 300 ms. This paper includes validation tests of unirradiated and prehydrided ZIRLO cladding tubes. A ZIRLO cladding sample with a hydrogen content of 168 wt. ppm showed ductile behavior and failed at the maximum limits of the test setup with hoop strain to failuremore » greater than 9.2%. ZIRLO samples showed high resistance to failure even at very high hydrogen contents (1,466 wt. ppm). When the hydrogen content was increased to 1,554 wt. ppm, brittle-like behavior was observed at a hoop strain of 2.5%. Preliminary scoping tests at room temperature with FeCrAl tubes were conducted to imitate the pulse behavior of transient test reactors during integral tests. The preliminary FeCrAl tests are informative from the perspective of characterizing the test rig and supporting the design of integral tests for current and potentially accident tolerant cladding materials.« less

A pulse-controlled modified-burst test instrument for accident-tolerant fuel cladding

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

Cinbiz, M. Nedim; Brown, Nicholas R.; Terrani, Kurt A.

Pellet-cladding mechanical interaction due to thermal expansion of nuclear fuel pellets during a reactivity-initiated accident (RIA) is a potential mechanism for failure of nuclear fuel cladding. To investigate the mechanical behavior of cladding during an RIA, we developed a mechanical pulse-controlled modified burst test instrument that simulates transient events with a pulse width from 10 to 300 ms. This paper includes validation tests of unirradiated and prehydrided ZIRLO cladding tubes. A ZIRLO cladding sample with a hydrogen content of 168 wt. ppm showed ductile behavior and failed at the maximum limits of the test setup with hoop strain to failuremore » greater than 9.2%. ZIRLO samples showed high resistance to failure even at very high hydrogen contents (1,466 wt. ppm). When the hydrogen content was increased to 1,554 wt. ppm, brittle-like behavior was observed at a hoop strain of 2.5%. Preliminary scoping tests at room temperature with FeCrAl tubes were conducted to imitate the pulse behavior of transient test reactors during integral tests. The preliminary FeCrAl tests are informative from the perspective of characterizing the test rig and supporting the design of integral tests for current and potentially accident tolerant cladding materials.« less

Application of subsize specimens in nuclear plant life extension

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

Rosinski, S.T.; Kumar, A.S.; Cannon, S.C.

1991-01-01

The US Department of Energy is sponsoring a research effort through Sandia National Laboratories and the University of Missour-Rolla to test a correlation for the upper shelf energy (USE) values obtained from the impact testing of subsize Charpy V-notch specimens to those obtained from the testing of full size samples. The program involves the impact testing of unirradiated and irradiated full, half, and third size Charpy V-notch specimens. To verify the applicability of the correlation on LWR materials unirradiated and irradiated full, half, and third size Charpy V-notch specimens of a commercial pressure vessel steel (ASTM A533 Grade B) willmore » be tested. This paper will provide details of the program and present results obtained from the application of the developed correlation methodology to the impact testing of the unirradiated full, half, and third size A533 Grade B Charpy V-notch specimens.« less

Application of subsize specimens in nuclear plant life extension

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

Rosinski, S.T.; Kumar, A.S.; Cannon, S.C.

1991-12-31

The US Department of Energy is sponsoring a research effort through Sandia National Laboratories and the University of Missour-Rolla to test a correlation for the upper shelf energy (USE) values obtained from the impact testing of subsize Charpy V-notch specimens to those obtained from the testing of full size samples. The program involves the impact testing of unirradiated and irradiated full, half, and third size Charpy V-notch specimens. To verify the applicability of the correlation on LWR materials unirradiated and irradiated full, half, and third size Charpy V-notch specimens of a commercial pressure vessel steel (ASTM A533 Grade B) willmore » be tested. This paper will provide details of the program and present results obtained from the application of the developed correlation methodology to the impact testing of the unirradiated full, half, and third size A533 Grade B Charpy V-notch specimens.« less

Modeling and testing miniature torsion specimens for SiC joining development studies for fusion

DOE PAGES

Henager, Jr., C. H.; Nguyen, Ba N.; Kurtz, Richard J.; ...

2015-08-05

The international fusion community has designed a miniature torsion specimen for neutron irradiation studies of joined SiC and SiC/SiC composite materials. For this research, miniature torsion joints based on this specimen design were fabricated using displacement reactions between Si and TiC to produce Ti 3SiC 2 + SiC joints with SiC and tested in torsion-shear prior to and after neutron irradiation. However, many miniature torsion specimens fail out-of-plane within the SiC specimen body, which makes it problematic to assign a shear strength value to the joints and makes it difficult to compare unirradiated and irradiated strengths to determine irradiation effects.more » Finite element elastic damage and elastic–plastic damage models of miniature torsion joints are developed that indicate shear fracture is more likely to occur within the body of the joined sample and cause out-of-plane failures for miniature torsion specimens when a certain modulus and strength ratio between the joint material and the joined material exists. The model results are compared and discussed with regard to unirradiated and irradiated test data for a variety of joint materials. The unirradiated data includes Ti 3SiC 2 + SiC/CVD-SiC joints with tailored joint moduli, and includes steel/epoxy and CVD-SiC/epoxy joints. Finally, the implications for joint data based on this sample design are discussed.« less

Is It Time To Consider Global Sharing of Integral Physics Data?

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

Harold F. McFarlane

The innocent days of the Atoms for Peace program vanished with the suicide attack on the World Trade Center in New York City that occurred while the GLOBAL 2001 international nuclear fuel cycle conference was convened in Paris. Today’s reality is that maintaining an inventory of unirradiated highly enriched uranium or plutonium for critical experiments requires a facility to accept substantial security cost and intrusion. In the context of a large collection of benchmark integral experiments collected over several decades and the ongoing rapid advances in computer modeling and simulation, there seems to be ample incentive to reduce both themore » number of facilities and material inventory quantities worldwide. As a result of ongoing nonproliferation initiatives, there are viable programs that will accept highly enriched uranium for down blending into commercial fuel. Nevertheless, there are formidable hurdles to overcome before national institutions will voluntarily give up existing nuclear research capabilities. GLOBAL 2005 was the appropriate forum to begin fostering a new spirit of cooperation that could lead to improved international security and better use of precious research and development resources, while ensuring access to existing and future critical experiment data.« less

Identification of irradiated spices by the use of thermoluminescence method (TL)

NASA Astrophysics Data System (ADS)

Sharifzadeh, M.; Sohrabpour, M.

1993-07-01

In this paper the results of the investigations of identification of irradiated spices by the use of thermoluminescence method is reported. The materials used were black and red peppers, turmeric, cinnamon, and garlic powder. Gamma Cell 220 was used for irradiating samples at dose values of 2.5, 5, 7.5 and 10 kGy respectively. The TL intensity of the unirradiated spices as well as the fading characteristics of the irradiated samples having received a dose of 10 kGy have been measured. Post-irradiation temperature treatment of the irradiated (10 kGy) and unirradiated samples at 60°C and 100°C for 24 hours have been also performed. The results show that the TL intensities of unirradiated and irradiated samples from different batches of each spice are fairly distributed. A reasonable TL intensity versus dose has been observed in nearly all cases. Based on the observations made it is possible to distinguish irradiated spices after (4-9) months post-irradiation.

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

Weaver, Jordan S.; Pathak, Siddhartha; Reichardt, Ashley

Experimentally quantifying the mechanical effects of radiation damage in reactor materials is necessary for the development and qualification of new materials for improved performance and safety. This can be achieved in a high-throughput fashion through a combination of ion beam irradiation and small scale mechanical testing in contrast to the high cost and laborious nature of bulk testing of reactor irradiated samples. The current paper focuses on using spherical nanoindentation stress-strain curves on unirradiated and proton irradiated (10 dpa at 360 °C) 304 stainless steel to quantify the mechanical effects of radiation damage. Spherical nanoindentation stress-strain measurements show a radiation-inducedmore » increase in indentation yield strength from 1.36 GPa to 2.72 GPa and a radiation-induced increase in indentation work hardening rate of 10 GPa–30 GPa. These measurements are critically compared against Berkovich nanohardness, micropillar compression, and micro-tension measurements on the same material and similar grain orientations. The ratio of irradiated to unirradiated yield strength increases by a similar factor of 2 when measured via spherical nanoindentation or Berkovich nanohardness testing. A comparison of spherical indentation stress-strain curves to uniaxial (micropillar and micro-tension) stress-strain curves was achieved using a simple scaling relationship which shows good agreement for the unirradiated condition and poor agreement in post-yield behavior for the irradiated condition. Finally, the disagreement between spherical nanoindentation and uniaxial stress-strain curves is likely due to the plastic instability that occurs during uniaxial tests but is absent during spherical nanoindentation tests.« less

SILICON CARBIDE GRAIN BOUNDARY DISTRIBUTIONS, IRRADIATION CONDITIONS, AND SILVER RETENTION IN IRRADIATED AGR-1 TRISO FUEL PARTICLES

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

Lillo, T. M.; Rooyen, I. J.; Aguiar, J. A.

Precession electron diffraction in the transmission electron microscope was used to map grain orientation and ultimately determine grain boundary misorientation angle distributions, relative fractions of grain boundary types (random high angle, low angle or coincident site lattice (CSL)-related boundaries) and the distributions of CSL-related grain boundaries in the SiC layer of irradiated TRISO-coated fuel particles. Two particles from the AGR-1 experiment exhibiting high Ag-110m retention (>80%) were compared to a particle exhibiting low Ag-110m retention (<19%). Irradiated particles with high Ag-110m retention exhibited a lower fraction of random, high angle grain boundaries compared to the low Ag-110m retention particle. Anmore » inverse relationship between the random, high angle grain boundary fraction and Ag-110m retention is found and is consistent with grain boundary percolation theory. Also, comparison of the grain boundary distributions with previously reported unirradiated grain boundary distributions, based on SEM-based EBSD for similarly fabricated particles, showed only small differences, i.e. a greater low angle grain boundary fraction in unirradiated SiC. It was, thus, concluded that SiC layers with grain boundary distributions susceptible to Ag-110m release were present prior to irradiation. Finally, irradiation parameters were found to have little effect on the association of fission product precipitates with specific grain boundary types.« less

Spherical nanoindentation of proton irradiated 304 stainless steel: A comparison of small scale mechanical test techniques for measuring irradiation hardening

DOE PAGES

Weaver, Jordan S.; Pathak, Siddhartha; Reichardt, Ashley; ...

2017-06-27

Experimentally quantifying the mechanical effects of radiation damage in reactor materials is necessary for the development and qualification of new materials for improved performance and safety. This can be achieved in a high-throughput fashion through a combination of ion beam irradiation and small scale mechanical testing in contrast to the high cost and laborious nature of bulk testing of reactor irradiated samples. The current paper focuses on using spherical nanoindentation stress-strain curves on unirradiated and proton irradiated (10 dpa at 360 °C) 304 stainless steel to quantify the mechanical effects of radiation damage. Spherical nanoindentation stress-strain measurements show a radiation-inducedmore » increase in indentation yield strength from 1.36 GPa to 2.72 GPa and a radiation-induced increase in indentation work hardening rate of 10 GPa–30 GPa. These measurements are critically compared against Berkovich nanohardness, micropillar compression, and micro-tension measurements on the same material and similar grain orientations. The ratio of irradiated to unirradiated yield strength increases by a similar factor of 2 when measured via spherical nanoindentation or Berkovich nanohardness testing. A comparison of spherical indentation stress-strain curves to uniaxial (micropillar and micro-tension) stress-strain curves was achieved using a simple scaling relationship which shows good agreement for the unirradiated condition and poor agreement in post-yield behavior for the irradiated condition. Finally, the disagreement between spherical nanoindentation and uniaxial stress-strain curves is likely due to the plastic instability that occurs during uniaxial tests but is absent during spherical nanoindentation tests.« less

Spherical nanoindentation of proton irradiated 304 stainless steel: A comparison of small scale mechanical test techniques for measuring irradiation hardening

NASA Astrophysics Data System (ADS)

Weaver, Jordan S.; Pathak, Siddhartha; Reichardt, Ashley; Vo, Hi T.; Maloy, Stuart A.; Hosemann, Peter; Mara, Nathan A.

2017-09-01

Experimentally quantifying the mechanical effects of radiation damage in reactor materials is necessary for the development and qualification of new materials for improved performance and safety. This can be achieved in a high-throughput fashion through a combination of ion beam irradiation and small scale mechanical testing in contrast to the high cost and laborious nature of bulk testing of reactor irradiated samples. The current work focuses on using spherical nanoindentation stress-strain curves on unirradiated and proton irradiated (10 dpa at 360 °C) 304 stainless steel to quantify the mechanical effects of radiation damage. Spherical nanoindentation stress-strain measurements show a radiation-induced increase in indentation yield strength from 1.36 GPa to 2.72 GPa and a radiation-induced increase in indentation work hardening rate of 10 GPa-30 GPa. These measurements are critically compared against Berkovich nanohardness, micropillar compression, and micro-tension measurements on the same material and similar grain orientations. The ratio of irradiated to unirradiated yield strength increases by a similar factor of 2 when measured via spherical nanoindentation or Berkovich nanohardness testing. A comparison of spherical indentation stress-strain curves to uniaxial (micropillar and micro-tension) stress-strain curves was achieved using a simple scaling relationship which shows good agreement for the unirradiated condition and poor agreement in post-yield behavior for the irradiated condition. The disagreement between spherical nanoindentation and uniaxial stress-strain curves is likely due to the plastic instability that occurs during uniaxial tests but is absent during spherical nanoindentation tests.

Spherical nanoindentation of proton irradiated 304 stainless steel: A comparison of small scale mechanical test techniques for measuring irradiation hardening

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

Weaver, Jordan S.; Pathak, Siddhartha; Reichardt, Ashley

Experimentally quantifying the mechanical effects of radiation damage in reactor materials is necessary for the development and qualification of new materials for improved performance and safety. This can be achieved in a high-throughput fashion through a combination of ion beam irradiation and small scale mechanical testing in contrast to the high cost and laborious nature of bulk testing of reactor irradiated samples. The current paper focuses on using spherical nanoindentation stress-strain curves on unirradiated and proton irradiated (10 dpa at 360 °C) 304 stainless steel to quantify the mechanical effects of radiation damage. Spherical nanoindentation stress-strain measurements show a radiation-inducedmore » increase in indentation yield strength from 1.36 GPa to 2.72 GPa and a radiation-induced increase in indentation work hardening rate of 10 GPa–30 GPa. These measurements are critically compared against Berkovich nanohardness, micropillar compression, and micro-tension measurements on the same material and similar grain orientations. The ratio of irradiated to unirradiated yield strength increases by a similar factor of 2 when measured via spherical nanoindentation or Berkovich nanohardness testing. A comparison of spherical indentation stress-strain curves to uniaxial (micropillar and micro-tension) stress-strain curves was achieved using a simple scaling relationship which shows good agreement for the unirradiated condition and poor agreement in post-yield behavior for the irradiated condition. Finally, the disagreement between spherical nanoindentation and uniaxial stress-strain curves is likely due to the plastic instability that occurs during uniaxial tests but is absent during spherical nanoindentation tests.« less

In-pile tests at Karlsruhe of LWR fuel-rod behavior during the heatup phase of a LOCA

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

Karb, E.H.

1980-01-01

In order to investigate the influence of a nuclar environment on the mechanisms of fuel-rod failure, in-pile tests simulating the heatup phase of a loss-of-coolant accident in a pressurized-water reactor are being conducted with irradiated and unirradiated short-length single rods in the FR2 reactor at Kernforschungszentrum karlsruhe (Karlsruhe Nuclear Reasearch Center), Federal Republic of Germany, within the Project Nuclear Safety. With nearly 70% of the scheduled tests completed, no such influences have been found. The in-pile burst and deformation data are in good agreement with results from nonnuclear tests with electrically heated fuel-rod simulators. The phenomenon of pellet disintegration, whichmore » has been observed in all tests with previously irradiated rods, needs further investigation.« less

Quasipermanent magnets of high temperature superconductor - Temperature dependence

NASA Technical Reports Server (NTRS)

Chen, In-Gann; Liu, Jianxiong; Ren, Yanru; Weinstein, Roy; Kozlowski, Gregory; Oberly, Charles E.

1993-01-01

We report on persistent field in quasi-permanent magnets of high temperature superconductors. Magnets composed of irradiated Y(1+)Ba2Cu3O7 trapped field Bt = 1.52 T at 77 K and 1.9 T at lower temperature. However, the activation magnet limited Bt at lower temperature. We present data on Jc(H,T) for unirradiated materials, and calculate Bt at various T. Based upon data at 65 K, we calculate Bt in unirradiated single grains at 20 K and find that 5.2 T will be trapped for grain diameter d about 1.2 cm, and 7.9 T for d = 2.3 cm. Irradiated grains will trap four times these values.

Stress corrosion cracking of Zircaloys in unirradiated and irradiated CsI

NASA Astrophysics Data System (ADS)

Cox, B.; Surette, B. A.; Wood, J. C.

1986-03-01

Unirradiated split-ring specimens of Zircaloy fuel cladding, coated with CsI, cracked when stressed at elevated temperatures. The specimens have been reexamined fractographically and metallographically in order to confirm that the cause of cracking was stress corrosion (SCC) and not delayed hydride cracking (DHC). Further specimens have been cracked at 350°C by a solution of CsI in a fused mixture of nitrates of rubidium, cesium, strontium and barium, by a similar mechanism. CsI dissolved in a fused molybdate melt was not stable at 400°C, and rapidly evolved iodine, leaving a melt that was incapable of causing SCC. Irradiation of stressed split-ring specimens of Zircaloy fuel cladding in a γ-irradiator of 10 6 R/h and in the U-5 loop in the NRU reactor at an estimated 10 9 R/h caused SCC when the specimens were packed in dry CsI powder. Care had to be taken to dry the CsI, otherwise cracking occurred by a DHC mechanism from hydrogen absorbed from residual moisture in the CsI. Fractography showed that the crack surfaces obtained with dry CsI were typical of iodine-induced SCC rather than cesium-induced metal vapour embrittlement. Thus, if a transport process is provided for the iodide to obtain access to the zirconium surface, CsI is capable of causing SCC of Zircaloy. This transport process might be ionic diffusion in a fission product oxide melt in the fuel-clad gap, however, radiolysis of CsI to form a volatile iodine species in a radiation field is the more probable explanation of PCI failures.

Texture evolution and mechanical behaviour of irradiated face-centred cubic metals

NASA Astrophysics Data System (ADS)

Chen, L. R.; Xiao, X. Z.; Yu, L.; Chu, H. J.; Duan, H. L.

2018-02-01

A physically based theoretical model is proposed to investigate the mechanical behaviour and crystallographic texture evolution of irradiated face-centred cubic metals. This model is capable of capturing the main features of irradiated polycrystalline materials including irradiation hardening, post-yield softening and plasticity localization. Numerical results show a good agreement with experimental data for both unirradiated and irradiated stress-strain relationships. The study of crystallographic texture reveals that the initial randomly distributed texture of unirradiated metals under tensile loading can evolve into a mixture of [111] and [100] textures. Regarding the irradiated case, crystallographic texture develops in a different way, and an extra part of [110] texture evolves into [100] and [111] textures. Thus, [100] and [111] textures become dominant more quickly compared with those of the unirradiated case for the reason that [100] and [111]-oriented crystals have higher strength, and their plastic deformation behaviours are more active than other oriented crystals. It can be concluded that irradiation-induced defects can affect both the mechanical behaviour and texture evolution of metals, both of which are closely related to irradiation hardening.

The effect of fission products on the rate of U3O8 formation in SIMFUEL oxidized in air at 250°C

NASA Astrophysics Data System (ADS)

Choi, Jong-Won; McEachern, Rod J.; Taylor, Peter; Wood, Donald D.

1996-06-01

The effect of fission products on the rate of U3O8 formation was investigated by oxidizing UO2-based SIMFUEL (simulated high burnup nuclear fuel) and unirradiated UO2 fuel specimens in air at 250°C for different times (1-317 days). The progress of oxidation was monitored by X-ray diffraction, revealing that the rate of U3O8 formation declines with increasing burnup. An expression was derived to describe quantitatively the time for U3O8 powder formation as a function of simulated burnup. These findings were supported by additional isochronal oxidation experiments conducted between 200 and 300°C.

Separation of actinides from irradiated An-Zr based fuel by electrorefining on solid aluminium cathodes in molten LiCl-KCl

NASA Astrophysics Data System (ADS)

Souček, P.; Murakami, T.; Claux, B.; Meier, R.; Malmbeck, R.; Tsukada, T.; Glatz, J.-P.

2015-04-01

An electrorefining process for metallic spent nuclear fuel treatment is being investigated in ITU. Solid aluminium cathodes are used for homogeneous recovery of all actinides within the process carried out in molten LiCl-KCl eutectic salt at a temperature of 500 °C. As the selectivity, efficiency and performance of solid Al has been already shown using un-irradiated An-Zr alloy based test fuels, the present work was focused on laboratory-scale demonstration of the process using irradiated METAPHIX-1 fuel composed of U67-Pu19-Zr10-MA2-RE2 (wt.%, MA = Np, Am, Cm, RE = Nd, Ce, Gd, Y). Different electrorefining techniques, conditions and cathode geometries were used during the experiment yielding evaluation of separation factors, kinetic parameters of actinide-aluminium alloy formation, process efficiency and macro-structure characterisation of the deposits. The results confirmed an excellent separation and very high efficiency of the electrorefining process using solid Al cathodes.

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

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

Rudisill, T. S.; Shehee, T. C.; Jones, D. H.

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 Pumore » 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.« less

Fracture mechanism maps in unirradiated and irradiated metals and alloys

NASA Astrophysics Data System (ADS)

Li, Meimei; Zinkle, S. J.

2007-04-01

This paper presents a methodology for computing a fracture mechanism map in two-dimensional space of tensile stress and temperature using physically-based constitutive equations. Four principal fracture mechanisms were considered: cleavage fracture, low temperature ductile fracture, transgranular creep fracture, and intergranular creep fracture. The methodology was applied to calculate fracture mechanism maps for several selected reactor materials, CuCrZr, 316 type stainless steel, F82H ferritic-martensitic steel, V4Cr4Ti and Mo. The calculated fracture maps are in good agreement with empirical maps obtained from experimental observations. The fracture mechanism maps of unirradiated metals and alloys were modified to include radiation hardening effects on cleavage fracture and high temperature helium embrittlement. Future refinement of fracture mechanism maps is discussed.

Perform Tests and Document Results and Analysis of Oxide Layer Effects and Comparisons

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

Collins, E. D.; DelCul, G. D.; Spencer, B. B.

2014-08-30

During the initial feasibility test using actual used nuclear fuel (UNF) cladding in FY 2012, an incubation period of 30–45 minutes was observed in the initial dry chlorination. The cladding hull used in the test had been previously oxidized in a dry air oxidation pretreatment prior to removal of the fuel. The cause of this incubation period was attributed to the resistance to chlorination of an oxide layer imparted by the dry oxidation pretreatment on the cladding. Subsequently in 2013, researchers at the Korea Atomic Energy Institute (KAERI) reported on their chlorination study [R1] on ~9-gram samples of unirradiated ZirloTMmore » cladding tubes that had been previously oxidized in air at 500oC for various time periods to impart oxide layers of varying thickness. In early 2014, discussions with Indefinite Delivery, Indefinite Quantity (IDIQ) contracted technical consultants from Westinghouse described their previous development (and patents) [R2] on methods of chemical washing to remove some or all of the hydrous oxide layer imparted on UNF cladding during irradiation in light water reactors (LWRs) . Thus, the Oak Ridge National Laboratory (ORNL) study, described herein, was planned to extend the KAERI study on the effects of anhydrous oxide layers, but on larger ~100-gram samples of unirradiated zirconium alloy cladding tubes, and to investigate the effects of various methods of chemical pretreatment prior to chlorination with 100% chlorine on the average reaction rates and Cl2 usage efficiencies.« less

Heat treatment effects on toughness of 9Cr-1MoVNb and 12Cr-1MoVW steels irradiated at 365°C

NASA Astrophysics Data System (ADS)

Klueh, R. L.; Alexander, D. J.

1992-09-01

The 9Cr-1MoVNb and 12Cr-1MoVW steels were austenitized at 1040 and 1100°C to produce different prior austenite grain sizes, after which they were given different tempering treatments (1 h at 760°C or 2.5 h at 780°C). Subsize Charpy impact specimens from these materials were irradiated at 365°C up to 5 dpa. For 9Cr-1MoVNb steel in the unirradiated condition, the smaller the prior austenite grain size and the higher the tempering temperature, the lower the ductile-brittle transition temperature (DBTT). Regardless of the DBTT in the unirradiated condition, however, the DBTT shift for 9Cr-1MoVNb steel due to irradiation was the same for all heat treatments. This means heat treatment can be used to ensure a lower DBTT before and after irradiation. The 12Cr-1MoVW steel showed little effect of heat treatment on DBTT in the unirradiated condition, and the shift in DBTT was relatively constant. Thus, it appears that heat treatment cannot be used to reduce the effect of irradiation on DBTT for this steel.

49 CFR 173.426 - Excepted packages for articles containing natural uranium or thorium.

Code of Federal Regulations, 2013 CFR

2013-10-01

... uranium or thorium. 173.426 Section 173.426 Transportation Other Regulations Relating to Transportation....426 Excepted packages for articles containing natural uranium or thorium. A manufactured article in which the sole Class 7 (radioactive) material content is natural uranium, unirradiated depleted uranium...

49 CFR 173.426 - Excepted packages for articles containing natural uranium or thorium.

Code of Federal Regulations, 2014 CFR

2014-10-01

... uranium or thorium. 173.426 Section 173.426 Transportation Other Regulations Relating to Transportation....426 Excepted packages for articles containing natural uranium or thorium. A manufactured article in which the sole Class 7 (radioactive) material content is natural uranium, unirradiated depleted uranium...

49 CFR 173.426 - Excepted packages for articles containing natural uranium or thorium.

Code of Federal Regulations, 2012 CFR

2012-10-01

... uranium or thorium. 173.426 Section 173.426 Transportation Other Regulations Relating to Transportation....426 Excepted packages for articles containing natural uranium or thorium. A manufactured article in which the sole Class 7 (radioactive) material content is natural uranium, unirradiated depleted uranium...

49 CFR 173.426 - Excepted packages for articles containing natural uranium or thorium.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Excepted packages for articles containing natural....426 Excepted packages for articles containing natural uranium or thorium. A manufactured article in which the sole Class 7 (radioactive) material content is natural uranium, unirradiated depleted uranium...

49 CFR 173.426 - Excepted packages for articles containing natural uranium or thorium.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 49 Transportation 2 2011-10-01 2011-10-01 false Excepted packages for articles containing natural....426 Excepted packages for articles containing natural uranium or thorium. A manufactured article in which the sole Class 7 (radioactive) material content is natural uranium, unirradiated depleted uranium...

A prospective, randomized, double-blind study, comparing unirradiated to irradiated white blood cell transfusions in acute leukemia patients

PubMed Central

Freireich, E J; Lichtiger, B; Mattiuzzi, G; Martinez, F; Reddy, V; Kyle Wathen, J

2013-01-01

A prospective, randomized double-blind study comparing the effects of irradiated and unirradiated white blood cells was conducted in 108 acute leukemia patients with life-threatening infections, refractory to antibiotics. The study demonstrated no significant improvement in 30-day survival or overall survival. Transfusion of unirradiated white cells did not compromise the patient's opportunity to undergo allogeneic stem cell transplant, nor the success rate or overall survival after allogeneic transplant. The important positive finding in this study was that the unirradiated white cells produced a significantly higher increment in circulating granulocytes and in a higher proportion of patients granulocyte count exceeded 1000 per microliter, approaching normal concentrations. The increase in the number and the improved survival of the unirradiated granulocytes suggest that this procedure might potentially be a method to improve the utility of granulocyte transfusions and merits further investigation. The study demonstrated non-inferiority for unirradiated white cells. There were no harmful effects such as graft-versus-host disease, indicating that such studies would be safe to conduct in the future. PMID:23072780

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

NASA Astrophysics Data System (ADS)

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.

RIA simulation tests using driver tube for ATF cladding

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

Cinbiz, Mahmut N.; Brown, N. R.; Lowden, R. R.

Pellet-cladding mechanical interaction (PCMI) is a potential failure mechanism for accident-tolerant fuel (ATF) cladding candidates during a reactivity-initiated accident (RIA). This report summarizes Fiscal Year (FY) 2017 research activities that were undertaken to evaluate the PCMI-like hoop-strain-driven mechanical response of ATF cladding candidates. To achieve various RIA-like conditions, a modified-burst test (MBT) device was developed to produce different mechanical pulses. The calibration of the MBT instrument was accomplished by performing mechanical tests on unirradiated Generation-I iron-chromium-aluminum (FeCrAl) alloy samples. Shakedown tests were also conducted in both FY 2016 and FY 2017 using unirradiated hydrided ZIRLO™ tube samples. This milestone reportmore » focuses on testing of ATF materials, but the benchmark tests with hydrided ZIRLO™ tube samples are documented in a recent journal article.a For the calibration and benchmark tests, the hoop strain was monitored using strain gauges attached to the sample surface in the hoop direction. A novel digital image correlation (DIC) system composed of a single high-speed camera and an array of six mirrors was developed for the MBT instrument to better resolve the failure behavior of samples and to provide useful data for validation of high-fidelity modeling and simulation tools. The DIC system enable a 360° view of a sample’s outer surface. This feature was added to the instrument to determine the precise failure location on a sample’s surface for strain predictions. The DIC system was tested on several silicon carbide fiber/silicon carbide matrix (SiC/SiC) composite tube samples at various pressurization rates of the driver tube (which correspond to the strain rates for the samples). The hoop strains for various loading conditions were determined for the SiC/SiC composite tube samples. Future work is planned to enhance understanding of the failure behavior of the ATF cladding candidates of age-hardened FeCrAl alloys and SiC/SiC composites in detail during RIA conditions informed by the computational studies performed under the US Department of Energy Office of Nuclear Energy Advanced Fuels Campaign. The testing instrument and the new DIC system will be further developed to reach different stress-state conditions and to perform tests at elevated temperatures.« less

Irradiation embrittlement characterization of the EUROFER 97 material

NASA Astrophysics Data System (ADS)

Kytka, M.; Brumovsky, M.; Falcnik, M.

2011-02-01

The paper summarizes original results of irradiation embrittlement study of EUROFER 97 material that has been proposed as one candidate of structural materials for future fusion energy systems and GEN IV. Test specimens were manufactured from base metal as well as from weld metal and tested in initial unirradiated condition and also after neutron irradiation. Irradiation embrittlement was characterized by testing of toughness properties at transition temperature region - static fracture toughness and dynamic fracture toughness properties, all in sub-size three-point bend specimens (27 × 4 × 3 mm 3). Testing and evaluation was performed in accordance with ASTM and ESIS standards, fracture toughness KJC and KJd data were also evaluated with the "Master curve" approach. Moreover, J- R dependencies were determined and analyzed. The paper compares unirradiated and irradiated properties as well as changes in transition temperature shifts of these material parameters. Discussion about the correlation between static and dynamic properties is also given. Results from irradiation of EUROFER 97 show that this steel - base metal as well as weld metal - is suitable as a structural material for reactor pressure vessels of innovative nuclear systems - fusion energy systems and GEN IV. Transition temperature shifts after neutron irradiation by 2.5 dpa dose show a good agreement in the case of EUROFER 97 base material for both static and dynamic fracture toughness tests. From the results it can be concluded that there is a low sensitivity of weld metal to neutron irradiation embrittlement in comparison with EUROFER 97 base metal.

Investigation of gamma radiation induced changes in local structure of borosilicate glass by TDPAC and EXAFS

NASA Astrophysics Data System (ADS)

Kumar, Ashwani; Nayak, C.; Rajput, P.; Mishra, R. K.; Bhattacharyya, D.; Kaushik, C. P.; Tomar, B. S.

2016-12-01

Gamma radiation induced changes in local structure around the probe atom (Hafnium) were investigated in sodium barium borosilicate (NBS) glass, used for immobilization of high level liquid waste generated from the reprocessing plant at Trombay, Mumbai. The (NBS) glass was doped with 181Hf as a probe for time differential perturbed angular correlation (TDPAC) spectroscopy studies, while for studies using extended X-ray absorption fine structure (EXAFS) spectroscopy, the same was doped with 0.5 and 2 % (mole %) hafnium oxide. The irradiated as well as un-irradiated glass samples were studied by TDPAC and EXAFS techniques to obtain information about the changes (if any) around the probe atom due to gamma irradiation. TDPAC spectra of unirradiated and irradiated glasses were similar and reminescent of amorphous materials, indicating negligible effect of gamma radiation on the microstructure around Hafnium probe atom, though the quaqdrupole interaction frequency ( ω Q) and asymmetry parameter ( η) did show a marginal decrease in the irradiated glass compared to that in the unirradiated glass. EXAFS measurements showed a slight decrease in the Hf-O bond distance upon gamma irradiation of Hf doped NBS glass indicating densification of the glass matrix, while the cordination number around hafnium remains unchanged.

Elastic Modulus Measurement of ORNL ATF FeCrAl Alloys

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

Thompson, Zachary T.; Terrani, Kurt A.; Yamamoto, Yukinori

2015-10-01

Elastic modulus and Poisson’s ratio for a number of wrought FeCrAl alloys, intended for accident tolerant fuel cladding application, are determined via resonant ultrasonic spectroscopy. The results are reported as a function of temperature from room temperature to 850°C. The wrought alloys were in the fully annealed and unirradiated state. The elastic modulus for the wrought FeCrAl alloys is at least twice that of Zr-based alloys over the temperature range of this study. The Poisson’s ratio of the alloys was 0.28 on average and increased very slightly with increasing temperature.

Electrical and thermoluminescence properties of γ-irradiated La2CuO4 crystals

NASA Astrophysics Data System (ADS)

El-Kolaly, M. A.; Abd El-Kader, H. I.; Kassem, M. E.

1994-12-01

Measurements of the electrical properties of unirradiated as well as ?-irradiated La2CuO4 crystals were carried out at different temperatures in the frequency range of 0.1-100 kHz. Thermoluminescence (TL) studies were also performed on such crystals in the temperature range of 300-600K. The conductivity of the unirradiated La2CuO4 crystals were found to obey the power law frequency dependence at each measured temperature below the transition temperature (Tc = 450K). The activation energies for conduction and dielectric relaxation time have been calculated. The TL response and the dc resistance were found to increase with ?-irradiation dose up to 9-10 kGy. The results showed that the ferroelastic domain walls of La2CuO4 crystal as well as its TL traps are sensitive to ?-raditaion. This material can be used in radiation measurements in the range 225 Gy-10 kGy.

Irradiation effects in beryllium exposed to high energy protons of the NuMI neutrino source

NASA Astrophysics Data System (ADS)

Kuksenko, V.; Ammigan, K.; Hartsell, B.; Densham, C.; Hurh, P.; Roberts, S.

2017-07-01

A beryllium primary vacuum-to-air beam 'window' of the "Neutrinos at the Main Injector" (NuMI) beamline at Fermi National Accelerator Laboratory (Fermilab), Batavia, Illinois, USA, has been irradiated by 120 GeV protons over 7 years, with a maximum integrated fluence at the window centre of 2.06 1022 p/cm2 corresponding to a radiation damage level of 0.48 dpa. The proton beam is pulsed at 0.5 Hz leading to an instantaneous temperature rise of 40 °C per pulse. The window is cooled by natural convection and is estimated to operate at an average of around 50 °C. The microstructure of this irradiated material was investigated by SEM/EBSD and Atom Probe Tomography, and compared to that of unirradiated regions of the beam window and that of stock material of the same PF-60 grade. Microstructural investigations revealed a highly inhomogeneous distribution of impurity elements in both unirradiated and irradiated conditions. Impurities were mainly localised in precipitates, and as segregations at grain boundary and dislocation lines. Low levels of Fe, Cu, Ni, C and O were also found to be homogeneously distributed in the beryllium matrix. In the irradiated materials, up to 440 appm of Li, derived from transmutation of beryllium was homogeneously distributed in solution in the beryllium matrix.

The interaction of molecular hydrogen with α-radiolytic oxidants on a (U,Pu)O2 surface

NASA Astrophysics Data System (ADS)

Bauhn, Lovisa; Hansson, Niklas; Ekberg, Christian; Fors, Patrik; Delville, Rémi; Spahiu, Kastriot

2018-07-01

In order to assess the impact of α-radiolysis of water on the oxidative dissolution of spent fuel, an un-irradiated, annealed MOX fuel pellet with high content of Pu (∼24 wt%), and a specific α-activity of 4.96 GBq/gMOX, was leached in carbonate-containing solutions of low ionic strength. The high Pu content in the pellet stabilizes the (U,Pu)O2(s) matrix towards oxidative dissolution, whereas the α-decays emitted from the surface are expected to produce ∼3.6 × 10-7 mol H2O2/day, contributing to the oxidative dissolution of the pellet. Two sets of leaching tests were conducted under different redox conditions: Ar gas atmosphere and deuterium gas atmosphere. A relatively slow increase of the U and Pu concentrations was observed in the Ar case, with U concentrations increasing from 1·10-6 M after 1 h to ∼7 × 10-5 M after 58 days. Leaching under an atmosphere starting at 1 MPa deuterium gas was undertaken in order to evaluate any effect of dissolved hydrogen on the radiolytic dissolution of the pellet, as well as to investigate any potential recombination of the α-radiolytic products with dissolved deuterium. For the latter purpose, isotopic analysis of the D/H content was carried out on solution samples taken during the leaching. Despite the continuous production of radiolytic oxidants, the concentrations of U and Pu remained quite constant at the level of ∼3 × 10-8 M during the first 30 days, i.e. as long as the deuterium pressure remained higher than 0.8 MPa. These data rule out any oxidative dissolution of the pellet during the first month. The un-irradiated MOX fuel does not contain metallic ε-particles, hence it is mainly the interaction of radiolytic oxidants and dissolved deuterium with the surface of the mixed actinide oxide that causes the neutralization of the oxidants. This conclusion is supported by the steadily increasing levels of HDO measured in the leachate samples.

Determination of He and D permeability of neutron-irradiated SiC tubes to examine the potential for release due to micro-cracking

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

Katoh, Yutai; Hu, Xunxiang; Koyanagi, Takaaki

Driven by the need to enlarge the safety margins of light water reactors in both design-basis and beyond-design-basis accident scenarios, the research and development of accident-tolerant fuel (ATF) has become an importance topic in the nuclear engineering and materials community. Continuous SiC fiber-reinforced SiC matrix ceramic composites are under consideration as a replacement for traditional zirconium alloy cladding owing to their high-temperature stability, chemical inertness, and exceptional irradiation resistance. Among the key technical feasibility issues, potential failure of the fission product containment due to probabilistic penetrating cracking has been identified as one of the two most critical feasibility issues, togethermore » with the radiolysisassisted hydrothermal corrosion of SiC. The experimental capability to evaluate the hermeticity of SiC-based claddings is an urgent need. In this report, we present the development of a comprehensive permeation testing station established in the Low Activation Materials Development and Analysis laboratory at Oak Ridge National Laboratory. Preliminary results for the hermeticity evaluation of un-irradiated monolithic SiC tubes, uncoated and coated SiC/SiC composite tubes, and neutron-irradiated monolithic SiC tubes at room temperature are exhibited. The results indicate that this new permeation testing station is capable of evaluating the hermeticity of SiC-based tubes by determining the helium and deuterium permeation flux as a function of gas pressure at a high resolution of 8.07 x 10 -12 atm-cc/s for helium and 2.83 x 10 -12 atm-cc/s for deuterium, respectively. The detection limit of this system is sufficient to evaluate the maximum allowable helium leakage rate of lab-scale tubular samples, which is linearly extrapolated from the evaluation standard used for a commercial as-manufactured light water reactor fuel rod at room temperature. The un-irradiated monolithic SiC tube is hermetic, as is manifested by the un-detectable deuterium permeation flux at various feeding gas pressures. A large helium leakage rate was detected for the uncoated SiC/SiC composite tube exposed to atmosphere, indicating it is inherently not hermetic. The hermeticity of coated SiC/SiC composite tubes is strongly dependent on the coating materials and the preparation of the substrate SiC/SiC composite samples. To simulate the practical application environment, monolithic CVD SiC tubes were exposed to neutron irradiation at the High Flux Isotope Reactor under high heat flux from the internal surface to the external surface. Although finite element analysis and resonant ultrasound spectroscopy measurement indicated that the combined neutron irradiation and high heat flux gave rise to a high probability of cracking within the sample, the hermeticity evaluation of the tested sample still exhibited gas tightness, emphasizing that SiC cracking is inherently a statistical phenomenon. The developed permeation testing station is capable of measuring the gas permeation flux in the range of interest with full confidence based on the presented results. It is considered a critical pre- /post-irradiation examination technique to characterize SiC-based cladding materials in asreceived and irradiated states to aid the research and development of ATF.« less

Influence of gamma radiation on the colour strength and fastness properties of fabric using turmeric (Curcuma longa L .) as natural dye

NASA Astrophysics Data System (ADS)

Bhatti, Ijaz A.; Adeel, Shahid; Jamal, M. Asghar; Safdar, Muhammad; Abbas, Muhammad

2010-05-01

The effect of gamma radiation on the dyeing of cotton with extract of turmeric ( Curcuma longa L.) powder has been investigated. Cotton fabric and turmeric powder were irradiated to absorbed doses of 1, 2, 3, 4 and 6 kGy using Co-60 gamma irradiator. Dyeing parameters such as temperature, pH and mordant concentration were optimized. Dyeing was performed using un-irradiated and irradiated cotton with the extracts of un-irradiated and irradiated turmeric powder in order to investigate the effect of radiation treatment on the colour strength of dyed fabric. The reported data of un-irradiated and irradiated fabrics dyed with un-irradiated and irradiated dyes were obtained using the spectraflash SF-650. The colourfastness to light, rubbing- and washing-fastness properties showed that gamma irradiation has improved the dyeing characteristics from fair to good.

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

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

Shott, Gregory

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 conditionmore » that the total uranium-233 ( 233U) inventory be limited to 2.7E13 Bq (7.2E2 Ci).« less

Reflector and Shield Material Properties for Project Prometheus

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

J. Nash

2005-11-02

This letter provides updated reflector and shield preliminary material property information to support reactor design efforts. The information provided herein supersedes the applicable portions of Revision 1 to the Space Power Program Preliminary Reactor Design Basis (Reference (a)). This letter partially answers the request in Reference (b) to provide unirradiated and irradiated material properties for beryllium, beryllium oxide, isotopically enriched boron carbide ({sup 11}B{sub 4}C) and lithium hydride. With the exception of {sup 11}B{sub 4}C, the information is provided in Attachments 1 and 2. At the time of issuance of this document, {sup 11}B{sub 4}C had not been studied.

Azimuthally anisotropic hydride lens structures in Zircaloy 4 nuclear fuel cladding: High-resolution neutron radiography imaging and BISON finite element analysis

NASA Astrophysics Data System (ADS)

Lin, Jun-Li; Zhong, Weicheng; Bilheux, Hassina Z.; Heuser, Brent J.

2017-12-01

High-resolution neutron radiography has been used to image bulk circumferential hydride lens particles in unirradiated Zircaloy 4 tubing cross section specimens. Zircaloy 4 is a common light water nuclear reactor (LWR) fuel cladding; hydrogen pickup, hydride formation, and the concomitant effect on the mechanical response are important for LWR applications. Ring cross section specimens with three hydrogen concentrations (460, 950, and 2830 parts per million by weight) and an as-received reference specimen were imaged. Azimuthally anisotropic hydride lens particles were observed at 950 and 2830 wppm. The BISON finite element analysis nuclear fuel performance code was used to model the system elastic response induced by hydride volumetric dilatation. The compressive hoop stress within the lens structure becomes azimuthally anisotropic at high hydrogen concentrations or high hydride phase fraction. This compressive stress anisotropy matches the observed lens anisotropy, implicating the effect of stress on hydride formation as the cause of the observed lens azimuthal asymmetry. The cause and effect relation between compressive stress and hydride lens anisotropy represents an indirect validation of a key BISON output, the evolved hoop stress associated with hydride formation.

Properties of HIPed stainless steel powder

NASA Astrophysics Data System (ADS)

Dellis, Ch.; Le Marois, G.; Gentzbittel, J. M.; Robert, G.; Moret, F.

1996-10-01

In the current design of ITER primary wall, 316LN stainless steel is the reference structural material. Austenitic stainless steel is used for water-cooling channels and structures. As material data on hot isostatic pressed (HIP) 316LN were not available in open literature and from powder producers, the main properties of unirradiated samples have been measured in CEA/CEREM. Fully dense material without any porosity is obtained when appropriate HIP parameters are applied. Microstructural examination and mechanical properties are confirmed that the HIPed 316LN material is equivalent to a very good fine-grain, isotropic and uniformly forged 316LN. Moreover, ultrasonic inspection showed that this fine and uniform microstructure produced a remarkably low noise, which allow the use of transverse waves at very high frequencies (4 MHz). Defects undetectable in forged material will be easily detected in HIPed material.

Processing Plan for Potentially Reactive/Ignitable Remote Handled Transuranic Waste at the Idaho Cleanup Project - 12090

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

Troescher, Patrick D.; Hobbes, Tammy L.; Anderson, Scott A.

Remote Handle Transuranic (RH-TRU) Waste generated at Argonne National Laboratory - East, from the examination of irradiated and un-irradiated fuel pins and other reactor materials requires a detailed processing plan to ensure reactive/ignitable material is absent to meet WIPP Waste Acceptance Criteria prior to shipping and disposal. The Idaho Cleanup Project (ICP) approach to repackaging Lot 2 waste and how we ensure prohibited materials are not present in waste intended for disposal at Waste Isolation Pilot Plant 'WIPP' uses an Argon Repackaging Station (ARS), which provides an inert gas blanket. Opening of the Lot 2 containers under an argon gasmore » blanket is proposed to be completed in the ARS. The ARS is an interim transition repackaging station that provides a mitigation technique to reduce the chances of a reoccurrence of a thermal event prior to rendering the waste 'Safe'. The consequences, should another thermal event be encountered, (which is likely) is to package the waste, apply the reactive and or ignitable codes to the container, and store until the future treatment permit and process are available. This is the same disposition that the two earlier containers in the 'Thermal Events' were assigned. By performing the initial handling under an inert gas blanket, the waste can sorted and segregate the fines and add the Met-L-X to minimize risk before it is exposed to air. The 1-gal cans that are inside the ANL-E canister will be removed and each can is moved to the ARS for repackaging. In the ARS, the 1-gal can is opened in the inerted environment. The contained waste is sorted, weighed, and visually examined for non compliant items such as unvented aerosol cans and liquids. The contents of the paint cans are transferred into a sieve and manipulated to allow the fines, if any, to be separated into the tray below. The fines are weighed and then blended with a minimum 5:1 mix of Met-L-X. Other debris materials found are segregated from the cans into containers for later packaging. Recoverable fissile waste material (Fuel and fuel-like pieces) suspected of containing sodium bonded pieces) are segregated and will remain in the sieve or transferred to a similar immersion basket in the ARS. The fuel like pieces will be placed into a container with sufficient water to cover the recoverable fissile waste. If a 'reactive characteristic' is present the operator will be able to observe the formation of 'violent' hydrogen gas bubbles. When sodium bonded fuel-like pieces are placed in water the expected reaction is a non-violent reaction that does not meet the definition of reactivity. It is expected that there will be a visible small stream of bubbles present if there is any sodium-bonded fuel-like piece placed in the water. The test will be completed when there is no reaction or the expected reaction is observed..At that point, the fuel like pieces complete the processing cycle in preparation for characterization and shipment to WIPP. If a violent reaction occurs, the fuel-like pieces will be removed from the water, split into the required fissile material content, placed into a screened basket in a 1 gallon drum and drummed out of the hot cell with appropriate RCRA codes applied and placed into storage until sodium treatment is available. These 'violent' reactions will be evidenced by gas bubbles being evolved at the specimen surface where sodium metal is present. The operators will be trained to determine if the reaction is 'violent' or 'mild'. If a 'violent' reaction occurs, the sieve will be immediately removed from the water, placed in a 1 gallon paint can, canned in the argon cover gas and removed from the hot cell to await a future treatment. If the reaction is 'mild', the sieve will then be removed from the water; the material weighed for final packaging and allowed to dry by air exposure. Lot 2 waste cans can be opened, sorted, processed, and weighed while mitigating the potential of thermal events that could occur prior to exposing to air. Exposure to air is a WIPP compliance step demonstrating the absence of reactive or ignitable characteristics. (authors)« less

Investigations on neutron irradiated 3D carbon fibre reinforced carbon composite material

NASA Astrophysics Data System (ADS)

Venugopalan, Ramani; Alur, V. D.; Patra, A. K.; Acharya, R.; Srivastava, D.

2018-04-01

As against conventional graphite materials carbon-carbon (C/C) composite materials are now being contemplated as the promising candidate materials for the high temperature and fusion reactor owing to their high thermal conductivity and high thermal resistance, better mechanical/thermal properties and irradiation stability. The current need is for focused research on novel carbon materials for future new generation nuclear reactors. The advantage of carbon-carbon composite is that the microstructure and the properties can be tailor made. The present study encompasses the irradiation of 3D carbon composite prepared by reinforcement using PAN carbon fibers for nuclear application. The carbon fiber reinforced composite was subjected to neutron irradiation in the research reactor DHRUVA. The irradiated samples were characterized by Differential Scanning Calorimetry (DSC), small angle neutron scattering (SANS), XRD and Raman spectroscopy. The DSC scans were taken in argon atmosphere under a linear heating program. The scanning was carried out at temperature range from 30 °C to 700 °C at different heating rates in argon atmosphere along with reference as unirradiated carbon composite. The Wigner energy spectrum of irradiated composite showed two peaks corresponding to 200 °C and 600 °C. The stored energy data for the samples were in the range 110-170 J/g for temperature ranging from 30 °C to 700 °C. The Wigner energy spectrum of irradiated carbon composite did not indicate spontaneous temperature rise during thermal annealing. Small angle neutron scattering (SANS) experiments have been carried out to investigate neutron irradiation induced changes in porosity of the composite samples. SANS data were recorded in the scattering wave vector range of 0.17 nm-1 to 3.5 nm-1. Comparison of SANS profiles of irradiated and unirradiated samples indicates significant change in pore morphology. Pore size distributions of the samples follow power law size distribution with different exponent. Narrowing of SANS profile of the irradiated sample indicates creation of significant number of larger pores due to neutron irradiation.

Isoconversional approach for non-isothermal decomposition of un-irradiated and photon-irradiated 5-fluorouracil.

PubMed

Mohamed, Hala Sh; Dahy, AbdelRahman A; Mahfouz, Refaat M

2017-10-25

Kinetic analysis for the non-isothermal decomposition of un-irradiated and photon-beam-irradiated 5-fluorouracil (5-FU) as anti-cancer drug, was carried out in static air. Thermal decomposition of 5-FU proceeds in two steps. One minor step in the temperature range of (270-283°C) followed by the major step in the temperature range of (285-360°C). The non-isothermal data for un-irradiated and photon-irradiated 5-FU were analyzed using linear (Tang) and non-linear (Vyazovkin) isoconversional methods. The results of the application of these free models on the present kinetic data showed quite a dependence of the activation energy on the extent of conversion. For un-irradiated 5-FU, the non-isothermal data analysis indicates that the decomposition is generally described by A3 and A4 modeles for the minor and major decomposition steps, respectively. For a photon-irradiated sample of 5-FU with total absorbed dose of 10Gy, the decomposition is controlled by A2 model throughout the coversion range. The activation energies calculated in case of photon-irradiated 5-FU were found to be lower compared to the values obtained from the thermal decomposition of the un-irradiated sample probably due to the formation of additional nucleation sites created by a photon-irradiation. The decomposition path was investigated by intrinsic reaction coordinate (IRC) at the B3LYP/6-311++G(d,p) level of DFT. Two transition states were involved in the process by homolytic rupture of NH bond and ring secession, respectively. Published by Elsevier B.V.

Topical Hazard Evaluation Program of Candidate Insect Repellent AI3-30180-c, US Department of Agriculture Proprietary Chemical, April 1982 - September 1984.

