Demonstration of optimum fuel-to-moderator ratio in a PWR unit fuel cell

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

Feltus, M.A.; Pozsgai, C.

1992-01-01

Nuclear engineering students at The Pennsylvania State University develop scaled-down [[approx]350 MW(thermal)] pressurized water reactors (PWRs) using actual plants as references. The design criteria include maintaining the clad temperature below 2200[degree]F, fuel temperature below melting point, sufficient departure from nucleate boiling ratio (DNBR) margin, a beginning-of-life boron concentration that yields a negative moderator temperature coefficient, an adequate cycle power production (330 effective full-power days), and a batch loading scheme that is economical. The design project allows for many degrees of freedom (e.g., assembly number, pitch and height and batch enrichments) so that each student's result is unique. The iterative naturemore » of the design process is stressed in the course. The LEOPARD code is used for the unit cell depletion, critical boron, and equilibrium xenon calculations. Radial two-group diffusion equations are solved with the TWIDDLE-DEE code. The steady-state ZEBRA thermal-hydraulics program is used for calculating DNBR. The unit fuel cell pin radius and pitch (fuel-to-moerator ratio) for the scaled-down design, however, was set equal to the already optimized ratio for the reference PWR. This paper describes an honors project that shows how the optimum fuel-to-moderator ratio is found for a unit fuel cell shown in terms of neutron economics. This exercise illustrates the impact of fuel-to-moderator variations on fuel utilization factor and the effect of assuming space and energy separability.« less

Waterside corrosion of Zircaloy-clad fuel rods in a PWR environment

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

Garzarolli, F.; Jorde, D.; Manzel, R.

A data base of Zircaloy corrosion behavior under PWR operating conditions has been established from previously published reports as well as from new Kraftwerk Union (KWU) fuel examinations. The data show that the reactor environment increases the corrosion. ZrO/sub 2/ film thermal conductivity is another major factor that influences corrosion behavior. It was inferred from KWU film thickness data that the oxide film thermal conductivity may decrease once circumferential cracks develop in the layer. 57 refs.

Monte Carlo characterization of PWR spent fuel assemblies to determine the detectability of pin diversion

NASA Astrophysics Data System (ADS)

Burdo, James S.

This research is based on the concept that the diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies is feasible by a careful comparison of spontaneous fission neutron and gamma levels in the guide tube locations of the fuel assemblies. The goal is to be able to determine whether some of the assembly fuel pins are either missing or have been replaced with dummy or fresh fuel pins. It is known that for typical commercial power spent fuel assemblies, the dominant spontaneous neutron emissions come from Cm-242 and Cm-244. Because of the shorter half-life of Cm-242 (0.45 yr) relative to that of Cm-244 (18.1 yr), Cm-244 is practically the only neutron source contributing to the neutron source term after the spent fuel assemblies are more than two years old. Initially, this research focused upon developing MCNP5 models of PWR fuel assemblies, modeling their depletion using the MONTEBURNS code, and by carrying out a preliminary depletion of a ¼ model 17x17 assembly from the TAKAHAMA-3 PWR. Later, the depletion and more accurate isotopic distribution in the pins at discharge was modeled using the TRITON depletion module of the SCALE computer code. Benchmarking comparisons were performed with the MONTEBURNS and TRITON results. Subsequently, the neutron flux in each of the guide tubes of the TAKAHAMA-3 PWR assembly at two years after discharge as calculated by the MCNP5 computer code was determined for various scenarios. Cases were considered for all spent fuel pins present and for replacement of a single pin at a position near the center of the assembly (10,9) and at the corner (17,1). Some scenarios were duplicated with a gamma flux calculation for high energies