DTIC Science & Technology

1985-01-10

irritation photochemical chemical and 10 percent reaction under test con- irritation in humans. (wlv) Oil of Bergamot ditions. 2 * - Study No. 75-51-0367-85...control (oil of Bergamot ), than unirradiated skin areas. a and diluent were applied to additional skin areas to serve as unirradiated control sites

Detection of Irradiation Treatment of Foods Using DNA `Comet Assay'

NASA Astrophysics Data System (ADS)

Khan, Hasan M.; Delincée, Henry

1998-06-01

Microgel electrophoresis of single cells (DNA comet assay) has been investigated to detect irradiation treatment of some food samples. These samples of fresh and frozen rainbow trout, red lentil, gram and sliced almonds were irradiated to 1 or 2 kGy using 10 MeV electron beam from a linear accelerator. Rainbow trout samples yielded good results with samples irradiated to 1 or 2 kGy showing fragmentation of DNA and, therefore, longer comets with no intact cells. Unirradiated samples showed shorter comets with a significant number of intact cells. For rainbow trout stored in a freezer for 11 days the irradiated samples can still be discerned by electrophoresis from unirradiated samples, however, the unirradiated trouts also showed some longer comets besides some intact cells. Radiation treatment of red lentils can also be detected by this method, i.e. no intact cells in 1 or 2 kGy irradiated samples and shorter comets and some intact cells in unirradiated samples. However, the results for gram and sliced almond samples were not satisfactory since some intact DNA cells were observed in irradiated samples as well. Probably, incomplete lysis has led to these deviating results.

Advanced 3D Characterization and Reconstruction of Reactor Materials FY16 Final Report

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

Fromm, Bradley; Hauch, Benjamin; Sridharan, Kumar

2016-12-01

A coordinated effort to link advanced materials characterization methods and computational modeling approaches is critical to future success for understanding and predicting the behavior of reactor materials that operate at extreme conditions. The difficulty and expense of working with nuclear materials have inhibited the use of modern characterization techniques on this class of materials. Likewise, mesoscale simulation efforts have been impeded due to insufficient experimental data necessary for initialization and validation of the computer models. The objective of this research is to develop methods to integrate advanced materials characterization techniques developed for reactor materials with state-of-the-art mesoscale modeling and simulationmore » tools. Research to develop broad-ion beam sample preparation, high-resolution electron backscatter diffraction, and digital microstructure reconstruction techniques; and methods for integration of these techniques into mesoscale modeling tools are detailed. Results for both irradiated and un-irradiated reactor materials are presented for FY14 - FY16 and final remarks are provided.« less

Computational Modeling of Radiation Phenomenon in SiC for Nuclear Applications

NASA Astrophysics Data System (ADS)

Ko, Hyunseok

Silicon carbide (SiC) material has been investigated for promising nuclear materials owing to its superior thermo-mechanical properties, and low neutron cross-section. While the interest in SiC has been increasing, the lack of fundamental understanding in many radiation phenomena is an important issue. More specifically, these phenomena in SiC include the fission gas transport, radiation induced defects and its evolution, radiation effects on the mechanical stability, matrix brittleness of SiC composites, and low thermal conductivities of SiC composites. To better design SiC and SiC composite materials for various nuclear applications, understanding each phenomenon and its significance under specific reactor conditions is important. In this thesis, we used various modeling approaches to understand the fundamental radiation phenomena in SiC for nuclear applications in three aspects: (a) fission product diffusion through SiC, (b) optimization of thermodynamic stable self-interstitial atom clusters, (c) interface effect in SiC composite and their change upon radiation. In (a) fission product transport work, we proposed that Ag/Cs diffusion in high energy grain boundaries may be the upper boundary in unirradiated SiC at relevant temperature, and radiation enhanced diffusion is responsible for fast diffusion measured in post-irradiated fuel particles. For (b) the self-interstitial cluster work, thermodynamically stable clusters are identified as a function of cluster size, shape, and compositions using a genetic algorithm. We found that there are compositional and configurational transitions for stable clusters as the cluster size increases. For (c) the interface effect in SiC composite, we investigated recently proposed interface, which is CNT reinforced SiC composite. The analytical model suggests that CNT/SiC composites have attractive mechanical and thermal properties, and these fortify the argument that SiC composites are good candidate materials for the cladding. We used grand canonical monte carlo to optimize the interface, as a part of the stepping stone for further study using the interface.

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

PubMed

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.

Evaluation of the limulus amoebocyte lysate test in conjunction with a gram negative bacterial plate count for detecting irradiation of chicken

NASA Astrophysics Data System (ADS)

Scotter, Susan L.; Wood, Roger; McWeeny, David J.

A study to evaluate the potential of the Limulus amoebocyte lysate (LAL) test in conjuction with a Gram negative bacteria (GNB) plate count for detecting the irradiation of chicken is described. Preliminary studies demonstrated that chickens irradiated at an absorbed dose of 2.5 kGy could be differentiated from unirradiated birds by measuring levels of endotoxin and of numbers of GNB on chicken skin. Irradiated birds were found to have endotoxin levels similar to those found in unirradiated birds but significantly lower numbers of GNB. In a limited study the test was found to be applicable to birds from different processors. The effect of temperature abuse on the microbiological profile, and thus the efficacy of the test, was also investigated. After temperature abuse, the irradiated birds were identifiable at worst up to 3 days after irradiation treatment at the 2.5 kGy level and at best some 13 days after irradiation. Temperature abuse at 15°C resulted in rapid recovery of surviving micro-organisms which made differentiation of irradiated and unirradiated birds using this test unreliable. The microbiological quality of the bird prior to irradiation treatment also affected the test as large numbers of GNB present on the bird prior to irradiation treatment resulted in larger numbers of survivors. In addition, monitoring the developing flora after irradiation treatment and during subsequent chilled storage also aided differentiation of irradiated and unirradiated birds. Large numbers of yeasts and Gram positive cocci were isolated from irradiated carcasses whereas Gram negative oxidative rods were the predominant spoilage flora on unirradiated birds.

Cell cycle tracking for irradiated and unirradiated bystander cells in a single colony with exposure to a soft X-ray microbeam.

PubMed

Kaminaga, Kiichi; Noguchi, Miho; Narita, Ayumi; Hattori, Yuya; Usami, Noriko; Yokoya, Akinari

2016-11-01

To establish a new experimental technique to explore the photoelectric and subsequent Auger effects on the cell cycles of soft X-ray microbeam-irradiated cells and unirradiated bystander cells in a single colony. Several cells located in the center of a microcolony of HeLa-Fucci cells consisting of 20-80 cells were irradiated with soft X-ray (5.35 keV) microbeam using synchrotron radiation as a light source. All cells in the colony were tracked for 72 h by time-lapse microscopy imaging. Cell cycle progression, division, and death of each cell in the movies obtained were analyzed by pedigree assay. The number of cell divisions in the microcolony was also determined. The fates of these cells were clarified by tracking both irradiated and unirradiated bystander cells. Irradiated cells showed significant cell cycle retardation, explosive cell death, or cell fusion after a few divisions. These serious effects were also observed in 15 and 26% of the bystander cells for 10 and 20 Gy irradiation, respectively, and frequently appeared in at least two daughter or granddaughter cells from a single-parent cell. We successfully tracked the fates of microbeam-irradiated cells and unirradiated bystander cells with live cell recordings, which have revealed the dynamics of soft X-ray irradiated and unirradiated bystander cells for the first time. Notably, cell deaths or cell cycle arrests frequently arose in closely related cells. These details would not have been revealed by a conventional immunostaining imaging method. Our approach promises to reveal the dynamic cellular effects of soft X-ray microbeam irradiation and subsequent Auger processes from various endpoints in future studies.

Stress corrosion crack initiation of Zircaloy-4 cladding tubes in an iodine vapor environment during creep, relaxation, and constant strain rate tests

NASA Astrophysics Data System (ADS)

Jezequel, T.; Auzoux, Q.; Le Boulch, D.; Bono, M.; Andrieu, E.; Blanc, C.; Chabretou, V.; Mozzani, N.; Rautenberg, M.

2018-02-01

During accidental power transient conditions with Pellet Cladding Interaction (PCI), the synergistic effect of the stress and strain imposed on the cladding by thermal expansion of the fuel, and corrosion by iodine released as a fission product, may lead to cladding failure by Stress Corrosion Cracking (SCC). In this study, internal pressure tests were conducted on unirradiated cold-worked stress-relieved Zircaloy-4 cladding tubes in an iodine vapor environment. The goal was to investigate the influence of loading type (constant pressure tests, constant circumferential strain rate tests, or constant circumferential strain tests) and test temperature (320, 350, or 380 °C) on iodine-induced stress corrosion cracking (I-SCC). The experimental results obtained with different loading types were consistent with each other. The apparent threshold hoop stress for I-SCC was found to be independent of the test temperature. SEM micrographs of the tested samples showed many pits distributed over the inner surface, which tended to coalesce into large pits in which a microcrack could initiate. A model for the time-to-failure of a cladding tube was developed using finite element simulations of the viscoplastic mechanical behavior of the material and a modified Kachanov's damage growth model. The times-to-failure predicted by this model are consistent with the experimental data.

Modernization at the Y-12 National Security Complex: A Case for Additional Experimental Benchmarks

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

Thornbury, M. L.; Juarez, C.; Krass, A. W.

Efforts are underway at the Y-12 National Security Complex (Y-12) to modernize the recovery, purification, and consolidation of un-irradiated, highly enriched uranium metal. Successful integration of advanced technology such as Electrorefining (ER) eliminates many of the intermediate chemistry systems and processes that are the current and historical basis of the nuclear fuel cycle at Y-12. The cost of operations, the inventory of hazardous chemicals, and the volume of waste are significantly reduced by ER. It also introduces unique material forms and compositions related to the chemistry of chloride salts for further consideration in safety analysis and engineering. The work hereinmore » briefly describes recent investigations of nuclear criticality for 235UO2Cl2 (uranyl chloride) and 6LiCl (lithium chloride) in aqueous solution. Of particular interest is the minimum critical mass of highly enriched uranium as a function of the molar ratio of 6Li to 235U. The work herein also briefly describes recent investigations of nuclear criticality for 235U metal reflected by salt mixtures of 6LiCl or 7LiCl (lithium chloride), KCl (potassium chloride), and 235UCl3 or 238UCl3 (uranium tri-chloride). Computational methods for analysis of nuclear criticality safety and published nuclear data are employed in the absence of directly relevant experimental criticality benchmarks.« less

Microstructural Characterization of Irradiated U0.7ZrH1.6 Using Ultrasonic Techniques

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

Ramuhalli, Pradeep; Jacob, Richard E.; MacFarlan, Paul J.

In recent years, there has been an increased level of effort to understand the changes in microstructure that occur due to irradiation of nuclear fuel. The primary driver for this increased effort is the potential for designing new fuels that are safer and more reliable, in turn enabling new and improved reactor technologies. Much of the data on microstructural change in irradiated fuels is generated through a host of post irradiation examination techniques such as optical microscopy (OM), scanning electron microscopy (SEM), and transmission electron microscopy (TEM) to determine grain structure, porosity, crack geometry, etc. in irradiated fuels. Such “traditional”more » examination techniques were recently used to characterize a novel new fuel consisting of U0.17ZrH1.6 pellets bonded to zircaloy-2 cladded with lead-bismuth eutectic before and after irradiation. However, alternative methods such as ultrasonic inspection can provide an opportunity for nondestructively assessing microstructure in both in-pile and post-irradiation examinations. In this paper, we briefly describe initial results of ultrasonic examination of the U0.17ZrH1.6 pellets (unirradiated and irradiated), in a post-irradiation examination study. Data indicate some correlation with microstructural changes due to irradiation; however, it is not clear what the specific microstructural changes are that are influencing the ultrasonic measurements. Interestingly, specimens with nominally identical burnup show differences in ultrasonic signatures, indicating apparent microstructural differences between these specimens. A summary of the experimental study, preliminary data and findings are presented in this short paper. Additional details of the analysis will be included in the presentation.« less

$gamma$-RAY PROCESSING OF RICE. (IV.) STUDIES ON THE AVAILABILITY OF $gamma$-IRRADIATED RICE FOR KOJI

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

Nishimura, A.; Takaoka, K.

The availability of rice, gamma -irradiated up to 4.6 x 10/sup 4/ and 3.5 x 10/sup 5/r for koji was studied. The enzyme activities of koji of the steamed samples were stronger than the unirradiated rice in amylase and protease. The sensory test on the once-steamed irradiated rice was almost the same as the twice-steamed unirradiated rice. (OID)

Use of UO 2 films for electrochemical studies

NASA Astrophysics Data System (ADS)

Miserque, F.; Gouder, T.; Wegen, D. H.; Bottomley, P. D. W.

2001-10-01

UO 2 films have been prepared by dc reactive sputtering of a uranium metal target in an Ar/O 2 atmosphere. We have used the films deposited on gold substrates as working electrodes for electrochemical investigations as simulating the surfaces of fuel pellets. Film composition was determined by photoelectron spectroscopy (XPS and UPS) and X-ray diffraction (XRD). The oxide stoichiometry as a function of deposition conditions was determined and the appropriate conditions for UO 2.0 formation established. AC impedance and cyclic voltammetry measurements were performed. A double RC electrical equivalent circuit was used to fit the data from impedance measurements, similar to those used in unirradiated UO 2 or spent fuel pellets. However due to the porosity or adhesion defects on the thin films that permitted a direct contact between the solution and the gold substrate, we were obliged to add a contribution simulating the water-gold system. Cyclic voltammetry measurements show the influence of pH on the dissolution mechanism. Alkaline solutions permit the formation of an oxidised layer (UO 2.33) which is not present in the acidic solutions. In both pH=2 and pH=6 solutions, a U VI species layer is formed.

The non-targeted effects of radiation are perpetuated by exosomes.

PubMed

Al-Mayah, Ammar; Bright, Scott; Chapman, Kim; Irons, Sarah; Luo, Ping; Carter, David; Goodwin, Edwin; Kadhim, Munira

2015-02-01

Exosomes contain cargo material from endosomes, cytosol, plasma membrane and microRNA molecules, they are released by a number of non-cancer and cancer cells into both the extracellular microenvironment and body fluids such as blood plasma. Recently we demonstrated radiation-induced non-targeted effects [NTE: genomic instability (GI) and bystander effects (BE)] are partially mediated by exosomes, particularly the RNA content. However the mechanistic role of exosomes in NTE is yet to be fully understood. The present study used MCF7 cells to characterise the longevity of exosome-induced activity in the progeny of irradiated and unirradiated bystander cells. Exosomes extracted from conditioned media of irradiated and bystander progeny were added to unirradiated cells. Analysis was carried out at 1 and 20/24 population doublings following medium/exosome transfer for DNA/chromosomal damage. Results confirmed exosomes play a significant role in mediating NTE of ionising radiation (IR). This effect was remarkably persistent, observed >20 doublings post-irradiation in the progeny of bystander cells. Additionally, cell progeny undergoing a BE were themselves capable of inducing BE in other cells via exosomes they released. Furthermore we investigated the role of exosome cargo. Culture media from cells exposed to 2 Gy X-rays was subjected to ultracentrifugation and four inoculants prepared, (a) supernatants with exosomes removed, and pellets with (b) exosome proteins denatured, (c) RNA degraded, and (d) a combination of protein-RNA inactivation. These were added to separate populations of unirradiated cells. The BE was partially inhibited when either exosome protein or exosome RNA were inactivated separately, whilst combined RNA-protein inhibition significantly reduced or eliminated the BE. These results demonstrate that exosomes are associated with long-lived signalling of the NTE of IR. Both RNA and protein molecules of exosomes work in a synergistic manner to initiate NTE, spread these effects to naïve cells, and perpetuate GI in the affected cells. Copyright © 2015. Published by Elsevier B.V.

Deep cytoplasmic rearrangements in ventralized Xenopus embryos

NASA Technical Reports Server (NTRS)

Brown, E. E.; Denegre, J. M.; Danilchik, M. V.

1993-01-01

Following fertilization in Xenopus, dramatic rearrangements of the egg cytoplasm relocalize maternally synthesized egg components. During the first cell cycle the vegetal yolk mass rotates relative to the egg surface, toward the sperm entry point (SEP) (J. P. Vincent, G. F. Oster, and J. C. Gerhart, 1986, Dev. Biol. 113, 484-500), while concomitant deep cytoplasmic rearrangements occur in the animal hemisphere (M. V. Danilchik and J. M. Denegre, 1991, Development 111, 845-856). In this paper we examine the role of vegetal yolk mass rotation in producing the animal cytoplasmic rearrangements. We inhibited rotation by uv-irradiating embryos during the first cell cycle, a treatment that yields an extremely ventralized phenotype. Both uv-irradiated embryos and unirradiated control embryos show cytoplasmic rearrangements in the animal hemisphere during the first cell cycle. Cytoplasmic rearrangements on the SEP side of the embryo associated with the path of the sperm pronucleus, plus a swirl on the anti-SEP (dorsal) side, are seen, whether or not yolk mass rotation has occurred. This result suggests a role for the expanding sperm aster in directing animal hemisphere cytoplasmic movements. In unirradiated control embryos the anti-SEP (dorsal) swirl is larger than that in uv-irradiated embryos and often extends into the vegetal hemisphere, consistent with the animal cytoplasm having been pulled dorsally and vegetally by the sliding vegetal yolk mass. Thus the yolk mass rotation may normally enhance the dorsalward cytoplasmic movement, begun by the sperm aster, enough to induce normal axis formation. We extended our observations of unirradiated control and uv-irradiated embryos through early cleavages. The vegetal extent of the anti-SEP (dorsal) swirl pattern seen in control embryos persists through the early cleavage period, such that labeled animal cytoplasm extends deep into dorsal third-tier blastomeres at the 32-cell stage. Significantly, in uv-irradiated embryos, which have not undergone vegetal rotation, most of this labeled material remains more equatorial.

Investigation of high-energy ion-irradiated MA957 using synchrotron radiation under in-situ tension

DOE PAGES

Mo, Kun; Yun, Di; Miao, Yinbin; ...

2016-01-02

In this paper, an MA957 oxide dispersion-strengthened (ODS) alloy was irradiated with high-energy ions in the Argonne Tandem Linac Accelerator System. Fe ions at an energy of 84 MeV bombarded MA957 tensile specimens, creating a damage region similar to 7.5 μm in depth; the peak damage (similar to 40 dpa) was estimated to be at similar to 7 μm from the surface. Following the irradiation, in-situ high-energy X-ray diffraction measurements were performed at the Advanced Photon Source in order to study the dynamic deformation behavior of the specimens after ion irradiation damage. In-situ X-ray measurements taken during tensile testing ofmore » the ion-irradiated MA957 revealed a difference in loading behavior between the irradiated and un-irradiated regions of the specimen. At equivalent applied stresses, lower lattice strains were found in the radiation-damaged region than those in the un-irradiated region. This might be associated with a higher level of Type II stresses as a result of radiation hardening. The study has demonstrated the feasibility of combining high-energy ion radiation and high-energy synchrotron X-ray diffraction to study materials' radiation damage in a dynamic manner.« less

Bystander effects in unicellular organisms.

PubMed

DeVeaux, Linda C; Durtschi, Lynn S; Case, Jonathan G; Wells, Douglas P

2006-05-11

Radiation-induced bystander effects have been seen in mammalian cells from diverse origins. These effects can be transmitted through the medium to cells not present at the time of irradiation. We have developed an assay for detecting bystander effects in the unicellular eukaryote, the fission yeast Schizosaccharomyces pombe. This assay allows maximal exposure of unirradiated cells to cells that have received electron beam irradiation. S. pombe cells were irradiated with 16-18 MeV electrons from a pulsed electron LINAC. When survival of the irradiated cells decreased to approximately 50%, forward-mutation to 2-deoxy-d-glucose resistance increased in the unirradiated bystander cells. Further increase in dose had no additional effect on this increase. In order to detect this response, it was necessary for the irradiated cell/unirradiated cell ratio to be high. Other cellular stresses, such as heat treatment, UV irradiation, and bleomycin exposure, also caused a detectable response in untreated cells grown with the treated cells. We discuss evolutionary implications of these results.

Progress in the Assessment of Waste-forms for the Immobilisation of UK Civil Plutonium

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

Harrison, M.T.; Scales, C.R.; Maddrell, E.R.

The alternatives for the disposition of the UK's civil plutonium stocks are currently being investigated by Nexia Solutions Ltd. on behalf of the Nuclear Decommissioning Authority (NDA). A number of scenarios are currently being considered depending on the strategic requirements of the UK. The two main disposition options are: re-use as MOX (Mixed Oxide) fuel in reactors, or immobilisation in the event of any material being declared surplus to requirements. The amount of Pu which will require immobilisation will depend on future UK nuclear strategy, along with the extent of any stocks deemed unsuitable for re-use. However, it is likelymore » that some portion will have to be immobilised and therefore three credible waste-forms are under consideration; ceramic, glass and 'immobilisation' MOX. These are currently being developed and assessed in a systematic programme that involves periodic evaluation against a range of criteria. In this way, by down-selecting on the basis of robust and technical review, the most appropriate option for immobilising surplus civil plutonium in the UK can be recommended. The latest results from the immobilisation experimental programme are presented following the de-selection of the least favourable glass and ceramic candidates. The main criteria for this decision were waste loading, durability, processability, criticality and proliferation resistance. In addition, the durability of unirradiated MOX fuel is being examined to determine its potential as a wasteform for Pu, and recent leach test data is discussed. The current evaluation comprises not only a comparison of the relevant physical properties of the various waste-forms, but also key processing parameters, e.g. glass viscosity and melter technology, ceramic fabrication routes, and criticality issues. Other important aspects of the long-term behaviour of the waste-forms under consideration in a potential repository environment, such as radiation damage, criticality control and the properties of any neutron poisons present, are also included. (authors)« less

RH-TRU Waste Characterization by Acceptable Knowledge at the Idaho National Engineering and Environmental Laboratory

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

Schulz, C.; Givens, C.; Bhatt, R.

2003-02-24

Idaho National Engineering and Environmental Laboratory (INEEL) is conducting an effort to characterize approximately 620 drums of remote-handled (RH-) transuranic (TRU) waste currently in its inventory that were generated at the Argonne National Laboratory-East (ANL-E) Alpha Gamma Hot Cell Facility (AGHCF) between 1971 and 1995. The waste was generated at the AGHCF during the destructive examination of irradiated and unirradiated fuel pins, targets, and other materials from reactor programs at ANL-West (ANL-W) and other Department of Energy (DOE) reactors. In support of this effort, Shaw Environmental and Infrastructure (formerly IT Corporation) developed an acceptable knowledge (AK) collection and management programmore » based on existing contact-handled (CH)-TRU waste program requirements and proposed RH-TRU waste program requirements in effect in July 2001. Consistent with Attachments B-B6 of the Waste Isolation Pilot Plant (WIPP) Hazardous Waste Facility Permit (HWFP) and th e proposed Class 3 permit modification (Attachment R [RH-WAP] of this permit), the draft AK Summary Report prepared under the AK procedure describes the waste generating process and includes determinations in the following areas based on AK: physical form (currently identified at the Waste Matrix Code level); waste stream delineation; applicability of hazardous waste numbers for hazardous waste constituents; and prohibited items. In addition, the procedure requires and the draft summary report contains information supporting determinations in the areas of defense relationship and radiological characterization.« less

Radiation induced corrosion of copper for spent nuclear fuel storage

NASA Astrophysics Data System (ADS)

Björkbacka, Åsa; Hosseinpour, Saman; Johnson, Magnus; Leygraf, Christofer; Jonsson, Mats

2013-11-01

The long term safety of repositories for radioactive waste is one of the main concerns for countries utilizing nuclear power. The integrity of engineered and natural barriers in such repositories must be carefully evaluated in order to minimize the release of radionuclides to the biosphere. One of the most developed concepts of long term storage of spent nuclear fuel is the Swedish KBS-3 method. According to this method, the spent fuel will be sealed inside copper canisters surrounded by bentonite clay and placed 500 m down in stable bedrock. Despite the importance of the process of radiation induced corrosion of copper, relatively few studies have been reported. In this work the effect of the total gamma dose on radiation induced corrosion of copper in anoxic pure water has been studied experimentally. Copper samples submerged in water were exposed to a series of total doses using three different dose rates. Unirradiated samples were used as reference samples throughout. The copper surfaces were examined qualitatively using IRAS and XPS and quantitatively using cathodic reduction. The concentration of copper in solution after irradiation was measured using ICP-AES. The influence of aqueous radiation chemistry on the corrosion process was evaluated based on numerical simulations. The experiments show that the dissolution as well as the oxide layer thickness increase upon radiation. Interestingly, the evaluation using numerical simulations indicates that aqueous radiation chemistry is not the only process driving the corrosion of copper in these systems.

Investigation of radiosterilization feasibility of sulfamethoxazole by ESR spectroscopy

NASA Astrophysics Data System (ADS)

Çolak, Şeyda

2017-12-01

In the present study, the spectroscopic features of the radiolytic intermediates that were produced in gamma-irradiated (5, 10, 25 and 50 kGy) sulfamethoxazole (SMX) have been investigated by electron spin resonance (ESR) spectroscopy and the radiation sterilization feasibility of SMX by ionizing radiation was examined. Gamma-irradiated SMX exhibited a complex ESR spectrum consisting of 13 resonance lines where spectral parameters for the central resonance line were found to be g = 2.0062 and ΔHpp = 0.6 mT. The radiation yield of SMX was calculated to be relatively low (G = 0.1) by ESR spectroscopy and no meaningful difference was observed in the comparison of unirradiated and 50 kGy gamma irradiated SMX by the Fourier transform infrared (FT-IR) technique, confirming that SMX is a radioresistive material. Although SMX could not be accepted to be a good dosimetric material, the identification of irradiated SMX from the unirradiated sample was possible even for the low absorbed radiation doses and for a relatively long time (three months) after the irradiation process. Decay activation energy of the radical species, which is mostly responsible for the central intense resonance line, is calculated to be 45.15 kJ/mol by using the signal intensity decay data derived from annealing studies. Four radical species with different spectroscopic properties were accepted to be responsible for the ESR spectra of gamma-irradiated SMX, by simulation calculations. It is concluded that SMX and SMX-containing drugs can be sterilized by gamma radiation and ESR spectroscopy is an appropriate technique for the characterization of these induced radical intermediates during the gamma irradiation process of SMX. Toxicology tests should also be done for its safe usage.

Physically-based biodosimetry using in vivo EPR of teeth in patients undergoing total body irradiation

PubMed Central

Williams, Benjamin B.; Dong, Ruhong; Nicolalde, Roberto J.; Matthews, Thomas P.; Gladstone, David J.; Demidenko, Eugene; Zaki, Bassem I.; Salikhov, Ildar K.; Lesniewski, Piotr N.; Swartz, Harold M.

2014-01-01

Purpose The ability to estimate individual exposures to radiation following a large attack or incident has been identified as a necessity for rational and effective emergency medical response. In vivo electron paramagnetic resonance (EPR) spectroscopy of tooth enamel has been developed to meet this need. Materials and methods A novel transportable EPR spectrometer, developed to facilitate tooth dosimetry in an emergency response setting, was used to measure upper incisors in a model system, in unirradiated subjects, and in patients who had received total body doses of 2 Gy. Results A linear dose response was observed in the model system. A statistically significant increase in the intensity of the radiation-induced EPR signal was observed in irradiated versus unirradiated subjects, with an estimated standard error of dose prediction of 0.9 + 0.3 Gy. Conclusions These results demonstrate the current ability of in vivo EPR tooth dosimetry to distinguish between subjects who have not been irradiated and those who have received exposures that place them at risk for acute radiation syndrome. Procedural and technical developments to further increase the precision of dose estimation and ensure reliable operation in the emergency setting are underway. With these developments EPR tooth dosimetry is likely to be a valuable resource for triage following potential radiation exposure of a large population. PMID:21696339

Fungal decontamination and enhancement of shelf life of edible split beans of wild legume Canavalia maritima by the electron beam irradiation

NASA Astrophysics Data System (ADS)

Supriya, P.; Sridhar, K. R.; Ganesh, S.

2014-03-01

Ripened split beans of the coastal sand dune wild legume Canavalia maritima serve as one of the traditional nutritional sources of the coastal dwellers in Southwest coast of India. Nine fungi were isolated from the unirradiated dry beans by plating on the potato dextrose agar medium. Toxigenic fungus Aspergillus niger showed the highest incidence (33-50%) followed by Aspergillus flavus (14-20%) and Penicillium chrysogenum (7-13%). Unirradiated dry beans and irradiated dry beans with electron beam doses 2.5, 5, 10 and 15 kGy were monitored for occurrence of fungal species and their incidence during 0, 3 and 6 months storage period under laboratory conditions. Irradiation resulted in dose-dependent decrease in fungal species (5-7, 4-6, 3-6 and 0 on irradiation at 0, 2.5, 5 and 10 or 15 kGy, respectively) as well as incidence (80-99, 19-46, 13-21 and 0%, respectively). Although aflatoxins (B1 and B2) were found below detectable level (<2 ng/g) in 0, 3 and 6 months stored unirradiated and irradiated beans (2.5 and 5 kGy), they were not present in beans irradiated with 10 and 15 kGy. In spite of occurrence of toxigenic fungus Aspergillus ochraceus in unirradiated and irradiated beans (2.5 and 5 kGy) stored for 3 and 6 months, the beans were devoid of ochratoxin-A. Electron beam irradiation dose 10 kGy could be recommended for fungal decontamination and improvement of shelf life of C. maritima ripened dry split beans.

Infrared analysis of polyethylene wear specimens using attenuated total reflection spectroscopy. [effects of radiation on the surface properties of materials for total joint protheses

NASA Technical Reports Server (NTRS)

Jones, W. R.; Lauer, J. L.

1979-01-01

Attenuated total reflection infrared spectroscopy was used to analyze ultrahigh molecular weight polyethylene wear test specimens. Three different specimens were analyzed. One specimen was gamma irradiated to a dose of 5.0 MRad, another to a dose of 2.5 MRad, and the final specimen was unirradiated. There was no conclusive evidence of chemical changes (i.e., unsaturation or oxidation) in the surface regions of any of the polyethylene samples. Therefore, it was concluded that the gamma irradiation sterilization procedure shoud not alter the boundary lubricating properties of the polyethylene.

FUEL-BREEDER FUEL ELEMENT FOR NUCLEAR REACTOR

DOEpatents

Abbott, W.E.; Balent, R.

1958-09-16

A fuel element design to facilitate breeding reactor fuel is described. The fuel element is comprised of a coatainer, a central core of fertile material in the container, a first bonding material surrounding the core, a sheet of fissionable material immediately surrounding the first bonding material, and a second bonding material surrounding the fissionable material and being in coniact with said container.

Conductive polymer layers to limit transfer of fuel reactants to catalysts of fuel cells to reduce reactant crossover

DOEpatents

Stanis, Ronald J.; Lambert, Timothy N.

2016-12-06

An apparatus of an aspect includes a fuel cell catalyst layer. The fuel cell catalyst layer is operable to catalyze a reaction involving a fuel reactant. A fuel cell gas diffusion layer is coupled with the fuel cell catalyst layer. The fuel cell gas diffusion layer includes a porous electrically conductive material. The porous electrically conductive material is operable to allow the fuel reactant to transfer through the fuel cell gas diffusion layer to reach the fuel cell catalyst layer. The porous electrically conductive material is also operable to conduct electrons associated with the reaction through the fuel cell gas diffusion layer. An electrically conductive polymer material is coupled with the fuel cell gas diffusion layer. The electrically conductive polymer material is operable to limit transfer of the fuel reactant to the fuel cell catalyst layer.

33 CFR 183.538 - Metallic fuel line materials.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Metallic fuel line materials. 183... (CONTINUED) BOATING SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.538 Metallic fuel line materials. Each metallic fuel line connecting the fuel tank with the fuel inlet connection on...

Fuel pin

DOEpatents

Christiansen, D.W.; Karnesky, R.A.; Leggett, R.D.; Baker, R.B.

1987-11-24

A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

Fuel pin

DOEpatents

Christiansen, David W.; Karnesky, Richard A.; Leggett, Robert D.; Baker, Ronald B.

1989-10-03

A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

Fuel pin

DOEpatents

Christiansen, David W.; Karnesky, Richard A.; Leggett, Robert D.; Baker, Ronald B.

1989-01-01

A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

The use of unirradiated and γ-irradiated zinc oxide nanoparticles as a preservative in cosmetic preparations

PubMed Central

Hosny, Alaa El-Dien MS; Kashef, Mona T; Taher, Hadeer A; El-Bazza, Zeinab E

2017-01-01

Purpose Microbial contamination of different cosmetic preparations, as a result of preservative failure, presents a major public health threat. Also, most of the known preservatives have serious consumer side effects. The antimicrobial activity of zinc oxide nanoparticles (ZnO NP) is well documented. Therefore, we aimed to determine the possible use of unirradiated and γ-irradiated ZnO NP as a cosmetic preservative. Methods The possible use of ZnO NP as a preservative was tested and compared to commonly used preservatives using a challenge test. Their activity was tested in six different types of preparations. The effect of γ radiation on the antimicrobial activity of ZnO NP was tested through determination of the obtained zone diameters against different microorganisms and the total aerobic microbial count in tested preparations. The antimicrobial activity, of unirradiated and γ-irradiated ZnO NP during storage was also determined. Results ZnO NP were superior to other commonly used preservatives in all tested cosmetic preparations. They pass the challenge test in all types of tested preparations. γ irradiation enhanced their antimicrobial activity in all tested preparations. The irradiation causes a reduction in NP sizes that is directly proportional to the applied radiation dose. Upon storage, ZnO NP were effective in maintaining the microbial count of the product within the acceptable range. Their activity in stored products was enhanced by γ irradiation. Conclusion Unirradiated and γ-irradiated ZnO NP can be used as effective preservatives. They are compatible with the components of all tested products. γ irradiation enhanced the antimicrobial activity of ZnO NP. PMID:28979119

Direct methanol feed fuel cell and system

NASA Technical Reports Server (NTRS)

Surampudi, Subbarao (Inventor); Kindler, Andrew (Inventor); Halpert, Gerald (Inventor); Frank, Harvey A. (Inventor); Narayanan, Sekharipuram R. (Inventor); Chun, William (Inventor); Jeffries-Nakamura, Barbara (Inventor)

2009-01-01

Improvements to non acid methanol fuel cells include new formulations for materials. The platinum and ruthenium are more exactly mixed together. Different materials are substituted for these materials. The backing material for the fuel cell electrode is specially treated to improve its characteristics. A special sputtered electrode is formed which is extremely porous. The fuel cell system also comprises a fuel supplying part including a meter which meters an amount of fuel which is used by the fuel cell, and controls the supply of fuel based on said metering.

Measurement and Simulation of Thermal Conductivity of Hafnium-Aluminum Thermal Neutron Absorber Material

DOE PAGES

Guillen, Donna Post; Harris, William H.

2016-05-11

A metal matrix composite (MMC) material comprised of hafnium aluminide (Al3Hf) intermetallic particles in an aluminum matrix has been identified as a promising material for fast-flux irradiation testing applications. This material can filter thermal neutrons while simultaneously providing high rates of conductive cooling for experiment capsules. Our purpose is to investigate effects of Hf-Al material composition and neutron irradiation on thermophysical properties, which were measured before and after irradiation. When performing differential scanning calorimetry (DSC) on the irradiated specimens, a large exotherm corresponding to material annealment was observed. Thus, a test procedure was developed to perform DSC and laser flashmore » analysis (LFA) to obtain the specific heat and thermal diffusivity of pre- and post-annealment specimens. This paper presents the thermal properties for three states of the MMC material: (1) unirradiated, (2) as-irradiated, and (3) irradiated and annealed. Microstructure-property relationships were obtained for the thermal conductivity. These relationships are useful for designing components from this material to operate in irradiation environments. Furthermore, the ability of this material to effectively conduct heat as a function of temperature, volume fraction Al 3Hf, radiation damage and annealing is assessed using the MOOSE suite of computational tools.« less

The tensile and fatigue properties of DIN 1.4914 martensitic stainless steel after 590 MeV proton irradiation

NASA Astrophysics Data System (ADS)

Marmy, P.; Victoria, M.

1992-09-01

Tensile and low cycle fatigue subsize specimens of DIN 1.4914 martensitic steel (MANET) have been irradiated with 590 MeV protons to doses up to 1 dpa and at temperatures between 363 and 703 K. The helium produced by spallation reactions was measured as 130 appm/dpa. A strong radiation hardening is found, which decreases as the irradiation temperature increases. The tensile elongation is reduced after irradiation, but the fracture mode is always ductile and transgranular. The radition hardening produced at low irradiation temperatures is recovered after annealing at higher temperatures. Continous softening is observed during low cycle fatigue testing. The rate of softening of the irradiated material is stonger than that of the unirradiated material and tends to reach the saturation level of the latter. The irradiation badly affects the fatigue life, particularly in the temperature domain of dynamic strain ageing between 553 and 653 K.

Low temperature embrittlement behaviour of different ferritic-martensitic alloys for fusion applications

NASA Astrophysics Data System (ADS)

Rieth, M.; Dafferner, B.

1996-10-01

In the last few years a lot of different low activation CrWVTa steels have been developed world-wide. Without irradiation some of these alloys show clearly a better low temperature embrittlement behaviour than commercial CrNiMoV(Nb) alloys. Within the MANITU project a study was carried out to compare, prior to the irradiation program, the embrittlement behaviour of different alloys in the unirradiated condition performing instrumented Charpy impact bending tests with sub-size specimens. The low activation materials (LAM) considered were different OPTIFER alloys (Forschungszentrum Karlsruhe), F82H (JAERI), 9Cr2WVTa (ORNL), and GA3X (PNL). The modified commercial 10-11% CrNiMoVNb steels were MANET and OPTIMAR. A meaningful comparison between these alloys could be drawn, since the specimens of all materials were manufactured and tested under the same conditions.

Effects of ATR-2 Irradiation to High Fluence on Nine RPV Surveillance Materials

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

Nanstad, Randy K.; Odette, George R.; Almirall, Nathan

2017-05-01

The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely degraded, with the degree of toughnessmore » loss dependent on the radiation sensitivity of the materials. The available embrittlement predictive models and our present understanding of radiation damage are not fully quantitative, and do not treat all potentially significant variables and issues, particularly considering extension of operation to 80y.« less

gamma-Irradiation effects on the thermal decomposition behaviour and IR absorption spectra of piperacillin

NASA Astrophysics Data System (ADS)

Mahfouz, R. M.; Gaffar, M. A.; Abu El-Fadl, A.; Hamad, Ar. G. K.

2003-11-01

The thermal decomposition behaviour of unirradiated and pre-gamma-irradiated piperacillin (pipril) as a semi-synthetic penicillin antibiotic has been studied in the temperature range of (273-1072 K). The decomposition was found to proceed through three major steps both for unirradiated and gamma-irradiated samples. Neither appearance nor disappearance of new bands in the IR spectrum of piperacillin was recorded as a result of gamma-irradiation but only a decrease in the intensity of most bands was observed. A degradation mechanism was suggested to explain the bond rupture and the decrease in the intensities of IR bands of gamma-irradiated piperacillin.

EPR investigation of some traditional oriental irradiated spices

NASA Astrophysics Data System (ADS)

Duliu, Octavian G.; Georgescu, Rodica; Ali, Shaban Ibrahim

2007-06-01

The 9.50 GHz electron paramagnetic resonance (EPR) spectra of unirradiated and 60Co γ-ray irradiated cardamom ( Elettaria cardamomum L. Maton, Zingiberaceae), ginger (( Zingiber officinale Rosc., Zingiberaceae), and saffron ( Crocus sativus L., Iridaceae) have been investigated at room temperature. All unirradiated spices presented a weak resonance line with g-factors around free-electron ones. After γ-ray irradiation at an absorbed dose of up to 11.3 kGy, the presence of EPR spectra whose amplitude increase monotonously with the absorbed dose has been noticed with all spices. A 100 °C isothermal annealing of 11.3 kGy irradiated samples has shown a differential reduction of amplitude of various components that compose initial spectra, but even after 3.6 h of thermal treatment, the remaining amplitude represents no less then 30% of the initial ones. The same peculiarities have been noticed after 83 days storage at room temperature but after 340 days storage at ambient conditions only irradiated ginger displays a weak signal that differs from those of unirradiated sample. All these factors could be taken into account in establishing at which extent the EPR is suitable to evidence any irradiation treatment applied to these spices.

NUCLEAR FUEL MATERIAL

DOEpatents

Goeddel, W.V.

1962-06-26

An improved method is given for making the carbides of nuclear fuel material. The metal of the fuel material, which may be a fissile and/or fertile material, is transformed into a silicide, after which the silicide is comminuted to the desired particle size. This silicide is then carburized at an elevated temperature, either above or below the melting point of the silicide, to produce an intimate mixture of the carbide of the fuel material and the carbide of silicon. This mixture of the fuel material carbide and the silicon carbide is relatively stable in the presence of moisture and does not exhibit the highly reactive surface condition which is observed with fuel material carbides made by most other known methods. (AEC)

Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

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

Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

1976-04-01

Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

49 CFR 173.230 - Fuel cell cartridges containing hazardous material.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Fuel cell cartridges containing hazardous material... Than Class 1 and Class 7 § 173.230 Fuel cell cartridges containing hazardous material. (a) Requirements for Fuel Cell Cartridges. Fuel cell cartridges, including when contained in or packed with equipment...

ALD coating of nuclear fuel actinides materials

DOEpatents

Yacout, A. M.; Pellin, Michael J.; Yun, Di; Billone, Mike

2017-09-05

The invention provides a method of forming a nuclear fuel pellet of a uranium containing fuel alternative to UO.sub.2, with the steps of obtaining a fuel form in a powdered state; coating the fuel form in a powdered state with at least one layer of a material; and sintering the powdered fuel form into a fuel pellet. Also provided is a sintered nuclear fuel pellet of a uranium containing fuel alternative to UO.sub.2, wherein the pellet is made from particles of fuel, wherein the particles of fuel are particles of a uranium containing moiety, and wherein the fuel particles are coated with at least one layer between about 1 nm to about 4 nm thick of a material using atomic layer deposition, and wherein the at least one layer of the material substantially surrounds each interfacial grain barrier after the powdered fuel form has been sintered.

MOX fuel assembly design

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

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

1992-02-18

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

49 CFR 392.51 - Reserve fuel; materials of trade.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 49 Transportation 5 2012-10-01 2012-10-01 false Reserve fuel; materials of trade. 392.51 Section... COMMERCIAL MOTOR VEHICLES Fueling Precautions § 392.51 Reserve fuel; materials of trade. Small amounts of...) may be designated as materials of trade (see 49 CFR 171.8). (a) The aggregate gross weight of all...

75 FR 26025 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard Program

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-10

... information whose disclosure is restricted by statute. Certain other material, such as copyrighted material, will be publicly available only in hard copy. Publicly available docket materials are available either... materials, as provided in 40 CFR part 2. IV. Renewable Fuel Standard (RFS2) Program Amendments EPA is taking...

46 CFR 58.50-15 - Alternate material for construction of independent fuel tanks.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Alternate material for construction of independent fuel...) MARINE ENGINEERING MAIN AND AUXILIARY MACHINERY AND RELATED SYSTEMS Independent Fuel Tanks § 58.50-15 Alternate material for construction of independent fuel tanks. (a) Materials other than those specifically...

Mass, energy and material balances of SRF production process. Part 3: solid recovered fuel produced from municipal solid waste.

PubMed

Nasrullah, Muhammad; Vainikka, Pasi; Hannula, Janne; Hurme, Markku; Kärki, Janne

2015-02-01

This is the third and final part of the three-part article written to describe the mass, energy and material balances of the solid recovered fuel production process produced from various types of waste streams through mechanical treatment. This article focused the production of solid recovered fuel from municipal solid waste. The stream of municipal solid waste used here as an input waste material to produce solid recovered fuel is energy waste collected from households of municipality. This article presents the mass, energy and material balances of the solid recovered fuel production process. These balances are based on the proximate as well as the ultimate analysis and the composition determination of various streams of material produced in a solid recovered fuel production plant. All the process streams are sampled and treated according to CEN standard methods for solid recovered fuel. The results of the mass balance of the solid recovered fuel production process showed that 72% of the input waste material was recovered in the form of solid recovered fuel; 2.6% as ferrous metal, 0.4% as non-ferrous metal, 11% was sorted as rejects material, 12% as fine faction and 2% as heavy fraction. The energy balance of the solid recovered fuel production process showed that 86% of the total input energy content of input waste material was recovered in the form of solid recovered fuel. The remaining percentage (14%) of the input energy was split into the streams of reject material, fine fraction and heavy fraction. The material balances of this process showed that mass fraction of paper and cardboard, plastic (soft) and wood recovered in the solid recovered fuel stream was 88%, 85% and 90%, respectively, of their input mass. A high mass fraction of rubber material, plastic (PVC-plastic) and inert (stone/rock and glass particles) was found in the reject material stream. © The Author(s) 2014.

Hot Hydrogen Testing of Tungsten-Uranium Dioxide (W-UO2) CERMET Fuel Materials for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie

2014-01-01

CERMET fuel materials are being developed at the NASA Marshall Space Flight Center for a Nuclear Cryogenic Propulsion Stage. Recent work has resulted in the development and demonstration of a Compact Fuel Element Environmental Test (CFEET) System that is capable of subjecting depleted uranium fuel material samples to hot hydrogen. A critical obstacle to the development of an NCPS engine is the high-cost and safety concerns associated with developmental testing in nuclear environments. The purpose of this testing capability is to enable low-cost screening of candidate materials, fabrication processes, and further validation of concepts. The CERMET samples consist of depleted uranium dioxide (UO2) fuel particles in a tungsten metal matrix, which has been demonstrated on previous programs to provide improved performance and retention of fission products1. Numerous past programs have utilized hot hydrogen furnace testing to develop and evaluate fuel materials. The testing provides a reasonable simulation of temperature and thermal stress effects in a flowing hydrogen environment. Though no information is gained about radiation damage, the furnace testing is extremely valuable for development and verification of fuel element materials and processes. The current work includes testing of subscale W-UO2 slugs to evaluate fuel loss and stability. The materials are then fabricated into samples with seven cooling channels to test a more representative section of a fuel element. Several iterations of testing are being performed to evaluate fuel mass loss impacts from density, microstructure, fuel particle size and shape, chemistry, claddings, particle coatings, and stabilizers. The fuel materials and forms being evaluated on this effort have all been demonstrated to control fuel migration and loss. The objective is to verify performance improvements of the various materials and process options prior to expensive full scale fabrication and testing. Post test analysis will include weight percent fuel loss, microscopy, dimensional tolerance, and fuel stability.

Engineered glass seals for solid-oxide fuel cells

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

Surdoval, Wayne; Lara-Curzio, Edgar; Stevenson, Jeffry

2017-02-07

A seal for a solid oxide fuel cell includes a glass matrix having glass percolation therethrough and having a glass transition temperature below 650.degree. C. A deformable second phase material is dispersed in the glass matrix. The second phase material can be a compliant material. The second phase material can be a crushable material. A solid oxide fuel cell, a precursor for forming a seal for a solid oxide fuel cell, and a method of making a seal for a solid oxide fuel cell are also disclosed.

The effect of ion irradiation on the dissolution of UO 2 and UO 2 -based simulant fuel

DOE PAGES

Popel, Aleksej J.; Wietsma, Thomas W.; Engelhard, Mark H.; ...

2017-11-21

Our aim is to study the separate effect of fission fragment damage on the dissolution of simulant UK advanced gas-cooled reactor nuclear fuel in water. Plain UO 2 and UO 2 samples, doped with inactive fission products to simulate 43 GWd/tU of burn-up, were fabricated. A set of these samples were then irradiated with 92 MeV 129Xe 23+ ions to a fluence of 4.8 × 10 15 ions/cm 2 to simulate the fission damage that occurs within nuclear fuels. The primary effect of the irradiation on the UO 2 samples, observed by scanning electron microscopy, was to induce a smootheningmore » of the surface features and formation of hollow blisters, which was attributed to multiple overlap of ion tracks. Dissolution experiments were conducted in single-pass flow-through (SPFT) mode under anoxic conditions (<0.1 O 2 ppm in Ar) to study the effect of the induced irradiation damage on the dissolution of the UO 2 matrix with data collection capturing six minute intervals for several hours. These time-resolved data showed that the irradiated samples showed a higher initial release of uranium than unirradiated samples, but that the uranium concentrations converged towards ~10 -9 mol/l after a few hours. And apart from the initial spike in uranium concentration, attributed to irradiation induced surficial micro-structural changes, no noticeable difference in uranium chemistry as measured by X-ray electron spectroscopy or ‘effective solubility’ was observed between the irradiated, doped and undoped samples in this work. Some secondary phase formation was observed on the surface of UO 2 samples after the dissolution experiment.« less

The effect of ion irradiation on the dissolution of UO 2 and UO 2 -based simulant fuel

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

Popel, Aleksej J.; Wietsma, Thomas W.; Engelhard, Mark H.

Our aim is to study the separate effect of fission fragment damage on the dissolution of simulant UK advanced gas-cooled reactor nuclear fuel in water. Plain UO 2 and UO 2 samples, doped with inactive fission products to simulate 43 GWd/tU of burn-up, were fabricated. A set of these samples were then irradiated with 92 MeV 129Xe 23+ ions to a fluence of 4.8 × 10 15 ions/cm 2 to simulate the fission damage that occurs within nuclear fuels. The primary effect of the irradiation on the UO 2 samples, observed by scanning electron microscopy, was to induce a smootheningmore » of the surface features and formation of hollow blisters, which was attributed to multiple overlap of ion tracks. Dissolution experiments were conducted in single-pass flow-through (SPFT) mode under anoxic conditions (<0.1 O 2 ppm in Ar) to study the effect of the induced irradiation damage on the dissolution of the UO 2 matrix with data collection capturing six minute intervals for several hours. These time-resolved data showed that the irradiated samples showed a higher initial release of uranium than unirradiated samples, but that the uranium concentrations converged towards ~10 -9 mol/l after a few hours. And apart from the initial spike in uranium concentration, attributed to irradiation induced surficial micro-structural changes, no noticeable difference in uranium chemistry as measured by X-ray electron spectroscopy or ‘effective solubility’ was observed between the irradiated, doped and undoped samples in this work. Some secondary phase formation was observed on the surface of UO 2 samples after the dissolution experiment.« less

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-05-01

A nuclear reactor fuel element comprising high density ceramic fissionable material enclosed in a tubular cladding of corrosion-resistant material is described. The fissionable material is in the form of segments of a tube which have cooperating tapered interfaces which produce outward radial displacement when the segments are urged axially together. A resilient means is provided within the tubular housing to constantly urge the fuel segments axially. This design maintains the fuel material in tight contacting engagement against the inner surface of the outer cladding tube to eliminate any gap therebetween which may be caused by differential thermal expansion between the fuel material and the material of the tube.

AN EVALUATION OF POTENTIAL LINER MATERIALS FOR ELIMINATING FCCI IN IRRADIATED METALLIC NUCLEAR FUEL ELEMENTS

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

D. D. Keiser; J. I. Cole

2007-09-01

Metallic nuclear fuels are being looked at as part of the Global Nuclear Energy Program for transmuting longlive transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. In order to optimize the performance of these fuels, the concept of using liners to eliminate the fuel/cladding chemical interactions that can occur during irradiation of a fuel element has been investigated. The potential liner materials Zr and V have been tested using solid-solid diffusion couples, consisting of liner materials butted against fuel alloys and against cladding materials. The couples were annealed at the relatively high temperature of 700°C. Thismore » temperature would be the absolute maximum temperature present at the fuel/cladding interface for a fuel element in-reactor. Analysis was performed using a scanning electron microscope equipped with energy-dispersive and wavelengthdispersive spectrometers (SEM/EDS/WDS) to evaluate any developed diffusion structures. At 700°C, minimal interaction was observed between the metallic fuels and either Zr or V. Similarly, limited interaction was observed between the Zr and V and the cladding materials. The best performing liner material appeared to be the V, based on amounts of interaction.« less

Electrolytes for solid oxide fuel cells

NASA Astrophysics Data System (ADS)

Fergus, Jeffrey W.

The high operating temperature of solid oxide fuel cells (SOFCs), as compared to polymer electrolyte membrane fuel cells (PEMFCs), improves tolerance to impurities in the fuel, but also creates challenges in the development of suitable materials for the various fuel cell components. In response to these challenges, intermediate temperature solid oxide fuel cells (IT-SOFCs) are being developed to reduce high-temperature material requirements, which will extend useful lifetime, improve durability and reduce cost, while maintaining good fuel flexibility. A major challenge in reducing the operating temperature of SOFCs is the development of solid electrolyte materials with sufficient conductivity to maintain acceptably low ohmic losses during operation. In this paper, solid electrolytes being developed for solid oxide fuel cells, including zirconia-, ceria- and lanthanum gallate-based materials, are reviewed and compared. The focus is on the conductivity, but other issues, such as compatibility with electrode materials, are also discussed.

Compatibility Studies of Hydrogen Peroxide and a New Hypergolic Fuel Blend

NASA Technical Reports Server (NTRS)

Baldridge, Jennifer; Villegas, Yvonne

2002-01-01

Several preliminary materials compatibility studies have been conducted to determine the practicality of a new hypergolic fuel system. Hypergolic fuel ignites spontaneously as the oxidizer decomposes and releases energy in the presence of the fuel. The bipropellant system tested consists of high-test hydrogen peroxide (HTP) and a liquid fuel blend consisting of a hydrocarbon fuel, an ignition enhancer and a transition metal catalyst. In order for further testing of the new fuel blend to take place, some basic materials compatibility and HTP decomposition studies must be accomplished. The thermal decomposition rate of HTP was tested using gas evolution and isothermal microcalorimetry (IMC). Materials were analyzed for compatibility with hydrogen peroxide including a study of the affect welding has on stainless steel elemental composition and its relation to HTP decomposition. Compatibility studies of valve materials in the fuel blend were performed to determine the corrosion resistance of the materials.

ESR detection of irradiated carob pods (Ceratoniasiliqua L) and its dosimetric feature

NASA Astrophysics Data System (ADS)

Tuner, Hasan; Polat, Mustafa

2017-12-01

Un-irradiated carob powder exhibited a weak ESR singlet at g = 2.0041 ± 0.0006 with peak-to-peak linewidth (ΔHpp) of 0.33 ± 0.01 mT. Irradiated carob powder exhibited an ESR spectrum consisting many resonance lines and similar to ESR spectrum of sugar in all aspects. A linear function of the absorbed radiation dose was found to describe best the dose-response curves of the ESR signal intensity Ipp. It is concluded that due to the similarity of carob powder ESR spectrum to the irradiated sugar and the fact that it is widely consumed, carob powder has the potential to be used as a retrospective and/or accidental dosimetric material.

Applications of polymer coatings for the fabrication of copper-based containers for the ultimate disposal of Canada's spent nuclear fuel

NASA Astrophysics Data System (ADS)

Mortley, Aba

Oxygen-free, phosphorous doped copper containers have been proposed for the storage of the used nuclear fuel bundles as a part of Canada's multi-barrier, adaptive phased management procedure for long term storage of spent nuclear fuel bundles. The spent nuclear fuel disposal system proposed for Canada has been engineered based on the multi-barrier approach intended to minimize the risk that the radioactive materials enter the biosphere. Copper is known to be susceptible to corrosion and it is thought that the simultaneous exposure to aggressive ionizing radiation field and residual heat produced by the spent nuclear fuel and the surrounding groundwater would all challenge the container's integrity. The goal of the present work is to reduce the impact of corrosion in the early stages of emplacement with the addition of a protective coating. Specifically, castor oil based polyurethanes were assessed as coatings and their ability to act as an additional physical barrier in the multi-barrier system mentioned previously. The novelty of this work stems from the use of a naturally derived non-petroleum based material in the form of castor oil as the polyol component. Two types of castor oil polyurethanes were investigated, one based on an aliphatic hexamethylene diisocyanate (HMDI), and the other based on an aromatic 2,4-toluene diisocyanate (TDI). Radiation and saturation tests were conducted using varying conditions. Mixed field ionizing radiation was provided by a SLOWPOKE-2 pool-type nuclear research reactor, up to accumulated doses of 6 MGy at dose rates of 37 kGy h-1 and 55.5 kGy h-1. Weight gain immersion studies, at temperatures of 25° C, 50° C, 70° C, were used to determine the mass uptake of several different solutions. The solutions utilized in the present work included hydrochloric acids of varying pHs, distilled water, and buffered solutions, which simulated chloride and sulphide rich calcium-sodium bicarbonate waters. After being exposed to radiation and saturation individually and in several combinations, the polymer samples were tested using a battery of tests. These tests were used to determine how the physico-mechanical properties of the materials, such as their strength, and their ability to be deformed and stretched were affected by radiation. The gist of these tests being that, if the sample showed significant variations in the physico-mechanical properties, then the material would be deemed as unacceptable for use in combined radiation-temperature-pH environments of the deep geological repository (DGR). The tests used included: Fourier transform infrared spectroscopy (FTIR), Matrix-assisted laser desorption/ionization (MALDI) spectroscopy, electronic ionization mass spectroscopy (EI MS), wide angle x-ray scattering (WAXS), solid-state nuclear magnetic resonance ( 13C-NMR), dynamic mechanical analysis (DMA), and differential scanning calorimetry (DSC). Radiation dose increases from 0 MGy to 2.0 MGy showed a four-fold increase from 0.930 MPa to 4.365 MPa in the initial modulus and 0.149 MPa to 0.747 MPa in the tensile strength values of the polyurethanes. Above 2.0 MGy, the modulus and the tensile strength values exhibited a plateau at values above those of the unirradiated samples. The sorption and diffusion experiments demonstrated a mass uptake of less than 2 %, regardless of what polar solution was used. As a reflection of the more rigid polymer matrix provided by the aromatic ring, the aromatic polyurethanes generally exhibited higher modulus and tensile strength values and lower solution mass uptake and diffusivities. The diffusion values ranged from 8.5 x 10-7 cm2 s-1 to 3.31 X 10-6 cm2 s -1 for the aliphatic polyurethanes and 8.8 x 10-7 cm 2 s-1 to 1.60 x 10-6 cm2 S-1 for the aromatic polyurethanes. The order in which the experimental conditions were applied, Le. sequentially or simultaneously, had definite effects on the finished product. However, regardless of the manner in which the radiation and saturation were combined, in a sequential or simultaneous manner, the end results showed that the moduli values remained above those of the virgin samples (> 2 MPa). The results of the present work indicate that the castor oil based polyurethanes may indeed be used as a viable material where the end-use conditions include combined radiation-thermal-pH environments such as part of the container to store used nuclear fuel. Future development on this work can look at how the adhesive qualities between the castor oil based polyurethanes and the metal copper container may change in such environments.

Findings and Recommendations from the NIST Workshop on Alternative Fuels and Materials: Biocorrosion

PubMed Central

Mansfield, Elisabeth; Sowards, Jeffrey W.; Crookes-Goodson, Wendy J.

2015-01-01

In 2013, the Applied Chemicals and Materials Division of the National Institute of Standards and Technology (NIST) hosted a workshop to identify and prioritize research needs in the area of biocorrosion. Materials used to store and distribute alternative fuels have experienced an increase in corrosion due to the unique conditions caused by the presence of microbes and the chemistry of biofuels and biofuel precursors. Participants in this workshop, including experts from the microbiological, fuel, and materials communities, delved into the unique materials and chemical challenges that occur with production, transport, and storage of alternative fuels. Discussions focused on specific problems including: a) the changing composition of “drop-in” fuels and the impact of that composition on materials; b) the influence of microbial populations on corrosion and fuel quality; and c) state-of-the-art measurement technologies for monitoring material degradation and biofilm formation. PMID:26958436

Findings and Recommendations from the NIST Workshop on Alternative Fuels and Materials: Biocorrosion.

PubMed

Mansfield, Elisabeth; Sowards, Jeffrey W; Crookes-Goodson, Wendy J

2015-01-01

In 2013, the Applied Chemicals and Materials Division of the National Institute of Standards and Technology (NIST) hosted a workshop to identify and prioritize research needs in the area of biocorrosion. Materials used to store and distribute alternative fuels have experienced an increase in corrosion due to the unique conditions caused by the presence of microbes and the chemistry of biofuels and biofuel precursors. Participants in this workshop, including experts from the microbiological, fuel, and materials communities, delved into the unique materials and chemical challenges that occur with production, transport, and storage of alternative fuels. Discussions focused on specific problems including: a) the changing composition of "drop-in" fuels and the impact of that composition on materials; b) the influence of microbial populations on corrosion and fuel quality; and c) state-of-the-art measurement technologies for monitoring material degradation and biofilm formation.

NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT

DOEpatents

Currier, E.L. Jr.; Nicklas, J.H.

1962-08-14

A fuel-breeder fuel element was developed for a nuclear reactor wherein discrete particles of fissionable material are dispersed in a matrix of fertile breeder material. The fuel element combines the advantages of a dispersion type and a breeder-type. (AEC)

Direct induction of molecular alignment in liquid crystal polymer network film by photopolymerization

NASA Astrophysics Data System (ADS)

Hisano, K.; Aizawa, M.; Ishizu, M.; Kurata, Y.; Shishido, A.

2016-09-01

Liquid crystal (LC) is the promising material for the fabrication of high-performance soft, flexible devices. The fascinating and useful properties arise from their cooperative effect that inherently allows the macroscopic integration and control of molecular alignment through various external stimuli. To date, light-matter interaction is the most attractive stimuli and researchers developed photoalignment through photochemical or photophysical reactions triggered by linearly polarized light. Here we show the new choice based on molecular diffusion by photopolymerization. We found that photopolymerization of a LC monomer and a crosslinker through a photomask enables to direct molecular alignment in the resultant LC polymer network film. The key generating the molecular alignment is molecular diffusion due to the difference of chemical potentials between irradiated and unirradiated regions. This concept is applicable to various shapes of photomask and two-dimensional molecular alignments can be fabricated depending on the spatial design of photomask. By virtue of the inherent versatility of molecular diffusion in materials, the process would shed light on the fabrication of various high-performance flexible materials with molecular alignment having controlled patterns.

Nuclear fuel element with axially aligned fuel pellets and fuel microspheres therein

DOEpatents

Sease, J.D.; Harrington, F.E.

1973-12-11

Elongated single- and multi-region fuel elements are prepared by replacing within a cladding container a coarse fraction of fuel material which includes plutonium and uranium in the appropriate regions of the fuel element and then infiltrating with vibration a fine-sized fraction of uranium-containing microspheres throughout all interstices in the coarse material in a single loading. The fine, rigid material defines a thin annular layer between the coarse fraction and the cladding to reduce adverse mechanical and chemical interactions. (Official Gazette)

33 CFR 183.512 - Fuel tanks: Prohibited materials.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 33 Navigation and Navigable Waters 2 2012-07-01 2012-07-01 false Fuel tanks: Prohibited materials... tanks: Prohibited materials. (a) A fuel tank must not be constructed from terneplate. (b) Unless it has an inorganic sacrificial galvanic coating on the inside and outside of the tank, a fuel tank must not...

Nuclear reactor fuel element having improved heat transfer

DOEpatents

Garnier, J.E.; Begej, S.; Williford, R.E.; Christensen, J.A.

1982-03-03

A nuclear reactor fuel element having improved heat transfer between fuel material and cladding is described. The element consists of an outer cladding tube divided into an upper fuel section containing a central core of fissionable or mixed fissionable and fertile fuel material, slightly smaller in diameter than the inner surface of the cladding tube and a small lower accumulator section, the cladding tube being which is filled with a low molecular weight gas to transfer heat from fuel material to cladding during irradiation. A plurality of essentially vertical grooves in the fuel section extend downward and communicate with the accumulator section. The radial depth of the grooves is sufficient to provide a thermal gradient between the hot fuel surface and the relatively cooler cladding surface to allow thermal segregation to take place between the low molecular weight heat transfer gas and high molecular weight fission product gases produced by the fuel material during irradiation.

Laser ablation based fuel ignition

DOEpatents

Early, J.W.; Lester, C.S.

1998-06-23

There is provided a method of fuel/oxidizer ignition comprising: (a) application of laser light to a material surface which is absorptive to the laser radiation; (b) heating of the material surface with the laser light to produce a high temperature ablation plume which emanates from the heated surface as an intensely hot cloud of vaporized surface material; and (c) contacting the fuel/oxidizer mixture with the hot ablation cloud at or near the surface of the material in order to heat the fuel to a temperature sufficient to initiate fuel ignition. 3 figs.

Laser ablation based fuel ignition

DOEpatents

Early, James W.; Lester, Charles S.

1998-01-01

There is provided a method of fuel/oxidizer ignition comprising: (a) application of laser light to a material surface which is absorptive to the laser radiation; (b) heating of the material surface with the laser light to produce a high temperature ablation plume which emanates from the heated surface as an intensely hot cloud of vaporized surface material; and (c) contacting the fuel/oxidizer mixture with the hot ablation cloud at or near the surface of the material in order to heat the fuel to a temperature sufficient to initiate fuel ignition.

Low Activation Joining of SiC/SiC Composites for Fusion Applications: Modeling Miniature Torsion Tests with Elastic and Elastic-Plastic Models

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

Henager, Charles H.; Nguyen, Ba Nghiep; Kurtz, Richard J.

2015-03-01

The use of SiC and SiC-composites in fission or fusion environments requires joining methods for assembling systems. The international fusion community designed miniature torsion specimens for joint testing and irradiation in test reactors with limited irradiation volumes. These torsion specimens fail out-of-plane when joints are strong and when elastic moduli are within a certain range compared to SiC, which causes difficulties in determining shear strengths for joints or for comparing unirradiated and irradiated joints. A finite element damage model was developed that indicates fracture is likely to occur within the joined pieces to cause out-of-plane failures for miniature torsion specimensmore » when a certain modulus and strength ratio between the joint material and the joined material exists. The model was extended to treat elastic-plastic joints such as SiC/epoxy and steel/epoxy joints tested as validation of the specimen design.« less

Enhanced mutagenesis parallels enhanced reactivation of herpes virus in a human cell line.

PubMed Central

Lytle, C D; Knott, D C

1982-01-01

U.v. irradiation of human NB-E cells results in enhanced mutagenesis and enhanced reactivation of u.v.-irradiated H-1 virus grown in those cells ( Cornelis et al., 1982). This paper reports a similar study using herpes simplex virus (HSV) in NB-E cells. The mutation frequency of HSV (resistance of virus plaque formation to 40 micrograms/ml iododeoxycytidine ) increased approximately linearly with exposure of the virus to u.v. radiation. HSV grown in unirradiated cells gave a slope of 1.8 X 10(-5)m2/J, with 3.2 X 10(-5)m2/J for HSV grown in cells irradiated (3 J/m2) 24 h before infection. There was no evidence for mutagenesis of unirradiated virus by irradiated cells, as seen with H-1 virus. Enhanced reactivation of irradiated HSV in parallel cultures increased virus survival, manifested as a change in slope of the final component of the two-component survival curve from a D0 of 27 J/m2 in unirradiated cells to 45 J/m2 in irradiated cells. Thus, enhanced mutagenesis and enhanced reactivation occurred for irradiated HSV in NB-E cells. The difference in the enhanced mutagenesis of HSV (dependent on damaged DNA sites) and of H-1 virus (primarily independent of damaged DNA sites) is discussed in terms of differences in DNA polymerases. PMID:6329698

Fuel cell electrode interconnect contact material encapsulation and method

DOEpatents

Derose, Anthony J.; Haltiner, Jr., Karl J.; Gudyka, Russell A.; Bonadies, Joseph V.; Silvis, Thomas W.

2016-05-31

A fuel cell stack includes a plurality of fuel cell cassettes each including a fuel cell with an anode and a cathode. Each fuel cell cassette also includes an electrode interconnect adjacent to the anode or the cathode for providing electrical communication between an adjacent fuel cell cassette and the anode or the cathode. The interconnect includes a plurality of electrode interconnect protrusions defining a flow passage along the anode or the cathode for communicating oxidant or fuel to the anode or the cathode. An electrically conductive material is disposed between at least one of the electrode interconnect protrusions and the anode or the cathode in order to provide a stable electrical contact between the electrode interconnect and the anode or cathode. An encapsulating arrangement segregates the electrically conductive material from the flow passage thereby, preventing volatilization of the electrically conductive material in use of the fuel cell stack.

Direct methanol feed fuel cell and system

NASA Technical Reports Server (NTRS)

Kindler, Andrew (Inventor); Halpert, Gerald (Inventor); Frank, Harvey A. (Inventor); Chun, William (Inventor); Jeffries-Nakamura, Barbara (Inventor); Surampudi, Subbarao (Inventor); Narayanan, Sekharipuram R. (Inventor)

2008-01-01

Improvements to non acid methanol fuel cells include new formulations for materials. The platinum and ruthenium are more exactly mixed together. Different materials are substituted for these materials. The backing material for the fuel cell electrode is specially treated to improve its characteristics. A special sputtered electrode is formed which is extremely porous.

Direct methanol feed fuel cell and system

NASA Technical Reports Server (NTRS)

Frank, Harvey A. (Inventor); Narayanan, Sekharipuram R. (Inventor); Chun, William (Inventor); Jeffries-Nakamura, Barbara (Inventor); Halpert, Gerald (Inventor); Surampudi, Subbarao (Inventor); Kindler, Andrew (Inventor)

2004-01-01

Improvements to non acid methanol fuel cells include new formulations for materials. The platinum and ruthenium are more exactly mixed together. Different materials are substituted for these materials. The backing material for the fuel cell electrode is specially treated to improve its characteristics. A special sputtered electrode is formed which is extremely porous.

Direct methanol feed fuel cell and system

NASA Technical Reports Server (NTRS)

Frank, Harvey A. (Inventor); Narayanan, Sekharipuram R. (Inventor); Chun, William (Inventor); Jeffries-Nakamura, Barbara (Inventor); Halpert, Gerald (Inventor); Surampudi, Subbarao (Inventor); Kindler, Andrew (Inventor)

2000-01-01

Improvements to non-acid methanol fuel cells include new formulations for materials. The platinum and ruthenium are more exactly mixed together. Different materials are substituted for these materials. The backing material for the fuel cell electrode is specially treated to improve its characteristics. A special sputtered electrode is formed which is extremely porous.

Direct methanol feed fuel cell and system

NASA Technical Reports Server (NTRS)

Surampudi, Subbarao (Inventor); Frank, Harvey A. (Inventor); Narayanan, Sekharipuram R. (Inventor); Chun, William (Inventor); Halpert, Gerald (Inventor); Jeffries-Nakamura, Barbara (Inventor); Kindler, Andrew (Inventor)

2001-01-01

Improvements to non acid methanol fuel cells include new formulations for materials. The platinum and ruthenium are more exactly mixed together. Different materials are substituted for these materials. The backing material for the fuel cell electrode is specially treated to improve its characteristics. A special sputtered electrode is formed which is extremely porous.

78 FR 67223 - Proposed Guidance for Fuel Cycle Facility; Material Control and Accounting Plans and Completing...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-08

..., 72, et al. Proposed Guidance for Fuel Cycle Facility; Material Control and Accounting Plans and Completing NRC Form 327 and Amendments to Material Control and Accounting Regulations; Proposed Rules #0;#0... Guidance for Fuel Cycle Facility; Material Control and Accounting Plans and Completing NRC Form 327 AGENCY...

Summary report on transportation of nuclear fuel materials in Japan : transportation infrastructure, threats identified in open literature, and physical protection regulations.

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

Cochran, John Russell; Ouchi, Yuichiro; Furaus, James Phillip

2008-03-01

This report summarizes the results of three detailed studies of the physical protection systems for the protection of nuclear materials transport in Japan, with an emphasis on the transportation of mixed oxide fuel materials1. The Japanese infrastructure for transporting nuclear fuel materials is addressed in the first section. The second section of this report presents a summary of baseline data from the open literature on the threats of sabotage and theft during the transport of nuclear fuel materials in Japan. The third section summarizes a review of current International Atomic Energy Agency, Japanese and United States guidelines and regulations concerningmore » the physical protection for the transportation of nuclear fuel materials.« less

Analysis on fuel breeding capability of FBR core region based on minor actinide recycling doping

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

Permana, Sidik; Novitrian,; Waris, Abdul

Nuclear fuel breeding based on the capability of fuel conversion capability can be achieved by conversion ratio of some fertile materials into fissile materials during nuclear reaction processes such as main fissile materials of U-233, U-235, Pu-239 and Pu-241 and for fertile materials of Th-232, U-238, and Pu-240 as well as Pu-238. Minor actinide (MA) loading option which consists of neptunium, americium and curium will gives some additional contribution from converted MA into plutonium such as conversion Np-237 into Pu-238 and it's produced Pu-238 converts to Pu-239 via neutron capture. Increasing composition of Pu-238 can be used to produce fissilemore » material of Pu-239 as additional contribution. Trans-uranium (TRU) fuel (Mixed fuel loading of MOX (U-Pu) and MA composition) and mixed oxide (MOX) fuel compositions are analyzed for comparative analysis in order to show the effect of MA to the plutonium productions in core in term of reactor criticality condition and fuel breeding capability. In the present study, neptunium (Np) nuclide is used as a representative of MAin trans-uranium (TRU) fuel composition as Np-MOX fuel type. It was loaded into the core region gives significant contribution to reduce the excess reactivity in comparing to mixed oxide (MOX) fuel and in the same time it contributes to increase nuclear fuel breeding capability of the reactor. Neptunium fuel loading scheme in FBR core region gives significant production of Pu-238 as fertile material to absorp neutrons for reducing excess reactivity and additional contribution for fuel breeding.« less

Revolutionary opportunities for materials and structures study

NASA Technical Reports Server (NTRS)

Schweiger, F. A.

1987-01-01

The revolutionary opportunities for materials and structures study was performed to provide Government and Industry focus for advanced materials technology. Both subsonic and supersonic engine studies and aircraft fuel burn and DOC evaluation are examined. Year 2010 goal materials were used in the advanced engine studies. These goal materials and improved component aero yielded subsonic fuel burn and DOC improvements of 13.4 percent and 5 percent, respectively and supersonic fuel burn and DOC improvements of 21.5 percent and 18 percent, respectively. Conclusions are that the supersonic study engine yielded fuel burn and DOC improvements well beyond the program goals; therefore, it is appropriate that advanced material programs be considered.

Materials technology for an advanced space power nuclear reactor concept: Program summary

NASA Technical Reports Server (NTRS)

Gluyas, R. E.; Watson, G. K.

1975-01-01

The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

FUEL ELEMENT CONSTRUCTION

DOEpatents

Zumwalt, L.R.

1961-08-01

Fuel elements having a solid core of fissionable material encased in a cladding material are described. A conversion material is provided within the cladding to react with the fission products to form stable, relatively non- volatile compounds thereby minimizing the migration of the fission products into the coolant. The conversion material is preferably a metallic fluoride, such as lead difluoride, and may be in the form of a coating on the fuel core or interior of the cladding, or dispersed within the fuel core. (AEC)

Nuclear fuel elements made from nanophase materials

DOEpatents

Heubeck, Norman B.

1998-01-01

A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000.degree. F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics.

Nuclear fuel elements made from nanophase materials

DOEpatents

Heubeck, N.B.

1998-09-08

A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000 F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics. 5 figs.

An electron spin resonance study of some gamma-irradiated fruits

NASA Astrophysics Data System (ADS)

Maloney, Darren R.; Tabner, Brian J.; Tabner, Vivienne A.

The ESR spectra of the seeds, skins and stalks of unirradiated and γ-irradiated Chilean white grapes have been obtained and the results compared to those previously reported for Cape black grapes. The high degree of reproducibility of the spectra obtained from the stalks of different varieties of grapes suggest that ESR spectroscopy could form the basis of a viable test to determine their irradiation history. The condition of the stalk prior to irradiation has been found to have little effect on the resulting spectra. The spectra from the stalks, skins and seeds of unirradiated and γ-irradiated apples, peers and cherries have also been examined. Although most of the spectra from irradiated components exhibit extra features, they are sometimes short-lived and restrict the development of ESR as a viable test.

Thermal analysis applied to irradiated propolis

NASA Astrophysics Data System (ADS)

Matsuda, Andrea Harumi; Machado, Luci Brocardo; del Mastro, Nélida Lucia

2002-03-01

Propolis is a resinous hive product, collected by bees. Raw propolis requires a decontamination procedure and irradiation appears as a promising technique for this purpose. The valuable properties of propolis for food and pharmaceutical industries have led to increasing interest in its technological behavior. Thermal analysis is a chemical analysis that gives information about changes on heating of great importance for technological applications. Ground propolis samples were 60Co gamma irradiated with 0 and 10 kGy. Thermogravimetry curves shown a similar multi-stage decomposition pattern for both irradiated and unirradiated samples up to 600°C. Similarly, through differential scanning calorimetry , a coincidence of melting point of irradiated and unirradiated samples was found. The results suggest that the irradiation process do not interfere on the thermal properties of propolis when irradiated up to 10 kGy.

Effect of ultrasonic irradiation on mammalian cells and chromosomes in vitro

NASA Technical Reports Server (NTRS)

Roseboro, J. A.; Buchanan, P.; Norman, A.; Stern, R.

1978-01-01

Human peripheral blood and HeLa cells were irradiated in vitro at the ultrasonic frequency of 65 kHz. The whole blood and HeLa cell suspensions were exposed to continuous and pulsed ultrasonic power levels of 0.12, 0.16, 0.72, 1.12 and 2.24 W for a period of one minute. The method of ultrasonic irradiation was carried out with the whole blood or HeLa cell suspensions coupled directly to a cylindrical transducer while heating of the cell suspensions in excess of 41 C was avoided. Irradiated and unirradiated peripheral blood lymphocyte chromosome cultures were prepared and scored for selected numerical and morphological aberrations. There was no significant difference in the frequency of chromosomal aberrations between irradiated and unirradiated cells.

A Comparison of Materials Issues for Cermet and Graphite-Based NTP Fuels

NASA Technical Reports Server (NTRS)

Stewart, Mark E.; Schnitzler, Bruce G.

2013-01-01

This paper compares material issues for cermet and graphite fuel elements. In particular, two issues in NTP fuel element performance are considered here: ductile to brittle transition in relation to crack propagation, and orificing individual coolant channels in fuel elements. Their relevance to fuel element performance is supported by considering material properties, experimental data, and results from multidisciplinary fluid/thermal/structural simulations. Ductile to brittle transition results in a fuel element region prone to brittle fracture under stress, while outside this region, stresses lead to deformation and resilience under stress. Poor coolant distribution between fuel element channels can increase stresses in certain channels. NERVA fuel element experimental results are consistent with this interpretation. An understanding of these mechanisms will help interpret fuel element testing results.

Fabrication technological development of the oxide dispersion strengthened alloy MA957 for fast reactor applications

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

ML Hamilton; DS Gelles; RJ Lobsinger

A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material, in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report. The alloy is a ferritic stainless steel developed by International Nickel Company specifically for structural reactor applications. It is strengthened by a very fine, uniformly distributed yttria dispersoid. Its fabrication involves a mechanical alloying process and subsequent extrusion, which ultimately results in a highly elongated grain structure. While the presence ofmore » the dispersoid produces a material with excellent strength, the body centered cubic structure inherent to the material coupled with the high aspect ratio that results from processing operations produces some difficulties with ductility. The alloy is very sensitive to variations in a number of processing parameters, and if the high strength is once lost during fabrication, it cannot be recovered. The microstructural evolution of the alloy under irradiation falls into two regimes. Below about 550 C, dislocation development, {alpha}{prime} precipitation and void evolution in the matrix are observed, while above about 550 C damage appears to be restricted to cavity formation within oxide particles. The thermal expansion of the alloy is very similar to that of HT9 up to the temperature where HT9 undergoes a phase transition to austenitic. Pulse magnetic welding of end caps onto MA957 tubing can be accomplished in a manner similar to that in which it is performed on HT9, although the welding parameters appear to be very sensitive to variations in the tubing that result from small changes in fabrication conditions. The tensile and stress rupture behavior of the alloy are acceptable in the unirradiated condition, being comparable to HT9 below about 700 C and exceeding those of HT9 at higher temperatures. Neither tensile nor rupture strength appear to be degraded by irradiation to fast fluencies on the order of 8 x 10{sup 22} n/cm{sup 2} in the range of 370--760 C, although some loss of ductility has been observed. The impact resistance of the alloy is very poor in the unirradiated condition, and is significantly degraded by irradiation.« less

U.S. Department of Energy-Funded Performance Validation of Fuel Cell Material Handling Equipment (Presentation)

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

Kurtz, J.; Sprik, S.; Ramsden, T.

2013-11-01

This webinar presentation to the UK Hydrogen and Fuel Cell Association summarizes how the U.S. Department of Energy is enabling early fuel cell markets; describes objectives of the National Fuel Cell Technology Evaluation Center; and presents performance status of fuel cell material handling equipment.

Materials and Fuels Complex Tour

ScienceCinema

Miley, Don

2017-12-11

The Materials and Fuels Complex at Idaho National Laboratory is home to several facilities used for the research and development of nuclear fuels. Stops include the Fuel Conditioning Facility, the Hot Fuel Examination Facility (post-irradiation examination), and the Space and Security Power System Facility, where radioisotope thermoelectric generators (RTGs) are assembled for deep space missions.

METHOD AND APPARATUS FOR IMPROVING PERFORMANCE OF A FAST REACTOR

DOEpatents

Koch, L.J.

1959-01-20

A specific arrangement of the fertile material and fissionable material in the active portion of a fast reactor to achieve improvement in performance and to effectively lower the operating temperatures in the center of the reactor is described. According to this invention a group of fuel elements containing fissionable material are assembled to form a hollow fuel core. Elements containing a fertile material, such as depleted uranium, are inserted into the interior of the fuel core to form a central blanket. Additional elemenis of fertile material are arranged about the fuel core to form outer blankets which in tunn are surrounded by a reflector. This arrangement of fuel core and blankets results in substantial flattening of the flux pattern.

Space Exploration Initiative Fuels, Materials and Related Nuclear Propulsion Technologies Panel

NASA Technical Reports Server (NTRS)

Bhattacharyya, S. K.; Olsen, C.; Cooper, R.; Matthews, R. B.; Walter, C.; Titran, R. J.

1993-01-01

This report was prepared by members of the Fuels, Materials and Related Technologies Panel, with assistance from a number of industry observers as well as laboratory colleagues of the panel members. It represents a consensus view of the panel members. This report was not subjected to a thorough review by DOE, NASA or DoD, and the opinions expressed should not be construed to represent the official position of these organizations, individually or jointly. Topics addressed include: requirement for fuels and materials development for nuclear thermal propulsion (NTP) and nuclear electric propulsion (NEP); overview of proposed concepts; fuels technology development plan; materials technology development plan; other reactor technology development; and fuels and materials requirements for advanced propulsion concepts.

40 CFR 60.41 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-07-01

... PERFORMANCE FOR NEW STATIONARY SOURCES Standards of Performance for Fossil-Fuel-Fired Steam Generators for..., matrix material, clay, and other organic and inorganic material. Fossil fuel means natural gas, petroleum, coal, and any form of solid, liquid, or gaseous fuel derived from such materials for the purpose of...

40 CFR 60.41 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-07-01

... PERFORMANCE FOR NEW STATIONARY SOURCES Standards of Performance for Fossil-Fuel-Fired Steam Generators for..., matrix material, clay, and other organic and inorganic material. Fossil fuel means natural gas, petroleum, coal, and any form of solid, liquid, or gaseous fuel derived from such materials for the purpose of...

Screening of advanced cladding materials and UN-U3Si5 fuel

NASA Astrophysics Data System (ADS)

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

2015-07-01

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

Dynamic, Hot Surface Ignition of Aircraft Fuels and Hydraulic Fluids

DTIC Science & Technology

1980-10-01

fuels on a heated stainless steel surface. Higher local surface air speeds necessitated higher surface temperatures for ignition of an applied fluid._-7...Aircraft Fuels ( stainless steel surface) 8. Air Speed and Surface Material Effects on Hot Surface 21 Ignition Temperature of Aircraft Fuels (Titanium...Material Effects on Hot Surface 26 Ignition Temperature of Aircraft Hydraulic Fluids ( Stainless steel surface) 11. Air Speed and Surface Material

Mechanistic materials modeling for nuclear fuel performance

DOE PAGES

Tonks, Michael R.; Andersson, David; Phillpot, Simon R.; ...

2017-03-15

Fuel performance codes are critical tools for the design, certification, and safety analysis of nuclear reactors. However, their ability to predict fuel behavior under abnormal conditions is severely limited by their considerable reliance on empirical materials models correlated to burn-up (a measure of the number of fission events that have occurred, but not a unique measure of the history of the material). In this paper, we propose a different paradigm for fuel performance codes to employ mechanistic materials models that are based on the current state of the evolving microstructure rather than burn-up. In this approach, a series of statemore » variables are stored at material points and define the current state of the microstructure. The evolution of these state variables is defined by mechanistic models that are functions of fuel conditions and other state variables. The material properties of the fuel and cladding are determined from microstructure/property relationships that are functions of the state variables and the current fuel conditions. Multiscale modeling and simulation is being used in conjunction with experimental data to inform the development of these models. Finally, this mechanistic, microstructure-based approach has the potential to provide a more predictive fuel performance capability, but will require a team of researchers to complete the required development and to validate the approach.« less

Metals and Ceramics Division annual progress report, October 1, 1978-June 30, 1979

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

Peterson, S.

Research is reported concerning: (1) engineering materials including materials compatibility, mechanical properties, nondestructive testing, pressure vessel technology, and welding and brazing; (2) fuels and processes consisting of ceramic technology, fuel cycle technology, fuels evaluation, fuels fabrication and metals processing; and (3) materials science which includes, ceramic studies, physical metallurgy and properties, radiation effects and microstructural analysis, metastable and superconducting materials, structure and properties of surfaces, theoretical research, and x-ray research and applications. Highlights of the work of the metallographic group and the current status of the High-Temperature Materials Laboratory (HTML) and the Materials and Structures Technology Management Center (MSTMC) aremore » presented. (FS)« less

Deep desulfurization of hydrocarbon fuels

DOEpatents

Song, Chunshan [State College, PA; Ma, Xiaoliang [State College, PA; Sprague, Michael J [Calgary, CA; Subramani, Velu [State College, PA

2012-04-17

The invention relates to processes for reducing the sulfur content in hydrocarbon fuels such as gasoline, diesel fuel and jet fuel. The invention provides a method and materials for producing ultra low sulfur content transportation fuels for motor vehicles as well as for applications such as fuel cells. The materials and method of the invention may be used at ambient or elevated temperatures and at ambient or elevated pressures without the need for hydrogen.

Pretreatment microRNA Expression Impacting on Epithelial-to-Mesenchymal Transition Predicts Intrinsic Radiosensitivity in Head and Neck Cancer Cell Lines and Patients.

PubMed

de Jong, Monique C; Ten Hoeve, Jelle J; Grénman, Reidar; Wessels, Lodewyk F; Kerkhoven, Ron; Te Riele, Hein; van den Brekel, Michiel W M; Verheij, Marcel; Begg, Adrian C

2015-12-15

Predominant causes of head and neck cancer recurrence after radiotherapy are rapid repopulation, hypoxia, fraction of cancer stem cells, and intrinsic radioresistance. Currently, intrinsic radioresistance can only be assessed by ex vivo colony assays. Besides being time-consuming, colony assays do not identify causes of intrinsic resistance. We aimed to identify a biomarker for intrinsic radioresistance to be used before start of treatment and to reveal biologic processes that could be targeted to overcome intrinsic resistance. We analyzed both microRNA and mRNA expression in a large panel of head and neck squamous cell carcinoma (HNSCC) cell lines. Expression was measured on both irradiated and unirradiated samples. Results were validated using modified cell lines and a series of patients with laryngeal cancer. miRs, mRNAs, and gene sets that correlated with resistance could be identified from expression data of unirradiated cells. The presence of epithelial-to-mesenchymal transition (EMT) and low expression of miRs involved in the inhibition of EMT were important radioresistance determinants. This finding was validated in two independent cell line pairs, in which the induction of EMT reduced radiosensitivity. Moreover, low expression of the most important miR (miR-203) was shown to correlate with local disease recurrence after radiotherapy in a series of patients with laryngeal cancer. These findings indicate that EMT and low expression of EMT-inhibiting miRs, especially miR-203, measured in pretreatment material, causes intrinsic radioresistance of HNSCC, which could enable identification and treatment modification of radioresistant tumors. Clin Cancer Res; 21(24); 5630-8. ©2015 AACR. ©2015 American Association for Cancer Research.

Computational Modeling of Micrometastatic Breast Cancer Radiation Dose Response

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

Smith, Daniel L.; Debeb, Bisrat G.; Morgan Welch Inflammatory Breast Cancer Research Program and Clinic, The University of Texas MD Anderson Cancer Center, Houston, Texas

Purpose: Prophylactic cranial irradiation (PCI) involves giving radiation to the entire brain with the goals of reducing the incidence of brain metastasis and improving overall survival. Experimentally, we have demonstrated that PCI prevents brain metastases in a breast cancer mouse model. We developed a computational model to expand on and aid in the interpretation of our experimental results. Methods and Materials: MATLAB was used to develop a computational model of brain metastasis and PCI in mice. Model input parameters were optimized such that the model output would match the experimental number of metastases per mouse from the unirradiated group. Anmore » independent in vivo–limiting dilution experiment was performed to validate the model. The effect of whole brain irradiation at different measurement points after tumor cells were injected was evaluated in terms of the incidence, number of metastases, and tumor burden and was then compared with the corresponding experimental data. Results: In the optimized model, the correlation between the number of metastases per mouse and the experimental fits was >95. Our attempt to validate the model with a limiting dilution assay produced 99.9% correlation with respect to the incidence of metastases. The model accurately predicted the effect of whole-brain irradiation given 3 weeks after cell injection but substantially underestimated its effect when delivered 5 days after cell injection. The model further demonstrated that delaying whole-brain irradiation until the development of gross disease introduces a dose threshold that must be reached before a reduction in incidence can be realized. Conclusions: Our computational model of mouse brain metastasis and PCI correlated strongly with our experiments with unirradiated mice. The results further suggest that early treatment of subclinical disease is more effective than irradiating established disease.« less

EFFECTS OF IRRADIATION ON BRAIN VASCULATURE USING AN IN SITU TUMOR MODEL

PubMed Central

Zawaski, Janice A.; Gaber, M. Waleed; Sabek, Omaima M.; Wilson, Christy M.; Duntsch, Christopher D.; Merchant, Thomas E.

2013-01-01

Purpose Damage to normal tissue is a limiting factor in clinical radiotherapy (RT). We tested the hypothesis that the presence of tumor alters the response of normal tissues to irradiation using a rat in situ brain tumor model. Methods and Materials Intravital microscopy was used with a rat cranial window to assess the in situ effect of rat C6 glioma on peritumoral tissue with and without RT. The RT regimen included 40 Gy at 8 Gy/day starting Day 5 after tumor implant. Endpoints included blood–brain barrier permeability, clearance index, leukocyte-endothelial interactions and staining for vascular endothelial growth factor (VEGF) glial fibrillary acidic protein, and apoptosis. To characterize the system response to RT, animal survival and tumor surface area and volume were measured. Sham experiments were performed on similar animals implanted with basement membrane matrix absent of tumor cells. Results The presence of tumor alone increases permeability but has little effect on leukocyte–endothelial interactions and astrogliosis. Radiation alone increases tissue permeability, leukocyte-endothelial interactions, and astrogliosis. The highest levels of permeability and cell adhesion were seen in the model that combined tumor and irradiation; however, the presence of tumor appeared to reduce the volume of rolling leukocytes. Unirradiated tumor and peritumoral tissue had poor clearance. Irradiated tumor and peritumoral tissue had a similar clearance index to irradiated and unirradiated sham-implanted animals. Radiation reduces the presence of VEGF in peritumoral normal tissues but did not affect the amount of apoptosis in the normal tissue. Apoptosis was identified in the tumor tissue with and without radiation. Conclusions We developed a novel approach to demonstrate that the presence of the tumor in a rat intracranial model alters the response of normal tissues to irradiation. PMID:22197233

New Materials for Biological Fuel Cells

DTIC Science & Technology

2012-04-01

polymer electrolyte membrane ( PEM ), to the membrane-less biological fuel cell (center figure) where the two electrodes are submerged in the same... PEM . MT15_4p166_173.indd 171 4/10/2012 3:46:31 PM REVIEW New materials for biological fuel cells APRIL 2012 | VOLUME 15 | NUMBER 4172 These...ISSN:1369 7021 © Elsevier Ltd 2012APRIL 2012 | VOLUME 15 | NUMBER 4166 New materials for biological fuel cells Over the last decade, there has

Nanocrystalline cerium oxide materials for solid fuel cell systems

DOEpatents

Brinkman, Kyle S

2015-05-05

Disclosed are solid fuel cells, including solid oxide fuel cells and PEM fuel cells that include nanocrystalline cerium oxide materials as a component of the fuel cells. A solid oxide fuel cell can include nanocrystalline cerium oxide as a cathode component and microcrystalline cerium oxide as an electrolyte component, which can prevent mechanical failure and interdiffusion common in other fuel cells. A solid oxide fuel cell can also include nanocrystalline cerium oxide in the anode. A PEM fuel cell can include cerium oxide as a catalyst support in the cathode and optionally also in the anode.

Hydrocarbon-fuel/combustion-chamber-liner materials compatibility

NASA Technical Reports Server (NTRS)

Gage, Mark L.

1990-01-01

Results of material compatibility experiments using hydrocarbon fuels in contact with copper-based combustion chamber liner materials are presented. Mil-Spec RP-1, n- dodecane, propane, and methane fuels were tested in contact with OFHC, NASA-Z, and ZrCu coppers. Two distinct test methods were employed. Static tests, in which copper coupons were exposed to fuel for long durations at constant temperature and pressure, provided compatibility data in a precisely controlled environment. Dynamic tests, using the Aerojet Carbothermal Test Facility, provided fuel and copper compatibility data under realistic booster engine service conditions. Tests were conducted using very pure grades of each fuel and fuels to which a contaminant, e.g., ethylene or methyl mercaptan, was added to define the role played by fuel impurities. Conclusions are reached as to degradation mechanisms and effects, methods for the elimination of these mechanisms, selection of copper alloy combustion chamber liners, and hydrocarbon fuel purchase specifications.

75 FR 44817 - Notice of Availability of Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-29

... Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services, National... Enrichment Branch, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear Material Safety and... Enrichment Branch, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear Material Safety and...

SOLID SOLUTION CARBIDES ARE THE KEY FUELS FOR FUTURE NUCLEAR THERMAL PROPULSION

NASA Technical Reports Server (NTRS)

Panda, Binayak; Hickman, Robert R.; Shah, Sandeep

2005-01-01

Nuclear thermal propulsion uses nuclear energy to directly heat a propellant (such as liquid hydrogen) to generate thrust for space transportation. In the 1960 s, the early Rover/Nuclear Engine for Rocket Propulsion Application (NERVA) program showed very encouraging test results for space nuclear propulsion but, in recent years, fuel research has been dismal. With NASA s renewed interest in long-term space exploration, fuel researchers are now revisiting the RoverMERVA findings, which indicated several problems with such fuels (such as erosion, chemical reaction of the fuel with propellant, fuel cracking, and cladding issues) that must be addressed. It is also well known that the higher the temperature reached by a propellant, the larger the thrust generated from the same weight of propellant. Better use of fuel and propellant requires development of fuels capable of reaching very high temperatures. Carbides have the highest melting points of any known material. Efforts are underway to develop carbide mixtures and solid solutions that contain uranium carbide, in order to achieve very high fuel temperatures. Binary solid solution carbides (U, Zr)C have proven to be very effective in this regard. Ternary carbides such as (U, Zr, X) carbides (where X represents Nb, Ta, W, and Hf) also hold great promise as fuel material, since the carbide mixtures in solid solution generate a very hard and tough compact material. This paper highlights past experience with early fuel materials and bi-carbides, technical problems associated with consolidation of the ingredients, and current techniques being developed to consolidate ternary carbides as fuel materials.

Method and device for fabricating dispersion fuel comprising fission product collection spaces

DOEpatents

Shaber, Eric L; Fielding, Randall S

2015-05-05

A method of fabricating a nuclear fuel comprising a fissile material, one or more hollow microballoons, a phenolic resin, and metal matrix. The fissile material, phenolic resin and the one or more hollow microballoons are combined. The combined fissile material, phenolic resin and the hollow microballoons are heated sufficiently to form at least some fissile material carbides creating a nuclear fuel particle. The resulting nuclear fuel particle comprises one or more fission product collection spaces. In a preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by forming the fissile material into microspheres. The fissile material microspheres are then overcoated with the phenolic resin and microballoon. In another preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by overcoating the microballoon with the fissile material, and phenolic resin.

In situ high-energy X-ray diffraction study of tensile deformation of neutron-irradiated polycrystalline Fe-9%Cr alloy

DOE PAGES

Zhang, Xuan; Li, Meimei; Park, Jun -Sang; ...

2016-12-30

The effect of neutron irradiation on tensile deformation of a Fe-9wt.%Cr alloy was investigated using in situ high-energy synchrotron X-ray diffraction during room-temperature uniaxial tensile tests. New insights into the deformation mechanisms were obtained through the measurements of lattice strain evolution and the analysis of diffraction peak broadening using the modified Williamson-Hall method. Two neutron-irradiated specimens, one irradiated at 300 °C to 0.01 dpa and the other at 450 °C to 0.01dpa, were tested along with an unirradiated specimen. The macroscopic stress–strain curves of the irradiated specimens showed increased strength, reduced ductility and work-hardening exponent compared to the unirradiated specimen.more » The evolutions of the lattice strain, the dislocation density and the coherent scattering domain size in the deformation process revealed different roles of the submicroscopic defects in the 300°C/0.01 dpa specimen and the TEM-visible nanometer-sized dislocation loops in the 450°C/0.01 dpa specimen: submicroscopic defects extended the linear work hardening stage (stage II) to a higher strain, while irradiation-induced dislocation loops were more effective in dislocation pinning. Lastly, while the work hardening rate of stage II was unaffected by irradiation, significant dynamic recovery in stage III in the irradiated specimens led to the early onset of necking without stage IV as observed in the unirradiated specimen.« less

Modification of radiation carcinogenesis by marihuana

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

Montour, J.L.; Dutz, W.; Harris, L.S.

1981-03-15

Male, female, and ovariectomized female Sprague-Dawley rats were irradiated with 400 rads, 150 rads, or 300 rads, respectively, of /sup 60/Co gamma rays when they were between 40 and 50 days of age. The animals were injected three times weekly with either marihuana extract or with alcohol-emulphor carrier. Comparable unirradiated groups were similarly injected. Mean survival time in males was significantly shorter in the 400 rad + marihuana group compared with the three other groups whose mean survival times did not differ. Through the 546 days that the males were observed, the total number of tumors other than fibrosarcomas wasmore » significantly greater following radiation and marihuana (22) than radiation alone (6). Fifteen of the tumors were of breast or endocrine tissues. No differences were seen in the unirradiated groups. In the females, which were observed for 635 days, the total number of breast tumors was greater with the combined treatment (38) compared with radiation alone (22). This was entirely due to a marked difference in the adenocarcinoma incidence, which was 21 (radiation + marihuana) compared with four (radiation alone). The number of adenofibromas was similar in the two groups. In the unirradiated female groups the breast adenocarcinoma incidence was eight in the marihuana group and two in the control group. Ovariectomy resulted in a lower breast tumor incidence in all groups. Nonbreast tumors were more frequent in the ovariectomized-irradiated groups. Radiation plus marihuana produced more nonbreast tumors (25) than radiation alone (17) in the ovariectomized females.« less

Modification of radiation carcinogenesis by marijuana

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

Montour, J.L.; Dutz, W.; Harris, L.S.

Male, female, and ovariectomized female Sprague-Dawley rats were irradiated with 400 rads, 150 rads, or 300 rads, respectively, of /sup 60/Co gamma rays when they were between 40 and 50 days of age. The animals were injected three times weekly with either marihuana extract or with alcohol-emulphor carrier. Comparable unirradiated groups were similarly injected. Mean survival time in males was significantly shorter in the 400 rad + marihuana group compared with the three other groups whose mean survival times did not differ. Through the 546 days that the males were observed, the total number of tumors other than fibrosarcomas wasmore » significantly greater following radiation and marihuana (22) than radiation alone (6). Fifteen of the tumors were of breast or endocrine tissues. No differences were seen in the unirradiated groups. In the females, which were observed for 635 days, the total number of breast tumors was greater with the combined treatment (38) compared with radiation alone (22). This was entirely due to a marked difference in the adenocarcinoma incidence, which was 21 (radiation + marihuana) compared with four (radiation alone). The number of adenofibromas was similar in the two groups. In the unirradiated female groups the breast adenocarcinoma incidence was eight in the marihuana group and two in the control group. Ovariectomy resulted in a lower breast tumor incidence in all groups. Nonbreast tumors were more frequent in the ovariectomized-irradiated groups. Radiation plus marihuana produced more nonbreast tumors (25) than radiation alone (17) in the ovariectomized females.« less

In situ measurements of a homogeneous to heterogeneous transition in the plastic response of ion-irradiated <111> Ni microspecimens

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

Zhao, Xinyu; Strickland, Daniel J.; Derlet, Peter M.

We report on the use of quantitative in situ microcompression experiments in a scanning electron microscope to systematically investigate the effect of self-ion irradiation damage on the full plastic response of <111> Ni. In addition to the well-known irradiationinduced increases in the yield and flow strengths with increasing dose, we measure substantial changes in plastic flow intermittency behavior, manifested as stress drops accompanying energy releases as the driven material transits critical states. At low irradiation doses, the magnitude of stress drops reduces relative to the unirradiated material and plastic slip proceeds on multiple slip systems, leading to quasi-homogeneous plastic flow.more » In contrast, highly irradiated specimens exhibit pronounced shear localization on parallel slip planes, which we ascribe to the onset of defect free channels normally seen in bulk irradiated materials. Our in situ testing system and approach allows for a quantitative study of the energy release and dynamics associated with defect free channel formation and subsequent localization. As a result, this study provides fundamental insight to the nature of interactions between mobile dislocations and irradiation-mediated and damage-dependent defect structures.« less

In situ measurements of a homogeneous to heterogeneous transition in the plastic response of ion-irradiated <111> Ni microspecimens

DOE PAGES

Zhao, Xinyu; Strickland, Daniel J.; Derlet, Peter M.; ...

2015-02-11

We report on the use of quantitative in situ microcompression experiments in a scanning electron microscope to systematically investigate the effect of self-ion irradiation damage on the full plastic response of <111> Ni. In addition to the well-known irradiationinduced increases in the yield and flow strengths with increasing dose, we measure substantial changes in plastic flow intermittency behavior, manifested as stress drops accompanying energy releases as the driven material transits critical states. At low irradiation doses, the magnitude of stress drops reduces relative to the unirradiated material and plastic slip proceeds on multiple slip systems, leading to quasi-homogeneous plastic flow.more » In contrast, highly irradiated specimens exhibit pronounced shear localization on parallel slip planes, which we ascribe to the onset of defect free channels normally seen in bulk irradiated materials. Our in situ testing system and approach allows for a quantitative study of the energy release and dynamics associated with defect free channel formation and subsequent localization. As a result, this study provides fundamental insight to the nature of interactions between mobile dislocations and irradiation-mediated and damage-dependent defect structures.« less

A Study of the Oxidation Behaviour of Pile Grade A (PGA) Nuclear Graphite Using Thermogravimetric Analysis (TGA), Scanning Electron Microscopy (SEM) and X-Ray Tomography (XRT).

PubMed

Payne, Liam; Heard, Peter J; Scott, Thomas B

2015-01-01

Pile grade A (PGA) graphite was used as a material for moderating and reflecting neutrons in the UK's first generation Magnox nuclear power reactors. As all but one of these reactors are now shut down there is a need to understand the residual state of the material prior to decommissioning of the cores, in particular the location and concentration of key radio-contaminants such as 14C. The oxidation behaviour of unirradiated PGA graphite was studied, in the temperature range 600-1050°C, in air and nitrogen using thermogravimetric analysis, scanning electron microscopy and X-ray tomography to investigate the possibility of using thermal degradation techniques to examine 14C distribution within irradiated material. The thermal decomposition of PGA graphite was observed to follow the three oxidation regimes historically identified by previous workers with limited, uniform oxidation at temperatures below 600°C and substantial, external oxidation at higher temperatures. This work demonstrates that the different oxidation regimes of PGA graphite could be developed into a methodology to characterise the distribution and concentration of 14C in irradiated graphite by thermal treatment.

A Study of the Oxidation Behaviour of Pile Grade A (PGA) Nuclear Graphite Using Thermogravimetric Analysis (TGA), Scanning Electron Microscopy (SEM) and X-Ray Tomography (XRT)

PubMed Central

Payne, Liam; Heard, Peter J.; Scott, Thomas B.

2015-01-01

Pile grade A (PGA) graphite was used as a material for moderating and reflecting neutrons in the UK’s first generation Magnox nuclear power reactors. As all but one of these reactors are now shut down there is a need to understand the residual state of the material prior to decommissioning of the cores, in particular the location and concentration of key radio-contaminants such as 14C. The oxidation behaviour of unirradiated PGA graphite was studied, in the temperature range 600–1050°C, in air and nitrogen using thermogravimetric analysis, scanning electron microscopy and X-ray tomography to investigate the possibility of using thermal degradation techniques to examine 14C distribution within irradiated material. The thermal decomposition of PGA graphite was observed to follow the three oxidation regimes historically identified by previous workers with limited, uniform oxidation at temperatures below 600°C and substantial, external oxidation at higher temperatures. This work demonstrates that the different oxidation regimes of PGA graphite could be developed into a methodology to characterise the distribution and concentration of 14C in irradiated graphite by thermal treatment. PMID:26575374

Multidimensional Fuel Performance Code: BISON

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

BISON is a finite element based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO fuel particles, and metallic rod and plate fuel (Refs. [a, b, c]). It solves the fully-coupled equations of thermomechanics and species diffusion and includes important fuel physics such as fission gas release and material property degradation with burnup. BISON is based on the MOOSE framework (Ref. [d]) and can therefore efficiently solve problems on 1-, 2- or 3-D meshes using standard workstations or large high performance computers. BISON is also coupled to a MOOSE-based mesoscale phasemore » field material property simulation capability (Refs. [e, f]). As described here, BISON includes the code library named FOX, which was developed concurrent with BISON. FOX contains material and behavioral models that are specific to oxide fuels.« less

Affordable Development and Optimization of CERMET Fuels for NTP Ground Testing

NASA Technical Reports Server (NTRS)

Hickman, Robert R.; Broadway, Jeramie W.; Mireles, Omar R.

2014-01-01

CERMET fuel materials for Nuclear Thermal Propulsion (NTP) are currently being developed at NASA's Marshall Space Flight Center. The work is part of NASA's Advanced Space Exploration Systems Nuclear Cryogenic Propulsion Stage (NCPS) Project. The goal of the FY12-14 project is to address critical NTP technology challenges and programmatic issues to establish confidence in the affordability and viability of an NTP system. A key enabling technology for an NCPS system is the fabrication of a stable high temperature nuclear fuel form. Although much of the technology was demonstrated during previous programs, there are currently no qualified fuel materials or processes. The work at MSFC is focused on developing critical materials and process technologies for manufacturing robust, full-scale CERMET fuels. Prototypical samples are being fabricated and tested in flowing hot hydrogen to understand processing and performance relationships. As part of this initial demonstration task, a final full scale element test will be performed to validate robust designs. The next phase of the project will focus on continued development and optimization of the fuel materials to enable future ground testing. The purpose of this paper is to provide a detailed overview of the CERMET fuel materials development plan. The overall CERMET fuel development path is shown in Figure 2. The activities begin prior to ATP for a ground reactor or engine system test and include materials and process optimization, hot hydrogen screening, material property testing, and irradiation testing. The goal of the development is to increase the maturity of the fuel form and reduce risk. One of the main accomplishmens of the current AES FY12-14 project was to develop dedicated laboratories at MSFC for the fabrication and testing of full length fuel elements. This capability will enable affordable, near term development and optimization of the CERMET fuels for future ground testing. Figure 2 provides a timeline of the development and optimization tasks for the AES FY15-17 follow on program.

System for determining biofuel concentration

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

Huff, Shean P.; Janke, Christopher James; Kass, Michael D.

2016-09-13

A measurement device or system configured to measure the content of biofuels within a fuel blend. By measuring a state of a responsive material within a fuel blend, a biofuel content of the fuel blend may be measured. For example, the solubility of a responsive material to biofuel content within a fuel blend, may affect a property of the responsive material, such as shape, dimensional size, or electrical impedance, which may be measured and used as a basis for determining biofuel content.

Stuck fuel rod capping sleeve

DOEpatents

Gorscak, Donald A.; Maringo, John J.; Nilsen, Roy J.

1988-01-01

A stuck fuel rod capping sleeve to be used during derodding of spent fuel assemblies if a fuel rod becomes stuck in a partially withdrawn position and, thus, has to be severed. The capping sleeve has an inner sleeve made of a lower work hardening highly ductile material (e.g., Inconel 600) and an outer sleeve made of a moderately ductile material (e.g., 304 stainless steel). The inner sleeve may be made of an epoxy filler. The capping sleeve is placed on a fuel rod which is then severed by using a bolt cutter device. Upon cutting, the capping sleeve deforms in such a manner as to prevent the gross release of radioactive fuel material

PEM fuel cell bipolar plate material requirements for transportation applications

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

Borup, R.L.; Stroh, K.R.; Vanderborgh, N.E.

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.

Renewable synthetic diesel fuel from triglycerides and organic waste materials

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

Hillard, J.C.; Strassburger, R.S.

1986-03-01

A renewable, synthetic diesel fuel has been developed that employs ethanol and organic waste materials. These organic materials, such as soybean oil or animal fats, are hydrolized to yield a mixture of solid soap like materials and glycerol. These soaps, now soluble in ethanol, are blended with ethanol; the glycerol is nitrated and added as well as castor oil when necessary. The synthetic fuel is tailored to match petroleum diesel fuel in viscosity, lubricity and cetane quality and, therefore, does not require any engine modifications. Testing in a laboratory engine and in a production Oldsmobile Cutlass has revealed that thismore » synthetic fuel is superior to petroleum diesel fuel in vehicle efficiency, cetane quality, combustion noise, cold start characteristics, exhaust odor and emissions. Performance characteristics are indistinguishable from those of petroleum diesel fuel. These soaps are added to improve the calorific value, lubricity and cetane quality of the ethanol. The glycerol from the hydrolysis process is nitrated and added to the ethanol as an additional cetane quality improver. Caster oil is added to the fuel when necessary to match the viscosity and lubricity of petroleum diesel fuel as well as to act as a corrosion inhibitor, thereby, precluding any engine modifications. The cetane quality of the synthetic fuel is better than that of petroleum diesel as the fuel carries its own oxygen. The synthetic fuel is also completely miscible with petroleum diesel.« less

Molten carbonate fuel cell separator

DOEpatents

Nickols, Richard C.

1986-09-02

In a stacked array of molten carbonate fuel cells, a fuel cell separator is positioned between adjacent fuel cells to provide isolation as well as a conductive path therebetween. The center portion of the fuel cell separator includes a generally rectangular, flat, electrical conductor. Around the periphery of the flat portion of the separator are positioned a plurality of elongated resilient flanges which form a gas-tight seal around the edges of the fuel cell. With one elongated flange resiliently engaging a respective edge of the center portion of the separator, the sealing flanges, which are preferably comprised of a noncorrosive material such as an alloy of yttrium, iron, aluminum or chromium, form a tight-fitting wet seal for confining the corrosive elements of the fuel cell therein. This arrangement permits a good conductive material which may be highly subject to corrosion and dissolution to be used in combination with a corrosion-resistant material in the fuel cell separator of a molten carbonate fuel cell for improved fuel cell conductivity and a gas-tight wet seal.

Molten carbonate fuel cell separator

DOEpatents

Nickols, R.C.

1984-10-17

In a stacked array of molten carbonate fuel cells, a fuel cell separator is positioned between adjacent fuel cells to provide isolation as well as a conductive path therebetween. The center portion of the fuel cell separator includes a generally rectangular, flat, electrical conductor. Around the periphery of the flat portion of the separator are positioned a plurality of elongated resilient flanges which form a gas-tight seal around the edges of the fuel cell. With one elongated flange resiliently engaging a respective edge of the center portion of the separator, the sealing flanges, which are preferably comprised of a noncorrosive material such as an alloy of yttrium, iron, aluminum or chromium, form a tight-fitting wet seal for confining the corrosive elements of the fuel cell therein. This arrangement permits a good conductive material which may be highly subject to corrosion and dissolution to be used in combination with a corrosion-resistant material in the fuel cell separator of a molten carbonate fuel cell for improved fuel cell conductivity and a gas-tight wet seal.

Hydrogen storage and integrated fuel cell assembly

DOEpatents

Gross, Karl J.

2010-08-24

Hydrogen is stored in materials that absorb and desorb hydrogen with temperature dependent rates. A housing is provided that allows for the storage of one or more types of hydrogen-storage materials in close thermal proximity to a fuel cell stack. This arrangement, which includes alternating fuel cell stack and hydrogen-storage units, allows for close thermal matching of the hydrogen storage material and the fuel cell stack. Also, the present invention allows for tailoring of the hydrogen delivery by mixing different materials in one unit. Thermal insulation alternatively allows for a highly efficient unit. Individual power modules including one fuel cell stack surrounded by a pair of hydrogen-storage units allows for distribution of power throughout a vehicle or other electric power consuming devices.

Nuclear reactor fuel element

DOEpatents

Johnson, Carl E.; Crouthamel, Carl E.

1980-01-01

A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of oxygen gettering material on the inner surface of the cladding. The gettering material reacts with oxygen released by the fissionable material during irradiation of the core thereby preventing the oxygen from reacting with and corroding the cladding. Also described is an improved method for coating the inner surface of the cladding with a layer of gettering material.

An analysis of international nuclear fuel supply options

NASA Astrophysics Data System (ADS)

Taylor, J'tia Patrice

As the global demand for energy grows, many nations are considering developing or increasing nuclear capacity as a viable, long-term power source. To assess the possible expansion of nuclear power and the intricate relationships---which cover the range of economics, security, and material supply and demand---between established and aspirant nuclear generating entities requires models and system analysis tools that integrate all aspects of the nuclear enterprise. Computational tools and methods now exist across diverse research areas, such as operations research and nuclear engineering, to develop such a tool. This dissertation aims to develop methodologies and employ and expand on existing sources to develop a multipurpose tool to analyze international nuclear fuel supply options. The dissertation is comprised of two distinct components: the development of the Material, Economics, and Proliferation Assessment Tool (MEPAT), and analysis of fuel cycle scenarios using the tool. Development of MEPAT is aimed for unrestricted distribution and therefore uses publicly available and open-source codes in its development when possible. MEPAT is built using the Powersim Studio platform that is widely used in systems analysis. MEPAT development is divided into three modules focusing on: material movement; nonproliferation; and economics. The material movement module tracks material quantity in each process of the fuel cycle and in each nuclear program with respect to ownership, location and composition. The material movement module builds on techniques employed by fuel cycle models such as the Verifiable Fuel Cycle Simulation (VISION) code developed at the Idaho National Laboratory under the Advanced Fuel Cycle Initiative (AFCI) for the analysis of domestic fuel cycle. Material movement parameters such as lending and reactor preference, as well as fuel cycle parameters such as process times and material factors are user-specified through a Microsoft Excel(c) data spreadsheet. The material movement module is the largest of the three, and the two other modules that assess nonproliferation and economics of the options are dependent on its output. Proliferation resistance measures from literature are modified and incorporated in MEPAT. The module to assess the nonproliferation of the supply options allows the user to specify defining attributes for the fuel cycle processes, and determines significant quantities of materials as well as measures of proliferation resistance. The measure is dependent on user-input and material information. The economics module allows the user to specify costs associated with different processes and other aspects of the fuel cycle. The simulation tool then calculates economic measures that relate the cost of the fuel cycle to electricity production. The second part of this dissertation consists of an examination of four scenarios of fuel supply option using MEPAT. The first is a simple scenario illustrating the modules and basic functions of MEPAT. The second scenario recreates a fuel supply study reported earlier in literature, and compares MEPAT results with those reported earlier for validation. The third, and a rather realistic, scenario includes four nuclear programs with one program entering the nuclear energy market. The fourth scenario assesses the reactor options available to the Hashemite Kingdom of Jordan, which is currently assessing available options to introduce nuclear power in the country. The methodology developed and implemented in MEPAT to analyze the material, proliferation and economics of nuclear fuel supply options is expected to help simplify and assess different reactor and fuel options available to utilities, government agencies and international organizations.

High performance, high durability non-precious metal fuel cell catalysts

DOEpatents

Wood, Thomas E.; Atanasoski, Radoslav; Schmoeckel, Alison K.

2016-03-15

This invention relates to non-precious metal fuel cell cathode catalysts, fuel cells that contain these catalysts, and methods of making the same. The fuel cell cathode catalysts are highly nitrogenated carbon materials that can contain a transition metal. The highly nitrogenated carbon materials can be supported on a nanoparticle substrate.

Nuclear fuel elements and method of making same

DOEpatents

Schweitzer, Donald G.

1992-01-01

A nuclear fuel element for a high temperature gas nuclear reactor that has an average operating temperature in excess of 2000.degree. C., and a method of making such a fuel element. The fuel element is characterized by having fissionable fuel material localized and stabilized within pores of a carbon or graphite member by melting the fissionable material to cause it to chemically react with the carbon walls of the pores. The fissionable fuel material is further stabilized and localized within the pores of the graphite member by providing one or more coatings of pyrolytic carbon or diamond surrounding the porous graphite member so that each layer defines a successive barrier against migration of the fissionable fuel from the pores, and so that the outermost layer of pyrolytic carbon or diamond forms a barrier between the fissionable material and the moderating gases used in an associated high temperature gas reactor. The method of the invention provides for making such new elements either as generally spherically elements, or as flexible filaments, or as other relatively small-sized fuel elements that are particularly suited for use in high temperature gas reactors.

78 FR 71532 - Amendments to Material Control and Accounting Regulations and Proposed Guidance for Fuel Cycle...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-29

... Accounting Regulations and Proposed Guidance for Fuel Cycle Facility Material Control and Accounting Plans... material control and accounting (MC&A) of special nuclear material (SNM) and the proposed guidance...

Corrosion of graphite composites in phosphoric acid fuel cells

NASA Technical Reports Server (NTRS)

Christner, L. G.; Dhar, H. P.; Farooque, M.; Kush, A. K.

1986-01-01

Polymers, polymer-graphite composites and different carbon materials are being considered for many of the fuel cell stack components. Exposure to concentrated phosphoric acid in the fuel cell environment and to high anodic potential results in corrosion. Relative corrosion rates of these materials, failure modes, plausible mechanisms of corrosion and methods for improvement of these materials are investigated.

40 CFR 241.3 - Standards and procedures for identification of non-hazardous secondary materials that are solid...

Code of Federal Regulations, 2011 CFR

2011-07-01

... identification of non-hazardous secondary materials that are solid wastes when used as fuels or ingredients in...) SOLID WASTES SOLID WASTES USED AS FUELS OR INGREDIENTS IN COMBUSTION UNITS Identification of Non-Hazardous Secondary Materials That Are Solid Wastes When Used as Fuels or Ingredients in Combustion Units...

Space Reflector Materials for Prometheus Application

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

J. Nash; V. Munne; LL Stimely

2006-01-31

The two materials studied in depth which appear to have the most promise in a Prometheus reflector application are beryllium (Be) and beryllium oxide (BeO). Three additional materials, magnesium oxide (MgO), alumina (Al{sub 2}O{sub 3}), and magnesium aluminate spinel (MgAl{sub 2}O{sub 4}) were also recently identified to be of potential interest, and may have promise in a Prometheus application as well, but are expected to be somewhat higher mass than either a Be or BeO based reflector. Literature review and analysis indicates that material properties for Be are largely known, but there are gaps in the properties of Be0 relativemore » to the operating conditions for a Prometheus application. A detailed preconceptual design information document was issued providing material properties for both materials (Reference (a)). Beryllium oxide specimens were planned to be irradiated in the JOY0 Japanese test reactor to partially fill the material property gaps, but more testing in the High Flux Isotope Reactor (HFIR) test reactor at Oak Ridge National Laboratory (ORNL) was expected to be needed. A key issue identified for BeO was obtaining material for irradiation testing with an average grain size of {approx}5 micrometers, reminiscent of material for which prior irradiation test results were promising. Current commercially available material has an average grain size of {approx}10 micrometers. The literature indicated that improved irradiation performance could be expected (e.g., reduced irradiation-induced swelling) with the finer grain size material. Confirmation of these results would allow the use of historic irradiated materials test results from the literature, reducing the extent of required testing and therefore the cost of using this material. Environmental, safety and health (ES&H) concerns associated with manufacturing are significant but manageable for Be and BeO. Although particulate-generating operations (e.g., machining, grinding, etc.) involving Be-bearing materials require significant controls, handling of clean, finished products requires only modest controls. Neither material was initially considered to be viable as a structural material, however, based on improved understanding of its unirradiated properties, Be should be evaluated due to having potentially acceptable structural properties in the unirradiated condition, i. e., during launch, when loads might be most limiting. All three of the alternative materials are non-hazardous, and thus do not engender the ES&H concerns associated with use of Be or BeO. Aluminum oxide is a widely available ceramic material with well characterized physical properties and well developed processing practices. Although the densest (3.97 g/cm{sup 3} versus Be: 1.85, BeO: 3.01, MgO: 3.58, and MgAl{sub 2}O{sub 4}: 3.60, all theoretical density), and therefore the heaviest, of all the materials considered for this application, its ease of fabrication, mechanical properties, availability and neutronic characteristics warrant its evaluation. Similarly, MgO is widely used in the refractory materials industry and has a large established manufacturing base while being lighter than Al{sub 2}O{sub 3}. Most of the commercially available MgO products incorporate additives or a second phase to avoid the formation of Mg(OH){sub 2} due to spontaneous reaction with ambient humidity. The hygroscopicity of MgO makes it a more difficult material to work with than Al{sub 2}O{sub 3} or MgAl{sub 2}O{sub 4}. Magnesium aluminate spinel, although not as widely available as either Al{sub 2}O{sub 3} or MgO, has the advantage of a density almost as low as MgO without being hygroscopic, and shares comparable neutronic performance characteristics in the reflector application.« less

Fuels from Biomass: Integration with Food and Materials Systems

ERIC Educational Resources Information Center

Lipinsky, E. S.

1978-01-01

The development of fuels from biomass can lead naturally to dispersed facilities that incorporate food or materials production (or both) with fuel production. The author analyzes possible systems based on sugarcane, corn, and guayule. (Author/MA)

Fuel cell generator with fuel electrodes that control on-cell fuel reformation

DOEpatents

Ruka, Roswell J [Pittsburgh, PA; Basel, Richard A [Pittsburgh, PA; Zhang, Gong [Murrysville, PA

2011-10-25

A fuel cell for a fuel cell generator including a housing including a gas flow path for receiving a fuel from a fuel source and directing the fuel across the fuel cell. The fuel cell includes an elongate member including opposing first and second ends and defining an interior cathode portion and an exterior anode portion. The interior cathode portion includes an electrode in contact with an oxidant flow path. The exterior anode portion includes an electrode in contact with the fuel in the gas flow path. The anode portion includes a catalyst material for effecting fuel reformation along the fuel cell between the opposing ends. A fuel reformation control layer is applied over the catalyst material for reducing a rate of fuel reformation on the fuel cell. The control layer effects a variable reformation rate along the length of the fuel cell.

Permeation of gasoline, diesel, bioethanol (E85), and biodiesel (B20) fuels through six glove materials.

PubMed

Chin, Jo-Yu; Batterman, Stuart A

2010-07-01

Biofuels and conventional fuels differ in terms of their evaporation rates, permeation rates, and exhaust emissions, which can alter exposures of workers, especially those in the fuel refining and distribution industries. This study investigated the permeation of biofuels (bioethanol 85%, biodiesel 20%) and conventional petroleum fuels (gasoline and diesel) through gloves used in occupational settings (neoprene, nitrile, and Viton) and laboratories (latex, nitrile, and vinyl), as well as a standard reference material (neoprene sheet). Permeation rates and breakthrough times were measured using the American Society for Testing and Materials F739-99 protocol, and fuel and permeant compositions were measured by gas chromatography/mass spectrometry. In addition, we estimated exposures for three occupational scenarios and recommend chemical protective clothing suitable for use with motor fuels. Permeation rates and breakthrough times depended on the fuel-glove combination. Gasoline had the highest permeation rate among the four fuels. Bioethanol (85%) had breakthrough times that were two to three times longer than gasoline through neoprene, nitrile Sol-Vex, and the standard reference materials. Breakthrough times for biodiesel (20%) were slightly shorter than for diesel for the latex, vinyl, nitrile examination, and the standard neoprene materials. The composition of permeants differed from neat fuels, e.g., permeants were significantly enriched in the lighter aromatics including benzene. Viton was the best choice among the tested materials for the four fuels tested. Among the scenarios, fuel truck drivers had the highest uptake via inhalation based on the personal measurements available in the literature, and gasoline station attendants had highest uptake via dermal exposure if gloves were not worn. Appropriate selection and use of gloves can protect workers from dermal exposures; however, current recommendations from the National Institute for Occupational Safety and Health should be revised to account for contemporary fuel formulations that routinely contain ethanol.

Re-weldability tests of irradiated 316L(N) stainless steel using laser welding technique

NASA Astrophysics Data System (ADS)

Yamada, Hirokazu; Kawamura, Hiroshi; Tsuchiya, Kunihiko; Kalinin, George; Kohno, Wataru; Morishima, Yasuo

2002-12-01

SS316L(N)-IG is the candidate material for the in-vessel and ex-vessel components of fusion reactors such as ITER (International Thermonuclear Experimental Reactor). This paper describes a study on re-weldability of un-irradiated and/or irradiated SS316L(N)-IG and the effect of helium generation on the mechanical properties of the weld joint. The laser welding process is used for re-welding of the water cooling branch pipeline repairs. It is clarified that re-welding of SS316L(N)-IG irradiated up to about 0.2 dpa (3.3 appm He) can be carried out without a serious deterioration of tensile properties due to helium accumulation. Therefore, repair of the ITER blanket cooling pipes can be performed by the laser welding process.

Current conducting end plate of fuel cell assembly

DOEpatents

Walsh, Michael M.

1999-01-01

A fuel cell assembly has a current conducting end plate with a conductive body formed integrally with isolating material. The conductive body has a first surface, a second surface opposite the first surface, and an electrical connector. The first surface has an exposed portion for conducting current between a working section of the fuel cell assembly and the electrical connector. The isolating material is positioned on at least a portion of the second surface. The conductive body can have support passage(s) extending therethrough for receiving structural member(s) of the fuel cell assembly. Isolating material can electrically isolate the conductive body from the structural member(s). The conductive body can have service passage(s) extending therethrough for servicing one or more fluids for the fuel cell assembly. Isolating material can chemically isolate the one or more fluids from the conductive body. The isolating material can also electrically isolate the conductive body from the one or more fluids.

Partial replacement of non renewable fossil fuels energy by the use of waste materials as alternative fuels

NASA Astrophysics Data System (ADS)

Indrawati, V.; Manaf, A.; Purwadi, G.

2009-09-01

This paper reports recent investigations on the use of biomass like rice husk, palm kernel shell, saw dust and municipal waste to reduce the use of fossil fuels energy in the cement production. Such waste materials have heat values in the range approximately from 2,000 to 4,000 kcal/kg. These are comparable to the average value of 5800 kcal/kg from fossil materials like coals which are widely applied in many industrial processing. Hence, such waste materials could be used as alternative fuels replacing the fossil one. It is shown that replacement of coals with such waste materials has a significant impact on cost effectiveness as well as sustainable development. Variation in moisture content of the waste materials, however should be taken into account because this is one of the parameter that could not be controlled. During fuel combustion, some amount of the total energy is used to evaporate the water content and thus the net effective heat value is less.

An assessment of the attractiveness of material associated with thorium/uranium and uranium closed fuel cycles from a safeguards perspective

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

Bathke, Charles Gary; Wallace, Richard K; Hase, Kevin R

2010-01-01

This paper reports the continued evaluation of the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with various proposed nuclear fuel cycles. Specifically, this paper examines two closed fuel cycles. The first fuel cycle examined is a thorium fuel cycle in which a pressurized heavy water reactor (PHWR) is fueled with mixtures of plutonium/thorium and {sup 233}U/thorium. The used fuel is then reprocessed using the THOREX process and the actinides are recycled. The second fuel cycle examined consists of conventional light water reactors (LWR) whose fuel is reprocessed for actinides that are then fed to and recycled untilmore » consumed in fast-spectrum reactors: fast reactors and accelerator driven systems (ADS). As reprocessing of LWR fuel has already been examined, this paper will focus on the reprocessing of the scheme's fast-spectrum reactors' fuel. This study will indicate what is required to render these materials as having low utility for use in nuclear weapons. Nevertheless, the results of this paper suggest that all reprocessing products evaluated so far need to be rigorously safeguarded and provided high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE). The methodology and key findings will be presented.« less

A Rapid Survey of the Compatibility of Selected Seal Materials with Conventional and Semi-Synthetic JP-8

NASA Technical Reports Server (NTRS)

Graham, John L.; Striebich, Richard C.; Minus, Donald K.; Harrison, William E., III

2007-01-01

Since the synthesis of a liquid hydrocarbon fuel from coal by Franz Fischer and Hans Tropsch in 1923, there has been cyclic interest in developing this fuel for military and commercial applications. In recent years the U.S. Department of Defense has taken interest in producing a unified battlespace fuel using the Fischer Tropsch (FT) process for a variety of reasons including cost, quality, and logistics. In the past year there has been a particular emphasis on moving quickly to demonstrate that an FT fuel can be used in the form of a blend with conventional petroleum-derived jet fuel. The initial objective is to employ this semi-synthetic fuel with blend ratios as high as 50 percent FT with longer range goals to use even high blend ratios and ultimately a fully synthetic jet fuel. A significant concern associated with the use of a semi-synthetic jet fuel with high FT blend ratios is the effect these low aromatic fuels will have on fuel-wetted polymeric materials, most notably seals and sealants. These materials typically swell and soften to some degree when exposed to jet fuel and the aromatic content of these fuels contribute to this effect. Semi-synthetic jet fuels with very low aromatic contents may cause seals and sealants to shrink and harden leading to acute or chronic failure. Unfortunately, most of the material qualification tests are more concerned with excessive swelling than shrinkage and there is little guidance offered as to an acceptable level of shrinkage or other changes in physical properties related to low aromatic content. Given the pressing need for guidance data, a program was developed to rapidly survey the volume swell of selected fuel-wetted materials in a range of conventional and semi-synthetic jet fuels and through a statistical analysis to make a determination as to whether there was a basis to be concerned about using fuels with FT blend ratios as high as 50 percent. Concurrent with this analysis data was obtained as to the composition of the fuel absorbed in fuel-wetted materials through the use of GC-MS analysis of swollen samples as well as other supporting data. In this presentation the authors will present a summary of the results of the volume swell and fuel absorbed by selected O-rings and sealants as well as a description of the measurement protocols developed for this program.

40 CFR 86.099-17 - Emission control diagnostic system for 1999 and later light-duty vehicles and light-duty trucks.

Code of Federal Regulations, 2011 CFR

2011-07-01

... Model Year New Gasoline Fueled, Natural Gas-Fueled, Liquefied Petroleum Gas-Fueled and Methanol-Fueled..._locations.html. (1) SAE material. Copies of these materials may be obtained from the Society of Automotive... may be obtained from the International Organization for Standardization, Case Postale 56, CH-1211...

FUEL ELEMENT

DOEpatents

Fortescue, P.; Zumwalt, L.R.

1961-11-28

A fuel element was developed for a gas cooled nuclear reactor. The element is constructed in the form of a compacted fuel slug including carbides of fissionable material in some cases with a breeder material carbide and a moderator which slug is disposed in a canning jacket of relatively impermeable moderator material. Such canned fuel slugs are disposed in an elongated shell of moderator having greater gas permeability than the canning material wherefore application of reduced pressure to the space therebetween causes gas diffusing through the exterior shell to sweep fission products from the system. Integral fission product traps and/or exterior traps as well as a fission product monitoring system may be employed therewith. (AEC)

Solid recovered fuels in the steel industry.

PubMed

Kepplinger, Werner L; Tappeiner, Tamara

2012-04-01

By using waste materials as alternative fuels in metallurgical plants it is possible to minimize the traditionally used reducing agents, such as coke, coal, oil or natural gas. Moreover, by using waste materials in the metallurgical industry it is feasible to recover these materials as far as possible. This also represents another step towards environmental protection because carbon dioxide emissions can be reduced, if the H(2) content of the waste material is greater in comparison with that of the substituted fuel and the effects of global warming can therefore be reduced. In the present article various solid recovered fuels and their applications in the metallurgical industry are detailed.

NEUTRONIC REACTOR COUNTER METHOD AND SYSTEM

DOEpatents

Graham, C.B.; Spiewak, I.

1960-05-31

An improved method is given for controlling the rate of fission in circulating-fuel neutronic reactors in which the fuel is a homogeneous liquid containing fissionable material and a neutron moderator. A change in the rate of flssion is effected by preferentially retaining apart from the circulating fuel a variable amount of either fissionable material or moderator, thereby varying the concentration of fissionable material in the fuel. In the case of an aqueous fuel solution a portion of the water may be continuously vaporized from the circulating solution and the amount of condensate, or condensate plus make-up water, returned to the solution is varied to control the fission rate.

Injector nozzle for molten salt destruction of energetic waste materials

DOEpatents

Brummond, William A.; Upadhye, Ravindra S.

1996-01-01

An injector nozzle has been designed for safely injecting energetic waste materials, such as high explosives, propellants, and rocket fuels, into a molten salt reactor in a molten salt destruction process without premature detonation or back burn in the injection system. The energetic waste material is typically diluted to form a fluid fuel mixture that is injected rapidly into the reactor. A carrier gas used in the nozzle serves as a carrier for the fuel mixture, and further dilutes the energetic material and increases its injection velocity into the reactor. The injector nozzle is cooled to keep the fuel mixture below the decomposition temperature to prevent spontaneous detonation of the explosive materials before contact with the high-temperature molten salt bath.

Injector nozzle for molten salt destruction of energetic waste materials

DOEpatents

Brummond, W.A.; Upadhye, R.S.

1996-02-13

An injector nozzle has been designed for safely injecting energetic waste materials, such as high explosives, propellants, and rocket fuels, into a molten salt reactor in a molten salt destruction process without premature detonation or back burn in the injection system. The energetic waste material is typically diluted to form a fluid fuel mixture that is injected rapidly into the reactor. A carrier gas used in the nozzle serves as a carrier for the fuel mixture, and further dilutes the energetic material and increases its injection velocity into the reactor. The injector nozzle is cooled to keep the fuel mixture below the decomposition temperature to prevent spontaneous detonation of the explosive materials before contact with the high-temperature molten salt bath. 2 figs.

Aircraft Wing Fuel Tank Environmental Simulator Tests for Evaluation of Antimisting Fuels.

DTIC Science & Technology

1984-10-01

C.*: % _ _ _.__ _ o During boost pump operation, strands of a gel-like, semi-transparent material were observed on the free surface of the fuel and...Boeing Materials Technology (BMT) laboratory to measure the water content of the fuel samples is described in appendix C. 2.5.3 Water Ingestion Results...Jet A pump at 8 gpm 32 .. . . ... . . . . . . . -%tr. go*7 .*.**.*.*..* -*.... * . . recuroed for each fueling increment. From these data a height

Delivery of Fuel and Construction Materials to South Pole Station

DTIC Science & Technology

1993-07-01

AID-A270 431 Delivery of Fuel and Construction Materials to South Pole Station Stephen L. DenHartog and George L. Blaisdell July 993 DTIC ELECT S OCT...South Pole Station, ideally with minimal impact on the current science and operational program. The new station will require the delivery of massive...amounts of construction materials to this remote site. The existing means of delivering material and fuel to the South Pole include the use of specialized

Fiber-modified polyurethane foam for ballistic protection

NASA Technical Reports Server (NTRS)

Fish, R. H.; Parker, J. A.; Rosser, R. W.

1975-01-01

Closed-cell, semirigid, fiber-loaded, self-extinguishing polyurethane foam material fills voids around fuel cells in aircraft. Material prevents leakage of fuel and spreading of fire in case of ballistic incendiary impact. It also protects fuel cell in case of exterior fire.

Performance and breakdown characteristics of irradiated vertical power GaN P-i-N diodes

DOE PAGES

King, M. P.; Armstrong, A. M.; Dickerson, J. R.; ...

2015-10-29

Electrical performance and defect characterization of vertical GaN P-i-N diodes before and after irradiation with 2.5 MeV protons and neutrons is investigated. Devices exhibit increase in specific on-resistance following irradiation with protons and neutrons, indicating displacement damage introduces defects into the p-GaN and n- drift regions of the device that impact on-state device performance. The breakdown voltage of these devices, initially above 1700 V, is observed to decrease only slightly for particle fluence <; 10 13 cm -2. Furthermore, the unipolar figure of merit for power devices indicates that while the on-resistance and breakdown voltage degrade with irradiation, vertical GaNmore » P-i-Ns remain superior to the performance of the best available, unirradiated silicon devices and on-par with unirradiated modern SiC-based power devices.« less

Application of DNA comet assay for detection of radiation treatment of grams and pulses.

PubMed

Khan, Hasan M; Khan, Ashfaq A; Khan, Sanaullah

2011-12-01

Several types of whole pulses (green lentils, red lentils, yellow lentils, chickpeas, green peas, cowpeas and yellow peas) and grams (black grams, red grams and white grams) have been investigated for the identification of radiation treatment using microgel electrophoresis of single cells (DNA comet assay). Pulses and grams were exposed to the radiation doses of 0.5, 1.0 and 5 kGy covering the legalized commercial dose range for protection from insect/pest infestations. All irradiated samples showed comet like stretching of fragmented DNA toward anode, which is expected for irradiated samples. Unirradiated samples showed many intact cells/nuclei in form of round stains or with short faint tails, which is typical for unirradiated food samples. The study shows that DNA comet assay can be used as a rapid, inexpensive and highly effective screening test for the detection of radiation treatment of foods, like pulses and grams.

Comparative viability of unirradiated and gamma irradiated bacterial cells

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

Maxcy, R.B.

1977-01-01

Gamma radiation injured Escherichia coli, Salmonella typhimurium, and Moraxella sp. were studied under various environmental stresses to determine their fate relative to the parent population. Irradiated cultures formed smaller colonies on surface plates with fewer cells per colony. Unirradiated cultures had a shorter lag phase than irradiated cultures in broth and duration of lag increased as a result of increasing the radiation dose. Repeated irradiation and subculture progressively retarded growth rate. Multiple radiation of highly resistant Moraxella sp. showed radiation injured cells to be more sensitive than uninjured cells. With the three species studied, irradiation raised the lower limits ofmore » growth temperature, increased the sensitivity to freezing and thawing, and increased the susceptibility to lowered water activity. This work indicated that the production of a bizarre, resistant strain of bacteria through recycling in a food processing operation is highly unlikely.« less

Proton irradiation studies on Al and Al5083 alloy

NASA Astrophysics Data System (ADS)

Bhattacharyya, P.; Gayathri, N.; Bhattacharya, M.; Gupta, A. Dutta; Sarkar, Apu; Dhar, S.; Mitra, M. K.; Mukherjee, P.

2017-10-01

The change in the microstructural parameters and microhardness values in 6.5 MeV proton irradiated pure Al and Al5083 alloy samples have been evaluated using different model based techniques of X-ray diffraction Line Profile Analysis (XRD) and microindendation techniques. The detailed line profile analysis of the XRD data showed that the domain size increases and saturates with irradiation dose both in the case of Al and Al5083 alloy. The corresponding microstrain values did not show any change with irradiation dose in the case of the pure Al but showed an increase at higher irradiation doses in the case of Al5083 alloy. The microindendation results showed that unirradiated Al5083 alloy has higher hardness value compared to that of unirradiated pure Al. The hardness increased marginally with irradiation dose in the case of Al5083, whereas for pure Al, there was no significant change with dose.

NUCLEAR MATERIAL ATTRACTIVENESS: AN ASSESSMENT OF MATERIAL FROM PHWR'S IN A CLOSED THORIUM FUEL CYCLE

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

Sleaford, B W; Collins, B A; Ebbinghaus, B B

2010-04-26

This paper examines the attractiveness of material mixtures containing special nuclear materials (SNM) associated with reprocessing and the thorium-based LWR fuel cycle. This paper expands upon the results from earlier studies that examined the attractiveness of SNM associated with the reprocessing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR. This study shows that {sup 233}U that is produced in thorium-based fuel cycles is very attractive for weapons use. Consistent with other studies, these results also show that all fuel cycles examined to date needmore » to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of 'attractiveness levels' that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities. The methodology and key findings will be presented.« less

Nuclear Material Attractiveness: An Assessment of Material from PHWR's in a Closed Thorium Fuel Cycle

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

Sleaford, Brad W.; Ebbinghaus, B. B.; Bradley, Keith S.

2010-06-11

This paper examines the attractiveness of material mixtures containing special nuclear materials (SNM) associated with reprocessing and the thorium-based LWR fuel cycle. This paper expands upon the results from earlier studies [ , ] that examined the attractiveness of SNM associated with the reprocessing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR. This study shows that 233U that is produced in thorium-based fuel cycles is very attractive for weapons use. Consistent with other studies, these results also show that all fuel cycles examined tomore » date need to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of "attractiveness levels" that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities [ ]. The methodology and key findings will be presented.« less

Numerical study of slip system activity and crystal lattice rotation under wedge nanoindents in tungsten single crystals

NASA Astrophysics Data System (ADS)

Volz, T.; Schwaiger, R.; Wang, J.; Weygand, S. M.

2018-05-01

Tungsten is a promising material for plasma facing components in future nuclear fusion reactors. In the present work, we numerically investigate the deformation behavior of unirradiated tungsten (a body-centered cubic (bcc) single crystal) underneath nanoindents. A finite element (FE) model is presented to simulate wedge indentation. Crystal plasticity finite element (CPFE) simulations were performed for face-centered and body-centered single crystals accounting for the slip system family {110} <111> in the bcc crystal system and the {111} <110> slip family in the fcc system. The 90° wedge indenter was aligned parallel to the [1 ¯01 ]-direction and indented the crystal in the [0 1 ¯0 ]-direction up to a maximum indentation depth of 2 µm. In both, the fcc and bcc single crystals, the activity of slip systems was investigated and compared. Good agreement with the results from former investigations on fcc single crystals was observed. Furthermore, the in-plane lattice rotation in the material underneath an indent was determined and compared for the fcc and bcc single crystals.

Low Activation Joining of SiC/SiC Composites for Fusion Applications: Modeling Miniature Torsion Tests

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

Henager, Charles H.; Nguyen, Ba Nghiep; Kurtz, Richard J.

2014-06-30

The use of SiC and SiC-composites in fission or fusion environments appears to require joining methods for assembling systems. The international fusion community has designed miniature torsion specimens for joint testing and for irradiation in HFIR. Therefore, miniature torsion joints were fabricated using displacement reactions between Si and TiC to produce Ti3SiC2 + SiC joints with CVD-SiC that were tested in shear prior to and after HFIR irradiation. However, these torsion specimens fail out-of-plane, which causes difficulties in determining a shear strength for the joints or for comparing unirradiated and irradiated joints. A finite element damage model has been developedmore » that indicates fracture is likely to occur within the joined pieces to cause out-of-plane failures for miniature torsion specimens when a certain modulus and strength ratio between the joint material and the joined material exists. The implications for torsion shear joint data based on this sample design are discussed.« less

Radiation effects on beta 10.6 of pure and europium doped KCl

NASA Technical Reports Server (NTRS)

Grimes, H. H.; Maisel, J. E.; Hartford, R. H.

1975-01-01

Changes in the optical absorption coefficient as a result of X-ray and electron bombardment of pure KCl (monocrystalline and polycrystalline), and divalent europium doped polycrystalline KCl were determined. The optical absorption coefficients were measured by a constant heat flow calorimetric method. Both 300 KV X-irradiation and 2 MeV electron irradiation produced significant increases in beta 10.6, measured at room temperature. The X-irradiation of pure moncrystalline KCl increased beta 10.6 by 0.005/cm for a 113 MR dose. For an equivalent dose, 2 MeV electrons were found less efficient in changing beta 10.6. However, electron irradiation of pure and Eu-doped polycrystalline KCl produced marked increases in adsorption. Beta increased to over 0.25/cm in Eu-doped material for a 30 x 10 to the 14th power electrons/sq cm dose, a factor of 20 increase over unirradiated material. Moreover, bleaching the electron irradiated doped KCl with 649 m light produced and additional factor of 1.5 increase. These findings will be discussed in light of known defect-center properties in KCl.

Modeling plastic deformation of post-irradiated copper micro-pillars

NASA Astrophysics Data System (ADS)

Crosby, Tamer; Po, Giacomo; Ghoniem, Nasr M.

2014-12-01

We present here an application of a fundamentally new theoretical framework for description of the simultaneous evolution of radiation damage and plasticity that can describe both in situ and ex situ deformation of structural materials [1]. The theory is based on the variational principle of maximum entropy production rate; with constraints on dislocation climb motion that are imposed by point defect fluxes as a result of irradiation. The developed theory is implemented in a new computational code that facilitates the simulation of irradiated and unirradiated materials alike in a consistent fashion [2]. Discrete Dislocation Dynamics (DDD) computer simulations are presented here for irradiated fcc metals that address the phenomenon of dislocation channel formation in post-irradiated copper. The focus of the simulations is on the role of micro-pillar boundaries and the statistics of dislocation pinning by stacking-fault tetrahedra (SFTs) on the onset of dislocation channel and incipient surface crack formation. The simulations show that the spatial heterogeneity in the distribution of SFTs naturally leads to localized plastic deformation and incipient surface fracture of micro-pillars.

Principles of Fuel and Fuel Systems, 8-4. Military Curriculum Materials for Vocational and Technical Education.

ERIC Educational Resources Information Center

Army Ordnance Center and School, Aberdeen Proving Ground, MD.

This volume of student materials for a secondary/postsecondary level course in principles of fuel and fuel systems is one of a number of military-developed curriculum packages selected for adaptation to vocational instruction and curriculum development in a civilian setting. The purpose of the individualized, self-paced course is to provide the…

Advanced ceramic materials for next-generation nuclear applications

NASA Astrophysics Data System (ADS)

Marra, John

2011-10-01

The nuclear industry is at the eye of a 'perfect storm' with fuel oil and natural gas prices near record highs, worldwide energy demands increasing at an alarming rate, and increased concerns about greenhouse gas (GHG) emissions that have caused many to look negatively at long-term use of fossil fuels. This convergence of factors has led to a growing interest in revitalization of the nuclear power industry within the United States and across the globe. Many are surprised to learn that nuclear power provides approximately 20% of the electrical power in the US and approximately 16% of the world-wide electric power. With the above factors in mind, world-wide over 130 new reactor projects are being considered with approximately 25 new permit applications in the US. Materials have long played a very important role in the nuclear industry with applications throughout the entire fuel cycle; from fuel fabrication to waste stabilization. As the international community begins to look at advanced reactor systems and fuel cycles that minimize waste and increase proliferation resistance, materials will play an even larger role. Many of the advanced reactor concepts being evaluated operate at high-temperature requiring the use of durable, heat-resistant materials. Advanced metallic and ceramic fuels are being investigated for a variety of Generation IV reactor concepts. These include the traditional TRISO-coated particles, advanced alloy fuels for 'deep-burn' applications, as well as advanced inert-matrix fuels. In order to minimize wastes and legacy materials, a number of fuel reprocessing operations are being investigated. Advanced materials continue to provide a vital contribution in 'closing the fuel cycle' by stabilization of associated low-level and high-level wastes in highly durable cements, ceramics, and glasses. Beyond this fission energy application, fusion energy will demand advanced materials capable of withstanding the extreme environments of high-temperature plasma systems. Fusion reactors will likely depend on lithium-based ceramics to produce tritium that fuels the fusion plasma, while high-temperature alloys or ceramics will contain and control the hot plasma. All the while, alloys, ceramics, and ceramic-related processes continue to find applications in the management of wastes and byproducts produced by these processes.

Solid oxide fuel cell with single material for electrodes and interconnect

DOEpatents

McPheeters, Charles C.; Nelson, Paul A.; Dees, Dennis W.

1994-01-01

A solid oxide fuel cell having a plurality of individual cells. A solid oxide fuel cell has an anode and a cathode with electrolyte disposed therebetween, and the anode, cathode and interconnect elements are comprised of substantially one material.

Method and apparatus for close packing of nuclear fuel assemblies

DOEpatents

Newman, Darrell F.

1993-01-01

The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

Method and apparatus for close packing of nuclear fuel assemblies

DOEpatents

Newman, D.F.

1993-03-30

The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

Fuel containment and damage tolerance for large composite primary aircraft structures. Phase 1: Testing

NASA Technical Reports Server (NTRS)

Sandifer, J. P.

1983-01-01

Technical problems associated with fuel containment and damage tolerance of composite material wings for transport aircraft were identified. The major tasks are the following: (1) the preliminary design of damage tolerant wing surface using composite materials; (2) the evaluation of fuel sealing and lightning protection methods for a composite material wing; and (3) an experimental investigation of the damage tolerant characteristics of toughened resin graphite/epoxy materials. The test results, the test techniques, and the test data are presented.

The Role of Ceramics in a Resurgent Nuclear Industry

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

Marra, J

2006-02-28

With fuel oil and natural gas prices near record highs and worldwide energy demands increasing at an alarming rate, there is growing interest in revitalization of the nuclear power industry within the United States and across the globe. Ceramic materials have long played a very important part in the commercial nuclear industry with applications throughout the entire fuel cycle; from fuel fabrication to waste stabilization. As the international community begins to look at advanced fuel cycles that minimize waste and increase proliferation resistance, ceramic materials will play an even larger role. Many of the advanced reactor concepts being evaluated operatemore » at high-temperature requiring the use of durable, heat-resistant materials. Ceramic fuels are being investigated for a variety of Generation IV reactor concepts. These include the traditional TRISO-coated particles as well as advanced inert-matrix fuels. In order to minimize wastes and legacy materials, ceramic processes are also being applied to fuel reprocessing operations. Ceramic materials continue to provide a vital contribution in ''closing the fuel cycle'' by stabilization of associated low-level and high-level wastes in highly durable grout, ceramics, and glass. In the next five years, programs that are currently in the conceptual phase will begin laboratory- and engineering-scale demonstrations. This will require production-scale demonstrations of several ceramic technologies from fuel form development to advanced stabilization methods. Within the next five to ten years, these demonstrations will move to even larger scales and will also include radioactive demonstrations of these advanced technologies. These radioactive demonstrations are critical to program success and will require advances in ceramic materials associated with nuclear energy applications.« less

Mating competitiveness of male Anopheles arabiensis mosquitoes irradiated with a partially or fully sterilizing dose in small and large laboratory cages.

PubMed

Helinski, M E H; Knols, B G J

2008-07-01

Male mating competitiveness is a crucial parameter in many genetic control programs including the sterile insect technique (SIT). We evaluated competitiveness of male Anopheles arabiensis Patton as a function of three experimental variables: (1) small or large cages for mating, (2) the effects of either a partially sterilizing (70 Gy) or fully sterilizing (120 Gy) dose, and (3) pupal or adult irradiation. Irradiated males competed for females with an equal number of unirradiated males. Competitiveness was determined by measuring hatch rates of individually laid egg batches. In small cages, pupal irradiation with the high dose resulted in the lowest competitiveness, whereas adult irradiation with the low dose gave the highest, with the latter males being equal in competitiveness to unirradiated males. In the large cage, reduced competitiveness of males irradiated in the pupal stage was more pronounced compared with the small cage; the males irradiated as adults at both doses performed similarly to unirradiated males. Unexpectedly, males irradiated with the high dose performed better in a large cage than in a small one. A high proportion of intermediate hatch rates was observed for eggs collected in the large cage experiments with males irradiated at the pupal stage. It is concluded that irradiation of adult An. arabiensis with the partially sterilizing dose results in the highest competitiveness for both cage designs. Cage size affected competitiveness for some treatments; therefore, competitiveness determined in laboratory experiments must be confirmed by releases into simulated field conditions. The protocols described are readily transferable to evaluate male competitiveness for other genetic control techniques.

Coal liquefaction process wherein jet fuel, diesel fuel and/or ASTM No. 2 fuel oil is recovered

DOEpatents

Bauman, Richard F.; Ryan, Daniel F.

1982-01-01

An improved process for the liquefaction of coal and similar solid carbonaceous materials wherein a hydrogen donor solvent or diluent derived from the solid carbonaceous material is used to form a slurry of the solid carbonaceous material and wherein the naphthenic components from the solvent or diluent fraction are separated and used as jet fuel components. The extraction increases the relative concentration of hydroaromatic (hydrogen donor) components and as a result reduces the gas yield during liquefaction and decreases hydrogen consumption during said liquefaction. The hydrogenation severity can be controlled to increase the yield of naphthenic components and hence the yield of jet fuel and in a preferred embodiment jet fuel yield is maximized while at the same time maintaining solvent balance.

Fuel Performance Calculations for FeCrAl Cladding in BWRs

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

George, Nathan; Sweet, Ryan; Maldonado, G. Ivan

2015-01-01

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

Fabrication and Testing of CERMET Fuel Materials for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie; Mireles, Omar

2012-01-01

A first generation Nuclear Cryogenic Propulsion Stage (NCPS) based on Nuclear Thermal Propulsion (NTP) is currently being developed for Advanced Space Exploration Systems. The overall goal of the project is to address critical NTP technology challenges and programmatic issues to establish confidence in the affordability and viability of NTP systems. The current technology roadmap for NTP identifies the development of a robust fuel form as a critical near term need. The lack of a qualified nuclear fuel is a significant technical risk that will require a considerable fraction of program resources to mitigate. Due to these risks and the cost for qualification, the development and selection of a primary fuel must begin prior to Authority to Proceed (ATP) for a specific mission. The fuel development is a progressive approach to incrementally reduce risk, converge the fuel materials, and mature the design and fabrication process of the fuel element. A key objective of the current project is to advance the maturity of CERMET fuels. The work includes fuel processing development and characterization, fuel specimen hot hydrogen screening, and prototypic fuel element testing. Early fuel materials development is critical to help validate requirements and fuel performance. The purpose of this paper is to provide an overview and status of the work at Marshall Space Flight Center (MSFC).

Nuclear fuels for very high temperature applications

NASA Astrophysics Data System (ADS)

Lundberg, L. B.; Hobbins, R. R.

The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO2 or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures.

10 CFR 32.2 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Definitions. 32.2 Section 32.2 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL... disposal, or nuclear material contained in any fuel assembly, subassembly, fuel rod, or fuel pellet...

10 CFR 32.2 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Definitions. 32.2 Section 32.2 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL... disposal, or nuclear material contained in any fuel assembly, subassembly, fuel rod, or fuel pellet...

Three approaches to fuels from fatty compounds

USDA-ARS?s Scientific Manuscript database

Biodiesel, the alkyl esters, usually methyl esters, of vegetable oils, animal fats, or other triacylglycerol-containing materials, are the most common approach to producing a fuel from the mentioned materials. This fuel is obtained by transesterifying the oil or fat with an alcohol, usually methanol...

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

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

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

1995-09-01

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

Designed porosity materials in nuclear reactor components

DOEpatents

Yacout, A. M.; Pellin, Michael J.; Stan, Marius

2016-09-06

A nuclear fuel pellet with a porous substrate, such as a carbon or tungsten aerogel, on which at least one layer of a fuel containing material is deposited via atomic layer deposition, and wherein the layer deposition is controlled to prevent agglomeration of defects. Further, a method of fabricating a nuclear fuel pellet, wherein the method features the steps of selecting a porous substrate, depositing at least one layer of a fuel containing material, and terminating the deposition when the desired porosity is achieved. Also provided is a nuclear reactor fuel cladding made of a porous substrate, such as silicon carbide aerogel or silicon carbide cloth, upon which layers of silicon carbide are deposited.

Effect of Radiation Exposure on the Retention of Commercial NAND Flash Memory

NASA Technical Reports Server (NTRS)

Oldham, Timothy R.; Chen, D.; Friendlich, M.; Carts, M. A.; Seidleck, C. M.; LaBel, K. A.

2011-01-01

We have compared the data retention of irradiated commercial NAND flash memories with that of unirradiated controls. Under some circumstanc es, radiation exposure has a significant effect on the retention of f lash memories.

Temperature characteristics for PTC material heating diesel fuel

NASA Astrophysics Data System (ADS)

Gu, Lefeng; Li, Xiaolu; Wang, Jun; Li, Ying; Li, Ming

2010-08-01

This paper gives a way which utilizes the PTC (Positive Temperature Coefficient) material to preheat diesel fuel in the injector in order to improve the cold starting and emissions of engine. A new injector is also designed. In order to understand the preheating process in this new injector, a dynamic temperature testing system combined with the MSP430F149 data acquisition system is developed for PTC material heating diesel fuel. Especially, the corresponding software and hardware circuits are explained. The temperature of diesel fuel preheating by PTC ceramics is measured under different voltages and distances, which Curie point is 75 °C. Diesel fuel is heated by self-defined temperature around the Curie point of PTC ceramics. The diesel fuel temperature rises rapidly in 2 minutes of the beginning, then can reach 60 °C within 5 minutes as its distance is 5mm away from the surface of PTC ceramics. However, there are a lot of fundamental studies and technology to be resolved in order to apply PTC material in the injector successfully.

Nuclear fuel elements having a composite cladding

DOEpatents

Gordon, Gerald M.; Cowan, II, Robert L.; Davies, John H.

1983-09-20

An improved nuclear fuel element is disclosed for use in the core of nuclear reactors. The improved nuclear fuel element has a composite cladding of an outer portion forming a substrate having on the inside surface a metal layer selected from the group consisting of copper, nickel, iron and alloys of the foregoing with a gap between the composite cladding and the core of nuclear fuel. The nuclear fuel element comprises a container of the elongated composite cladding, a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container and a nuclear fuel material retaining means positioned in the cavity. The metal layer of the composite cladding prevents perforations or failures in the cladding substrate from stress corrosion cracking or from fuel pellet-cladding interaction or both. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-11-21

A fuel element is designed which is particularly adapted for reactors of high power density used to generate steam for the production of electricity. The fuel element consists of inner and outer concentric tubes forming an annular chamber within which is contained fissionable fuel pellet segments, wedge members interposed between the fuel segments, and a spring which, acting with wedge members, urges said fuel pellets radially into contact against the inner surface of the outer tube. The wedge members may be a fertile material convertible into fissionable fuel material by absorbing neutrons emitted from the fissionable fuel pellet segments. The costly grinding of cylindrical fuel pellets to close tolerances for snug engagement is reduced because the need to finish the exact size is eliminated. (AEC)

LIFE Materials: Overview of Fuels and Structural Materials Issues Volume 1

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

Farmer, J

2008-09-08

The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spentmore » nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission blanket in a fusion-fission hybrid system is subcritical, a LIFE engine can burn any fertile or fissile nuclear material, including un-enriched natural or depleted U and SNF, and can extract a very high percentage of the energy content of its fuel resulting in greatly enhanced energy generation per metric ton of nuclear fuel, as well as nuclear waste forms with vastly reduced concentrations of long-lived actinides. LIFE engines could thus provide the ability to generate vast amounts of electricity while greatly reducing the actinide content of any existing or future nuclear waste and extending the availability of low cost nuclear fuels for several thousand years. LIFE also provides an attractive pathway for burning excess weapons Pu to over 99% FIMA (fission of initial metal atoms) without the need for fabricating or reprocessing mixed oxide fuels (MOX). Because of all of these advantages, LIFE engines offer a pathway toward sustainable and safe nuclear power that significantly mitigates nuclear proliferation concerns and minimizes nuclear waste. An important aspect of a LIFE engine is the fact that there is no need to extract the fission fuel from the fission blanket before it is burned to the desired final level. Except for fuel inspection and maintenance process times, the nuclear fuel is always within the core of the reactor and no weapons-attractive materials are available outside at any point in time. However, an important consideration when discussing proliferation concerns associated with any nuclear fuel cycle is the ease with which reactor fuel can be converted to weapons usable materials, not just when it is extracted as waste, but at any point in the fuel cycle. Although the nuclear fuel remains in the core of the engine until ultra deep actinide burn up is achieved, soon after start up of the engine, once the system breeds up to full power, several tons of fissile material is present in the fission blanket. However, this fissile material is widely dispersed in millions of fuel pebbles, which can be tagged as individual accountable items, and thus made difficult to divert in large quantities. Several topical reports are being prepared on the materials and processes required for the LIFE engine. Specific materials of interest include: (1) Baseline TRISO Fuel (TRISO); (2) Inert Matrix Fuel (IMF) & Other Alternative Solid Fuels; (3) Beryllium (Be) & Molten Lead Blankets (Pb/PbLi); (4) Molten Salt Coolants (FLIBE/FLiNaBe/FLiNaK); (5) Molten Salt Fuels (UF4 + FLIBE/FLiNaBe); (6) Cladding Materials for Fuel & Beryllium; (7) ODS FM Steel (ODS); (8) Solid First Wall (SFW); and (9) Solid-State Tritium Storage (Hydrides).« less

Zirconia-magnesia inert matrix fuel and waste form: Synthesis, characterization and chemical performance in an advanced fuel cycle

NASA Astrophysics Data System (ADS)

Holliday, Kiel Steven

There is a significant buildup in plutonium stockpiles throughout the world, because of spent nuclear fuel and the dismantling of weapons. The radiotoxicity of this material and proliferation risk has led to a desire for destroying excess plutonium. To do this effectively, it must be fissioned in a reactor as part of a uranium free fuel to eliminate the generation of more plutonium. This requires an inert matrix to volumetrically dilute the fissile plutonium. Zirconia-magnesia dual phase ceramic has been demonstrated to be a favorable material for this task. It is neutron transparent, zirconia is chemically robust, magnesia has good thermal conductivity and the ceramic has been calculated to conform to current economic and safety standards. This dissertation contributes to the knowledge of zirconia-magnesia as an inert matrix fuel to establish behavior of the material containing a fissile component. First, the zirconia-magnesia inert matrix is synthesized in a dual phase ceramic containing a fissile component and a burnable poison. The chemical constitution of the ceramic is then determined. Next, the material performance is assessed under conditions relevant to an advanced fuel cycle. Reactor conditions were assessed with high temperature, high pressure water. Various acid solutions were used in an effort to dissolve the material for reprocessing. The ceramic was also tested as a waste form under environmental conditions, should it go directly to a repository as a spent fuel. The applicability of zirconia-magnesia as an inert matrix fuel and waste form was tested and found to be a promising material for such applications.

Permeability of Impacted Coated Composite Laminates

NASA Technical Reports Server (NTRS)

Johnson, W. S.; Findley, Benjamin

2002-01-01

Composite materials are being considered for use on future generations of Reusable Launch Vehicles (RLVs) for both fuel tanks and fuel feedlines. Through the use of composite materials NASA can reduce the overall weight of the vehicle dramatically. This weight savings can then be translated into an increase in the weight of payload sent into orbit, reducing the cost per pound of payload. It is estimated that by switching to composite materials for fuel tanks the weight of the tanks can be reduced by 40 percent, which translates to a total vehicle weight savings of 14 percent. In this research, carbon/epoxy composites were studied for fuel feedline applications. There are concerns about using composite materials for feedlines and fuel tanks because these materials are extremely vulnerable to impact in the form of inadvertent bumping or dropped tools both during installation and maintenance. Additionally, it has been found that some of the sample feedlines constructed have had leaks, and thus there may be a need to seal preexisting leaks in the composite prior to usage.

An overview—Functional nanomaterials for lithium rechargeable batteries, supercapacitors, hydrogen storage, and fuel cells

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

Liu, Hua Kun, E-mail: hua@uow.edu.au

2013-12-15

Graphical abstract: Nanomaterials play important role in lithium ion batteries, supercapacitors, hydrogen storage and fuel cells. - Highlights: • Nanomaterials play important role for lithium rechargeable batteries. • Nanostructured materials increase the capacitance of supercapacitors. • Nanostructure improves the hydrogenation/dehydrogenation of hydrogen storage materials. • Nanomaterials enhance the electrocatalytic activity of the catalysts in fuel cells. - Abstract: There is tremendous worldwide interest in functional nanostructured materials, which are the advanced nanotechnology materials with internal or external dimensions on the order of nanometers. Their extremely small dimensions make these materials unique and promising for clean energy applications such as lithiummore » ion batteries, supercapacitors, hydrogen storage, fuel cells, and other applications. This paper will highlight the development of new approaches to study the relationships between the structure and the physical, chemical, and electrochemical properties of functional nanostructured materials. The Energy Materials Research Programme at the Institute for Superconducting and Electronic Materials, the University of Wollongong, has been focused on the synthesis, characterization, and applications of functional nanomaterials, including nanoparticles, nanotubes, nanowires, nanoporous materials, and nanocomposites. The emphases are placed on advanced nanotechnology, design, and control of the composition, morphology, nanostructure, and functionality of the nanomaterials, and on the subsequent applications of these materials to areas including lithium ion batteries, supercapacitors, hydrogen storage, and fuel cells.« less

Investigation of 14.5mm API Self-Sealing/Crashworthy Fuel Tank Material

DTIC Science & Technology

1974-09-01

describes the results of a f-rogram for a crashworthy, 14.5mm API tolerant fuel cell construction developed and subjected co qualification testing. The...Paragraphs 4.6.6.4 and 4.6.6.5), which were not required by contract. Two fuel tanks were built of a construction designated by The Goodyear Tire & Rubber...TABLES 3 INTRODUCTION 4 FUEL TANK MATERIAL DESIGN STUDY (TASK I) 4 QUALIFIC/.TION OF CONSTRUCTION (TASK 11) ........ 5 FUEL TANK GUNFIRE 12

Solar Energy for Transportation Fuel (LBNL Science at the Theater)

ScienceCinema

Lewis, Nate

2018-05-25

Nate Lewis' talk looks at the challenge of capturing solar energy and storing it as an affordable transportation fuel - all on a scale necessary to reduce global warming. Overcoming this challenge will require developing new materials that can use abundant and inexpensive elements rather than costly and rare materials. He discusses the promise of new materials in the development of carbon-free alternatives to fossil fuel.

LMFBR fuel assembly design for HCDA fuel dispersal

DOEpatents

Lacko, Robert E.; Tilbrook, Roger W.

1984-01-01

A fuel assembly for a liquid metal fast breeder reactor having an upper axial blanket region disposed in a plurality of zones within the fuel assembly. The characterization of a zone is dependent on the height of the axial blanket region with respect to the active fuel region. The net effect of having a plurality of zones is to establish a dispersal flow path for the molten materials resulting during a core meltdown accident. Upward flowing molten material can escape from the core region and/or fuel assembly without solidifying on the surface of fuel rods due to the heat sink represented by blanket region pellets.

Solid oxide fuel cell with single material for electrodes and interconnect

DOEpatents

McPheeters, C.C.; Nelson, P.A.; Dees, D.W.

1994-07-19

A solid oxide fuel cell is described having a plurality of individual cells. A solid oxide fuel cell has an anode and a cathode with electrolyte disposed there between, and the anode, cathode and interconnect elements are comprised of substantially one material. 9 figs.

Progress In Developing Laser Based Post Irradiation Examination Infrastructure

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

Smith, James A.; Scott, Clark L.; Benefiel, Brad C.

To be able to understand the performance of reactor fuels and materials, irradiated materials must be characterized effectively and efficiently in a high rad environment. The characterization work must be performed remotely and in an environment hostile to instrumentation. Laser based characterization techniques provide the ability to be remote and robust in a hot-cell environment. Laser based instrumentation also can provide high spatial resolution suitable for scanning and imaging large areas. The INL is currently developing three laser based Post Irradiation Examination (PIE) stations for the Hot Fuel Examination Facility at the INL. These laser based systems will characterize irradiatedmore » materials and fuels. The characterization systems are the following: Laser Shock Laser based ultrasonic C-scan system Gas Assay, Sample, and Recharge system (GASR, up-grade to an existing system). The laser shock technique will characterize material properties and failure loads/mechanisms in various materials such as LWR fuel, plate fuel, and next generation fuel forms, for PIE in high radiation areas. The laser shock-technique induces large amplitude shock waves to mechanically characterize interfaces such as the fuel-clad bond. The shock wave travels as a compression wave through the material to the free (unconfined) back surface and reflects back through the material under test as a rarefaction (tensile) wave. This rarefaction wave is the physical mechanism that produces internal de-lamination failure. As part of the laser shock system, a laser-based ultrasonic C-scan system will be used to detect and characterize debonding caused by the laser shock technique. The laser ultrasonic system will be fully capable of performing classical non-destructive evaluation testing and imaging functions such as microstructure characterization, flaw detection and dimensional metrology in complex components. The purpose of the GASR is to measure the pressure/volume of the plenum of an irradiated fuel element and obtain fission gas samples for analysis. The study of pressure and volume in the plenum of an irradiated fuel element and the analysis of fission gases released from the fuel is important to understanding the performance of reactor fuels and materials. This system may also be used to measure the pressure/volume of other components (such as control blades) and obtain gas samples from these components for analysis. The main function of the laser in this application is to puncture the fuel element to allow the fission gas to escape and if necessary to weld the spot close. The GASR station will have the inherent capability to perform cutting welding and joining functions within a hot-cell.« less

Multi-stage fuel cell system method and apparatus

DOEpatents

George, Thomas J.; Smith, William C.

2000-01-01

A high efficiency, multi-stage fuel cell system method and apparatus is provided. The fuel cell system is comprised of multiple fuel cell stages, whereby the temperatures of the fuel and oxidant gas streams and the percentage of fuel consumed in each stage are controlled to optimize fuel cell system efficiency. The stages are connected in a serial, flow-through arrangement such that the oxidant gas and fuel gas flowing through an upstream stage is conducted directly into the next adjacent downstream stage. The fuel cell stages are further arranged such that unspent fuel and oxidant laden gases too hot to continue within an upstream stage because of material constraints are conducted into a subsequent downstream stage which comprises a similar cell configuration, however, which is constructed from materials having a higher heat tolerance and designed to meet higher thermal demands. In addition, fuel is underutilized in each stage, resulting in a higher overall fuel cell system efficiency.

The TMI regenerable solid oxide fuel cell

NASA Technical Reports Server (NTRS)

Cable, Thomas L.

1995-01-01

Energy storage and production in space requires rugged, reliable hardware which minimizes weight, volume, and maintenance while maximizing power output and usable energy storage. These systems generally consist of photovoltaic solar arrays which operate during sunlight cycles to provide system power and regenerate fuel (hydrogen) via water electrolysis; during dark cycles, hydrogen is converted by the fuel cell into system. The currently preferred configuration uses two separate systems (fuel cell and electrolyzer) in conjunction with photovoltaic cells. Fuel cell/electrolyzer system simplicity, reliability, and power-to-weight and power-to-volume ratios could be greatly improved if both power production (fuel cell) and power storage (electrolysis) functions can be integrated into a single unit. The Technology Management, Inc. (TMI), solid oxide fuel cell-based system offers the opportunity to both integrate fuel cell and electrolyzer functions into one unit and potentially simplify system requirements. Based an the TMI solid oxide fuel cell (SOPC) technology, the TMI integrated fuel cell/electrolyzer utilizes innovative gas storage and operational concepts and operates like a rechargeable 'hydrogen-oxygen battery'. Preliminary research has been completed on improved H2/H2O electrode (SOFC anode/electrolyzer cathode) materials for solid oxide, regenerative fuel cells. Improved H2/H2O electrode materials showed improved cell performance in both fuel cell and electrolysis modes in reversible cell tests. ln reversible fuel cell/electrolyzer mode, regenerative fuel cell efficiencies (ratio of power out (fuel cell mode) to power in (electrolyzer model)) improved from 50 percent (using conventional electrode materials) to over 80 percent. The new materials will allow the TMI SOFC system to operate as both the electrolyzer and fuel cell in a single unit. Preliminary system designs have also been developed which indicate the technical feasibility of using the TMI SOFC technology for space applications with high energy storage efficiencies and high specific energy. Development of small space systems would also have potential dual-use, terrestrial applications.

The TMI regenerable solid oxide fuel cell

NASA Astrophysics Data System (ADS)

Cable, Thomas L.

1995-04-01

Energy storage and production in space requires rugged, reliable hardware which minimizes weight, volume, and maintenance while maximizing power output and usable energy storage. These systems generally consist of photovoltaic solar arrays which operate during sunlight cycles to provide system power and regenerate fuel (hydrogen) via water electrolysis; during dark cycles, hydrogen is converted by the fuel cell into system. The currently preferred configuration uses two separate systems (fuel cell and electrolyzer) in conjunction with photovoltaic cells. Fuel cell/electrolyzer system simplicity, reliability, and power-to-weight and power-to-volume ratios could be greatly improved if both power production (fuel cell) and power storage (electrolysis) functions can be integrated into a single unit. The Technology Management, Inc. (TMI), solid oxide fuel cell-based system offers the opportunity to both integrate fuel cell and electrolyzer functions into one unit and potentially simplify system requirements. Based an the TMI solid oxide fuel cell (SOPC) technology, the TMI integrated fuel cell/electrolyzer utilizes innovative gas storage and operational concepts and operates like a rechargeable 'hydrogen-oxygen battery'. Preliminary research has been completed on improved H2/H2O electrode (SOFC anode/electrolyzer cathode) materials for solid oxide, regenerative fuel cells. Improved H2/H2O electrode materials showed improved cell performance in both fuel cell and electrolysis modes in reversible cell tests. ln reversible fuel cell/electrolyzer mode, regenerative fuel cell efficiencies (ratio of power out (fuel cell mode) to power in (electrolyzer model)) improved from 50 percent (using conventional electrode materials) to over 80 percent. The new materials will allow the TMI SOFC system to operate as both the electrolyzer and fuel cell in a single unit. Preliminary system designs have also been developed which indicate the technical feasibility of using the TMI SOFC technology for space applications with high energy storage efficiencies and high specific energy. Development of small space systems would also have potential dual-use, terrestrial applications.

Facile and gram-scale synthesis of metal-free catalysts: toward realistic applications for fuel cells.

PubMed

Kim, Ok-Hee; Cho, Yong-Hun; Chung, Dong Young; Kim, Min Jeong; Yoo, Ji Mun; Park, Ji Eun; Choe, Heeman; Sung, Yung-Eun

2015-03-02

Although numerous reports on nonprecious metal catalysts for replacing expensive Pt-based catalysts have been published, few of these studies have demonstrated their practical application in fuel cells. In this work, we report graphitic carbon nitride and carbon nanofiber hybrid materials synthesized by a facile and gram-scale method via liquid-based reactions, without the use of toxic materials or a high pressure-high temperature reactor, for use as fuel cell cathodes. The resulting materials exhibited remarkable methanol tolerance, selectivity, and stability even without a metal dopant. Furthermore, these completely metal-free catalysts exhibited outstanding performance as cathode materials in an actual fuel cell device: a membrane electrode assembly with both acidic and alkaline polymer electrolytes. The fabrication method and remarkable performance of the single cell produced in this study represent progressive steps toward the realistic application of metal-free cathode electrocatalysts in fuel cells.

Facile and Gram-scale Synthesis of Metal-free Catalysts: Toward Realistic Applications for Fuel Cells

PubMed Central

Kim, Ok-Hee; Cho, Yong-Hun; Chung, Dong Young; Kim, Min Jeong; Yoo, Ji Mun; Park, Ji Eun; Choe, Heeman; Sung, Yung-Eun

2015-01-01

Although numerous reports on nonprecious metal catalysts for replacing expensive Pt-based catalysts have been published, few of these studies have demonstrated their practical application in fuel cells. In this work, we report graphitic carbon nitride and carbon nanofiber hybrid materials synthesized by a facile and gram-scale method via liquid-based reactions, without the use of toxic materials or a high pressure-high temperature reactor, for use as fuel cell cathodes. The resulting materials exhibited remarkable methanol tolerance, selectivity, and stability even without a metal dopant. Furthermore, these completely metal-free catalysts exhibited outstanding performance as cathode materials in an actual fuel cell device: a membrane electrode assembly with both acidic and alkaline polymer electrolytes. The fabrication method and remarkable performance of the single cell produced in this study represent progressive steps toward the realistic application of metal-free cathode electrocatalysts in fuel cells. PMID:25728910

78 FR 79328 - Amendments to Material Control and Accounting Regulations and Proposed Guidance for Fuel Cycle...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-30

... Accounting Regulations and Proposed Guidance for Fuel Cycle Facility Material Control and Accounting Plans... accounting (MC&A) of special nuclear material (SNM). The public meeting has been rescheduled for January 9...

49 CFR 392.51 - Reserve fuel; materials of trade.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 5 2010-10-01 2010-10-01 false Reserve fuel; materials of trade. 392.51 Section 392.51 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL MOTOR CARRIER SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION FEDERAL MOTOR CARRIER SAFETY REGULATIONS DRIVING OF COMMERCIAL MOTOR VEHICLES Fueling Precautions...

49 CFR 392.51 - Reserve fuel; materials of trade.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 49 Transportation 5 2013-10-01 2013-10-01 false Reserve fuel; materials of trade. 392.51 Section 392.51 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL MOTOR CARRIER SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION FEDERAL MOTOR CARRIER SAFETY REGULATIONS DRIVING OF COMMERCIAL MOTOR VEHICLES Fueling Precautions...

49 CFR 392.51 - Reserve fuel; materials of trade.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 49 Transportation 5 2014-10-01 2014-10-01 false Reserve fuel; materials of trade. 392.51 Section 392.51 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL MOTOR CARRIER SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION FEDERAL MOTOR CARRIER SAFETY REGULATIONS DRIVING OF COMMERCIAL MOTOR VEHICLES Fueling Precautions...

49 CFR 392.51 - Reserve fuel; materials of trade.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 49 Transportation 5 2011-10-01 2011-10-01 false Reserve fuel; materials of trade. 392.51 Section 392.51 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL MOTOR CARRIER SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION FEDERAL MOTOR CARRIER SAFETY REGULATIONS DRIVING OF COMMERCIAL MOTOR VEHICLES Fueling Precautions...

PRELIMINARY EVALUATION OF FeCrAl CLADDING AND U-Si FUEL FOR ACCIDENT TOLERANT FUEL CONCEPTS

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

Hales, J. D.; Gamble, K. A.

2015-09-01

Since the accident at the Fukushima Daiichi Nuclear Power Station, enhancing the accident tolerance of light water reactors (LWRs) has become an important research topic. In particular, the community is actively developing enhanced fuels and cladding for LWRs to improve safety in the event of accidents in the reactor or spent fuel pools. Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system, can tolerate loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normalmore » operations and operational transients. This paper presents early work in developing thermal and mechanical models for two materials that may have promise: U-Si for fuel, and FeCrAl for cladding. These materials would not necessarily be used together in the same fuel system, but individually have promising characteristics. BISON, the finite element-based fuel performance code in development at Idaho National Laboratory, was used to compare results from normal operation conditions with Zr-4/UO2 behavior. In addition, sensitivity studies are presented for evaluating the relative importance of material parameters such as ductility and thermal conductivity in FeCrAl and U-Si in order to provide guidance on future experiments for these materials.« less

Materials challenges for nuclear systems

DOE PAGES

Allen, Todd; Busby, Jeremy; Meyer, Mitch; ...

2010-11-26

The safe and economical operation of any nuclear power system relies to a great extent, on the success of the fuel and the materials of construction. During the lifetime of a nuclear power system which currently can be as long as 60 years, the materials are subject to high temperature, a corrosive environment, and damage from high-energy particles released during fission. The fuel which provides the power for the reactor has a much shorter life but is subject to the same types of harsh environments. This article reviews the environments in which fuels and materials from current and proposed nuclearmore » systems operate and then describes how the creation of the Advanced Test Reactor National Scientific User Facility is allowing researchers from across the U.S. to test their ideas for improved fuels and materials.« less

Technical Assistance to Developers

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

Rockward, Tommy; Borup, Rodney L.; Garzon, Fernando H.

2012-07-17

This task supports the allowance of technical assistance to fuel-cell component and system developers as directed by the DOE. This task includes testing of novel materials and participation in the further development and validation of single cell test protocols. This task also covers technical assistance to DOE Working Groups, the U.S. Council for Automotive Research (USCAR) and the USCAR/DOE Driving Research and Innovation for Vehicle efficiency and Energy sustainability (U.S. Drive) Fuel Cell Technology Team. Assistance includes technical validation of new fuel cell materials and methods, single cell fuel cell testing to support the development of targets and test protocols,more » and regular advisory participation in other working groups and reviews. This assistance is made available to PEM fuel cell developers by request and DOE Approval. The objectives are to: (1) Support technically, as directed by DOE, fuel cell component and system developers; (2) Assess fuel cell materials and components and give feedback to developers; (3) Assist the DOE Durability Working Group with the development of various new material durability Testing protocols; and (4) Provide support to the U.S. Council for Automotive Research (USCAR) and the USCAR/DOE Fuel Cell Technology Team. FY2012 specific technical objectives are: (1) Evaluate novel MPL materials; (2) Develop of startup/ shutdown protocol; (3) Test the impact of hydrophobic treatment on graphite bi-polar plates; (4) Perform complete diagnostics on metal bi-polar plates for corrosion; and (5) Participate and lead efforts in the DOE Working Groups.« less

FISSILE MATERIAL AND FUEL ELEMENTS FOR NEUTRONIC REACTORS

DOEpatents

Shaner, B.E.

1961-08-15

The fissile material consists of about 64 to 70% (weight) zirconium dioxide, 15 to 19% uranium dioxide, and 8 to 17% calcium oxide. The fissile material is formed into sintered composites which are disposed in a compartmented fuel element, comprising essentially a flat filler plate having a plurality of compartments therein, enclosed in cladding plates of the same material as the filler plate. The resultant fuel has good resistance to corrosion in high temperature pressurized water, good dimensional stability to elevated temperatures, and good resistance to thermal shock. (AEC)

Strategies for Carbon and Sulfur Tolerant Solid Oxide Fuel Cell Materials, Incorporating Lessons from Heterogeneous Catalysis.

PubMed

Boldrin, Paul; Ruiz-Trejo, Enrique; Mermelstein, Joshua; Bermúdez Menéndez, José Miguel; Ramı Rez Reina, Tomás; Brandon, Nigel P

2016-11-23

Solid oxide fuel cells (SOFCs) are a rapidly emerging energy technology for a low carbon world, providing high efficiency, potential to use carbonaceous fuels, and compatibility with carbon capture and storage. However, current state-of-the-art materials have low tolerance to sulfur, a common contaminant of many fuels, and are vulnerable to deactivation due to carbon deposition when using carbon-containing compounds. In this review, we first study the theoretical basis behind carbon and sulfur poisoning, before examining the strategies toward carbon and sulfur tolerance used so far in the SOFC literature. We then study the more extensive relevant heterogeneous catalysis literature for strategies and materials which could be incorporated into carbon and sulfur tolerant fuel cells.

First result of deuterium retention in neutron-irradiated tungsten exposed to high flux plasma in TPE

NASA Astrophysics Data System (ADS)

Shimada, Masashi; Hatano, Y.; Calderoni, P.; Oda, T.; Oya, Y.; Sokolov, M.; Zhang, K.; Cao, G.; Kolasinski, R.; Sharpe, J. P.

2011-08-01

With the Japan-US joint research project Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), an initial set of tungsten samples (99.99% purity, A.L.M.T. Co.) were irradiated by high flux neutrons at 323 K to 0.025 dpa in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Subsequently, one of the neutron-irradiated tungsten samples was exposed to a high-flux deuterium plasma (ion flux: 5 × 1021 m-2 s-1, ion fluence: 4 × 1025 m-2) in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory (INL). The deuterium retention in the neutron-irradiated tungsten was 40% higher in comparison to the unirradiated tungsten. The observed broad desorption spectrum from neutron-irradiated tungsten and associated TMAP modeling of the deuterium release suggest that trapping occurs in the bulk material at more than three different energy sites.

Automated X-Ray Diffraction of Irradiated Materials

DOE PAGES

Rodman, John; Lin, Yuewei; Sprouster, David; ...

2017-10-26

Synchrotron-based X-ray diffraction (XRD) and small-angle Xray scattering (SAXS) characterization techniques used on unirradiated and irradiated reactor pressure vessel steels yield large amounts of data. Machine learning techniques, including PCA, offer a novel method of analyzing and visualizing these large data sets in order to determine the effects of chemistry and irradiation conditions on the formation of radiation induced precipitates. In order to run analysis on these data sets, preprocessing must be carried out to convert the data to a usable format and mask the 2-D detector images to account for experimental variations. Once the data has been preprocessed, itmore » can be organized and visualized using principal component analysis (PCA), multi-dimensional scaling, and k-means clustering. In conclusion, from these techniques, it is shown that sample chemistry has a notable effect on the formation of the radiation induced precipitates in reactor pressure vessel steels.« less

Tests of several bearing materials lubricated by gasoline

NASA Technical Reports Server (NTRS)

Joachin, W F; Case, Harold W

1926-01-01

This investigation on the relative wear of several bearing materials lubricated by gasoline was conducted at the Langley Memorial Aeronautical Laboratory, as part of a general research on fuel injection engines for aircraft. The specific purpose of the work was to find a durable bearing material for gear pumps to be used for the delivery of gasoline and diesel engine fuel oil at moderate pressures to the high pressure pumps of fuel injection engines.

Combustible structural composites and methods of forming combustible structural composites

DOEpatents

Daniels, Michael A [Idaho Falls, ID; Heaps, Ronald J [Idaho Falls, ID; Steffler, Eric D [Idaho Falls, ID; Swank, William D [Idaho Falls, ID

2011-08-30

Combustible structural composites and methods of forming same are disclosed. In an embodiment, a combustible structural composite includes combustible material comprising a fuel metal and a metal oxide. The fuel metal is present in the combustible material at a weight ratio from 1:9 to 1:1 of the fuel metal to the metal oxide. The fuel metal and the metal oxide are capable of exothermically reacting upon application of energy at or above a threshold value to support self-sustaining combustion of the combustible material within the combustible structural composite. Structural-reinforcing fibers are present in the composite at a weight ratio from 1:20 to 10:1 of the structural-reinforcing fibers to the combustible material. Other embodiments and aspects are disclosed.

Combustible structural composites and methods of forming combustible structural composites

DOEpatents

Daniels, Michael A.; Heaps, Ronald J.; Steffler, Eric D.; Swank, W. David

2013-04-02

Combustible structural composites and methods of forming same are disclosed. In an embodiment, a combustible structural composite includes combustible material comprising a fuel metal and a metal oxide. The fuel metal is present in the combustible material at a weight ratio from 1:9 to 1:1 of the fuel metal to the metal oxide. The fuel metal and the metal oxide are capable of exothermically reacting upon application of energy at or above a threshold value to support self-sustaining combustion of the combustible material within the combustible structural composite. Structural-reinforcing fibers are present in the composite at a weight ratio from 1:20 to 10:1 of the structural-reinforcing fibers to the combustible material. Other embodiments and aspects are disclosed.

Methods and apparatuses for the development of microstructured nuclear fuels

DOEpatents

Jarvinen, Gordon D [Los Alamos, NM; Carroll, David W [Los Alamos, NM; Devlin, David J [Santa Fe, NM

2009-04-21

Microstructured nuclear fuel adapted for nuclear power system use includes fissile material structures of micrometer-scale dimension dispersed in a matrix material. In one method of production, fissile material particles are processed in a chemical vapor deposition (CVD) fluidized-bed reactor including a gas inlet for providing controlled gas flow into a particle coating chamber, a lower bed hot zone region to contain powder, and an upper bed region to enable powder expansion. At least one pneumatic or electric vibrator is operationally coupled to the particle coating chamber for causing vibration of the particle coater to promote uniform powder coating within the particle coater during fuel processing. An exhaust associated with the particle coating chamber and can provide a port for placement and removal of particles and powder. During use of the fuel in a nuclear power reactor, fission products escape from the fissile material structures and come to rest in the matrix material. After a period of use in a nuclear power reactor and subsequent cooling, separation of the fissile material from the matrix containing the embedded fission products will provide an efficient partitioning of the bulk of the fissile material from the fission products. The fissile material can be reused by incorporating it into new microstructured fuel. The fission products and matrix material can be incorporated into a waste form for disposal or processed to separate valuable components from the fission products mixture.

Void growth and coalescence in irradiated copper under deformation

NASA Astrophysics Data System (ADS)

Barrioz, P. O.; Hure, J.; Tanguy, B.

2018-04-01

A decrease of fracture toughness of irradiated materials is usually observed, as reported for austenitic stainless steels in Light Water Reactors (LWRs) or copper alloys for fusion applications. For a wide range of applications (e.g. structural steels irradiated at low homologous temperature), void growth and coalescence fracture mechanism has been shown to be still predominant. As a consequence, a comprehensive study of the effects of irradiation-induced hardening mechanisms on void growth and coalescence in irradiated materials is required. The effects of irradiation on ductile fracture mechanisms - void growth to coalescence - are assessed in this study based on model experiments. Pure copper thin tensile samples have been irradiated with protons up to 0.01 dpa. Micron-scale holes drilled through the thickness of these samples subjected to uniaxial loading conditions allow a detailed description of void growth and coalescence. In this study, experimental data show that physical mechanisms of micron-scale void growth and coalescence are similar between the unirradiated and irradiated copper. However, an acceleration of void growth is observed in the later case, resulting in earlier coalescence, which is consistent with the decrease of fracture toughness reported in irradiated materials. These results are qualitatively reproduced with numerical simulations accounting for irradiation macroscopic hardening and decrease of strain-hardening capability.

THE ATTRACTIVENESS OF MATERIAS ASSOCIATED WITH THORIUM-BASED NUCLEAR FUEL CYCLES FOR PHWRS

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

Prichard, Andrew W.; Niehus, Mark T.; Collins, Brian A.

2011-07-17

This paper reports the continued evaluation of the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with thorium based nuclear fuel cycles. Specifically, this paper examines a thorium fuel cycle in which a pressurized heavy water reactor (PHWR) is fueled with mixtures of natural uranium/233U/thorium. This paper uses a PHWR fueled with natural uranium as a base fuel cycle, and then compares material attractiveness of fuel cycles that use 233U/thorium salted with natural uranium. The results include the material attractiveness of fuel at beginning of life (BoL), end of life (EoL), and the number of fuel assemblies requiredmore » to collect a bare critical mass of plutonium or uranium. This study indicates what is required to render the uranium as having low utility for use in nuclear weapons; in addition, this study estimates the increased number of assemblies required to accumulate a bare critical mass of plutonium that has a higher utility for use in nuclear weapons. This approach identifies that some fuel cycles may be easier to implement the International Atomic Energy Agency (IAEA) safeguards approach and have a more effective safeguards by design outcome. For this study, approximately one year of fuel is required to be reprocessed to obtain one bare critical mass of plutonium. Nevertheless, the result of this paper suggests that all spent fuel needs to be rigorously safeguarded and provided with high levels of physical protection. This study was performed at the request of the United States Department of Energy /National Nuclear Security Administration (DOE/NNSA). The methodology and key findings will be presented.« less

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

Carmack, William Jonathan; Braase, Lori Ann

Fuel recovery from severe accidents requires careful planning and execution. The Idaho National Laboratory played a key role in the Three Mile Island (TMI) fuel and core recovery. This involved technology development to locate and handle the damaged fuel; characterization of fuel and debris; analysis of fuel interaction with structural components and materials; development of fuel drying technology for long-term storage. However, one of the critical activities from the TMI project was the extensive effort document all the activities and archive the reports and photos. A historical review of the TMI project at the INL leads to the identification ofmore » current applications and considerations for facility designs, fuel handling, robotic applications, material characterization, etc.« less

78 FR 79019 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Materials...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-27

... Subcommittee on Materials, Metallurgy & Reactor Fuels; Notice of Meeting The ACRS Subcommittee on Materials, Metallurgy & Reactor Fuels will hold a meeting on January 14, 2014, Room T-2B1, 11545 Rockville Pike... NRC's research activities in materials and metallurgy. The Subcommittee will hear presentations by and...

40 CFR 86.005-17 - On-board diagnostics.

Code of Federal Regulations, 2011 CFR

2011-07-01

... Petroleum Gas-Fueled and Methanol-Fueled Heavy-Duty Vehicles § 86.005-17 On-board diagnostics. (a) General...: (1) SAE material. Copies of these materials may be obtained from the Society of Automotive Engineers..., J1939-71, J1939-73, J1939-81). (2) ISO materials. Copies of these materials may be obtained from the...

40 CFR 86.005-17 - On-board diagnostics.

Code of Federal Regulations, 2010 CFR

2010-07-01

... Petroleum Gas-Fueled and Methanol-Fueled Heavy-Duty Vehicles § 86.005-17 On-board diagnostics. (a) General...: (1) SAE material. Copies of these materials may be obtained from the Society of Automotive Engineers..., J1939-71, J1939-73, J1939-81). (2) ISO materials. Copies of these materials may be obtained from the...

Stability of Materials in High Temperature Water Vapor: SOFC Applications

NASA Technical Reports Server (NTRS)

Opila, E. J.; Jacobson, N. S.

2010-01-01

Solid oxide fuel cell material systems require long term stability in environments containing high-temperature water vapor. Many materials in fuel cell systems react with high-temperature water vapor to form volatile hydroxides which can degrade cell performance. In this paper, experimental methods to characterize these volatility reactions including the transpiration technique, thermogravimetric analysis, and high pressure mass spectrometry are reviewed. Experimentally determined data for chromia, silica, and alumina volatility are presented. In addition, data from the literature for the stability of other materials important in fuel cell systems are reviewed. Finally, methods for predicting material recession due to volatilization reactions are described.

Estimating Fuel Bed Loadings in Masticated Areas

Treesearch

Sharon Hood; Ros Wu

2006-01-01

Masticated fuel treatments that chop small trees, shrubs, and dead woody material into smaller pieces to reduce fuel bed depth are used increasingly as a mechanical means to treat fuels. Fuel loading information is important to monitor changes in fuels. The commonly used planar intercept method however, may not correctly estimate fuel loadings because masticated fuels...

An early glimpse at long-term subsonic commercial turbofan technology requirements. [fuel conservation

NASA Technical Reports Server (NTRS)

Gray, D. E.; Dugan, J. F.

1975-01-01

This paper reports on the exploratory investigation and initial findings of the study of future turbofan concepts to conserve fuel. To date, these studies have indicated a potential reduction in cruise thrust specific fuel consumption in 1990 turbofans of approximately 15% relative to present day new engines through advances in internal aerodynamics, structure-mechanics, and materials. Advanced materials also offer the potential for fuel savings through engine weight reduction. Further studies are required to balance fuel consumption reduction with sound airlines operational economics.

HTGR fuels and core development program. Quarterly progress report for the period ending August 31, 1975

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

None

1975-09-30

Studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies are described. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and, where appropriate, the data are presented in tables, graphs, and photographs. (auth)

Fuel reduction and coarse woody debris dynamics with early season and late season prescribed fires in a Sierra Nevada mixed conifer forest

Treesearch

Eric E. Knapp; Jon E. Keeley; Elizabeth A. Ballenger; Teresa J. Brennan

2005-01-01

Fire exclusion has led to an unnatural accumulation and greater spatial continuity of organic material on the ground in many forests. This material serves both as potential fuel for forest fires and habitat for a large array of forest species. Managers must balance fuel reduction to reduce wildfire hazard with fuel retention targets to maintain other forest functions....

Processing fissile material mixtures containing zirconium and/or carbon

DOEpatents

Johnson, Michael Ernest; Maloney, Martin David

2013-07-02

A method of processing spent TRIZO-coated nuclear fuel may include adding fluoride to complex zirconium present in a dissolved TRIZO-coated fuel. Complexing the zirconium with fluoride may reduce or eliminate the potential for zirconium to interfere with the extraction of uranium and/or transuranics from fission materials in the spent nuclear fuel.

40 CFR 60.2875 - What definitions must I know?

Code of Federal Regulations, 2012 CFR

2012-07-01

... burn liquid wastes material and gas (Liquid/gas),” “Energy recovery unit designed to burn solid..., liquid fuel or gaseous fuels. Energy recovery unit designed to burn liquid waste material and gas (Liquid/gas) means an energy recovery unit that burns a liquid waste with liquid or gaseous fuels not combined...

40 CFR 60.2875 - What definitions must I know?

Code of Federal Regulations, 2011 CFR

2011-07-01

... burn liquid wastes material and gas (Liquid/gas),” “Energy recovery unit designed to burn solid..., liquid fuel or gaseous fuels. Energy recovery unit designed to burn liquid waste material and gas (Liquid/gas) means an energy recovery unit that burns a liquid waste with liquid or gaseous fuels not combined...

Global Spent Fuel Logistics Systems Study (GSFLS). Volume 2A. GSFLS visit findings (appendix). Interim report

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

None

1978-01-31

This appendix is a part of the interim report documentation for the Global Spent Fuel Logistics System (GSFLS) study. This appendix provides the legal/regulatory reference material, supportive of Volume 2 - GSFLS Visit Finding and Evaluations; and certain background material on British Nuclear Fuel Limited (BNFL).

Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

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

Luther, Erik; Rooyen, Isabella van; Leckie, Rafael

2015-03-01

In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabricationmore » must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.« less

Parametric scramjet analysis

NASA Astrophysics Data System (ADS)

Choi, Jongseong

The performance of a hypersonic flight vehicle will depend on existing materials and fuels; this work presents the performance of the ideal scramjet engine for three different combustion chamber materials and three different candidate fuels. Engine performance is explored by parametric cycle analysis for the ideal scramjet as a function of material maximum service temperature and the lower heating value of jet engine fuels. The thermodynamic analysis is based on the Brayton cycle as similarly employed in describing the performance of the ramjet, turbojet, and fanjet ideal engines. The objective of this work is to explore material operating temperatures and fuel possibilities for the combustion chamber of a scramjet propulsion system to show how they relate to scramjet performance and the seven scramjet engine parameters: specific thrust, fuel-to-air ratio, thrust-specific fuel consumption, thermal efficiency, propulsive efficiency, overall efficiency, and thrust flux. The information presented in this work has not been done by others in the scientific literature. This work yields simple algebraic equations for scramjet performance which are similar to that of the ideal ramjet, ideal turbojet and ideal turbofan engines.

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

Lee, Young-Ho; Byun, Thak Sang

Accident-tolerant fuels are expected to have considerably longer coping time to respond to the loss of active cooling under severe accidents and, at the same time, have comparable or improved fuel performance during normal operation. The wear resistance of accident tolerant fuels, therefore, needs to be examined to determine the applicability of these cladding candidates to the current operating PWRs because the most common failure of nuclear fuel claddings is still caused by grid-to-rod fretting during normal operations. In this study, reciprocating sliding wear tests on three kinds of cladding candidates for accident-tolerant fuels have been performed to investigate themore » tribological compatibilities of selfmated cladding candidates and to determine the direct applicability of conventional Zirconium-based alloys as supporting structural materials. The friction coefficients of the cladding candidates are strongly influenced by the test environments and coupled materials. The wear test results under water lubrication conditions indicate that the supporting structural materials for the cladding candidates of accident-tolerant fuels need to be replaced with the same cladding materials instead of using conventional Zirconium-based alloys.« less

Issues relating to spent nuclear fuel storage on the Oak Ridge Reservation

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

Klein, J.A.; Turner, D.W.

1994-12-31

Currently, about 2,800 metric tons of spent nuclear fuel (SNF) is stored in the US, 1,000 kg of SNF (or about 0.03% of the nation`s total) are stored at the US Department of Energy (DOE) complex in Oak Ridge, Tennessee. However small the total quantity of material stored at Oak Ridge, some of the material is quite singular in character and, thus, poses unique management concerns. The various types of SNF stored at Oak Ridge will be discussed including: (1) High-Flux Isotope Reactor (HFIR) and future Advanced Neutron Source (ANS) fuels; (2) Material Testing Reactor (MTR) fuels, including Bulk Shieldingmore » Reactor (BSR) and Oak Ridge Research Reactor (ORR) fuels; (3) Molten Salt Reactor Experiment (MSRE) fuel; (4) Homogeneous Reactor Experiment (HRE) fuel; (5) Miscellaneous SNF stored in Oak Ridge National Laboratory`s (ORNL`s) Solid Waste Storage Areas (SWSAs); (6) SNF stored in the Y-12 Plant 9720-5 Warehouse including Health. Physics Reactor (HPRR), Space Nuclear Auxiliary Power (SNAP-) 10A, and DOE Demonstration Reactor fuels.« less

Solid oxide MEMS-based fuel cells

DOEpatents

Jankowksi, Alan F.; Morse, Jeffrey D.

2007-03-13

A micro-electro-mechanical systems (MEMS) based thin-film fuel cells for electrical power applications. The MEMS-based fuel cell may be of a solid oxide type (SOFC), a solid polymer type (SPFC), or a proton exchange membrane type (PEMFC), and each fuel cell basically consists of an anode and a cathode separated by an electrolyte layer. The electrolyte layer can consist of either a solid oxide or solid polymer material, or proton exchange membrane electrolyte materials may be used. Additionally catalyst layers can also separate the electrodes (cathode and anode) from the electrolyte. Gas manifolds are utilized to transport the fuel and oxidant to each cell and provide a path for exhaust gases. The electrical current generated from each cell is drawn away with an interconnect and support structure integrated with the gas manifold. The fuel cells utilize integrated resistive heaters for efficient heating of the materials. By combining MEMS technology with thin-film deposition technology, thin-film fuel cells having microflow channels and full-integrated circuitry can be produced that will lower the operating temperature an will yield an order of magnitude greater power density than the currently known fuel cells.

Solid polymer MEMS-based fuel cells

DOEpatents

Jankowski, Alan F [Livermore, CA; Morse, Jeffrey D [Pleasant Hill, CA

2008-04-22

A micro-electro-mechanical systems (MEMS) based thin-film fuel cells for electrical power applications. The MEMS-based fuel cell may be of a solid oxide type (SOFC), a solid polymer type (SPFC), or a proton exchange membrane type (PEMFC), and each fuel cell basically consists of an anode and a cathode separated by an electrolyte layer. The electrolyte layer can consist of either a solid oxide or solid polymer material, or proton exchange membrane electrolyte materials may be used. Additionally catalyst layers can also separate the electrodes (cathode and anode) from the electrolyte. Gas manifolds are utilized to transport the fuel and oxidant to each cell and provide a path for exhaust gases. The electrical current generated from each cell is drawn away with an interconnect and support structure integrated with the gas manifold. The fuel cells utilize integrated resistive heaters for efficient heating of the materials. By combining MEMS technology with thin-film deposition technology, thin-film fuel cells having microflow channels and full-integrated circuitry can be produced that will lower the operating temperature an will yield an order of magnitude greater power density than the currently known fuel cells.

Potential radiological impact of tornadoes on the safety of Nuclear Fuel Services' West Valley Fuel Reprocessing Plant. 2. Reentrainment and discharge of radioactive materials

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

Davis, W Jr

1981-07-01

This report describes results of a parametric study of quantities of radioactive materials that might be discharged by a tornado-generated depressurization on contaminated process cells within the presently inoperative Nuclear Fuel Services' (NFS) fuel reprocessing facility near West Valley, New York. The study involved the following tasks: determining approximate quantities of radioactive materials in the cells and characterizing particle-size distribution; estimating the degree of mass reentrainment from particle-size distribution and from air speed data presented in Part 1; and estimating the quantities of radioactive material (source term) released from the cells to the atmosphere. The study has shown that improperlymore » sealed manipulator ports in the Process Mechanical Cell (PMC) present the most likely pathway for release of substantial quantities of radioactive material in the atmosphere under tornado accident conditions at the facility.« less

Serially connected solid oxide fuel cells having monolithic cores

DOEpatents

Herceg, Joseph E.

1987-01-01

A solid oxide fuel cell for electrochemically combining fuel and oxidant for generating galvanic output, wherein the cell core has an array of cell segments electrically serially connected in the flow direction, each segment consisting of electrolyte walls and interconnect that are substantially devoid of any composite inert materials for support. Instead, the core is monolithic, where each electrolyte wall consists of thin layers of cathode and anode materials sandwiching a thin layer of electrolyte material therebetween. Means direct the fuel to the anode-exposed core passageways and means direct the oxidant to the cathode-exposed core passageways; and means also direct the galvanic output to an exterior circuit. Each layer of the electrolyte composite materials is of the order of 0.002-0.01 cm thick; and each layer of the cathode and anode materials is of the order of 0.002-0.05 cm thick. Between 2 and 50 cell segments may be connected in series.

Solid oxide fuel cell having monolithic cross flow core and manifolding

DOEpatents

Poeppel, Roger B.; Dusek, Joseph T.

1984-01-01

This invention discloses a monolithic core construction having the flow passageways for the fuel and for the oxidant gases extended transverse to one another, whereby full face core manifolding can be achieved for these gases and their reaction products. The core construction provides that only anode material surround each fuel passageway and only cathode material surround each oxidant passageway, each anode and each cathode further sandwiching at spaced opposing sides electrolyte and interconnect materials to define electrolyte and interconnect walls. Webs of the cathode and anode material hold the electrolyte and interconnect walls spaced apart to define the flow passages. The composite anode and cathode wall structures are further alternately stacked on one another (with the separating electrolyte or interconnect material typically being a single common layer) whereby the fuel passageway and the oxidant passageways are disposed transverse to one another.

Solid oxide fuel cell having monolithic cross flow core and manifolding

DOEpatents

Poeppel, R.B.; Dusek, J.T.

1983-10-12

This invention discloses a monolithic core construction having the flow passageways for the fuel and for the oxidant gases extended transverse to one another, whereby full face core manifolding can be achieved for these gases and their reaction products. The core construction provides that only anode material surround each fuel passageway and only cathode material surround each oxidant passageway, each anode and each cathode further sandwiching at spaced opposing sides electrolyte and interconnect materials to define electrolyte and interconnect walls. Webs of the cathode and anode material hold the electrolyte and interconnect walls spaced apart to define the flow passages. The composite anode and cathode wall structures are further alternately stacked on one another (with the separating electrolyte or interconnect material typically being a single common layer) whereby the fuel passageways and the oxidant passageways are disposed transverse to one another.

Method and apparatus for conversion of carbonaceous materials to liquid fuel

DOEpatents

Lux, Kenneth W.; Namazian, Mehdi; Kelly, John T.

2015-12-01

Embodiments of the invention relates to conversion of hydrocarbon material including but not limited to coal and biomass to a synthetic liquid transportation fuel. The invention includes the integration of a non-catalytic first reaction scheme, which converts carbonaceous materials into a solid product that includes char and ash and a gaseous product; a non-catalytic second reaction scheme, which converts a portion of the gaseous product from the first reaction scheme to light olefins and liquid byproducts; a traditional gas-cleanup operations; and the third reaction scheme to combine the olefins from the second reaction scheme to produce a targeted fuel like liquid transportation fuels.

Fuel from Tobacco and Arundo Donax: Synthetic Crop for Direct Drop-in Biofuel Production through Re-routing the Photorespiration Intermediates and Engineering Terpenoid Pathways

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

None

PETRO Project: Biofuels offer renewable alternatives to petroleum-based fuels that reduce net greenhouse gas emissions to nearly zero. However, traditional biofuels production is limited not only by the small amount of solar energy that plants convert through photosynthesis into biological materials, but also by inefficient processes for converting these biological materials into fuels. Farm-ready, non-food crops are needed that produce fuels or fuel-like precursors at significantly lower costs with significantly higher productivity. To make biofuels cost-competitive with petroleum-based fuels, biofuels production costs must be cut in half.

Calculation of Distribution Dynamics of Inhomogeneous Temperature Field in Range of Fuel Elements by Using FreeFem++

NASA Astrophysics Data System (ADS)

Amosova, E. V.; Shishkin, A. V.

2017-11-01

This article introduces the result of studying the heat exchange in the fuel element of the nuclear reactor fuel magazine. Fuel assemblies are completed as a bundle of cylindrical fuel elements located at the tops of a regular triangle. Uneven distribution of fuel rods in a nuclear reactor’s core forms the inhomogeneity of temperature fields. This article describes the developed method for heat exchange calculation with the account for impact of an inhomogeneous temperature field on the thermal-physical properties of materials and unsteady effects. The acquired calculation results are used for evaluating the tolerable temperature levels in protective case materials.

Enzyme-activatable imaging probe reveals enhanced neutrophil elastase activity in tumors following photodynamic therapy

PubMed Central

Modi, Kshitij D.; Foster, Thomas H.

2013-01-01

Abstract. We demonstrate the use of an enzyme-activatable fluorogenic probe, Neutrophil Elastase 680 FAST (NE680), for in vivo imaging of neutrophil elastase (NE) activity in tumors subjected to photodynamic therapy (PDT). NE protease activity was assayed in SCC VII and EMT6 tumors established in C3H and BALB/c mice, respectively. Four nanomoles of NE680 was injected intravenously immediately following PDT irradiation. 5 h following administration of NE680, whole-mouse fluorescence imaging was performed. At this time point, levels of NE680 fluorescence were at least threefold greater in irradiated versus unirradiated SCC VII and EMT6 tumors sensitized with Photofrin. To compare possible photosensitizer-specific differences in therapy-induced elastase activity, EMT6 tumors were also subjected to 2-(1-hexyloxyethyl)-2-devinyl pyropheophorbide-a (HPPH)-PDT. NE levels measured in HPPH-PDT-treated tumors were twofold higher than in unirradiated controls. Ex vivo labeling of host cells using fluorophore-conjugated antibodies and confocal imaging were used to visualize Gr1+ cells in Photofrin-PDT-treated EMT6 tumors. These data were compared with recently reported analysis of Gr1+ cell accumulation in EMT6 tumors subjected to HPPH-PDT. The population density of infiltrating Gr1+ cells in treated versus unirradiated drug-only control tumors suggests that the differential in NE680 fold enhancement observed in Photofrin versus HPPH treatment may be attributed to the significantly increased inflammatory response induced by Photofrin-PDT. The in vivo imaging of NE680, which is a fluorescent reporter of NE extracellular release caused by neutrophil activation, demonstrates that PDT results in increased NE levels in treated tumors, and the accumulation of the cleaved probe tracks qualitatively with the intratumor Gr1+ cell population. PMID:23897439

Solid oxide fuel cell having monolithic core

DOEpatents

Ackerman, John P.; Young, John E.

1984-01-01

A solid oxide fuel cell for electrochemically combining fuel and oxidant for generating galvanic output, wherein the cell core has an array of electrolyte and interconnect walls that are substantially devoid of any composite inert materials for support. Instead, the core is monolithic, where each electrolyte wall consists of thin layers of cathode and anode materials sandwiching a thin layer of electrolyte material therebetween, and each interconnect wall consists of thin layers of the cathode and anode materials sandwiching a thin layer of interconnect material therebetween. The electrolyte walls are arranged and backfolded between adjacent interconnect walls operable to define a plurality of core passageways alternately arranged where the inside faces thereof have only the anode material or only the cathode material exposed. Means direct the fuel to the anode-exposed core passageways and means direct the oxidant to the cathode-exposed core passageway; and means also direct the galvanic output to an exterior circuit. Each layer of the electrolyte and interconnect materials is of the order of 0.002-0.01 cm thick; and each layer of the cathode and anode materials is of the order of 0.002-0.05 cm thick.

Solar energy for electricity and fuels.

PubMed

Inganäs, Olle; Sundström, Villy

2016-01-01

Solar energy conversion into electricity by photovoltaic modules is now a mature technology. We discuss the need for materials and device developments using conventional silicon and other materials, pointing to the need to use scalable materials and to reduce the energy payback time. Storage of solar energy can be achieved using the energy of light to produce a fuel. We discuss how this can be achieved in a direct process mimicking the photosynthetic processes, using synthetic organic, inorganic, or hybrid materials for light collection and catalysis. We also briefly discuss challenges and needs for large-scale implementation of direct solar fuel technologies.

Advanced catalyst supports for PEM fuel cell cathodes

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

Du, Lei; Shao, Yuyan; Sun, Junming

2016-11-01

Electrocatalyst support materials are key components for polymer exchange membrane (PEM) fuel cells, which play a critical role in determining electrocatalyst durability and activity, mass transfer and water management. The commonly-used supports, e.g. porous carbon black, cannot meet all the requirements under the harsh operation condition of PEM fuel cells. Great efforts have been made in the last few years in developing alternative support materials. In this paper, we selectively review recent progress on three types of important support materials: carbon, non-carbon and hybrid carbon-oxides nanocomposites. A perspective on future R&D of electrocatalyst support materials is also provided.

Materials Requirements for Advanced Energy Systems - New Fuels. Volume 3: Materials Research Needs in Advanced Energy Systems Using New Fuels

DTIC Science & Technology

1974-07-01

elec- Materials se: trode materials and associ- operational ated conductors. 2.5.1 General. H" (02) Materials resources Technoeconomic analysis - None...Advanced Energy Systems Using New Fnels VIII Correlation and Analysis of Materials Requirements IX Research Recommendations and Priorities The authois...of government and industrial organizal ions who gave us the benefit of their knowledge and experience. iv VIII CORRELATION ANU ANALYSIS OF MATERIALS

Air-Cooled Stack Freeze Tolerance Freeze Failure Modes and Freeze Tolerance Strategies for GenDriveTM Material Handling Application Systems and Stacks Final Scientific Report

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

Hancock, David, W.

2012-02-14

Air-cooled stack technology offers the potential for a simpler system architecture (versus liquid-cooled) for applications below 4 kilowatts. The combined cooling and cathode air allows for a reduction in part count and hence a lower cost solution. However, efficient heat rejection challenges escalate as power and ambient temperature increase. For applications in ambient temperatures below freezing, the air-cooled approach has additional challenges associated with not overcooling the fuel cell stack. The focus of this project was freeze tolerance while maintaining all other stack and system requirements. Through this project, Plug Power advanced the state of the art in technology formore » air-cooled PEM fuel cell stacks and related GenDrive material handling application fuel cell systems. This was accomplished through a collaborative work plan to improve freeze tolerance and mitigate freeze-thaw effect failure modes within innovative material handling equipment fuel cell systems designed for use in freezer forklift applications. Freeze tolerance remains an area where additional research and understanding can help fuel cells to become commercially viable. This project evaluated both stack level and system level solutions to improve fuel cell stack freeze tolerance. At this time, the most cost effective solutions are at the system level. The freeze mitigation strategies developed over the course of this project could be used to drive fuel cell commercialization. The fuel cell system studied in this project was Plug Power's commercially available GenDrive platform providing battery replacement for equipment in the material handling industry. The fuel cell stacks were Ballard's commercially available FCvelocity 9SSL (9SSL) liquid-cooled PEM fuel cell stack and FCvelocity 1020ACS (Mk1020) air-cooled PEM fuel cell stack.« less

The Manufacture of W-UO2 Fuel Elements for NTP Using the Hot Isostatic Pressing Consolidation Process

NASA Technical Reports Server (NTRS)

Broadway, Jeramie; Hickman, Robert; Mireles, Omar

2012-01-01

NTP is attractive for space exploration because: (1) Higher Isp than traditional chemical rockets (2)Shorter trip times (3) Reduced propellant mass (4) Increased payload. Lack of qualified fuel material is a key risk (cost, schedule, and performance). Development of stable fuel form is a critical path, long lead activity. Goals of this project are: Mature CERMET and Graphite based fuel materials and Develop and demonstrate critical technologies and capabilities.

Study and program plan for improved heavy duty gas turbine engine ceramic component development

NASA Technical Reports Server (NTRS)

Helms, H. E.

1977-01-01

Fuel economy in a commercially viable gas turbine engine was demonstrated through use of ceramic materials. Study results show that increased turbine inlet and generator inlet temperatures, through the use of ceramic materials, contribute the greatest amount to achieving fuel economy goals. Improved component efficiencies show significant additional gains in fuel economy.

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

Johnson, D.R.

The purpose of the Heavy Vehicle Propulsion System Materials Program is the development of materials: ceramics, intermetallics, metal alloys, and metal and ceramic coatings, to support the dieselization of class 1-3 trucks to realize a 35% fuel-economy improvement over current gasoline-fueled trucks and to support commercialization of fuel-flexible LE-55 low-emissions, high-efficiency diesel engines for class 7-8 trucks.

Solid oxide fuel cell having monolithic core

DOEpatents

Ackerman, J.P.; Young, J.E.

1983-10-12

A solid oxide fuel cell is described for electrochemically combining fuel and oxidant for generating galvanic output, wherein the cell core has an array of electrolyte and interconnect walls that are substantially devoid of any composite inert materials for support. Instead, the core is monolithic, where each electrolyte wall consists of thin layers of cathode and anode materials sandwiching a thin layer of electrolyte material therebetween. The electrolyte walls are arranged and backfolded between adjacent interconnect walls operable to define a plurality of core passageways alternately arranged where the inside faces thereof have only the anode material or only the cathode material exposed. Means direct the fuel to the anode-exposed core passageways and means direct the oxidant to the anode-exposed core passageways and means direct the oxidant to the cathode-exposed core passageway; and means also direct the galvanic output to an exterior circuit. Each layer of the electrolyte and interconnect materials is of the order of 0.002 to 0.01 cm thick; and each layer of the cathode and anode materials is of the order of 0.002 to 0.05 cm thick.

46 CFR 182.720 - Nonmetallic piping materials.

Code of Federal Regulations, 2013 CFR

2013-10-01

... braid; (iv) Flexible hose used for alcohol-gasoline blend fuels must meet the permeability requirements...) and (d) of this section. (b) Flexible nonmetallic materials (hose) may be used in vital and non-vital... gasoline or diesel fuel systems. Flexible nonmetallic materials (hose) may be used where permitted by...

46 CFR 182.720 - Nonmetallic piping materials.

Code of Federal Regulations, 2011 CFR

2011-10-01

... braid; (iv) Flexible hose used for alcohol-gasoline blend fuels must meet the permeability requirements...) and (d) of this section. (b) Flexible nonmetallic materials (hose) may be used in vital and non-vital... gasoline or diesel fuel systems. Flexible nonmetallic materials (hose) may be used where permitted by...

46 CFR 182.720 - Nonmetallic piping materials.

Code of Federal Regulations, 2012 CFR

2012-10-01

... braid; (iv) Flexible hose used for alcohol-gasoline blend fuels must meet the permeability requirements...) and (d) of this section. (b) Flexible nonmetallic materials (hose) may be used in vital and non-vital... gasoline or diesel fuel systems. Flexible nonmetallic materials (hose) may be used where permitted by...

46 CFR 182.720 - Nonmetallic piping materials.

Code of Federal Regulations, 2014 CFR

2014-10-01

... braid; (iv) Flexible hose used for alcohol-gasoline blend fuels must meet the permeability requirements...) and (d) of this section. (b) Flexible nonmetallic materials (hose) may be used in vital and non-vital... gasoline or diesel fuel systems. Flexible nonmetallic materials (hose) may be used where permitted by...

46 CFR 182.720 - Nonmetallic piping materials.

Code of Federal Regulations, 2010 CFR

2010-10-01

... braid; (iv) Flexible hose used for alcohol-gasoline blend fuels must meet the permeability requirements...) and (d) of this section. (b) Flexible nonmetallic materials (hose) may be used in vital and non-vital... gasoline or diesel fuel systems. Flexible nonmetallic materials (hose) may be used where permitted by...

Identification of gamma-irradiated foodstuffs by chemiluminescence measurements in Taiwan

NASA Astrophysics Data System (ADS)

Ma, Ming-Shia Chang; Chen, Li-Hsiang; Tsai, Zei-Tsan; Fu, Ying-Kai

In order to establish chemiluminescence (CL) measurements as an identification method for γ-irradiated foodstuffs in Taiwan, ten agricultural products including wheat flour, rice, ginger, potatoes, garlic, onions, red beans, mung beans, soy beans, xanthoxylon seeds and Japanese star anises have been tested to compare CL intensities between untreated samples and samples subject to a 10 kGy γ-irradiation dose. Amongst them, wheat flour is the most eligible product to be identified by CL measurements. The CL intensities of un-irradiated and irradiated flour have shown large differences associated with a significant dose-effect relationship. Effects of three different protein contents of flour, unsieved and sieved (100-200 mesh), the reproducibility and the storage experiment on CL intensities at various doses were investigated in this study. In addition, the white bulb part of onions has shown some CL in irradiated samples. The CL data obtained from the other eight agricultural products have shown large fluctuations and cannot be used to differentiate between irradiated and un-irradiated samples.

Some cell kinetic effects of combined injury with ionizing radiation and cyclophosphamide on mouse bladder urothelium.

PubMed

Reitan, J B

1985-01-01

Cyclophosphamide was given intraperitoneally to groups of eight female mice 48 h after local electron irradiation to the bladder with 0, 10 and 20 Gy respectively. The reactions in the urothelium were monitored by histology, incorporation of tritiated thymidine and flow cytometry. A wave of increased thymidine incorporation combined with an increase in the proportion of diploid S-phase cells was seen in the unirradiated bladders 24 h after the drug treatment, followed by normalization after 1 week. This response was significantly less pronounced in the irradiated animals. In the unirradiated animals a similar wave characterized by an increased proportion of octaploid cells was also seen, but this wave occurred later in the irradiated animals. Severe injury was observed in the rectum of the 20 Gy-irradiated animals. Irradiation prior to drug treatment led to only small effects, but a decreased ability for regenerative DNA synthesis after drug injury seems to persist. This affects both proliferation and the building up of polyploidy.

[Identification of irradiated abalone by ESR spectroscopy].

PubMed

Song, Yeping; Wang, Chuanxian; Yang, Zhenyu; Zhong, Weike; Geng, Jinpei; Lu, Di; Ding, Zhuoping

2012-05-01

To establish an analytical method for the detection and identification of irradiated abalone by electron spin resonance spectroscopy. Electron spin resonance (ESR) was used to study the spectral characteristics of abalone and the characteristic peak for quantitation. There were obvious different ESR spectra between unirradiated and irradiated abalone. The g factor for unirradiated abalone was 2.0055-2.0060, the g1 and g2 factor for irradiated abalone were (2.0027 +/- 0.0001) and (1.9994 +/- 0.0001), respectively. The ESR signal intensity of characteristic peak was positively correlated with absorbed dose in the range of 0.5 - 10 kGy, left peak was the characteristic peak for quantitation and the detection limit was < or = 0.5 kGy. It was difficult to quantitate when the absorbed dose was over 10 kGy. ESR characteristic peak and g factor were able to qualitatively determine the irradiation of abalone. ESR spectroscopy is an effective method to determine whether the abalone being irradiated or not.

RBS/Channeling Studies of Swift Heavy Ion Irradiated GaN Layers

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

Sathish, N.; Dhamodaran, S.; Pathak, A. P.

2009-03-10

Epitaxial GaN layers grown by MOCVD on c-plane sapphire substrates were irradiated with 150 MeV Ag ions at a fluence of 5x10{sup 12} ions/cm{sup 2}. Samples used in this study are 2 {mu}m thick GaN layers, with and without a thin AlN cap-layer. Energy dependent RBS/Channeling measurements have been carried out on both irradiated and unirradiated samples for defects characterization. Observed results are compared and correlated with previous HRXRD, AFM and optical studies. The {chi}{sub min} values for unirradiated samples show very high value and the calculated defect densities are of the order of 10{sup 10} cm{sup -2} as expectedmore » in these samples. Effects of irradiation on these samples are different as initial samples had different defect densities. Epitaxial reconstruction of GaN buffer layer has been attributed to the observed changes, which are generally grown to reduce the strain between GaN and Sapphire.« less

FY-16 Technology Gap Study Technical Report: Analysis of Undissolved Anode Materials of Mark-IV Electrorefiner

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

Yoo, Tae-Sic; Vaden, DeeEarl; Westphal, Brian Robert

2016-01-01

The Experimental Breeder Reactor II (EBR-II) is a sodium cooled fast reactor developed at Argonne National Laboratory (ANL). The used fuels from the EBR-II are currently being treated in the Fuel Conditioning Facility (FCF) at the Idaho National Laboratory (INL). The Mark IV (Mk-IV) electrorefiner (ER) is a unit process in the FCF, which is primarily assigned to treating the used driver fuels. The stainless steel anode baskets hold the chopped spent driver fuel segments. During electrorefining, the anode baskets are immersed into the electrolyte and the used fuel is dissolved electrochemically. Perforated sides and bottoms allow the flow ofmore » the electrolyte into and out of the anode baskets. The steel cathode is also immersed into the electrolyte and collects the reduced products. The active metal contents in the used fuel (e.g., Cs, Sr, lanthanides, Pu, etc.) reacts with uranium cations in the electrolyte and progressively reports to the electrolyte. Noble metals are mostly retained in the cladding hulls. Varying quantities of zirconium are retained in the cladding hulls depending on the operational conditions of the Mk-IV ER. The undissolved anode materials are removed from the anode baskets and stored for subsequent metal waste form processing. These undissolved materials typically include undissolved fuels, stainless steel cladding, and adhering electrolyte. A couple of hulls are retrieved for chemical analysis and used for estimating the composition of the entire undissolved anode materials. The mass balance attempt based on this practice of estimating the undissolved anode materials has been a challenge due to inherently high sampling errors associated with heterogeneous undissolved material compositions. Responding to the prescribed challenge, this report investigates chemical analysis data as a whole and finds noticeable trends in the compositions of undissolved anode material samples with respect to the mass of the whole undissolved anode materials. Based upon this discovery, an empirical model is proposed.« less

Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

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

Bradley K. Heath

2014-03-01

This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show thatmore » fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.« less

Next generation fuel irradiation capability in the High Flux Reactor Petten

NASA Astrophysics Data System (ADS)

Fütterer, Michael A.; D'Agata, Elio; Laurie, Mathias; Marmier, Alain; Scaffidi-Argentina, Francesco; Raison, Philippe; Bakker, Klaas; de Groot, Sander; Klaassen, Frodo

2009-07-01

This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.

The TMI Regenerative Solid Oxide Fuel Cell

NASA Technical Reports Server (NTRS)

Cable, Thomas L.; Ruhl, Robert C.; Petrik, Michael

1996-01-01

Energy storage and production in space requires rugged, reliable hardware which minimizes weight, volume, and maintenance while maximizing power output and usable energy storage. Systems generally consist of photovoltaic solar arrays which operate (during sunlight cycles) to provide system power and regenerate fuel (hydrogen) via water electrolysis and (during dark cycles) fuel cells convert hydrogen into electricity. Common configurations use two separate systems (fuel cell and electrolyzer) in conjunction with photovoltaic cells. Reliability, power to weight and power to volume ratios could be greatly improved if both power production (fuel cells) and power storage (electrolysis) functions can be integrated into a single unit. The solid oxide fuel cell (SOFC) based design integrates fuel cell and electrolyzer functions and potentially simplifies system requirements. The integrated fuel cell/electrolyzer design also utilizes innovative gas storage concepts and operates like a rechargeable 'hydrogen-oxygen battery'. Preliminary research has been completed on improved H2/H20 electrode (SOFC anode/electrolyzer cathode) materials for regenerative fuel cells. Tests have shown improved cell performance in both fuel and electrolysis modes in reversible fuel cell tests. Regenerative fuel cell efficiencies, ratio of power out (fuel cell mode) to power in (electrolyzer mode), improved from 50 percent using conventional electrode materials to over 80 percent. The new materials will allow a single SOFC system to operate as both the electolyzer and fuel cell. Preliminary system designs have also been developed to show the technical feasibility of using the design for space applications requiring high energy storage efficiencies and high specific energy. Small space systems also have potential for dual-use, terrestrial applications.

Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

NASA Astrophysics Data System (ADS)

Karahan, Aydın; Kazimi, Mujid S.

2013-10-01

The study evaluates the possible use of graphite foam as the bonding material between U-Pu-Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U-15Pu-6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600-660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

Alternative Fuels Data Center: Natural Gas Fuel Basics

Science.gov Websites

-derived natural gas, renewable natural gas-which is produced from decaying organic materials-must be on organic materials. Alternatively, renewable natural gas (RNG), also known as biomethane, is produced from organic materials-such as waste from landfills and livestock-through anaerobic digestion. RNG

JHR Project: a future Material Testing Reactor working as an International user Facility: The key-role of instrumentation in support to the development of modern experimental capacity

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

Bignan, G.; Gonnier, C.; Lyoussi, A.

2015-07-01

Research and development on fuel and material behaviour under irradiation is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. These needs mainly deal with a constant improvement of performances and safety in order to optimize the fuel cycle and hence to reach nuclear energy sustainable objectives. A sustainable nuclear energy requires a high level of performances in order to meet specific needs such as: - Pursuing improvement of the performances and safety of present and coming water cooled reactor technologies. This will require a continuous R and Dmore » support following a long-term trend driven by the plant life management, safety demonstration, flexibility and economics improvement. Experimental irradiations of structure materials are necessary to anticipate these material behaviours and will contribute to their optimisation. - Upgrading continuously nuclear fuel technology in present and future nuclear power plants to achieve better performances and to optimise the fuel cycle keeping the best level of safety. Fuel evolution for generation II, III and III+ is a key stake requiring developments, qualification tests and safety experiments to ensure the competitiveness and safety: experimental tests exploring the full range of fuel behaviour determine fuel stability limits and safety margins, as a major input for the fuel reliability analysis. To perform such accurate and innovative progress and developments, specific and ad hoc instrumentation, irradiation devices, measurement methods are necessary to be set up inside or beside the material testing reactor (MTR) core. These experiments require beforehand in situ and on line sophisticated measurements to accurately determine different key parameters such as thermal and fast neutron fluxes and nuclear heating in order to precisely monitor and control the conducted assays. The new Material Testing Reactor JHR (Jules Horowitz Reactor) currently under construction at CEA Cadarache research centre in the south of France will represent a major Research Infrastructure for scientific studies regarding material and fuel behavior under irradiation. It will also be devoted to medical isotopes production. Hence JHR will offer a real opportunity to perform R and D programs regarding needs above and hence will crucially contribute to the selection, optimization and qualification of these innovative materials and fuels. The JHR reactor objectives, principles and main characteristics associated to specific experimental devices associated to measurement techniques and methodology, their performances, their limitations and field of applications will be presented and discussed. (authors)« less

Laser-fusion targets for reactors

DOEpatents

Nuckolls, John H.; Thiessen, Albert R.

1987-01-01

A laser target comprising a thermonuclear fuel capsule composed of a centrally located quantity of fuel surrounded by at least one or more layers or shells of material for forming an atmosphere around the capsule by a low energy laser prepulse. The fuel may be formed as a solid core or hollow shell, and, under certain applications, a pusher-layer or shell is located intermediate the fuel and the atmosphere forming material. The fuel is ignited by symmetrical implosion via energy produced by a laser, or other energy sources such as an electron beam machine or ion beam machine, whereby thermonuclear burn of the fuel capsule creates energy for applications such as generation of electricity via a laser fusion reactor.

Molten salt destruction of energetic waste materials

DOEpatents

Brummond, W.A.; Upadhye, R.S.; Pruneda, C.O.

1995-07-18

A molten salt destruction process is used to treat and destroy energetic waste materials such as high explosives, propellants, and rocket fuels. The energetic material is pre-blended with a solid or fluid diluent in safe proportions to form a fluid fuel mixture. The fuel mixture is rapidly introduced into a high temperature molten salt bath. A stream of molten salt is removed from the vessel and may be recycled as diluent. Additionally, the molten salt stream may be pumped from the reactor, circulated outside the reactor for further processing, and delivered back into the reactor or cooled and circulated to the feed delivery system to further dilute the fuel mixture entering the reactor. 4 figs.

Molten salt destruction of energetic waste materials

DOEpatents

Brummond, William A.; Upadhye, Ravindra S.; Pruneda, Cesar O.

1995-01-01

A molten salt destruction process is used to treat and destroy energetic waste materials such as high explosives, propellants, and rocket fuels. The energetic material is pre-blended with a solid or fluid diluent in safe proportions to form a fluid fuel mixture. The fuel mixture is rapidly introduced into a high temperature molten salt bath. A stream of molten salt is removed from the vessel and may be recycled as diluent. Additionally, the molten salt stream may be pumped from the reactor, circulated outside the reactor for further processing, and delivered back into the reactor or cooled and circulated to the feed delivery system to further dilute the fuel mixture entering the reactor.

Technology readiness levels for advanced nuclear fuels and materials development

DOE PAGES

Carmack, W. J.; Braase, L. A.; Wigeland, R. A.; ...

2016-12-23

The Technology Readiness Level (TRL) process is used to quantitatively assess the maturity of a given technology. It was pioneered by the National Aeronautics and Space Administration (NASA) in the 1980s to develop and deploy new systems for space applications. The process was subsequently adopted by the Department of Defense (DoD) to develop and deploy new technology and systems for defense applications as well as the Department of Energy (DOE) to evaluate the maturity of new technologies in major construction projects. Advanced nuclear fuels and materials development is a critical technology needed for improving the performance and safety of currentmore » and advanced reactors, and ultimately closing the nuclear fuel cycle. Because deployment of new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the TRL concept to the advanced fuel development program is very useful as a management, communication and tracking tool. Furthermore, this article provides examples regarding the methods by which TRLs are currently used to assess the maturity of nuclear fuels and materials under development in the DOE Fuel Cycle Research and Development (FCRD) Program within the Advanced Fuels Campaign (AFC).« less

Technology readiness levels for advanced nuclear fuels and materials development

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

Carmack, W. J.; Braase, L. A.; Wigeland, R. A.

The Technology Readiness Level (TRL) process is used to quantitatively assess the maturity of a given technology. It was pioneered by the National Aeronautics and Space Administration (NASA) in the 1980s to develop and deploy new systems for space applications. The process was subsequently adopted by the Department of Defense (DoD) to develop and deploy new technology and systems for defense applications as well as the Department of Energy (DOE) to evaluate the maturity of new technologies in major construction projects. Advanced nuclear fuels and materials development is a critical technology needed for improving the performance and safety of currentmore » and advanced reactors, and ultimately closing the nuclear fuel cycle. Because deployment of new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the TRL concept to the advanced fuel development program is very useful as a management, communication and tracking tool. Furthermore, this article provides examples regarding the methods by which TRLs are currently used to assess the maturity of nuclear fuels and materials under development in the DOE Fuel Cycle Research and Development (FCRD) Program within the Advanced Fuels Campaign (AFC).« less

Overview of lower length scale model development for accident tolerant fuels regarding U3Si2 fuel and FeCrAl cladding

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

Zhang, Yongfeng

2016-09-01

U3Si2 and FeCrAl have been proposed as fuel and cladding concepts, respectively, for accident tolerance fuels with higher tolerance to accident scenarios compared to UO2. However, a lot of key physics and material properties regarding their in-pile performance are yet to be explored. To accelerate the understanding and reduce the cost of experimental studies, multiscale modeling and simulation are used to develop physics-based materials models to assist engineering scale fuel performance modeling. In this report, the lower-length-scale efforts in method and material model development supported by the Accident Tolerance Fuel (ATF) high-impact-problem (HIP) under the NEAMS program are summarized. Significantmore » progresses have been made regarding interatomic potential, phase field models for phase decomposition and gas bubble formation, and thermal conductivity for U3Si2 fuel, and precipitation in FeCrAl cladding. The accomplishments are very useful by providing atomistic and mesoscale tools, improving the current understanding, and delivering engineering scale models for these two ATF concepts.« less

Gas turbine critical research and advanced technology (CRT) support project

NASA Technical Reports Server (NTRS)

Furman, E. R.; Anderson, D. N.; Gedwill, M. A.; Lowell, C. E.; Schultz, D. F.

1982-01-01

The technical progress to provide a critical technology base for utility gas turbine systems capable of burning coal-derived fuels is summarized. Project tasks include the following: (1) combustion - to investigate the combustion of coal-derived fuels and the conversion of fuel-bound nitrogen to NOx; (2) materials - to understand and prevent the hot corrosion of turbine hot section materials; and (3) system studies - to integrate and guide the technological efforts. Technical accomplishments include: an extension of flame tube combustion testing of propane - Toluene Fuel Mixtures to vary H2 content from 9 to 18 percent by weight and the comparison of results with that predicted from a NASA Lewis General Chemical Kinetics Computer Code; the design and fabrication of combustor sector test section to test current and advanced combustor concepts; Testing of Catalytic combustors with residual and coal-derived liquid fuels; testing of high strength super alloys to evaluate their resistance to potential fuel impurities using doped clean fuels and coal-derived liquids; and the testing and evaluation of thermal barrier coatings and bond coatings on conventional turbine materials.

10 CFR 70.20b - General license for carriers of transient shipments of formula quantities of strategic special...

Code of Federal Regulations, 2010 CFR

2010-01-01

... significance, special nuclear material of low strategic significance, and irradiated reactor fuel. 70.20b..., special nuclear material of low strategic significance, and irradiated reactor fuel. (a) A general license... requirements of § 73.67 of this chapter. (3) Irradiated reactor fuel of the type and quantity subject to the...

10 CFR 70.20b - General license for carriers of transient shipments of formula quantities of strategic special...

Code of Federal Regulations, 2011 CFR

2011-01-01

... significance, special nuclear material of low strategic significance, and irradiated reactor fuel. 70.20b..., special nuclear material of low strategic significance, and irradiated reactor fuel. (a) A general license... requirements of § 73.67 of this chapter. (3) Irradiated reactor fuel of the type and quantity subject to the...

Analysis of Underground Storage Tanks System Materials to Increased Leak Potential Associated with E15 Fuel

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

Kass, Michael D; Theiss, Timothy J; Janke, Christopher James

2012-07-01

The Energy Independence and Security Act (EISA) of 2007 was enacted by Congress to move the nation toward increased energy independence by increasing the production of renewable fuels to meet its transportation energy needs. The law establishes a new renewable fuel standard (RFS) that requires the nation to use 36 billion gallons annually (2.3 million barrels per day) of renewable fuel in its vehicles by 2022. Ethanol is the most widely used renewable fuel in the US, and its production has grown dramatically over the past decade. According to EISA and RFS, ethanol (produced from corn as well as cellulosicmore » feedstocks) will make up the vast majority of the new renewable fuel requirements. However, ethanol use limited to E10 and E85 (in the case of flex fuel vehicles or FFVs) will not meet this target. Even if all of the E0 gasoline dispensers in the country were converted to E10, such sales would represent only about 15 billion gallons per year. If 15% ethanol, rather than 10% were used, the potential would be up to 22 billion gallons. The vast majority of ethanol used in the United States is blended with gasoline to create E10, that is, gasoline with up to 10% ethanol. The remaining ethanol is sold in the form of E85, a gasoline blend with as much as 85% ethanol that can only be used in FFVs. Although DOE remains committed to expanding the E85 infrastructure, that market will not be able to absorb projected volumes of ethanol in the near term. Given this reality, DOE and others have begun assessing the viability of using intermediate ethanol blends as one way to transition to higher volumes of ethanol. In October of 2010, the EPA granted a partial waiver to the Clean Air Act allowing the use of fuel that contains up to 15% ethanol for the model year 2007 and newer light-duty motor vehicles. This waiver represents the first of a number of actions that are needed to move toward the commercialization of E15 gasoline blends. On January 2011, this waiver was expanded to include model year 2001 light-duty vehicles, but specifically prohibited use in motorcycles and off-road vehicles and equipment. UST stakeholders generally consider fueling infrastructure materials designed for use with E0 to be adequate for use with E10, and there are no known instances of major leaks or failures directly attributable to ethanol use. It is conceivable that many compatibility issues, including accelerated corrosion, do arise and are corrected onsite and, therefore do not lead to a release. However, there is some concern that higher ethanol concentrations, such as E15 or E20, may be incompatible with current materials used in standard gasoline fueling hardware. In the summer of 2008, DOE recognized the need to assess the impact of intermediate blends of ethanol on the fueling infrastructure, specifically located at the fueling station. This includes the dispenser and hanging hardware, the underground storage tank, and associated piping. The DOE program has been co-led and funded by the Office of the Biomass Program and Vehicle Technologies Program with technical expertise from the Oak Ridge National Laboratory (ORNL) and the National Renewable Energy Laboratory (NREL). The infrastructure material compatibility work has been supported through strong collaborations and testing at Underwriters Laboratories (UL). ORNL performed a compatibility study investigating the compatibility of fuel infrastructure materials to gasoline containing intermediate levels of ethanol. These results can be found in the ORNL report entitled Intermediate Ethanol Blends Infrastructure Materials Compatibility Study: Elastomers, Metals and Sealants (hereafter referred to as the ORNL intermediate blends material compatibility study). These materials included elastomers, plastics, metals and sealants typically found in fuel dispenser infrastructure. The test fuels evaluated in the ORNL study were SAE standard test fuel formulations used to assess material-fuel compatibility within a relatively short timeframe. Initially, these material studies included test fuels of Fuel C, CE10a, CE17a, and CE25a. The CE17a test fuel was selected to represent E15 since surveys have shown that the actual ethanol upper limit can be as high as 17%. Later, CE50a and CE85a test fuels were added to the investigation and these results are being compiled for a follow-on report to be published in 2012. Fuel C was used as the baseline reference and is a 50:50 blend of isooctane and toluene. This particular composition was used to represent premium-grade gasoline and was also used as the base fuel for the ethanol blends, where it is denoted by 'C' in the fuel name. The level of ethanol is represented by the number following the letter E. Therefore a 10% blend of ethanol in Fuel C is written as CE10a, where 'a' represents an aggressive formulation of the ethanol that contains water, NaCl, acetic and sulfuric acids per the SAE J1681 protocol.« less

New materials for polymer electrolyte membrane fuel cell current collectors

NASA Astrophysics Data System (ADS)

Hentall, Philip L.; Lakeman, J. Barry; Mepsted, Gary O.; Adcock, Paul L.; Moore, Jon M.

Polymer Electrolyte Membrane Fuel cells for automotive applications need to have high power density, and be inexpensive and robust to compete effectively with the internal combustion engine. Development of membranes and new electrodes and catalysts have increased power significantly, but further improvements may be achieved by the use of new materials and construction techniques in the manufacture of the bipolar plates. To show this, a variety of materials have been fabricated into flow field plates, both metallic and graphitic, and single fuel cell tests were conducted to determine the performance of each material. Maximum power was obtained with materials which had lowest contact resistance and good electrical conductivity. The performance of the best material was characterised as a function of cell compression and flow field geometry.

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

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

Folsom, Charles; Xing, Changhu; Jensen, Colby

2015-03-01

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

FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Loeb, E.; Nicklas, J.H.

1959-02-01

A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

Low Activation Joining of SiC/SiC Composites for Fusion Applications: Modeling Miniature Torsion Tests with Elastic and Elastic-Plastic Models

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

Henager, Charles H.; Nguyen, Ba Nghiep; Kurtz, Richard J.

2015-06-30

The international fusion community designed miniature torsion specimens for joint testing and irradiation in test reactors with limited irradiation volumes since SiC and SiC-composites used in fission or fusion environments require joining methods for assembling systems. Torsion specimens fail out-of-plane when joints are strong and when elastic moduli are comparable to SiC, which causes difficulties in determining shear strengths for many joints or for comparing unirradiated and irradiated joints. A finite element damage model was developed to treat elastic joints such as SiC/Ti3SiC2+SiC and elastic-plastic joints such as SiC/epoxy and steel/epoxy. The model uses constitutive shear data and is validatedmore » using epoxy joint data. The elastic model indicates fracture is likely to occur within the joined pieces to cause out-of-plane failures for miniature torsion specimens when a certain modulus and strength ratio between the joint material and the joined material exists. Lower modulus epoxy joints always fail in plane and provide good model validation.« less

The Charpy impact properties of martensitic 10.6% Cr steel (MANET-1) before and after neutron exposure

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

Rieth, M.; Dafferner, B.; Rohrig, H.D.

1994-12-31

The MANET-I martensitic 10.6% Cr type of steel was developed as a potential structural material for the first wall and the blanket of a fusion device within the framework of the Nuclear Fusion Project. An extensive irradiation program (FRUST/SIENA) was elaborated to study the influence of radiation upon the Charpy impact characteristics. In addition to unirradiated reference specimens, 87 irradiated subsize Charpy specimens (3 x 4 x 27 mm{sup 3}) were examined under eight different heat treatments at irradiation temperatures between 287{degrees}C and 475{degrees}C and exposure doses of 5 dpa to 15 dpa. On the basis of the numerous testmore » results and their interpretation it is possible to describe radiation induced material embrittlement, and, consequently, the deterioration of the Charpy impact properties. The description is limited, on the one hand, by the variations in the test results and, on the other hand, by the gaps in the test matrix. Therefore, additional investigations, especially in the low irradiation temperature and low dose regimes will be the subject of further ongoing work.« less

Laser Plasma Microthruster Performance Evaluation

NASA Astrophysics Data System (ADS)

Luke, James R.; Phipps, Claude R.

2003-05-01

The micro laser plasma thruster (μLPT) is a sub-kilogram thruster that is capable of meeting the Air Force requirements for the Attitude Control System on a 100-kg class small satellite. The μLPT uses one or more 4W diode lasers to ablate a solid fuel, producing a jet of hot gas or plasma which creates thrust with a high thrust/power ratio. A pre-prototype continuous thrust experiment has been constructed and tested. The continuous thrust experiment uses a 505 mm long continuous loop fuel tape, which consists of a black laser-absorbing fuel material on a transparent plastic substrate. When the laser is operated continuously, the exhaust plume and thrust vector are steered in the direction of the tape motion. Thrust steering can be avoided by pulsing the laser. A torsion pendulum thrust stand has been constructed and calibrated. Many fuel materials and substrates have been tested. Best performance from a non-energetic fuel material was obtained with black polyvinyl chloride (PVC), which produced an average of 70 μN thrust and coupling coefficient (Cm) of 190 μN/W. A proprietary energetic material was also tested, in which the laser initiates a non-propagating detonation. This material produced 500 μN of thrust.

MEMS-based fuel cells with integrated catalytic fuel processor and method thereof

DOEpatents

Jankowski, Alan F [Livermore, CA; Morse, Jeffrey D [Martinez, CA; Upadhye, Ravindra S [Pleasanton, CA; Havstad, Mark A [Davis, CA

2011-08-09

Described herein is a means to incorporate catalytic materials into the fuel flow field structures of MEMS-based fuel cells, which enable catalytic reforming of a hydrocarbon based fuel, such as methane, methanol, or butane. Methods of fabrication are also disclosed.

Design and testing of digitally manufactured paraffin Acrylonitrile-butadiene-styrene hybrid rocket motors

NASA Astrophysics Data System (ADS)

McCulley, Jonathan M.

This research investigates the application of additive manufacturing techniques for fabricating hybrid rocket fuel grains composed of porous Acrylonitrile-butadiene-styrene impregnated with paraffin wax. The digitally manufactured ABS substrate provides mechanical support for the paraffin fuel material and serves as an additional fuel component. The embedded paraffin provides an enhanced fuel regression rate while having no detrimental effect on the thermodynamic burn properties of the fuel grain. Multiple fuel grains with various ABS-to-Paraffin mass ratios were fabricated and burned with nitrous oxide. Analytical predictions for end-to-end motor performance and fuel regression are compared against static test results. Baseline fuel grain regression calculations use an enthalpy balance energy analysis with the material and thermodynamic properties based on the mean paraffin/ABS mass fractions within the fuel grain. In support of these analytical comparisons, a novel method for propagating the fuel port burn surface was developed. In this modeling approach the fuel cross section grid is modeled as an image with white pixels representing the fuel and black pixels representing empty or burned grid cells.

Benchmark Evaluation of Fuel Effect and Material Worth Measurements for a Beryllium-Reflected Space Reactor Mockup

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

Marshall, Margaret A.; Bess, John D.

2015-02-01

The critical configuration of the small, compact critical assembly (SCCA) experiments performed at the Oak Ridge Critical Experiments Facility (ORCEF) in 1962-1965 have been evaluated as acceptable benchmark experiments for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The initial intent of these experiments was to support the design of the Medium Power Reactor Experiment (MPRE) program, whose purpose was to study “power plants for the production of electrical power in space vehicles.” The third configuration in this series of experiments was a beryllium-reflected assembly of stainless-steel-clad, highly enriched uranium (HEU)-O 2 fuel mockup of a potassium-cooledmore » space power reactor. Reactivity measurements cadmium ratio spectral measurements and fission rate measurements were measured through the core and top reflector. Fuel effect worth measurements and neutron moderating and absorbing material worths were also measured in the assembly fuel region. The cadmium ratios, fission rate, and worth measurements were evaluated for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The fuel tube effect and neutron moderating and absorbing material worth measurements are the focus of this paper. Additionally, a measurement of the worth of potassium filling the core region was performed but has not yet been evaluated Pellets of 93.15 wt.% enriched uranium dioxide (UO 2) were stacked in 30.48 cm tall stainless steel fuel tubes (0.3 cm tall end caps). Each fuel tube had 26 pellets with a total mass of 295.8 g UO 2 per tube. 253 tubes were arranged in 1.506-cm triangular lattice. An additional 7-tube cluster critical configuration was also measured but not used for any physics measurements. The core was surrounded on all side by a beryllium reflector. The fuel effect worths were measured by removing fuel tubes at various radius. An accident scenario was also simulated by moving outward twenty fuel rods from the periphery of the core so they were touching the core tank. The change in the system reactivity when the fuel tube(s) were removed/moved compared with the base configuration was the worth of the fuel tubes or accident scenario. The worth of neutron absorbing and moderating materials was measured by inserting material rods into the core at regular intervals or placing lids at the top of the core tank. Stainless steel 347, tungsten, niobium, polyethylene, graphite, boron carbide, aluminum and cadmium rods and/or lid worths were all measured. The change in the system reactivity when a material was inserted into the core is the worth of the material.« less

Hydrogen as fuel carrier in PEM fuelcell for automobile applications

NASA Astrophysics Data System (ADS)

Sk, Mudassir Ali; Venkateswara Rao, K.; Ramana Rao, Jagirdar V.

2015-02-01

The present work focuses the application of nanostructured materials for storing of hydrogen in different carbon materials by physisorption method. To market a hydrogen-fuel cell vehicle as competitively as the present internal combustion engine vehicles, there is a need for materials that can store a minimum of 6.5wt% of hydrogen. Carbon materials are being heavily investigated because of their promise to offer an economical solution to the challenge of safe storage of large hydrogen quantities. Hydrogen is important as a new source of energy for automotive applications. It is clear that the key challenge in developing this technology is hydrogen storage. Combustion of fossil fuels and their overuse is at present a serious concern as it is creates severe air pollution and global environmental problems; like global warming, acid rains, ozone depletion in stratosphere etc. This necessitated the search for possible alternative sources of energy. Though there are a number of primary energy sources available, such as thermonuclear energy, solar energy, wind energy, hydropower, geothermal energy etc, in contrast to the fossil fuels in most cases, these new primary energy sources cannot be used directly and thus they must be converted into fuels, that is to say, a new energy carrier is needed. Hydrogen fuel cells are two to three times more efficient than combustion engines. As they become more widely available, they will reduce dependence on fossil fuels. In a fuel cell, hydrogen and oxygen are combined in an electrochemical reaction that produces electricity and, as a byproduct, water.

Spent fuel canister for geological repository: Inner material requirements and candidates evaluation

NASA Astrophysics Data System (ADS)

Puig, Francesc; Dies, Javier; Pablo, Joan de; Martínez-Esparza, Aurora

2008-05-01

One of the key aspects in designing Spanish spent nuclear fuel canister for geological repository is selecting the inner material to be placed between the steel walls and the fuel assemblies. This material has to primarily avoid the possibility of a criticality event once the canister gets breached by corrosion and flooded by groundwater. A detailed set of requirements for a material to fulfil this role in that environment have been devised and presented in this paper. With these requirements in view, eight potentially interesting candidates were evaluated: cast iron or steel, borosilicate glass, spinel, depleted uranium, dehydrated zeolites, haematite, phosphates, and olivine. Among these, the first four materials or their families are found promising for this application.

Unmixed fuel processors and methods for using the same

DOEpatents

Kulkarni, Parag Prakash; Cui, Zhe

2010-08-24

Disclosed herein are unmixed fuel processors and methods for using the same. In one embodiment, an unmixed fuel processor comprises: an oxidation reactor comprising an oxidation portion and a gasifier, a CO.sub.2 acceptor reactor, and a regeneration reactor. The oxidation portion comprises an air inlet, effluent outlet, and an oxygen transfer material. The gasifier comprises a solid hydrocarbon fuel inlet, a solids outlet, and a syngas outlet. The CO.sub.2 acceptor reactor comprises a water inlet, a hydrogen outlet, and a CO.sub.2 sorbent, and is configured to receive syngas from the gasifier. The regeneration reactor comprises a water inlet and a CO.sub.2 stream outlet. The regeneration reactor is configured to receive spent CO.sub.2 adsorption material from the gasification reactor and to return regenerated CO.sub.2 adsorption material to the gasification reactor, and configured to receive oxidized oxygen transfer material from the oxidation reactor and to return reduced oxygen transfer material to the oxidation reactor.

Rational design of novel cathode materials in solid oxide fuel cells using first-principles simulations

NASA Astrophysics Data System (ADS)

Choi, YongMan; Lin, M. C.; Liu, Meilin

The search for clean and renewable sources of energy represents one of the most vital challenges facing us today. Solid oxide fuel cells (SOFCs) are among the most promising technologies for a clean and secure energy future due to their high energy efficiency and excellent fuel flexibility (e.g., direct utilization of hydrocarbons or renewable fuels). To make SOFCs economically competitive, however, development of new materials for low-temperature operation is essential. Here we report our results on a computational study to achieve rational design of SOFC cathodes with fast oxygen reduction kinetics and rapid ionic transport. Results suggest that surface catalytic properties are strongly correlated with the bulk transport properties in several material systems with the formula of La 0.5Sr 0.5BO 2.75 (where B = Cr, Mn, Fe, or Co). The predictions seem to agree qualitatively with available experimental results on these materials. This computational screening technique may guide us to search for high-efficiency cathode materials for a new generation of SOFCs.

Serially connected solid oxide fuel cells having monolithic cores

DOEpatents

Herceg, J.E.

1985-05-20

Disclosed is a solid oxide fuel cell for electrochemically combining fuel and oxidant for generating galvanic output. The cell core has an array of cell segments electrically serially connected in the flow direction, each segment consisting of electrolyte walls and interconnect that are substantially devoid of any composite inert materials for support. Instead, the core is monolithic, where each electrolyte wall consists of thin layers of cathode and anode materials sandwiching a thin layer of electrolyte material therebetween. Means direct the fuel to the anode-exposed core passageways and means direct the oxidant to the cathode-exposed core passageways; and means also direct the galvanic output to an exterior circuit. Each layer of the electrolyte composite materials is of the order of 0.002 to 0.01 cm thick; and each layer of the cathode and anode materials is of the order of 0.002 to 0.05 cm thick. Between 2 and 50 cell segments may be connected in series.

78 FR 31987 - Advisory Committee On Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials...

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2013-05-28

... NUCLEAR REGULATORY COMMISSION Advisory Committee On Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials, Metallurgy & Reactor Fuels; Notice of Meeting The ACRS Subcommittee on Materials, Metallurgy & Reactor Fuels will hold a meeting on June 4, 2013, Room T-2B1, 11545 Rockville Pike, Rockville...

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... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Materials, Metallurgy & Reactor Fuels; Notice of Meeting The ACRS Subcommittee on Materials, Metallurgy & Reactor Fuels will hold a meeting on June 23, 2011, Room T-2B3, 11545 Rockville Pike, Rockville...

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Solid oxide fuel cell having compound cross flow gas patterns

DOEpatents

Fraioli, A.V.

1983-10-12

A core construction for a fuel cell is disclosed having both parallel and cross flow passageways for the fuel and the oxidant gases. Each core passageway is defined by electrolyte and interconnect walls. Each electrolyte wall consists of cathode and anode materials sandwiching an electrolyte material. Each interconnect wall is formed as a sheet of inert support material having therein spaced small plugs of interconnect material, where cathode and anode materials are formed as layers on opposite sides of each sheet and are electrically connected together by the interconnect material plugs. Each interconnect wall in a wavy shape is connected along spaced generally parallel line-like contact areas between corresponding spaced pairs of generally parallel electrolyte walls, operable to define one tier of generally parallel flow passageways for the fuel and oxidant gases. Alternate tiers are arranged to have the passageways disposed normal to one another. Solid mechanical connection of the interconnect walls of adjacent tiers to the opposite sides of the common electrolyte wall therebetween is only at spaced point-like contact areas, 90 where the previously mentioned line-like contact areas cross one another.

Solid oxide fuel cell having compound cross flow gas patterns

DOEpatents

Fraioli, Anthony V.

1985-01-01

A core construction for a fuel cell is disclosed having both parallel and cross flow passageways for the fuel and the oxidant gases. Each core passageway is defined by electrolyte and interconnect walls. Each electrolyte wall consists of cathode and anode materials sandwiching an electrolyte material. Each interconnect wall is formed as a sheet of inert support material having therein spaced small plugs of interconnect material, where cathode and anode materials are formed as layers on opposite sides of each sheet and are electrically connected together by the interconnect material plugs. Each interconnect wall in a wavy shape is connected along spaced generally parallel line-like contact areas between corresponding spaced pairs of generally parallel electrolyte walls, operable to define one tier of generally parallel flow passageways for the fuel and oxidant gases. Alternate tiers are arranged to have the passageways disposed normal to one another. Solid mechanical connection of the interconnect walls of adjacent tiers to the opposite sides of the common electrolyte wall therebetween is only at spaced point-like contact areas, 90 where the previously mentioned line-like contact areas cross one another.

Alternative Fuels Compatibility with Army Equipment Testing - Alternative Fuels Material Compatibility Analysis

DTIC Science & Technology

2012-02-21

Testing and Materials °C Celsius DiEGME Diethylene Glycol Monomethyl Ether EPDM Ethylene Propylene Diene Monomer FARE Forward Area Refueling...urethane class AU, polyether urethane class EU, EPDM , Viton®, fluorosilicone class FQ, polytetrafluoroethylene (PTFE), polyolefin and polyester...sleeve Material not provided AAFARS 4720-00-540-1368 Hose, nonmetallic Material not provided AAFARS 4720-01-218-6958 Hose, preformed Rubber

CORROSION STUDIES FOR A FUSED SALT-LIQUID METAL EXTRACTION PROCESS FOR THE LIQUID METAL FUEL REACTOR

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

Susskind, H.; Hill, F.B.; Green, L.

1960-06-30

Corrosion screening tests were carried out on potential materials of construction for use in a fused salt-liquid metal extraction process plant. The corrodents of interest were NaCl--KCl-- MgCl/sub 2/ eutectic, LiCl--KCl eutectic, Bi-- U fuel, and BiCl/sub 3/, either separately or in various combinations. Screening tests to determine the resistance of a wide range of commercial alloys to the corrodents were performed in static and tilting-furnace capsules. Some ceramic materials were tested in static capsules. Largerscale tests of metallic materials were conducted in thermal convection loops and in a forced circulation loop. Some of the tests were conducted isothermally atmore » 500 deg C, and others were performed under 40 to 50 deg C temperature differences at roughly the same teinperature level. On the basis of metallographic examination of exposed test tabs and chemical analyses of corrodents, it was found that the binary and ternary eutectics by themselves produced little attack on any of the materials tested. A wide variety of materials including 1020 mild steel, 2 1/4 Cr--1 Mo alloy steel, types 304 (ELC), 310, 316, 347, 430, and 446 stainless steel, 16-1 Croloy, Inconel, Hastelloy C, Inor-8, Mo, and Ta is, therefore, available for further study. Corrosion by the ternary salt-fuel system was characteristic of that produced by the fuel alone. Alloys such as 1020 mild steel, and 1 1/4 Cr--1/ 2 Mo, and 2 1/4 Cr--1 Mo alloy steel, which are resistant to fuel, would be likely choices at present for container materials. BiCl/sub 3/ produced extensive attack on ternary salt-fuel containers when the fuel contained insufficient concentrations of oxidizable solutes. Au and Al/sub 2/O/sub 3/ were the only materials not attacked by BiCl/sub 3/ in ternary salt alone. (auth)« less

Transmutation Fuel Fabrication-Fiscal Year 2016

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

Fielding, Randall Sidney; Grover, Blair Kenneth

ABSTRACT Nearly all of the metallic fuel that has been irradiated and characterized by the Advanced Fuel Campaign, and its earlier predecessors, has been arc cast. Arc casting is a very flexible method of casting lab scale quantities of materials. Although the method offers flexibility, it is an operator dependent process. Small changes in parameter space or alloy composition may affect how the material is cast. This report provides a historical insight in how the casting process has been modified over the history of the advanced fuels campaign as well as the physical parameters of the fuels cast in fiscalmore » year 2016.« less

Compositions and methods for treating nuclear fuel

DOEpatents

Soderquist, Chuck Z; Johnsen, Amanda M; McNamara, Bruce K; Hanson, Brady D; Smith, Steven C; Peper, Shane M

2013-08-13

Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

Compositions and methods for treating nuclear fuel

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

Soderquist, Chuck Z; Johnsen, Amanda M; McNamara, Bruce K

Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

40 CFR 97.302 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-07-01

..., petroleum, coal, or any form of solid, liquid, or gaseous fuel derived from such material. Fossil-fuel-fired... right-of-way tree trimmings. Boiler means an enclosed fossil-or other-fuel-fired combustion device used... means a stationary, fossil-fuel-fired boiler or stationary, fossil-fuel-fired combustion turbine: (1...

40 CFR 97.102 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-07-01

..., petroleum, coal, or any form of solid, liquid, or gaseous fuel derived from such material. Fossil-fuel-fired... enclosed fossil- or other-fuel-fired combustion device used to produce heat and to transfer heat to... means a stationary, fossil-fuel-fired boiler or stationary, fossil-fuel-fired combustion turbine: (1...

40 CFR 96.102 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-07-01

..., petroleum, coal, or any form of solid, liquid, or gaseous fuel derived from such material. Fossil-fuel-fired... right-of-way tree trimmings. Boiler means an enclosed fossil- or other-fuel-fired combustion device used..., fossil-fuel-fired boiler or stationary, fossil-fuel-fired combustion turbine: (1) Having equipment used...

40 CFR 96.102 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-07-01

..., petroleum, coal, or any form of solid, liquid, or gaseous fuel derived from such material. Fossil-fuel-fired... right-of-way tree trimmings. Boiler means an enclosed fossil- or other-fuel-fired combustion device used..., fossil-fuel-fired boiler or stationary, fossil-fuel-fired combustion turbine: (1) Having equipment used...

40 CFR 97.102 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-07-01

..., petroleum, coal, or any form of solid, liquid, or gaseous fuel derived from such material. Fossil-fuel-fired... enclosed fossil- or other-fuel-fired combustion device used to produce heat and to transfer heat to... means a stationary, fossil-fuel-fired boiler or stationary, fossil-fuel-fired combustion turbine: (1...

40 CFR 96.102 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-07-01

..., petroleum, coal, or any form of solid, liquid, or gaseous fuel derived from such material. Fossil-fuel-fired... right-of-way tree trimmings. Boiler means an enclosed fossil- or other-fuel-fired combustion device used..., fossil-fuel-fired boiler or stationary, fossil-fuel-fired combustion turbine: (1) Having equipment used...

40 CFR 97.102 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-07-01

..., petroleum, coal, or any form of solid, liquid, or gaseous fuel derived from such material. Fossil-fuel-fired... enclosed fossil- or other-fuel-fired combustion device used to produce heat and to transfer heat to... means a stationary, fossil-fuel-fired boiler or stationary, fossil-fuel-fired combustion turbine: (1...

40 CFR 97.302 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-07-01

..., petroleum, coal, or any form of solid, liquid, or gaseous fuel derived from such material. Fossil-fuel-fired... right-of-way tree trimmings. Boiler means an enclosed fossil-or other-fuel-fired combustion device used... means a stationary, fossil-fuel-fired boiler or stationary, fossil-fuel-fired combustion turbine: (1...

40 CFR 97.302 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-07-01

..., petroleum, coal, or any form of solid, liquid, or gaseous fuel derived from such material. Fossil-fuel-fired... right-of-way tree trimmings. Boiler means an enclosed fossil-or other-fuel-fired combustion device used... means a stationary, fossil-fuel-fired boiler or stationary, fossil-fuel-fired combustion turbine: (1...

40 CFR 96.102 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-07-01

..., petroleum, coal, or any form of solid, liquid, or gaseous fuel derived from such material. Fossil-fuel-fired... right-of-way tree trimmings. Boiler means an enclosed fossil- or other-fuel-fired combustion device used..., fossil-fuel-fired boiler or stationary, fossil-fuel-fired combustion turbine: (1) Having equipment used...

40 CFR 97.102 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-07-01

..., petroleum, coal, or any form of solid, liquid, or gaseous fuel derived from such material. Fossil-fuel-fired... enclosed fossil- or other-fuel-fired combustion device used to produce heat and to transfer heat to... means a stationary, fossil-fuel-fired boiler or stationary, fossil-fuel-fired combustion turbine: (1...

40 CFR 96.102 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-07-01

..., petroleum, coal, or any form of solid, liquid, or gaseous fuel derived from such material. Fossil-fuel-fired... right-of-way tree trimmings. Boiler means an enclosed fossil- or other-fuel-fired combustion device used..., fossil-fuel-fired boiler or stationary, fossil-fuel-fired combustion turbine: (1) Having equipment used...

40 CFR 97.302 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-07-01

..., petroleum, coal, or any form of solid, liquid, or gaseous fuel derived from such material. Fossil-fuel-fired... right-of-way tree trimmings. Boiler means an enclosed fossil-or other-fuel-fired combustion device used... means a stationary, fossil-fuel-fired boiler or stationary, fossil-fuel-fired combustion turbine: (1...

40 CFR 97.302 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-07-01

..., petroleum, coal, or any form of solid, liquid, or gaseous fuel derived from such material. Fossil-fuel-fired... right-of-way tree trimmings. Boiler means an enclosed fossil-or other-fuel-fired combustion device used... means a stationary, fossil-fuel-fired boiler or stationary, fossil-fuel-fired combustion turbine: (1...

40 CFR 97.102 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-07-01

..., petroleum, coal, or any form of solid, liquid, or gaseous fuel derived from such material. Fossil-fuel-fired... enclosed fossil- or other-fuel-fired combustion device used to produce heat and to transfer heat to... means a stationary, fossil-fuel-fired boiler or stationary, fossil-fuel-fired combustion turbine: (1...

Evaluation of Fatty Acid Methyl Ester (FAME) Contamination on the Thermal Stability Characteristics of Military Jet Fuels (JP-8 And JP-5)

DTIC Science & Technology

2013-12-01

of increased contamination levels of FAME in Jet A, FAME material will likely be transported in the same conveyance as JP-5 – bringing with it the...is a blend of four common biodiesel (FAME) fuels from different feedstocks. All FAME contaminated fuels were prepared with this FAME material at a

REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

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

Nichols, T.; Beals, D.; Sternat, M.

2011-07-18

Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Manymore » research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.« less

Spray sealing: A breakthrough in integral fuel tank sealing technology

NASA Astrophysics Data System (ADS)

Richardson, Martin D.; Zadarnowski, J. H.

1989-11-01

In a continuing effort to increase readiness, a new approach to sealing integral fuel tanks is being developed. The technique seals potential leak sources by spraying elastomeric materials inside the tank cavity. Laboratory evaluations project an increase in aircraft supportability and reliability, an improved maintainability, decreasing acquisition and life cycle costs. Increased usable fuel volume and lower weight than conventional bladders improve performance. Concept feasibility was demonstrated on sub-scale aircraft fuel tanks. Materials were selected by testing sprayable elastomers in a fuel tank environment. Chemical stability, mechanical properties, and dynamic durability of the elastomer are being evaluated at the laboratory level and in sub-scale and full scale aircraft component fatigue tests. The self sealing capability of sprayable materials is also under development. Ballistic tests show an improved aircraft survivability, due in part to the elastomer's mechanical properties and its ability to damp vibrations. New application equipment, system removal, and repair methods are being investigated.

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

Braase, Lori

Develop advanced nuclear fuel cycle separation and waste management technologies that improve current fuel cycle performance and enable a sustainable fuel cycle, with minimal processing, waste generation, and potential for material diversion.

Advances in Materials and System Technology for Portable Fuel Cells

NASA Technical Reports Server (NTRS)

Narayanan, Sekharipuram R.

2007-01-01

This viewgraph presentation describes the materials and systems engineering used for portable fuel cells. The contents include: 1) Portable Power; 2) Technology Solution; 3) Portable Hydrogen Systems; 4) Direct Methanol Fuel Cell; 5) Direct Methanol Fuel Cell System Concept; 6) Overview of DMFC R&D at JPL; 7) 300-Watt Portable Fuel Cell for Army Applications; 8) DMFC units from Smart Fuel Cell Inc, Germany; 9) DMFC Status and Prospects; 10) Challenges; 11) Rapid Screening of Well-Controlled Catalyst Compositions; 12) Screening of Ni-Zr-Pt-Ru alloys; 13) Issues with New Membranes; 14) Membranes With Reduced Methanol Crossover; 15) Stacks; 16) Hybrid DMFC System; 17) Small Compact Systems; 18) Durability; and 19) Stack and System Parameters for Various Applications.

Imidazolium-Based Polymeric Materials as Alkaline Anion-Exchange Fuel Cell Membranes

NASA Technical Reports Server (NTRS)

Narayan, Sri R.; Yen, Shiao-Ping S.; Reddy, Prakash V.; Nair, Nanditha

2012-01-01

Polymer electrolyte membranes that conduct hydroxide ions have potential use in fuel cells. A variety of polystyrene-based quaternary ammonium hydroxides have been reported as anion exchange fuel cell membranes. However, the hydrolytic stability and conductivity of the commercially available membranes are not adequate to meet the requirements of fuel cell applications. When compared with commercially available membranes, polystyrene-imidazolium alkaline membrane electrolytes are more stable and more highly conducting. At the time of this reporting, this has been the first such usage for imidazolium-based polymeric materials for fuel cells. Imidazolium salts are known to be electrochemically stable over wide potential ranges. By controlling the relative ratio of imidazolium groups in polystyrene-imidazolium salts, their physiochemical properties could be modulated. Alkaline anion exchange membranes based on polystyrene-imidazolium hydroxide materials have been developed. The first step was to synthesize the poly(styrene-co-(1-((4-vinyl)methyl)-3- methylimidazolium) chloride through a free-radical polymerization. Casting of this material followed by in situ treatment of the membranes with sodium hydroxide solutions provided the corresponding hydroxide salts. Various ratios of the monomers 4-chloromoethylvinylbenzine (CMVB) and vinylbenzine (VB) provided various compositions of the polymer. The preferred material, due to the relative ease of casting the film, and its relatively low hygroscopic nature, was a 2:1 ratio of CMVB to VB. Testing confirmed that at room temperature, the new membranes outperformed commercially available membranes by a large margin. With fuel cells now in use at NASA and in transportation, and with defense potential, any improvement to fuel cell efficiency is a significant development.

Nanomaterials for Polymer Electrolyte Membrane Fuel Cells; Materials Challenges Facing Electrical Energy Storate

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

Gopal Rao, MRS Web-Editor; Yury Gogotsi, Drexel University; Karen Swider-Lyons, Naval Research Laboratory

Symposium T: Nanomaterials for Polymer Electrolyte Membrane Fuel Cells Polymer electrolyte membrane (PEM) fuel cells are under intense investigation worldwide for applications ranging from transportation to portable power. The purpose of this seminar is to focus on the nanomaterials and nanostructures inherent to polymer fuel cells. Symposium topics will range from high-activity cathode and anode catalysts, to theory and new analytical methods. Symposium U: Materials Challenges Facing Electrical Energy Storage Electricity, which can be generated in a variety of ways, offers a great potential for meeting future energy demands as a clean and efficient energy source. However, the use ofmore » electricity generated from renewable sources, such as wind or sunlight, requires efficient electrical energy storage. This symposium will cover the latest material developments for batteries, advanced capacitors, and related technologies, with a focus on new or emerging materials science challenges.« less

Analysis of cadmium in undissolved anode materials of Mark-IV electro-refiner

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

Yoo, Tae-Sic; Fredrickson, G.L.; Vaden, D.

2013-07-01

The Mark-IV electro-refiner (Mk-IV ER) is a unit process in the FCF (Fuel Conditioning Facility), which is primarily assigned to treating the used driver fuels. Mk-IV ER contains an electrolyte/molten cadmium system for refining uranium electrochemically. Typically, the anode of the Mk-IV ER consists of the chopped sodium-bonded metallic driver fuels, which have been primarily U-10Zr binary fuels. Chemical analysis of the residual anode materials after electrorefining indicates that a small amount of cadmium is removed from the Mk-IV ER along with the undissolved anode materials. Investigation of chemical analysis data indicates that the amount of cadmium in the undissolvedmore » anode materials is strongly correlated with the anode rotation speeds and the residence time of the anode in the Mk-IV ER. Discussions are given to explain the prescribed correlation. (authors)« less

Laccase/AuAg Hybrid Glucose Microfludic Fuel Cell

NASA Astrophysics Data System (ADS)

López-González, B.; Cuevas-Muñiz, F. M.; Guerra-Balcázar, M.; Déctor, A.; Arjona, N.; Ledesma-García, J.; Arriaga, L. G.

2013-12-01

In this work a hybrid microfluidic fuel cell was fabricated and evaluated with a AuAg/C bimetallic material for the anode and an enzymatic cathode. The cathodic catalyst was prepared adsorbing laccase and ABTS on Vulcan carbon (Lac-ABTS/C). This material was characterized by FTIR-ATR, the results shows the presence of absorption bands corresponding to the amide bounds. The electrochemical evaluation for the materials consisted in cyclic voltammetry (CV). The glucose electrooxidation reaction in AuAg/C occurs around - 0.3 V vs. NHE. Both electrocatalytic materials were placed in a microfluidic fuel cell. The fuel cell was fed with PBS pH 5 oxygen saturated solution in the cathodic compartment and 5 mM glucose + 0.3 M KOH in the anodic side. Several polarization curves were performed and the maximum power density obtained was 0.3 mWcm-2 .

CORN AND TUNA DIET--REPORT OF OBSERVATIONS ON NUTRITION: RADIATION AND STERILIZATION OF FOODS FOR HAZELTON LABORATORIES, FALLS CHURCH, VIRGINIA

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

Ross, M.A.

1963-09-15

No evidence was found to suggest that the consumption of irradiated corn or tuna fish produced lesions in rats significantly different in incidence or type than was found in animals that consumed the unirradiated products. (C.H.)

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

Kelly, Jarod C.; Sullivan, John L.; Burnham, Andrew

This study examines the vehicle-cycle impacts associated with substituting lightweight materials for those currently found in light-duty passenger vehicles. We determine part-based energy use and greenhouse gas (GHG) emission ratios by collecting material substitution data from both the literature and automotive experts and evaluating that alongside known mass-based energy use and GHG emission ratios associated with material pair substitutions. Several vehicle parts, along with full vehicle systems, are examined for lightweighting via material substitution to observe the associated impact on GHG emissions. Results are contextualized by additionally examining fuel-cycle GHG reductions associated with mass reductions relative to the baseline vehiclemore » during the use phase and also determining material pair breakeven driving distances for GHG emissions. The findings show that, while material substitution is useful in reducing vehicle weight, it often increases vehicle-cycle GHGs depending upon the material substitution pair. However, for a vehicle’s total life cycle, fuel economy benefits are greater than the increased burdens associated with the vehicle manufacturing cycle, resulting in a net total life-cycle GHG benefit. The vehicle cycle will become increasingly important in total vehicle life-cycle GHGs, since fuel-cycle GHGs will be gradually reduced as automakers ramp up vehicle efficiency to meet fuel economy standards.« less

Thermally and chemically responsive nanoporous materials for efficient capture of fission product gases.

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

Stroeve, Pieter; Faller, Roland

The objective of this project was to develop robust, high-efficiency materials for capture of fission product gases such as He, Xe and Kr in scenarios relevant for both reactor fuels and reprocessing operations. The relevant environments are extremely harsh, encompassing temperatures up to 1500 °C, high levels of radiation, as well as potential exposures to highly-reactive chemicals such as nitric acid and organic solvents such as kerosene. The requirement for nanostructured capture materials is driven in part by the very short (few micron) diffusion distances for product gases in nuclear fuel. We achieved synthesis, characterization and detailed modeling of themore » materials. Although not all materials reviewed in this report will be feasible for the ultimate goal of integration in nuclear fuel, nevertheless each material studied has particular properties which will enable an optimized material to be efficiently developed and characterized.« less

Method for fabrication of electrodes

DOEpatents

Jankowski, Alan F.; Morse, Jeffrey D.; Barksdale, Randy

2004-06-22

Described herein is a method to fabricate porous thin-film electrodes for fuel cells and fuel cell stacks. Furthermore, the method can be used for all fuel cell electrolyte materials which utilize a continuous electrolyte layer. An electrode layer is deposited on a porous host structure by flowing gas (for example, Argon) from the bottomside of the host structure while simultaneously depositing a conductive material onto the topside of the host structure. By controlling the gas flow rate through the pores, along with the process conditions and deposition rate of the thin-film electrode material, a film of a pre-determined thickness can be formed. Once the porous electrode is formed, a continuous electrolyte thin-film is deposited, followed by a second porous electrode to complete the fuel cell structure.

The effect of stress state on zirconium hydride reorientation

NASA Astrophysics Data System (ADS)

Cinbiz, Mahmut Nedim

Prior to storage in a dry-cask facility, spent nuclear fuel must undergo a vacuum drying cycle during which the spent fuel rods are heated up to elevated temperatures of ≤ 400°C to remove moisture the canisters within the cask. As temperature increases during heating, some of the hydride particles within the cladding dissolve while the internal gas pressure in fuel rods increases generating multi-axial hoop and axial stresses in the closed-end thin-walled cladding tubes. As cool-down starts, the hydrogen in solid solution precipitates as hydride platelets, and if the multiaxial stresses are sufficiently large, the precipitating hydrides reorient from their initial circumferential orientation to radial orientation. Radial hydrides can severely embrittle the spent nuclear fuel cladding at low temperature in response to hoop stress loading. Because the cladding can experience a range of stress states during the thermo-mechanical treatment induced during vacuum drying, this study has investigated the effect of stress state on the process of hydride reorientation during controlled thermo-mechanical treatments utilizing the combination of in situ X-ray diffraction and novel mechanical testing analyzed by the combination of metallography and finite element analysis. The study used cold worked and stress relieved Zircaloy-4 sheet containing approx. 180 wt. ppm hydrogen as its material basis. The failure behavior of this material containing radial hydrides was also studied over a range of temperatures. Finally, samples from reactor-irradiated cladding tubes were examined by X-ray diffraction using synchrotron radiation. To reveal the stress state effect on hydride reorientation, the critical threshold stress to reorient hydrides was determined by designing novel mechanical test samples which produce a range of stress states from uniaxial to "near-equibiaxial" tension when a load is applied. The threshold stress was determined after thermo-mechanical treatments by correlating the finite element stress-state results with the spatial distribution of hydride microstructures observed within the optical micrographs for each sample. Experiments showed that the hydride reorientation was enhanced as the stress biaxiality increased. The threshold stress decreased from 150 MPa to 80 MPa when stress biaxiality ratio increased from uniaxial tension to near-equibiaxial tension. This behavior was also predicted by classical nucleation theory based on the Gibbs free energy of transformation being assisted by the far-field stress. An analysis of in situ X-ray diffraction data obtained during a thermo-mechanical cycle typical of vacuum drying showed a complex lattice-spacing behavior of the hydride phase during the dissolution and precipitation. The in-plane hydrides showed bilinear lattice expansion during heating with the intrinsic thermal expansion rate of the hydrides being observed only at elevated temperatures as they dissolve. For radial hydrides that precipitate during cooling under stress, the spacing of the close-packed {111} planes oriented normal to the maximum applied stress was permanently higher than the corresponding {111} plane spacing in the other directions. This behavior is believed to be a result of a complex stress state within the precipitating plate-like hydrides that induces a strain component within the hydrides normal to its "plate" face (i.e., the applied stress direction) that exceeds the lattice spacing strains in the other directions. During heat-up, the lattice spacing of these same "plate" planes actually contract due to the reversion of the stress state within the plate-like hydrides as they dissolve. The presence of radial hydrides and their connectivity with in-plane hydrides was shown to increase the ductile-to-brittle transition temperature during tensile testing. This behavior can be understood in terms of the role of radial hydrides in promoting the initiation of a long crack that subsequently propagates under fracture mechanics conditions. Finally, the d-spacing of irradiated Zircaloy-4 and M5 cladding tubes was measured at room temperature and compared to that of unirradiated samples.

MARTINS: A foam/film flow model for molten material relocation in HWRs with U-Al-fueled multi-tube assemblies

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

Kalimullah

1994-03-01

Some special purpose heavy-water reactors (EM) are made of assemblies consisting of a number of coaxial aluminum-clad U-Al alloy fuel tubes and an outer Al sleeve surrounding the fuel tubes. The heavy water coolant flows in the annular gaps between the circular tubes. Analysis of severe accidents in such reactors requires a model for predicting the behavior of the fuel tubes as they melt and disrupt. This paper describes a detailed, mechanistic model for fuel tube heatup, melting, freezing, and molten material relocation, called MARTINS (Melting and Relocation of Tubes in Nuclear subassembly). The paper presents the modeling of themore » phenomena in MARTINS, and an application of the model to analysis of a reactivity insertion accident. Some models are being developed to compute gradual downward relocation of molten material at decay-heat power levels via candling along intact tubes, neglecting coolant vapor hydrodynamic forces on molten material. These models are inadequate for high power accident sequences involving significant hydrodynamic forces. These forces are included in MARTINS.« less

Implementation of focused ion beam (FIB) system in characterization of nuclear fuels and materials

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

A. Aitkaliyeva; J. W. Madden; B. D. Miller

2014-10-01

Beginning in 2007, a program was established at the Idaho National Laboratory to update key capabilities enabling microstructural and micro-chemical characterization of highly irradiated and/or radiologically contaminated nuclear fuels and materials at scales that previously had not been achieved for these types of materials. Such materials typically cannot be contact handled and pose unique hazards to instrument operators, facilities, and associated personnel. One of the first instruments to be acquired was a Dual Beam focused ion beam (FIB)-scanning electron microscope (SEM) to support preparation of transmission electron microscopy and atom probe tomography samples. Over the ensuing years, techniques have beenmore » developed and operational experience gained that has enabled significant advancement in the ability to characterize a variety of fuel types including metallic, ceramic, and coated particle fuels, obtaining insights into in-reactor degradation phenomena not obtainable by any other means. The following article describes insights gained, challenges encountered, and provides examples of unique results obtained in adapting Dual Beam FIB technology to nuclear fuels characterization.« less

Developmental status of thermionic materials.

NASA Technical Reports Server (NTRS)

Yang, L.; Chin, J.

1972-01-01

Description of the reference materials selected for the major components of the unit cell of a thermionic pile element (TFE), the out-of-pile and in-pile test results, and current efforts for improving the life and performance of thermionic fuel elements. The component materials are required to withstand the fuel burnup and fast neutron fluence dictated by the thermionic reactor system. Tungsten was selected as the cladding material because of its compatibility with both the carbide and the oxide fuel materials. Niobium was selected as the collector material because its thermal expansion coefficient matches closely with that of the thin aluminum oxide layer used to electrically insulate the collector from the TFE sheath. An unfueled converter has performed stably over 41,000 hr. Accelerated irradiation tests have attained burnups equivalent to that for 40,000 hr of the thermionic reactor under consideration.

MATERIALS SCIENCE: New Tigers in the Fuel Cell Tank.

PubMed

Service, R F

2000-06-16

After decades of incremental advances, a spurt of findings suggests that fuel cells that run on good old fossil fuels are almost ready for prime time. Although conventional ceramic cells, known as solid oxide fuel cells, require expensive heat-resistant materials, a new generation of SOFCs, including one featured on page 2031, converts hydrocarbons directly into electricity at lower temperatures. And a recent demonstration of a system of standard SOFCs large enough to light up more than 200 homes showed that it is the most efficient large-scale electrical generator ever designed.

Cathode for molten carbonate fuel cell

DOEpatents

Kaun, Thomas D.; Mrazek, Franklin C.

1990-01-01

A porous sintered cathode for a molten carbonate fuel cell and method of making same, the cathode including a skeletal structure of a first electronically conductive material slightly soluble in the electrolyte present in the molten carbonate fuel cell covered by fine particles of a second material of possibly lesser electronic conductivity insoluble in the electrolyte present in the molten carbonate fuel cell, the cathode having a porosity in the range of from about 60% to about 70% at steady-state cell operating conditions consisting of both macro-pores and micro-pores.

Radiolytic and Thermal Processes Relevant to Dry Storage of Spent Nuclear Fuels

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

Marschman, Steven C.; Madey,Theodore E.; Haustein, Peter E.

2000-06-01

The purpose of this project is to deliver pertinent information that can be used to make rational decisions about the safety and treatment issues associated with dry storage of spent nuclear fuel materials. In particular, we will establish an understanding of: (1) water interactions with failed-fuel rods and metal-oxide materials; (2) the role of thermal processes and radiolysis (solid-state and interfacial) in the generation of potentially explosive mixtures of gaseous H2 and O2; and (3) the potential role of radiation-assisted corrosion during fuel rod storage.

Computation of Material Demand in the Risk Assessment and Mitigation Framework for Strategic Materials (RAMF-SM) Process

DTIC Science & Technology

2015-08-01

Congress concerning requirements for the National Defense Stockpile (NDS) of strategic and critical non- fuel materials. 1 RAMF-SM, which was...critical non- fuel materials. The NDS was established in the World War II era and has been managed by the Department of Defense (DOD) since 1988. By...Department of the Interior. An alternative algorithm is used for materials with intensive defense demands. v Contents 1 .  Introduction

Wood fuel preparation

Treesearch

L. H. Reineke

1965-01-01

This report gives information on the preparation of wood fuel from wood residues and other wood raw materials. Types of wood fuel discussed are cordwood, stovewood, slabwood, kindling, chips, hogged fuel, sawdust and shavings, bark, charcoal, alcohol, and briquets. Related information is given on types of machinery for preparing wood fuel and on possible markets for...

Automotive Fuel and Exhaust Systems.

ERIC Educational Resources Information Center

Irby, James F.; And Others

Materials are provided for a 14-hour course designed to introduce the automotive mechanic to the basic operations of automotive fuel and exhaust systems incorporated on military vehicles. The four study units cover characteristics of fuels, gasoline fuel system, diesel fuel systems, and exhaust system. Each study unit begins with a general…

Impacts | Hydrogen and Fuel Cells | NREL

Science.gov Websites

Impacts Impacts Read about NREL's impacts on innovations in hydrogen and fuel cell research and -Splitting Electrodes NREL Shows How Cyanobacteria Build Hydrogen-Producing Enzyme Fuel Cell Systems R&D -Speed Scanner to Monitor Fuel Cell Material Defects Making Fuel Cells Cleaner, Better, and Cheaper GM

Intermittently-fed high-pressure gasifier process

DOEpatents

Bailey, J.M.; Zadoks, A.L.

1993-11-30

An improved gasifier is described which is adapted for gasifying a predetermined charge of non-gaseous fuel into fuel gas. Each charge of non-gaseous fuel, which may have optional conditioning materials added to it, is intermittently fed to a gasifier chamber where each charge is partially burned with high-pressure air supplied thereto. High-pressure and temperature fuel gas is produced which is cleansed prior to passing out of the gasifier chamber. After gasification of the charge of fuel is ended, the gasifier chamber is vented. The residue of the burned charge in the gasifier chamber is removed, along with the contaminated or reacted conditioning materials, and replaced by a fresh charge. The subject invention provides a feasible way of continuously fueling an internal combustion engine with gasified fuel and is compact enough to be practical for even mobile applications. 3 figures.

Intermittently-fed high-pressure gasifier process

DOEpatents

Bailey, John M.; Zadoks, Abraham L.

1993-11-30

An improved gasifier adapted for gasifying a predetermined charge of non-gaseous fuel into fuel gas. Each charge of non-gaseous fuel, which may have optional conditioning materials added to it, is intermittently fed to a gasifier chamber where each charge is partially burned with high-pressure air supplied thereto. High-pressure and temperature fuel gas is produced which is cleansed prior to passing out of the gasifier chamber. After gasification of the charge of fuel is is ended, the gasifier chamber is vented. The residue of the burned charge in the gasifier chamber is removed, along with the contaminated or reacted conditioning materials, and replaced by a fresh charge. The subject invention provides a feasible way of continuously fueling an internal combustion engine with gasified fuel and is compact enough to be practical for even mobile applications.

78 FR 49793 - Regulation of Fuels and Fuel Additives: 2013 Renewable Fuel Standards

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-15

... produced in plants using waste materials to displace 90% or more of fossil fuel use under the then... made to our approach in evaluating the information that forms the basis for our projection of...

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

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

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

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. Although the ultimate goal is to determine the grain boundary characteristics of fission product containing grain boundaries of neutron irradiated SiC, our work reports the effect of transmission electron microscope (TEM) lamella thickness on quality of data and establishes a baseline comparison on grain boundary characteristics determined previously using a conventional EBSD scanning electron microscope (SEM) based technique. In general, it was determined that the lamella thickness produced using the standardmore » FIB fabrication process, is sufficient to provide reliable PED measurements with thicker lamellae (~120 nm) produce higher quality orientation data. Analysis of 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 the same TRISO-coated particle as well as a SiC layer deposited at a slightly lower temperature. The fractions of high angle and 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 [12], depending on the fabrication parameters, and grain boundary fission product precipitates can be nano-sized, the TEM-based PED orientation data collection method is preferred to determine an accurate representation of the relative fractions of low angle, high angle and CSL-related grain boundaries. It was concluded that although the resolution of the PED data is better by more than an order of magnitude, data acquisition times may be significantly longer or the number of areas analyzed significantly larger than the SEM-based method to obtain a statistically relevant distribution. Also, grain size could be accurately determined but significantly larger analysis areas than those used in this study would be required.« less

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

DOE PAGES

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

2016-06-16

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. Although the ultimate goal is to determine the grain boundary characteristics of fission product containing grain boundaries of neutron irradiated SiC, our work reports the effect of transmission electron microscope (TEM) lamella thickness on quality of data and establishes a baseline comparison on grain boundary characteristics determined previously using a conventional EBSD scanning electron microscope (SEM) based technique. In general, it was determined that the lamella thickness produced using the standardmore » FIB fabrication process, is sufficient to provide reliable PED measurements with thicker lamellae (~120 nm) produce higher quality orientation data. Analysis of 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 the same TRISO-coated particle as well as a SiC layer deposited at a slightly lower temperature. The fractions of high angle and 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 [12], depending on the fabrication parameters, and grain boundary fission product precipitates can be nano-sized, the TEM-based PED orientation data collection method is preferred to determine an accurate representation of the relative fractions of low angle, high angle and CSL-related grain boundaries. It was concluded that although the resolution of the PED data is better by more than an order of magnitude, data acquisition times may be significantly longer or the number of areas analyzed significantly larger than the SEM-based method to obtain a statistically relevant distribution. Also, grain size could be accurately determined but significantly larger analysis areas than those used in this study would be required.« less

Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

NASA Technical Reports Server (NTRS)

Bradley, D. E.; Mireles, O. R.; Hickman, R. R.

2011-01-01

Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames.1,2 Conventional storable propellants produce average specific impulse. Nuclear thermal rockets capable of producing high specific impulse are proposed. Nuclear thermal rockets employ heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K), and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited.3 The primary concern is the mechanical failure of fuel elements that employ high-melting-point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. The purpose of the testing is to obtain data to assess the properties of the non-nuclear support materials, as-fabricated, and determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures. The fission process of the planned fissile material and the resulting heating performance is well known and does not therefore require that active fissile material be integrated in this testing. A small-scale test bed designed to heat fuel element samples via non-contact radio frequency heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

Life cycle analysis of vehicles powered by a fuel cell and by internal combustion engine for Canada

NASA Astrophysics Data System (ADS)

Zamel, Nada; Li, Xianguo

The transportation sector is responsible for a great percentage of the greenhouse gas emissions as well as the energy consumption in the world. Canada is the second major emitter of carbon dioxide in the world. The need for alternative fuels, other than petroleum, and the need to reduce energy consumption and greenhouse gases emissions are the main reasons behind this study. In this study, a full life cycle analysis of an internal combustion engine vehicle (ICEV) and a fuel cell vehicle (FCV) has been carried out. The impact of the material and fuel used in the vehicle on energy consumption and carbon dioxide emissions is analyzed for Canada. The data collected from the literature shows that the energy consumption for the production of 1 kg of aluminum is five times higher than that of 1 kg of steel, although higher aluminum content makes vehicles lightweight and more energy efficient during the vehicle use stage. Greenhouse gas regulated emissions and energy use in transportation (GREET) software has been used to analyze the fuel life cycle. The life cycle of the fuel consists of obtaining the raw material, extracting the fuel from the raw material, transporting, and storing the fuel as well as using the fuel in the vehicle. Four different methods of obtaining hydrogen were analyzed; using coal and nuclear power to produce electricity and extraction of hydrogen through electrolysis and via steam reforming of natural gas in a natural gas plant and in a hydrogen refueling station. It is found that the use of coal to obtain hydrogen generates the highest emissions and consumes the highest energy. Comparing the overall life cycle of an ICEV and a FCV, the total emissions of an FCV are 49% lower than an ICEV and the energy consumption of FCV is 87% lower than that of ICEV. Further, CO 2 emissions during the hydrogen fuel production in a central plant can be easily captured and sequestrated. The comparison carried out in this study between FCV and ICEV is extended to the use of recycled material. It is found that using 100% recycled material can reduce energy consumption by 45% and carbon dioxide emissions by 42%, mainly due to the reduced use of electricity during the manufacturing of the material.

Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

NASA Technical Reports Server (NTRS)

Bowles, K. J.; Gluyas, R. E.

1975-01-01

The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.

Preparation of Carbon-Platinum-Ceria and Carbon-Platinum-Cerium catalysts and its application in Polymer Electrolyte Fuel Cell: Hydrogen, Methanol, and Ethanol

NASA Astrophysics Data System (ADS)

Guzman Blas, Rolando Pedro

This thesis is focused on fuel cells using hydrogen, methanol and ethanol as fuel. Also, in the method of preparation of catalytic material for the anode: Supercritical Fluid Deposition (SFD) and impregnation method using ethylenediaminetetraacetic acid (EDTA) as a chelating agent. The first part of the thesis describes the general knowledge about Hydrogen Polymer Exchange Membrane Fuel Cell (HPEMFC),Direct Methanol Fuel Cell (DMFC) and Direct Ethanol Fuel Cell (DEFC), as well as the properties of Cerium and CeO2 (Ceria). The second part of the thesis describes the preparation of catalytic material by Supercritical Fluid Deposition (SFD). SFD was utilized to deposit Pt and ceria simultaneously onto gas diffusion layers. The Pt-ceria catalyst deposited by SFD exhibited higher methanol oxidation activity compared to the platinum catalyst alone. The linear sweep traces of the cathode made for the methanol cross over study indicate that Pt-Ceria/C as the anode catalyst, due to its better activity for methanol, improves the fuel utilization, minimizing the methanol permeation from anode to cathode compartment. The third and fourth parts of the thesis describe the preparation of material catalytic material Carbon-Platinum-Cerium by a simple and cheap impregnation method using EDTA as a chelating agent to form a complex with cerium (III). This preparation method allows the mass production of the material catalysts without additional significant cost. Fuel cell polarization and power curves experiments showed that the Carbon-Platinum-Cerium anode materials exhibited better catalytic activity than the only Vulcan-Pt catalysts for DMFC, DEFC and HPEMFC. In the case of Vulcan-20%Pt-5%w Cerium, this material exhibits better catalytic activity than the Vulcan-20%Pt in DMFC. In the case of Vulcan-40% Pt-doped Cerium, this material exhibits better catalytic activity than the Vulcan-40% Pt in DMFC, DEFC and HPEMFC. Finally, I propose a theory that explains the reason why the carbon-platinum-cerium has better catalytic activity than platinum-carbon. Due to the hybridization behavior of C and Ce could arise charge transfer, both carbon and cerium to the Platinum. Ce-C→Pt charge transfer could occur at the Ce-C/Pt interface. Thus, results in an increase in the catalytic activity of platinum-cerium-carbon when compared with carbon-platinum.

Fuel cells with doped lanthanum gallate electrolyte

NASA Astrophysics Data System (ADS)

Feng, Man; Goodenough, John B.; Huang, Keqin; Milliken, Christopher

Single cells with doped lanthanum gallate electrolyte material were constructed and tested from 600 to 800°C. Both ceria and the electrolyte material were mixed with NiO powder respectively to form composite anodes. Doped lanthanum cobaltite was used exclusively as the cathode material. While high power density from the solid oxide fuel cells at 800°C was achieved. our results clearly indicate that anode overpotential is the dominant factor in the power loss of the cells. Better anode materials and anode processing methods need to be found to fully utilize the high ionic conductivity of the doped lanthanum galiate and achieve higher power density at 800°C from solid oxide fuel cells.

SAVANNAH RIVER SITE'S H-CANYON FACILITY: IMPACTS OF FOREIGN OBLIGATIONS ON SPECIAL NUCLEAR MATERIAL DISPOSITION

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

Magoulas, V.

2013-06-03

The US has a non-proliferation policy to receive foreign and domestic research reactor returns of spent fuel materials of US origin. These spent fuel materials are returned to the Department of Energy (DOE) and placed in storage in the L-area spent fuel basin at the Savannah River Site (SRS). The foreign research reactor returns fall subject to the 123 agreements for peaceful cooperation. These “123 agreements” are named after section 123 of the Atomic Energy Act of 1954 and govern the conditions of nuclear cooperation with foreign partners. The SRS management of these foreign obligations while planning material disposition pathsmore » can be a challenge.« less

Methods for using novel cathode and electrolyte materials for solid oxide fuel cells and ion transport membranes

DOEpatents

Jacobson, Allan J.; Wang, Shuangyan; Kim, Gun Tae

2016-01-12

Methods using novel cathode, electrolyte and oxygen separation materials operating at intermediate temperatures for use in solid oxide fuel cells and ion transport membranes include oxides with perovskite related structures and an ordered arrangement of A site cations. The materials have significantly faster oxygen kinetics than in corresponding disordered perovskites.

Cathode and electrolyte materials for solid oxide fuel cells and ion transport membranes

DOEpatents

Jacobson, Allan J; Wang, Shuangyan; Kim, Gun Tae

2014-01-28

Novel cathode, electrolyte and oxygen separation materials are disclosed that operate at intermediate temperatures for use in solid oxide fuel cells and ion transport membranes based on oxides with perovskite related structures and an ordered arrangement of A site cations. The materials have significantly faster oxygen kinetics than in corresponding disordered perovskites.

76 FR 16016 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of The ACRS Subcommittee on Materials...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-22

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of The ACRS Subcommittee on Materials, Metallurgy And Reactor Fuels; Notice of Meeting The ACRS Subcommittee on Materials, Metallurgy and Reactor Fuels will hold a meeting on April 6, 2011, Room T-2B3, 11545 Rockville Pike...

METHOD OF MANUFACTURE OF METAL ENCASED CORE MATERIAL

DOEpatents

Peters, E.J.

1963-06-01

A method of making reactor fuel elements in which the fissionable material is encased and sealed in steel or aluminum cladding having an enclosed channel from the fissionable material to the surface of the cladding at one end is described. Heat and pressure sufficient to bond the assembled fuel element are applied in a nonoxidizing atmosphere. (AEC)

NETL- Severe Environment Corrosion Erosion Facility

ScienceCinema

None

2018-01-16

NETL's Severe Environment Corrosion Erosion Facility in Albany studies how new and old materials will stand up to new operating conditions. Work done in the lab supports NETL's oxy-fuel combustion oxidation work, refractory materials stability work, and the fuels program, in particular the hydrogen membrane materials stability work, to determine how best to upgrade existing power plants.

Viscous sealing glass compositions for solid oxide fuel cells

DOEpatents

Kim, Cheol Woon; Brow, Richard K.

2016-12-27

A sealant for forming a seal between at least two solid oxide fuel cell components wherein the sealant comprises a glass material comprising B.sub.2O.sub.3 as a principal glass former, BaO, and other components and wherein the glass material is substantially alkali-free and contains less than 30% crystalline material.

Energy conversion and storage program

NASA Astrophysics Data System (ADS)

Cairns, E. J.

1992-03-01

The Energy Conversion and Storage Program applies chemistry and materials science principles to solve problems in: (1) production of new synthetic fuels; (2) development of high-performance rechargeable batteries and fuel cells; (3) development of advanced thermochemical processes for energy conversion; (4) characterization of complex chemical processes; and (5) application of novel materials for energy conversion and transmission. Projects focus on transport-process principles, chemical kinetics, thermodynamics, separation processes, organic and physical chemistry, novel materials, and advanced methods of analysis. Electrochemistry research aims to develop advanced power systems for electric vehicle and stationary energy storage applications. Topics include identification of new electrochemical couples for advanced rechargeable batteries, improvements in battery and fuel-cell materials, and the establishment of engineering principles applicable to electrochemical energy storage and conversion. Chemical Applications research includes topics such as separations, catalysis, fuels, and chemical analyses. Included in this program area are projects to develop improved, energy-efficient methods for processing waste streams from synfuel plants and coal gasifiers. Other research projects seek to identify and characterize the constituents of liquid fuel-system streams and to devise energy-efficient means for their separation. Materials Applications research includes the evaluation of the properties of advanced materials, as well as the development of novel preparation techniques. For example, the use of advanced techniques, such as sputtering and laser ablation, are being used to produce high-temperature superconducting films.

Novel inorganic materials for polymer electrolyte and alkaline fuel cells

NASA Astrophysics Data System (ADS)

Tadanaga, Kiyoharu

2012-06-01

Inorganic materials with high ionic conductivity must have big advantages for the thermal and long term stability when the materials are used as the electrolyte of fuel cells. In the present paper, novel ionic conductive inorganic materials for polymer electrolyte fuel cells (PEFCs) and all solid state alkaline fuel cells (AFCs) that have been developed by our group have been reviewed. PEFCs which can operate in temperature range from 100 to 200 °C are intensively studied because of some advantages such as reduction of CO poisoning of Pt catalyst and acceleration of electrode reactions. We showed that the fuel cells using the composite membranes prepared from phosphosilicate gel powder and polyimide precursor can operate in the temperature range from 30 to 180 °C. We also found that the inorganic-organic hybrid membranes with acid-base pairs from 3-aminopropyl triethoxy silane and H2SO4 or H3PO4 show high proton conductivity under dry atmosphere, and the membranes are thermally stable at intermediate temperatures. On the other hand, because the use of noble platinum is the serious problem for the commercialization of PEFCs and because oxidation reactions are usually faster than those of acid-type fuel cells, alkaline type fuel cells, in which a nonplatinum catalyst can be used, are attractive. Recently, we have proposed an alkaline-type direct ethanol fuel cell (DEFC) using a natural clay electrolyte with non-platinum catalysts. So-called hydrotalcite clay, Mg-Al layered double hydroxide intercalated with CO32- (Mg-Al CO32- LDH), has been proved to be a hydroxide ion conductor. An alkalinetype DEFC using Mg-Al CO32- LDH as the electrolyte and aqueous solution of ethanol and potassium hydroxide as a source of fuel exhibited excellent electrochemical performance.

Spent Fuel Ratio Estimates from Numerical Models in ALE3D

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

Margraf, J. D.; Dunn, T. A.

Potential threat of intentional sabotage of spent nuclear fuel storage facilities is of significant importance to national security. Paramount is the study of focused energy attacks on these materials and the potential release of aerosolized hazardous particulates into the environment. Depleted uranium oxide (DUO 2) is often chosen as a surrogate material for testing due to the unreasonable cost and safety demands for conducting full-scale tests with real spent nuclear fuel. To account for differences in mechanical response resulting in changes to particle distribution it is necessary to scale the DUO 2 results to get a proper measure for spentmore » fuel. This is accomplished with the spent fuel ratio (SFR), the ratio of respirable aerosol mass released due to identical damage conditions between a spent fuel and a surrogate material like depleted uranium oxide (DUO 2). A very limited number of full-scale experiments have been carried out to capture this data, and the oft-questioned validity of the results typically leads to overly-conservative risk estimates. In the present work, the ALE3D hydrocode is used to simulate DUO 2 and spent nuclear fuel pellets impacted by metal jets. The results demonstrate an alternative approach to estimate the respirable release fraction of fragmented nuclear fuel.« less

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

Blaise Collin

The Idaho National Laboraroty (INL) PARFUME (particle fuel model) code was used to assess the overall fuel performance of uranium nitride (UN) tristructural isotropic (TRISO) ceramic fuel under irradiation conditions typical of a Light Water Reactor (LWR). The dimensional changes of the fuel particle layers and kernel were calculated, including the formation of an internal gap. The survivability of the UN TRISO particle was estimated depending on the strain behavior of the constituent materials at high fast fluence and burn up. For nominal cases, internal gas pressure and representative thermal profiles across the kernel and layers were determined along withmore » stress levels in the inner and outer pyrolytic carbon (IPyC/OPyC) and silicon carbide (SiC) layers. These parameters were then used to evaluate fuel particle failure probabilities. Results of the study show that the survivability of UN TRISO fuel under LWR irradiation conditions might only be guaranteed if the kernel and PyC swelling rates are limited at high fast fluence and burn up. These material properties have large uncertainties at the irradiation levels expected to be reached by UN TRISO fuel in LWRs. Therefore, a large experimental effort would be needed to establish material properties, including kernel and PyC swelling rates, under these conditions before definitive conclusions can be drawn on the behavior of UN TRISO fuel in LWRs.« less

BISON Fuel Performance Analysis of IFA-796 Rod 3 & 4 and Investigation of the Impact of Fuel Creep

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

Wirth, Brian; Terrani, Kurt A.; Sweet, Ryan T.

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromiumaluminum (FeCrAl) alloys because they exhibit slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and slow cladding consumption in the presence of high temperature steam. These alloys should also exhibit increased “coping time” in the event of an accident scenario by improving the mechanical performance at high temperatures, allowing greater flexibility to achieve core cooling.more » As a continuation of the development of these alloys, in-reactor irradiation testing of FeCrAl cladded fuel rods has started. In order to provide insight on the possible behavior of these fuel rods as they undergo irradiation in the Halden Boiling Water Reactor, engineering analysis has been performed using FeCrAl material models implemented into the BISON fuel performance code. This milestone report provides an update on the ongoing development of modeling capability to predict FeCrAl cladding fuel performance and to provide an early look at the possible behavior of planned in-reactor FeCrAl cladding experiments. In particular, this report consists of two separate analyses. The first analysis consists of fuel performance simulations of IFA-796 rod 4 and two segments of rod 3. These simulations utilize previously implemented material models for the C35M FeCrAl alloy and UO2 to provide a bounding behavior analysis corresponding to variation of the initial fuel cladding gap thickness within the fuel rod. The second analysis is an assessment of the fuel and cladding stress states after modification of the fuel creep model that is currently implemented in the BISON fuel performance code. Effects from modifying the fuel creep model were identified for the BISON simulations of the IFA-796 rod 4 experiment, but show that varying the creep model (within the range investigated here) only provide a minimal increase in the fuel radius and maximum cladding hoop stress. Continued investigation of fuel behavioral models will include benchmarking the modified fuel creep model against available experimental data, as well as an investigation of the role that fuel cracking will play in the compliance of the fuel. Correctly calculating stress evolution in the fuel is key to assessing fuel behavior up to gap closure and the subsequent deformation of the cladding due to PCMI. The inclusion of frictional contact should also be investigated to determine the axial elongation of the fuel rods for comparison with data from this experiment.« less

Modeling and analysis of UN TRISO fuel for LWR application using the PARFUME code

NASA Astrophysics Data System (ADS)

Collin, Blaise P.

2014-08-01

The Idaho National Laboratory (INL) PARFUME (PARticle FUel ModEl) code was used to assess the overall fuel performance of uranium nitride (UN) tristructural isotropic (TRISO) ceramic fuel under irradiation conditions typical of a Light Water Reactor (LWR). The dimensional changes of the fuel particle layers and kernel were calculated, including the formation of an internal gap. The survivability of the UN TRISO particle was estimated depending on the strain behavior of the constituent materials at high fast fluence and burn-up. For nominal cases, internal gas pressure and representative thermal profiles across the kernel and layers were determined along with stress levels in the inner and outer pyrolytic carbon (IPyC/OPyC) and silicon carbide (SiC) layers. These parameters were then used to evaluate fuel particle failure probabilities. Results of the study show that the survivability of UN TRISO fuel under LWR irradiation conditions might only be guaranteed if the kernel and PyC swelling rates are limited at high fast fluence and burn-up. These material properties have large uncertainties at the irradiation levels expected to be reached by UN TRISO fuel in LWRs. Therefore, a large experimental effort would be needed to establish material properties, including kernel and PyC swelling rates, under these conditions before definitive conclusions can be drawn on the behavior of UN TRISO fuel in LWRs.

Behavior or Nonmetallic Materials in Shale Oil Derived Jet Fuels and in High Aromatic and High Sulfur Petroleum Fuels

DTIC Science & Technology

1978-07-01

degrades thermal stability and forms undesirable sulfur dioxide emissions . Although the original premises for controlling total sulfur may not still...eliminate corrosive trace contamination, presence of surfactants which deactivate filter/ separators, carry-over of refinery processing materials, and...increase raw vapor emissions from ground fuel handling facilities and during refueling operations. Controlling raw vapor emissions is difficult at 3

Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

NASA Technical Reports Server (NTRS)

Bradley, David E.; Mireles, Omar R.; Hickman, Robert R.

2011-01-01

Deep space missions with large payloads require high specific impulse (Isp) and relatively high thrust in order to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average Isp. Nuclear thermal rockets (NTR) capable of high Isp thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high temperature hydrogen exposure on fuel elements is limited. The primary concern is the mechanical failure of fuel elements which employ high-melting-point metals, ceramics or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via non-contact RF heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

Experimental studies of thermal and chemical interactions between oxide and silicide nuclear fuels with water

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

farahani, A.A.; Corradini, M.L.

Given some transient power/cooling mismatch is a nuclear reactor and its inability to establish the necessary core cooling, energetic fuel-coolant interactions (FCI`s commonly called `vapor explosions`) could occur as a result of the core melting and coolant contact. Although a large number of studies have been done on energetic FCI`s, very few experiments have been performed with the actual fuel materials postulated to be produced in severe accidents. Because of the scarcity of well-characterized FCI data for uranium allows in noncommercial reactors (cermet and silicide fuels), we have conducted a series of experiments to provide a data base for themore » foregoing materials. An existing 1-D shock-tube facility was modified to handle depleted radioactive materials (U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al). Our objectives have been to determine the effects of the initial fuel composition and temperature and the driving pressure (triggering) on the explosion work output, dynamic pressures, transient temperatures, and the hydrogen production. Experimental results indicate limited energetics, mainly thermal interactions, for these fuel materials as compared to aluminum where more chemical reactions occur between the molten aluminum and water.« less

Compact Fuel Element Environment Test

NASA Technical Reports Server (NTRS)

Bradley, D. E.; Mireles, O. R.; Hickman, R. R.; Broadway, J. W.

2012-01-01

Deep space missions with large payloads require high specific impulse (I(sub sp)) and relatively high thrust to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average I(sub sp). Nuclear thermal rockets (NTRs) capable of high I(sub sp) thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3,000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements that employ high melting point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high-temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via noncontact radio frequency heating and expose samples to hydrogen for typical mission durations has been developed to assist in optimal material and manufacturing process selection without employing fissile material. This Technical Memorandum details the test bed design and results of testing conducted to date.

Removal of sulphur-containing odorants from fuel gases for fuel cell-based combined heat and power applications

NASA Astrophysics Data System (ADS)

de Wild, P. J.; Nyqvist, R. G.; de Bruijn, F. A.; Stobbe, E. R.

Natural gas (NG) and liquefied petroleum gas (LPG) are important potential feedstocks for the production of hydrogen for fuel cell-based (e.g. proton exchange membrane fuel cells (PEMFC) or solid oxide fuel Cells (SOFC) combined heat and power (CHP) applications. To prevent detrimental effects on the (electro)catalysts in fuel cell-based combined heat and power installations (FC-CHP), sulphur removal from the feedstock is mandatory. An experimental bench-marking study of adsorbents has identified several candidates for the removal of sulphur containing odorants at low temperature. Among these adsorbents a new material has been discovered that offers an economically attractive means to remove TetraHydroThiophene (THT), the main European odorant, from natural gas at ambient temperature. The material is environmentally benign, easy to use and possesses good activity (residual sulphur levels below 20 ppbv) and capacity for the common odorant THT in natural gas. When compared to state-of-the-art metal-promoted active carbon the new material has a THT uptake capacity that is up to 10 times larger, depending on temperature and pressure. Promoted versions of the new material have shown potential for the removal of THT at higher temperatures and/or for the removal of other odorants such as mercaptans from natural gas or from LPG.

Diffusive tunneling for alleviating Knudsen-layer reactivity reduction under hydrodynamic mix

NASA Astrophysics Data System (ADS)

Tang, Xianzhu; McDevitt, Chris; Guo, Zehua

2017-10-01

Hydrodynamic mix will produce small features for intermixed deuterium-tritium fuel and inert pusher materials. The geometrical characteristics of the mix feature have a large impact on Knudsen layer yield reduction. We considered two features. One is planar structure, and the other is fuel cells segmented by inert pusher material which can be represented by a spherical DT bubble enclosed by a pusher shell. The truly 3D fuel feature, the spherical bubble, has the largest degree of yield reduction, due to fast ions being lost in all directions. The planar fuel structure, which can be regarded as 1D features, has modest amount of potential for yield degradation. While the increasing yield reduction with increasing Knudsen number of the fuel region is straightforwardly anticipated, we also show, by a combination of direct simulation and simple model, that once the pusher materials is stretched sufficiently thin by hydrodynamic mix, the fast fuel ions diffusively tunnel through them with minimal energy loss, so the Knudsen layer yield reduction becomes alleviated. This yield recovery can occur in a chunk-mixed plasma, way before the far more stringent, asymptotic limit of an atomically homogenized fuel and pusher assembly. Work supported by LANL LDRD program.

Evaluation of a standard test method for screening fuels in soils

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

Sorini, S.S.; Schabron, J.F.

1996-12-31

A new screening method for fuel contamination in soils was recently developed as American Society for Testing and Materials (ASTM) Method D-5831-95, Standard Test Method for Screening Fuels in Soils. This method uses low-toxicity chemicals and can be sued to screen organic- rich soils, as well as being fast, easy, and inexpensive to perform. Fuels containing aromatic compounds, such as diesel fuel and gasoline, as well as other aromatic-containing hydrocarbon materials, such as motor oil, crude oil, and cola oil, can be determined. The screening method for fuels in soils was evaluated by conducting a Collaborative study on the method.more » In the Collaborative study, a sand and an organic soil spiked with various concentrations of diesel fuel were tested. Data from the Collaborative study were used to determine the reproducibility (between participants) and repeatability (within participants) precision of the method for screening the test materials. The Collaborative study data also provide information on the performance of portable field equipment (patent pending) versus laboratory equipment for performing the screening method and a comparison of diesel concentration values determined using the screening method versus a laboratory method.« less

A polycrystal plasticity model of strain localization in irradiated iron

NASA Astrophysics Data System (ADS)

Barton, Nathan R.; Arsenlis, Athanasios; Marian, Jaime

2013-02-01

At low to intermediate homologous temperatures, the degradation of structural materials performance in nuclear environments is associated with high number densities of nanometric defects produced in irradiation cascades. In polycrystalline ferritic materials, self-interstitial dislocations loops are a principal signature of irradiation damage, leading to a mechanical response characterized by increased yield strengths, decreased total strain to failure, and decreased work hardening as compared to the unirradiated behavior. Above a critical defect concentration, the material deforms by plastic flow localization, giving rise to strain softening in terms of the engineering stress-strain response. Flow localization manifests itself in the form of defect-depleted crystallographic channels, through which all dislocation activity is concentrated. In this paper, we describe the formulation of a crystal plasticity model for pure Fe embedded in a finite element polycrystal simulator and present results of uniaxial tensile deformation tests up to 10% strain. We use a tensorial damage descriptor variable to capture the evolution of the irradiation damage loop subpopulation during deformation. The model is parameterized with detailed dislocation dynamics simulations of tensile tests up to 1.5% deformation of systems containing various initial densities of irradiation defects. The coarse-grained simulations are shown to capture the essential details of the experimental stress response observed in ferritic alloys and steels. Our methodology provides an effective linkage between the defect scale, of the order of one nanometer, and the continuum scale involving multiple grain orientations.

Burnable absorber arrangement for fuel bundle

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

Crowther, R.L.; Townsend, D.B.

1986-12-16

This patent describes a boiling water reactor core whose operation is characterized by a substantial proportion of steam voids with concomitantly reduced moderation toward the top of the core when the reactor is in its hot operating condition. The reduced moderation leads to slower burnup and greater conversion ratio in an upper core region so that when the reactor is in its cold shut down condition the resulting relatively increased moderation in the upper core region is accompanied by a reactivity profile that peaks in the upper core region. A fuel assembly is described comprising; a component of fissile materialmore » distributed over a substantial axial extent of the fuel assembly; and a component of neutron absorbing material having an axial distribution characterized by an enhancement in an axial zone of the fuel assembly, designated the cold shutdown control zone, corresponding to at least a portion of the axial region of the core when the cold shutdown reactivity peaks. The aggregate amount of neutron absorbing material in the cold shutdown zone of the fuel assembly is greater than the aggregate amount of neutron absorbing material in the axial zones of the fuel assembly immediately above and immediately below the cold shutdown control zone whereby the cold shutdown reactivity peak is reduced relative to the cold shutdown reactivity in the zones immediately above and immediately below the cold shutdown control zone. The cold shutdown zone has an axial extent measured from the bottom of the fuel assembly in the range between 68-88 percent of the height of the fissile material in the fuel assembly.« less

Design, quality, and quality assurance of solid recovered fuels for the substitution of fossil feedstock in the cement industry.

PubMed

Sarc, R; Lorber, K E; Pomberger, R; Rogetzer, M; Sipple, E M

2014-07-01

This paper describes the requirements for the production, quality, and quality assurance of solid recovered fuels (SRF) that are increasingly used in the cement industry. Different aspects have to be considered before using SRF as an alternative fuel. Here, a study on the quality of SRF used in the cement industry is presented. This overview is completed by an investigation of type and properties of input materials used at waste splitting and SRF production plants in Austria. As a simplified classification, SRF can be divided into two classes: a fine, high-calorific SRF for the main burner, or coarser SRF material with low calorific value for secondary firing systems, such as precombustion chambers or similar systems. In the present study, SRFs coming from various sources that fall under these two different waste fuel classes are discussed. Both SRFs are actually fired in the grey clinker kiln of the Holcim (Slovensko) plant in Rohožnik (Slovakia). The fine premium-quality material is used in the main burner and the coarse regular-quality material is fed to a FLS Hotdisc combustion device. In general, the alternative fuels are used instead of their substituted fossil fuels. For this, chemical compositions and other properties of SRF were compared to hard coal as one of the most common conventional fuels in Europe. This approach allows to compare the heavy metal input from traditional and alternative fuels and to comment on the legal requirements on SRF that, at the moment, are under development in Europe. © The Author(s) 2014.

United States and Russian Cooperation on Issues of Nuclear Nonproliferation

DTIC Science & Technology

2005-06-01

Reactors ( RERTR ) This project works with Russia to facilitate conversion of its research and test reactors from highly enriched uranium (HEU) fuel...reactor fuel purchase, accelerated RERTR activities, and accelerated Material Conversion and Consolidation implementation. 89 j. Fissile Materials

Method of forming a package for MEMS-based fuel cell

DOEpatents

Morse, Jeffrey D; Jankowski, Alan F

2013-05-21

A MEMS-based fuel cell package and method thereof is disclosed. The fuel cell package comprises seven layers: (1) a sub-package fuel reservoir interface layer, (2) an anode manifold support layer, (3) a fuel/anode manifold and resistive heater layer, (4) a Thick Film Microporous Flow Host Structure layer containing a fuel cell, (5) an air manifold layer, (6) a cathode manifold support structure layer, and (7) a cap. Fuel cell packages with more than one fuel cell are formed by positioning stacks of these layers in series and/or parallel. The fuel cell package materials such as a molded plastic or a ceramic green tape material can be patterned, aligned and stacked to form three dimensional microfluidic channels that provide electrical feedthroughs from various layers which are bonded together and mechanically support a MEMS-based miniature fuel cell. The package incorporates resistive heating elements to control the temperature of the fuel cell stack. The package is fired to form a bond between the layers and one or more microporous flow host structures containing fuel cells are inserted within the Thick Film Microporous Flow Host Structure layer of the package.

Method of forming a package for mems-based fuel cell

DOEpatents

Morse, Jeffrey D.; Jankowski, Alan F.

2004-11-23

A MEMS-based fuel cell package and method thereof is disclosed. The fuel cell package comprises seven layers: (1) a sub-package fuel reservoir interface layer, (2) an anode manifold support layer, (3) a fuel/anode manifold and resistive heater layer, (4) a Thick Film Microporous Flow Host Structure layer containing a fuel cell, (5) an air manifold layer, (6) a cathode manifold support structure layer, and (7) a cap. Fuel cell packages with more than one fuel cell are formed by positioning stacks of these layers in series and/or parallel. The fuel cell package materials such as a molded plastic or a ceramic green tape material can be patterned, aligned and stacked to form three dimensional microfluidic channels that provide electrical feedthroughs from various layers which are bonded together and mechanically support a MEMOS-based miniature fuel cell. The package incorporates resistive heating elements to control the temperature of the fuel cell stack. The package is fired to form a bond between the layers and one or more microporous flow host structures containing fuel cells are inserted within the Thick Film Microporous Flow Host Structure layer of the package.

Thermal-mechanical performance modeling of thorium-plutonium oxide fuel and comparison with on-line irradiation data

NASA Astrophysics Data System (ADS)

Insulander Björk, Klara; Kekkonen, Laura

2015-12-01

Thorium-plutonium Mixed OXide (Th-MOX) fuel is considered for use in light water reactors fuel due to some inherent benefits over conventional fuel types in terms of neutronic properties. The good material properties of ThO2 also suggest benefits in terms of thermal-mechanical fuel performance, but the use of Th-MOX fuel for commercial power production demands that its thermal-mechanical behavior can be accurately predicted using a well validated fuel performance code. Given the scant operational experience with Th-MOX fuel, no such code is available today. This article describes the first phase of the development of such a code, based on the well-established code FRAPCON 3.4, and in particular the correlations reviewed and chosen for the fuel material properties. The results of fuel temperature calculations with the code in its current state of development are shown and compared with data from a Th-MOX test irradiation campaign which is underway in the Halden research reactor. The results are good for fresh fuel, whereas experimental complications make it difficult to judge the adequacy of the code for simulations of irradiated fuel.

Carbon nanotube-like materials in the exhaust from a diesel engine using gas oil/ethanol mixing fuel with catalysts and sulfur.

PubMed

Suzuki, Shunsuke; Mori, Shinsuke

2017-08-01

Particulate matter from a diesel engine, including soot and carbon nanomaterials, was collected on a sampling holder and the structure of the materials was studied by transmission electron microscopy (TEM) and scanning electron microscopy (SEM). As a result of employing gas oil/ethanol mixing fuel with sulfur and ferrocene/molybdenum as catalyst sources, formation of carbon nanotubes (CNT)-like materials in addition to soot was observed in the exhaust gas from a diesel engine. It was revealed that CNT-like materials were included among soot in our system only when the following three conditions were satisfied simultaneously: high ethanol fraction in fuel, high sulfur loading, and presence of catalyst sources in fuel. This study confirmed that if at least one of these three conditions was not satisfied, CNT-like materials were not observed in the exhaust from a diesel engine. These experimental results shown in this work provide insights into understanding CNT-like material formation mechanism in a diesel engine. Recent papers reported that carbon nanotube-like materials were included in the exhaust gas from engines, but conditions for carbon nanotube-like material formation have not been well studied. This work provides the required conditions for carbon nanotube-like material growth in a diesel engine, and this will be helpful for understanding the carbon nanotube-like material formation mechanism and taking countermeasures to preventing carbon nanotube-like material formation in a diesel engine.

Balanced program plan. Analysis for biomedical and environmental research

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

Not Available

1976-06-01

Major issues associated with the use of nuclear power are health hazards of exposure to radioactive materials; sources of radiation exposure; reactor accidents; sabotage of nuclear facilities; diversion of fissile material and its use for extortion; and the presence of plutonium in the environment. Fission fuel cycle technology is discussed with regard to milling, UF/sub 6/ production, uranium enrichment, plutonium fuel fabrication, power production, fuel processing, waste management, and fuel and waste transportation. The following problem areas of fuel cycle technology are briefly discussed: characterization, measurement, and monitoring; transport processes; health effects; ecological processes and effects; and integrated assessment. Estimatedmore » program unit costs are summarized by King-Muir Category. (HLW)« less

Overview of the DOE/SERI Biochemical Conversion Program

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

Wright, J D

1986-09-01

The Solar Energy Research Institute manages a program of research and development on the biochemical conversion of renewable lignocellulosic materials to liquid fuels for the Department of Energy's Biofuels and Municipal Waste Technology Division. The Biochemical Conversion Program is mission oriented so effort is concentrated on technologies which appear to have the greatest potential for being adopted by the private sector to economically convert lignocellulosic materials into high value liquid transportation fuels such as ethanol. The program is structured to supply the technology for such fuels to compete economically first as an octane booster or fuel additive, and, with additionalmore » improvements, as a neat fuel. 18 refs., 3 figs., 1 tab.« less

Materials for solar fuels and chemicals.

PubMed

Montoya, Joseph H; Seitz, Linsey C; Chakthranont, Pongkarn; Vojvodic, Aleksandra; Jaramillo, Thomas F; Nørskov, Jens K

2016-12-20

The conversion of sunlight into fuels and chemicals is an attractive prospect for the storage of renewable energy, and photoelectrocatalytic technologies represent a pathway by which solar fuels might be realized. However, there are numerous scientific challenges in developing these technologies. These include finding suitable materials for the absorption of incident photons, developing more efficient catalysts for both water splitting and the production of fuels, and understanding how interfaces between catalysts, photoabsorbers and electrolytes can be designed to minimize losses and resist degradation. In this Review, we highlight recent milestones in these areas and some key scientific challenges remaining between the current state of the art and a technology that can effectively convert sunlight into fuels and chemicals.

Pyroprocessing of Light Water Reactor Spent Fuels Based on an Electrochemical Reduction Technology

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

Ohta, Hirokazu; Inoue, Tadashi; Sakamura, Yoshiharu

A concept of pyroprocessing light water reactor (LWR) spent fuels based on an electrochemical reduction technology is proposed, and the material balance of the processing of mixed oxide (MOX) or high-burnup uranium oxide (UO{sub 2}) spent fuel is evaluated. Furthermore, a burnup analysis for metal fuel fast breeder reactors (FBRs) is conducted on low-decontamination materials recovered by pyroprocessing. In the case of processing MOX spent fuel (40 GWd/t), UO{sub 2} is separately collected for {approx}60 wt% of the spent fuel in advance of the electrochemical reduction step, and the product recovered through the rare earth (RE) removal step, which hasmore » the composition uranium:plutonium:minor actinides:fission products (FPs) = 76.4:18.4:1.7:3.5, can be applied as an ingredient of FBR metal fuel without a further decontamination process. On the other hand, the electroreduced alloy of high-burnup UO{sub 2} spent fuel (48 GWd/t) requires further decontamination of residual FPs by an additional process such as electrorefining even if RE FPs are removed from the alloy because the recovered plutonium (Pu) is accompanied by almost the same amount of FPs in addition to RE. However, the amount of treated materials in the electrorefining step is reduced to {approx}10 wt% of the total spent fuel owing to the prior UO{sub 2} recovery step. These results reveal that the application of electrochemical reduction technology to LWR spent oxide fuel is a promising concept for providing FBR metal fuel by a rationalized process.« less

Nuclear reactor fuel containment safety structure

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

Rosewell, M.P.

A nuclear reactor fuel containment safety structure is disclosed and is shown to include an atomic reactor fuel shield with a fuel containment chamber and exhaust passage means, and a deactivating containment base attached beneath the fuel reactor shield and having exhaust passages, manifold, and fluxing and control material and vessels. 1 claim, 8 figures.

40 CFR 52.1881 - Control strategy: Sulfur oxides (sulfur dioxide).

Code of Federal Regulations, 2013 CFR

2013-07-01

... sulfur oxides. (iii) Fossil fuel means natural gas, refinery fuel gas, coke oven gas, petroleum, coal and any form of solid, liquid, or gaseous fuel derived from such materials. (iv) Fossil fuel-fired steam generating unit means a furnace or boiler used in the process of burning fossil fuel for the purpose of...

40 CFR 52.1881 - Control strategy: Sulfur oxides (sulfur dioxide).

Code of Federal Regulations, 2014 CFR

2014-07-01

... sulfur oxides. (iii) Fossil fuel means natural gas, refinery fuel gas, coke oven gas, petroleum, coal and any form of solid, liquid, or gaseous fuel derived from such materials. (iv) Fossil fuel-fired steam generating unit means a furnace or boiler used in the process of burning fossil fuel for the purpose of...

40 CFR 52.1881 - Control strategy: Sulfur oxides (sulfur dioxide).

Code of Federal Regulations, 2010 CFR

2010-07-01

... sulfur oxides. (iii) Fossil fuel means natural gas, refinery fuel gas, coke oven gas, petroleum, coal and any form of solid, liquid, or gaseous fuel derived from such materials. (iv) Fossil fuel-fired steam generating unit means a furnace or boiler used in the process of burning fossil fuel for the purpose of...

40 CFR 52.1881 - Control strategy: Sulfur oxides (sulfur dioxide).

Code of Federal Regulations, 2012 CFR

2012-07-01

... sulfur oxides. (iii) Fossil fuel means natural gas, refinery fuel gas, coke oven gas, petroleum, coal and any form of solid, liquid, or gaseous fuel derived from such materials. (iv) Fossil fuel-fired steam generating unit means a furnace or boiler used in the process of burning fossil fuel for the purpose of...

40 CFR 52.1881 - Control strategy: Sulfur oxides (sulfur dioxide).

Code of Federal Regulations, 2011 CFR

2011-07-01

... sulfur oxides. (iii) Fossil fuel means natural gas, refinery fuel gas, coke oven gas, petroleum, coal and any form of solid, liquid, or gaseous fuel derived from such materials. (iv) Fossil fuel-fired steam generating unit means a furnace or boiler used in the process of burning fossil fuel for the purpose of...

3-Dimensional Computational Fluid Dynamics Modeling of Solid Oxide Fuel Cell Using Different Fuels

DTIC Science & Technology

2011-01-01

major types of fuel cells in practice are listed below: Polymer Electrolyte Membrane Fuel Cell ( PEMFC ) Alkaline Fuel cell (AFC) Phosphoric Acid...Material Operating Temperature (oC) Efficiency (%) PEMFC H2, Methanol, Formic Acid Hydrated Organic Polymer < 90 40-50 AFC Pure H2 Aqueous

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

DOE PAGES

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig; ...

2017-07-10

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

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

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

Properties and degradation of the gasket component of a proton exchange membrane fuel cell--a review.

PubMed

Basuli, Utpal; Jose, Jobin; Lee, Ran Hee; Yoo, Yong Hwan; Jeong, Kwang-Un; Ahn, Jou-Hyeon; Nah, Changwoon

2012-10-01

Proton exchange membrane (PEM) fuel cell stack requires gaskets and seals in each cell to keep the reactant gases within their respective regions. Gasket performance is integral to the successful long-term operation of a fuel cell stack. This review focuses on properties, performance and degradation mechanisms of the different polymer gasket materials used in PEM fuel cell under normal operating conditions. The different degradation mechanisms and their corresponding representative mitigation strategies are also presented here. Summary of various properties of elastomers and their advantages and disadvantages in fuel cell'environment are presented. By considering the level of chemical degradation, mechanical properties and cost effectiveness, it can be proposed that EPDM is one of the best choices for gasket material in PEM fuel cell. Finally, the challenges that remain in using rubber component as in PEM fuel cell, as well as the prospects for exploiting them in the future are discussed.

Hybrid fusion-fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

NASA Astrophysics Data System (ADS)

Shmelev, A. N.; Kulikov, G. G.; Kurnaev, V. A.; Salahutdinov, G. H.; Kulikov, E. G.; Apse, V. A.

2015-12-01

Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the 231Pa-232U-233U-Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of 232U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

Induction Heating Model of Cermet Fuel Element Environmental Test (CFEET)

NASA Technical Reports Server (NTRS)

Gomez, Carlos F.; Bradley, D. E.; Cavender, D. P.; Mireles, O. R.; Hickman, R. R.; Trent, D.; Stewart, E.

2013-01-01

Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames. Nuclear Thermal Rockets (NTR) are capable of producing a high specific impulse by employing heat produced by a fission reactor to heat and therefore accelerate hydrogen through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements due to large thermal gradients; therefore, high-melting-point ceramics-metallic matrix composites (cermets) are one of the fuels under consideration as part of the Nuclear Cryogenic Propulsion Stage (NCPS) Advance Exploration System (AES) technology project at the Marshall Space Flight Center. The purpose of testing and analytical modeling is to determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures and obtain data to assess the properties of the non-nuclear support materials. The fission process and the resulting heating performance are well known and do not require that active fissile material to be integrated in this testing. A small-scale test bed; Compact Fuel Element Environmental Tester (CFEET), designed to heat fuel element samples via induction heating and expose samples to hydrogen is being developed at MSFC to assist in optimal material and manufacturing process selection without utilizing fissile material. This paper details the analytical approach to help design and optimize the test bed using COMSOL Multiphysics for predicting thermal gradients induced by electromagnetic heating (Induction heating) and Thermal Desktop for radiation calculations.

Nuclear Fuels & Materials Spotlight Volume 5

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

Petti, David Andrew

2016-10-01

As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • Evaluation and modeling of light water reactor accident tolerant fuel concepts • Status and results of recent TRISO-coated particle fuel irradiations, post-irradiation examinations, high-temperature safety testing to demonstrate the accident performance of this fuel system, and advanced microscopy to improve the understanding of fission product transport in this fuel system.more » • Improvements in and applications of meso and engineering scale modeling of light water reactor fuel behavior under a range of operating conditions and postulated accidents (e.g., power ramping, loss of coolant accident, and reactivity initiated accidents) using the MARMOT and BISON codes. • Novel measurements of the properties of nuclear (actinide) materials under extreme conditions, (e.g. high pressure, low/high temperatures, high magnetic field) to improve the scientific understanding of these materials. • Modeling reactor pressure vessel behavior using the GRIZZLY code. • New methods using sound to sense temperature inside a reactor core. • Improved experimental capabilities to study the response of fusion reactor materials to a tritium plasma. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at Idaho National Laboratory, and hope that you find this issue informative.« less

High temperature fuel/emitter system for advanced thermionic fuel elements

NASA Astrophysics Data System (ADS)

Moeller, Helen H.; Bremser, Albert H.; Gontar, Alexander; Fiviesky, Evgeny

1997-01-01

Specialists in space applications are currently focusing on bimodal power systems designed to provide both electric power and thermal propulsion (Kennedy, 1994 and Houts, 1995). Our work showed that thermionics is a viable technology for nuclear bimodal power systems. We demonstrated that materials for a thermionic fuel-emitter combination capable of performing at operating temperatures of 2473 K are not only possible but available. The objective of this work, funded by the US Department of Energy, Office of Space and Defense Power Systems, was to evaluate the compatibility of fuel material consisting of an uranium carbide/tantalum carbide solid solution with an emitter material consisting of a monocrystalline tungsten-niobium alloy. The uranium loading of the fuel material was 70 mole% uranium carbide. The program was successfully accomplished by a B&W/SIA LUTCH team. Its workscope was integrated with tasks being performed at both Babcock & Wilcox, Lynchburg Research Center, Lynchburg, Virginia, and SIA LUTCH, Podolsk, Russia. Samples were fabricated by LUTCH and seven thermal tests were performed in a hydrogen atmosphere. The first preliminary test was performed at 2273 K by LUTCH, and the remaining six tests were performed At B&W. Three tests were performed at 2273 K, two at 2373 K, and the final test at 2473 K. The results showed that the fuel and emitter materials were compatible in the presence of hydrogen. No evidence of liquid formation, dissolution of the uranium carbide from the uranium carbide/tantalum carbide solid solution, or diffusion of the uranium into the monocrystalline tungsten alloy was observed. Among the highlights of the program was the successful export of the fuel samples from Russia and their import into the US by commercial transport. This paper will discuss the technical aspects of this work.

FY2016 Propulsion Materials Annual Progress Report

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

None, None

The Propulsion Materials Program actively supports the energy security and reduction of greenhouse emissions goals of VTO by investigating and identifying the materials properties that are most essential for continued development of cost-effective, highly efficient, and environmentally friendly next-generation heavy and light-duty powertrains. The technical approaches available to enhance propulsion systems focus on improvements in both vehicle efficiency and fuel substitution, both of which must overcome the performance limitations of the materials currently in use. Propulsion Materials Program activities work with national laboratories, industry experts, and VTO powertrain systems (e.g., Advanced Combustion Engines and Fuels) teams to develop strategies thatmore » overcome materials limitations in future powertrain performance. The technical maturity of the portfolio of funded projects ranges from basic science to subsystem prototype validation. Projects within a Propulsion Materials Program activity address materials concerns that directly impact critical technology barriers within each of the above programs, including barriers that impact fuel efficiency, thermal management, emissions reduction, improved reliability, and reduced manufacturing costs. The program engages only the barriers that result from material property limitations and represent fundamental, high-risk materials issues.« less

Heavy Duty and Industrial Alternative Fuel Applications. Forklift and Material Handling. Alternative Fuels Training.

ERIC Educational Resources Information Center

Eckert, Doug; Casto, Lori

This training manual is designed to lay the foundation for trainers and technicians by showing the steps to achieve and maintain good indoor air quality through use of cleaner-burning forklifts and materials handlers. The first part of the manual consists of nine units that provide informational material and diagrams on these topics: comparison of…

Stability of solid oxide fuel cell materials

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

Armstrong, T.R.; Bates, J.L.; Chick, L.A.

1996-04-01

Interconnection materials in a solid oxide fuel cell are exposed to both highly oxidizing conditions at the cathode and to highly reducing conditions at the anode. The thermal expansion characteristics of substituted lanthanum and yttrium chromite interconnect materials were evaluated by dilatometry as a function of oxygen partial pressures from 1 atm to 10{sup -18} atm, controlled using a carbon dioxide/hydrogen buffer.

Theoretical Studies in Enhancing the Efficiency of Cathode and Anode Materials in PEMFC (Proton Exchange Membrane Fuel Cells)

DTIC Science & Technology

2011-03-04

efficiency of cathode and anode materials in PEMFC (Proton Exchange Membrane Fuel Cells) 5a. CONTRACT NUMBER FA23861014012 5b. GRANT NUMBER 5c. PROGRAM...Rev. 8-98) Prescribed by ANSI Std Z39-18 Theoretical studies in enhancing the efficiency of cathode and anode materials in PEMFC (Proton Exchange

Characterization and Detailed Analysis of Regression Behavior for HTPB Solid Fuels Containing High Aluminum Loadings

NASA Technical Reports Server (NTRS)

Kibbey, Timothy P.; Cortopassi, Andrew C.; Boyer, Eric C.

2017-01-01

NASA Marshall Space Flight Center's Materials and Processes Department, with support from the Propulsion Systems Department, has renewed the development and maintenance of a hybrid test bed for exposing ablative thermal protection materials to an environment similar to that seen in solid rocket motors (SRM). The Solid Fuel Torch (SFT), operated during the Space Shuttle program, utilized gaseous oxygen for oxidizer and an aluminized hydroxyl-terminated polybutadiene (HTPB) fuel grain to expose a converging section of phenolic material to a 400 psi, 2-phase flow combustion environment. The configuration allows for up to a 2 foot long, 5 inch diameter fuel grain cartridge. Wanting to now test rubber insulation materials with a turn-back feature to mimic the geometry of an aft dome being impinged by alumina particles, the throat area has now been increased by several times to afford flow similarity. Combined with the desire to maintain a higher operating pressure, the oxidizer flow rate is being increased by a factor of 10. Out of these changes has arisen the need to characterize the fuel/oxidizer combination in a higher mass flux condition than has been previously tested at MSFC, and at which the literature has little to no reporting as well. For (especially) metalized fuels, hybrid references have pointed out possible dependence of fuel regression rate on a number of variables: mass flux, G - oxidizer only (G0), or - total mass flux (Gtot), Length, L, Pressure, P, and Diameter, D.

Electrochemistry for Energy Conversion

NASA Astrophysics Data System (ADS)

O'Hayre, Ryan

2010-10-01

Imagine a laptop computer that runs for 30 hours on a single charge. Imagine a world where you plug your house into your car and power lines are a distant memory. These dreams motivate today's fuel cell research. While some dreams (like powering your home with your fuel cell car) may be distant, others (like a 30-hour fuel cell laptop) may be closer than you think. If you are curious about fuel cells---how they work, when you might start seeing them in your daily life--- this talk is for you. Learn about the state-of-the art in fuel cells, and where the technology is likely to be headed in the next 20 years. You'll also be treated to several ``behind-the scenes'' glimpses of cutting-edge research projects under development in the Renewable Energy Materials Center at the Colorado School of Mines--- projects like an ``ionic transistor'' that works with protons instead of electrons, and a special ceramic membrane material that enables the ``uphill'' diffusion of steam. Associate Professor Ryan O'Hayre's laboratory at the Colorado School of Mines develops new materials and devices to enable alternative energy technologies including fuel cells and solar cells. Prof. O'Hayre and his students collaborate with the Colorado Fuel Cell Center, the Colorado Center for Advanced Ceramics, the Renewable Energy Materials Science and Engineering Center, and the National Renewable Energy Laboratory.[4pt] In collaboration with Ann Deml, Jianhua Tong, Svitlana Pylypenko, Archana Subramaniyan, Micahael Sanders, Jason Fish, Annette Bunge, Colorado School of Mines.

324 Building spent fuel segments pieces and fragments removal summary report

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

SMITH, C L

2003-01-09

As part of the 324 Building Deactivation Project, all Spent Nuclear Fuel (SNF) and Special Nuclear Material were removed. The removal entailed packaging the material into a GNS-12 cask and shipping it to the Central Waste Complex (CWC).

Alternative Fuels Data Center

Science.gov Websites

mixture containing at least 85% methanol, denatured ethanol, or other alcohols; natural gas, propane , hydrogen, or coal derived liquid fuels; or fuels derived from biological materials. PEVs are defined as

The myth of the ``proliferation-resistant'' closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Lyman, Edwin S.

2000-07-01

National nuclear energy programs that engage in reprocessing of spent nuclear fuel (SNF) and the development of "closed" nuclear fuel cycles based on the utilization of plutonium process and store large quantities of weapons-usable nuclear materials in forms vulnerable to diversion or theft by national or subnational groups. Proliferation resistance, an idea dating back at least as far as the International Fuel Cycle Evaluation (INFCE) of the late 1970s, is a loosely defined term referring to processes for chemical separation of SNF that do not extract weapons-usable materials in a purified form.

Multiphase transport in polymer electrolyte membrane fuel cells

NASA Astrophysics Data System (ADS)

Gauthier, Eric D.

Polymer electrolyte membrane fuel cells (PEMFCs) enable efficient conversion of fuels to electricity. They have enormous potential due to the high energy density of the fuels they utilize (hydrogen or alcohols). Power density is a major limitation to wide-scale introduction of PEMFCs. Power density in hydrogen fuel cells is limited by accumulation of water in what is termed fuel cell `flooding.' Flooding may occur in either the gas diffusion layer (GDL) or within the flow channels of the bipolar plate. These components comprise the electrodes of the fuel cell and balance transport of reactants/products with electrical conductivity. This thesis explores the role of electrode materials in the fuel cell and examines the fundamental connection between material properties and multiphase transport processes. Water is generated at the cathode catalyst layer. As liquid water accumulates it will utilize the largest pores in the GDL to go from the catalyst layer to the flow channels. Water collects to large pores via lateral transport at the interface between the GDL and catalyst layer. We have shown that water may be collected in these large pores from several centimeters away, suggesting that we could engineer the GDL to control flooding with careful placement and distribution of large flow-directing pores. Once liquid water is in the flow channels it forms slugs that block gas flow. The slugs are pushed along the channel by a pressure gradient that is dependent on the material wettability. The permeable nature of the GDL also plays a major role in slug growth and allowing bypass of gas between adjacent channels. Direct methanol fuel cells (DMFCs) have analogous multiphase flow issues where carbon dioxide bubbles accumulate, `blinding' regions of the fuel cell. This problem is fundamentally similar to water management in hydrogen fuel cells but with a gas/liquid phase inversion. Gas bubbles move laterally through the porous GDL and emerge to form large bubbles within the flow channel. We have compared the role of GDL materials in liquid drop and gas bubble formation and movement within fuel cells.

Ablation study of tungsten-based nuclear thermal rocket fuel

NASA Astrophysics Data System (ADS)

Smith, Tabitha Elizabeth Rose

The research described in this thesis has been performed in order to support the materials research and development efforts of NASA Marshall Space Flight Center (MSFC), of Tungsten-based Nuclear Thermal Rocket (NTR) fuel. The NTR was developed to a point of flight readiness nearly six decades ago and has been undergoing gradual modification and upgrading since then. Due to the simplicity in design of the NTR, and also in the modernization of the materials fabrication processes of nuclear fuel since the 1960's, the fuel of the NTR has been upgraded continuously. Tungsten-based fuel is of great interest to the NTR community, seeking to determine its advantages over the Carbide-based fuel of the previous NTR programs. The materials development and fabrication process contains failure testing, which is currently being conducted at MSFC in the form of heating the material externally and internally to replicate operation within the nuclear reactor of the NTR, such as with hot gas and RF coils. In order to expand on these efforts, experiments and computational studies of Tungsten and a Tungsten Zirconium Oxide sample provided by NASA have been conducted for this dissertation within a plasma arc-jet, meant to induce ablation on the material. Mathematical analysis was also conducted, for purposes of verifying experiments and making predictions. The computational method utilizes Anisimov's kinetic method of plasma ablation, including a thermal conduction parameter from the Chapman Enskog expansion of the Maxwell Boltzmann equations, and has been modified to include a tangential velocity component. Experimental data matches that of the computational data, in which plasma ablation at an angle shows nearly half the ablation of plasma ablation at no angle. Fuel failure analysis of two NASA samples post-testing was conducted, and suggestions have been made for future materials fabrication processes. These studies, including the computational kinetic model at an angle and the ablation of the NASA sample, could be applied to an atmospheric reentry body, reentering at a ballistic trajectory at hypersonic velocities.

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

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

Gussev, Maxim N.; Field, Kevin G.; Yamamoto, Yukinori

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 bestmore » 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.« less

Effect on fast neutron irradiation to 4 dpa at 400{degrees}C on the properties of V-(4-5)Cr-(4-5)Ti alloys

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

Zinkle, S.J.; Alexander, D.J.; Robertson, J.P.

1997-04-01

Tensile, Charpy impact and electrical resistivity measurements have been performed at ORNL on V-4Cr-4Ti and V-5Cr-5Ti specimens that were prepared at ANL and irradiated in the lithium-bonded X530 experiment in the EBR-II fast reactor. All of the specimens were irradiated to a damage level of about 4 dpa at a temperature of {approximately}400{degrees}C. A significant amount of radiation hardening was evident in both the tensile and Charpy impact tests. The irradiated V-4Cr-4Ti yield strength measured at {approximately}390{degrees}C was >800 MPa, which is more than three times as high as the unirradiated value. The uniform elongations of the irradiated tensile specimensmore » were typically {approximately}1%, with corresponding total elongations of 4-6%. The ductile to brittle transition temperature of the irradiated specimens was less than the unirradiated resistivity, which suggests that hardening associated with interstitial solute pickup was minimal.« less

Topological and morphological analysis of gamma rays irradiated chitosan-poly (vinyl alcohol) blends using atomic force microscopy

NASA Astrophysics Data System (ADS)

Bhatt, Rinkesh; Bisen, D. S.; Bajpai, R.; Bajpai, A. K.

2017-04-01

In the present communication, binary blends of poly (vinyl alcohol) (PVA) and chitosan (CS) were prepared by solution cast method and the roughness parameters of PVA, native CS and CS-PVA blend films were determined using atomic force microscopy (AFM). Moreover, the changes in the morphology of the samples were also investigated after irradiation of gamma rays at absorbed dose of 1 Mrad and 10 Mrad for the scanning areas of 5×5 μm2, 10×10 μm2 and 20×20 μm2. Amplitude, statistical and spatial parameters, including line, 3D and 2D image profiles of the experimental surfaces were examined and compared to un-irradiated samples. For gamma irradiated CS-PVA blends the larger waviness over the surface was found as compared to un-irradiated CS-PVA blends but the values of average roughness for both the films were found almost same. The coefficient of skewness was positive for gamma irradiated CS-PVA blends which revealed the presence of more peaks than valleys on the blend surfaces.

Fracture toughness of irradiated modified 9Cr-1Mo steel

NASA Astrophysics Data System (ADS)

Kim, Sung Ho; Yoon, Ji-Hyun; Ryu, Woo Seog; Lee, Chan Bock; Hong, Jun Hwa

2009-04-01

The effects of irradiation on fracture toughness of modified 9Cr-1Mo steel in the transition region were investigated. Half size precracked Charpy specimens were irradiated up to 1.2 × 10 21n/cm 2 ( E > 0.1 MeV) at 340 °C and 400 °C in the Korean research reactor. The irradiation induced transition temperature shift for a modified 9Cr-1Mo was evaluated by using the Master Curve methodology. The T0 temperature for the unirradiated specimens were measured as -67.7 °C and -72.4 °C from the tests with standard PCVN (precracked charpy V-notch) and half sized PCVN specimens, respectively. The T0 shifts of specimens after irradiation at 340 °C and 400 °C were 70.7 °C and 66.1 °C, respectively. The Weibull slopes for the fracture toughness data obtained from the unirradiated and irradiated modified 9Cr-1Mo steels were determined to confirm the applicability of master curve methodology to modified 9Cr-1Mo steel.

Energy Materials Center at Cornell: Final Report

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

Abruña, Héctor; Mutolo, Paul F

2015-01-02

The mission of the Energy Materials Center at Cornell (emc 2) was to achieve a detailed understanding, via a combination of synthesis of new materials, experimental and computational approaches, of how the nature, structure, and dynamics of nanostructured interfaces affect energy conversion and storage with emphasis on fuel cells, batteries and supercapacitors. Our research on these systems was organized around a full system strategy for; the development and improved performance of materials for both electrodes at which storage or conversion occurs; understanding their internal interfaces, such as SEI layers in batteries and electrocatalyst supports in fuel cells, and methods formore » structuring them to enable high mass transport as well as high ionic and electronic conductivity; development of ion-conducting electrolytes for batteries and fuel cells (separately) and other separator components, as needed; and development of methods for the characterization of these systems under operating conditions (operando methods) Generally, our work took industry and DOE report findings of current materials as a point of departure to focus on novel material sets for improved performance. In addition, some of our work focused on studying existing materials, for example observing battery solvent degradation, fuel cell catalyst coarsening or monitoring lithium dendrite growth, employing in operando methods developed within the center.« less

Carbonaceous material for production of hydrogen from low heating value fuel gases

DOEpatents

Koutsoukos, Elias P.

1989-01-01

A process for the catalytic production of hydrogen, from a wide variety of low heating value fuel gases containing carbon monoxide, comprises circulating a carbonaceous material between two reactors--a carbon deposition reactor and a steaming reactor. In the carbon deposition reactor, carbon monoxide is removed from a fuel gas and is deposited on the carbonaceous material as an active carbon. In the steaming reactor, the reactive carbon reacts with steam to give hydrogen and carbon dioxide. The carbonaceous material contains a metal component comprising from about 75% to about 95% cobalt, from about 5% to about 15% iron, and up to about 10% chromium, and is effective in suppressing the production of methane in the steaming reactor.

Accident-tolerant oxide fuel and cladding

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

Mariani, Robert D.

Systems and methods for accident tolerant oxide fuel. One or more disks can be placed between fuel pellets comprising UO.sub.2, wherein such disks possess a higher thermal conductivity material than that of the UO.sub.2 to provide enhanced heat rejection thereof. Additionally, a cladding coating comprising zircaloy coated with a material that provides stability and high melting capability can be provided. The pellets can be configured as annular pellets having an annulus filled with the higher thermal conductivity material. The material coating the zircaloy can be, for example, Zr.sub.5Si.sub.4 or another silicide such as, for example, a Zr-Silicide that limits corrosion.more » The aforementioned higher thermal conductivity material can be, for example, Si, Zr.sub.xSi.sub.y, Zr, or Al.sub.2O.sub.3.« less

Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

DTIC Science & Technology

2013-06-01

Densities ............................................................................................................ 21 2.3 Fuel Mass (Core Total...70 7.1 Geometry, Material Density, and Mass Summary for All Cores...21 Table 3: Fuel Rod Masses for Different Clads

Effect of gamma irradiation on the antimicrobial and free radical scavenging activities of Glycyrrhiza glabra root

NASA Astrophysics Data System (ADS)

Fatima Khattak, Khanzadi; James Simpson, Thomas

2010-04-01

The efficacy of gamma irradiation as a method of decontamination for food and herbal materials is well established. In the present study, Glycyrrhiza glabra roots were irradiated at doses 5, 10, 15, 20 and 25 kGy in a cobalt-60 irradiator. The irradiated and un-irradiated control samples were evaluated for phenolic contents, antimicrobial activities and DPPH scavenging properties. The result of the present study showed that radiation treatment up to 20 kGy does not affect the antifungal and antibacterial activity of the plant. While sample irradiated at 25 kGy does showed changes in the antibacterial activity against some selected pathogens. No significant differences in the phenolic contents were observed for control and samples irradiated at 5, 10 and 15 kGy radiation doses. However, phenolic contents increased in samples treated with 20 and 25 kGy doses. The DPPH scavenging activity significantly ( p<0.05) increased in all irradiated samples of the plant.

The deuterium depth profile in neutron-irradiated tungsten exposed to plasma

NASA Astrophysics Data System (ADS)

Shimada, Masashi; Cao, G.; Hatano, Y.; Oda, T.; Oya, Y.; Hara, M.; Calderoni, P.

2011-12-01

Tungsten samples (99.99% purity from A.L.M.T. Corp., 6 mm in diameter, 0.2 mm in thickness) were irradiated by high-flux neutrons at 50 °C to 0.025 dpa in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Subsequently, the neutron-irradiated tungsten samples were exposed to high-flux deuterium plasmas (ion flux: 1021-1022 m-2 s-1, ion fluence: 1025-1026 m-2) in the Tritium Plasma Experiment at Idaho National Laboratory. This paper reports the results of deuterium depth profiling in neutron-irradiated tungsten exposed to plasmas at 100, 200 and 500 °C via nuclear reaction analysis (NRA). The NRA measurements show that a significant amount of deuterium (>0.1 at.% D/W) remains trapped in the bulk material (up to 5 μm) at 500 °C. Tritium Migration Analysis Program simulation results using the NRA profiles indicate that different trapping mechanisms exist for neutron-irradiated and unirradiated tungsten.

Carbon Ion Irradiation Inhibits Glioma Cell Migration Through Downregulation of Integrin Expression

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

Rieken, Stefan, E-mail: Stefan.Rieken@med.uni-heidelberg.de; Habermehl, Daniel; Wuerth, Lena

2012-05-01

Purpose: To investigate the effect of carbon ion irradiation on glioma cell migration. Methods and Materials: U87 and Ln229 glioma cells were irradiated with photons and carbon ions. Migration was analyzed 24 h after irradiation. Fluorescence-activated cell sorting analysis was performed in order to quantify surface expression of integrins. Results: Single photon doses of 2 Gy and 10 Gy enhanced {alpha}{sub {nu}}{beta}{sub 3} and {alpha}{sub {nu}}{beta}{sub 5} integrin expression and caused tumor cell hypermigration on both vitronectin (Vn) and fibronectin (Fn). Compared to integrin expression in unirradiated cells, carbon ion irradiation caused decreased integrin expression and inhibited cell migration onmore » both Vn and Fn. Conclusion: Photon radiotherapy (RT) enhances the risk of tumor cell migration and subsequently promotes locoregional spread via photon induction of integrin expression. In contrast to photon RT, carbon ion RT causes decreased integrin expression and suppresses glioma cell migration on both Vn and Fn, thus promising improved local control.« less

Effect of helium to dpa ratio on fatigue behavior of austenitic stainless steel irradiated to 2 dpa

NASA Astrophysics Data System (ADS)

Ioka, I.; Yonekawa, M.; Miwa, Y.; Mimura, H.; Tsuji, H.; Hoshiya, T.

2000-12-01

The effect of helium due to nuclear transmutation reactions during neutron irradiation on low cycle fatigue life of type 304 stainless steel was investigated. The specimens were irradiated in spectrally tailored capsules in the Japan Materials Testing Reactor (JMTR) at a temperature of 823 K to a neutron fluence of approximately 1×1025 n/m2 (E>1 MeV) and helium levels of 0.8, 2.5 and 8.1 appm. The low cycle fatigue tests were performed in total axial strain ranges of 0.8-1.6% at 823 K. A laser extensometer was used for controlling the axial strain of a specimen under cyclic testing. The difference between unirradiated and irradiated specimens is quite clear and appears to be a reduction by a factor of 2-5 in fatigue life. The helium concentration of the specimen is not the main factor to shorten fatigue life in the present experimental condition.

A direct reading exposure monitor for radiation processing

NASA Astrophysics Data System (ADS)

Kantz, A. D.; Humpherys, K. C.

Various plastic films have been utilized to measure radiation fields. In general such films are rugged, easily handled, small enough to cause neligible perturbation on the radiation fields, and relatively inexpensive. The radiachromic materials have been shown to have advantages over other plastic fabrications in stability, reproducibility, equivalent response to electron and gamma ray processing fields, dose rate independence, and ready availability of calibration standards. Using a nylon matrix radiachromic detector, a system of direct read-out of absorbed dose has been developed to facilitate monitoring in the megarad region. When an exposed detector is inserted into the reader, the optical transmission signal is processed through an analog to digital converter. The digitized signal addresses a memory bank where the standard response curve is stored. The corresponding absorbed dose is displayed on a digital panel meter. The variation of relative sensitivity of detectors, the background of unirradiated detectors, environmental parameters, and the capacity of the memory bank are contributing factors to the total precision of the read-out system.

Overview of fuel inventory in JET with the ITER-like wall

NASA Astrophysics Data System (ADS)

Widdowson, A.; Coad, J. P.; Alves, E.; Baron-Wiechec, A.; Barradas, N. P.; Brezinsek, S.; Catarino, N.; Corregidor, V.; Heinola, K.; Koivuranta, S.; Krat, S.; Lahtinen, A.; Likonen, J.; Matthews, G. F.; Mayer, M.; Petersson, P.; Rubel, M.; Contributors, JET

2017-08-01

Post mortem analyses of JET ITER-Like-Wall tiles and passive diagnostics have been completed after each of the first two campaigns (ILW-1 and ILW-2). They show that the global fuel inventory is still dominated by co-deposition; hence plasma parameters and sputtering processes affecting material migration influence the distribution of retained fuel. In particular, differences between results from the two campaigns may be attributed to a greater proportion of pulses run with strike points in the divertor corners, and having about 300 discharges in hydrogen at the end of ILW-2. Recessed and remote areas can contribute to fuel retention due to the larger areas involved, e.g. recessed main chamber walls, gaps in castellated Be main chamber tiles and material migration to remote divertor areas. The fuel retention and material migration due to the bulk W Tile 5 during ILW-1 are presented. Overall these tiles account for only a small percentage of the global accountancy for ILW-1.

Integral manifolding structure for fuel cell core having parallel gas flow

DOEpatents

Herceg, Joseph E.

1984-01-01

Disclosed herein are manifolding means for directing the fuel and oxidant gases to parallel flow passageways in a fuel cell core. Each core passageway is defined by electrolyte and interconnect walls. Each electrolyte and interconnect wall consists respectively of anode and cathode materials layered on the opposite sides of electrolyte material, or on the opposite sides of interconnect material. A core wall projects beyond the open ends of the defined core passageways and is disposed approximately midway between and parallel to the adjacent overlaying and underlying interconnect walls to define manifold chambers therebetween on opposite sides of the wall. Each electrolyte wall defining the flow passageways is shaped to blend into and be connected to this wall in order to redirect the corresponding fuel and oxidant passageways to the respective manifold chambers either above or below this intermediate wall. Inlet and outlet connections are made to these separate manifold chambers respectively, for carrying the fuel and oxidant gases to the core, and for carrying their reaction products away from the core.

Integral manifolding structure for fuel cell core having parallel gas flow

DOEpatents

Herceg, J.E.

1983-10-12

Disclosed herein are manifolding means for directing the fuel and oxidant gases to parallel flow passageways in a fuel cell core. Each core passageway is defined by electrolyte and interconnect walls. Each electrolyte and interconnect wall consists respectively of anode and cathode materials layered on the opposite sides of electrolyte material, or on the opposite sides of interconnect material. A core wall projects beyond the open ends of the defined core passageways and is disposed approximately midway between and parallel to the adjacent overlaying and underlying interconnect walls to define manifold chambers therebetween on opposite sides of the wall. Each electrolyte wall defining the flow passageways is shaped to blend into and be connected to this wall in order to redirect the corresponding fuel and oxidant passageways to the respective manifold chambers either above or below this intermediate wall. Inlet and outlet connections are made to these separate manifold chambers respectively, for carrying the fuel and oxidant gases to the core, and for carrying their reaction products away from the core.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-07-11

Nuclear reactor fuel elements of the type in which the flssionsble material is in ceramic form, such as uranium dioxide, are described. The fuel element is comprised of elongated inner and outer concentric spaced tubular members providing an annular space therebetween for receiving the fissionable material, the annular space being closed at both ends and the inner tube being open at both ends. The fuel is in the form of compressed pellets of ceramic fissionsble material having the configuration of split bushings formed with wedge surfaces and arranged in seriated inner and outer concentric groups which are urged against the respective tubes in response to relative axial movement of the pellets in the direction toward each other. The pairs of pellets are axially urged together by a resilient means also enclosed within the annulus. This arrangement-permits relative axial displacement of the pellets during use dial stresses on the inner and outer tube members and yet maintains the fuel pellets in good thermal conductive relationship therewith.

Fuel: Logs, sticks, needles, duff, and much more

Treesearch

Russell T. Graham; Theresa Benevidez Jain; Alan E. Harvey

2000-01-01

Fuels burned by either prescribed or wildfires are complex and important components of forested ecosystems. Fine fuels consisting of fallen limbs, twigs, and leaves of shrubs and trees are rich in nutrients. If these fuels are not immediately burned, nutrients can leach from these materials into the forest floor, especially if they overwinter. Larger fuels consisting...

Feasibility of processing the experimental breeder reactor-II driver fuel from the Idaho National Laboratory through Savannah River Site's H-Canyon facility

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

Magoulas, V. E.

Savannah River National Laboratory (SRNL) was requested to evaluate the potential to receive and process the Idaho National Laboratory (INL) uranium (U) recovered from the Experimental Breeder Reactor II (EBR-II) driver fuel through the Savannah River Site’s (SRS) H-Canyon as a way to disposition the material. INL recovers the uranium from the sodium bonded metallic fuel irradiated in the EBR-II reactor using an electrorefining process. There were two compositions of EBR-II driver fuel. The early generation fuel was U-5Fs, which consisted of 95% U metal alloyed with 5% noble metal elements “fissium” (2.5% molybdenum, 2.0% ruthenium, 0.3% rhodium, 0.1% palladium,more » and 0.1% zirconium), while the later generation was U-10Zr which was 90% U metal alloyed with 10% zirconium. A potential concern during the H-Canyon nitric acid dissolution process of the U metal containing zirconium (Zr) is the explosive behavior that has been reported for alloys of these materials. For this reason, this evaluation was focused on the ability to process the lower Zr content materials, the U-5Fs material.« less

LCRE and SNAP 50-DR-1 programs. Engineering progress report, January 1, 1963--March 31, 1963

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

None

Declassified 5 Sep 1973. Information is presented concerning LCRE specifications, primary coolant circuit, aaxiliary systems, fuel elements, instrumentation, materials development, and fabrication; and SNAP-50DR-1 specifications, fuel elements, pumps, steam generator, and materials development. (DCC)

Oxygen electrodes for rechargeable alkaline fuel cells. II

NASA Technical Reports Server (NTRS)

Swette, L.; Kackley, N.

1990-01-01

The primary objective of this program is the investigation and development of electrocatalysts and supports for the positive electrode of moderate temperature, single-unit, rechargeable alkaline fuel cells. Approximately six support materials and five catalyst materials have been identified to date for further development.

Oxygen electrodes for rechargeable alkaline fuel cells-II

NASA Technical Reports Server (NTRS)

Swette, L.; Kackley, N.

1989-01-01

The primary objective of this program is the investigation and development of electrocatalysts and supports for the positive electrode of moderate temperature single-unit rechargeable alkaline fuel cells. Approximately six support materials and five catalyst materials have been identified to date for further development.

10 CFR 70.55 - Inspections.

Code of Federal Regulations, 2011 CFR

2011-01-01

... the case of fuel cycle facilities where nuclear reactor fuel is fabricated or processed each licensee... 10 Energy 2 2011-01-01 2011-01-01 false Inspections. 70.55 Section 70.55 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL Special Nuclear Material Control...

10 CFR 70.55 - Inspections.

Code of Federal Regulations, 2012 CFR

2012-01-01

... the case of fuel cycle facilities where nuclear reactor fuel is fabricated or processed each licensee... 10 Energy 2 2012-01-01 2012-01-01 false Inspections. 70.55 Section 70.55 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL Special Nuclear Material Control...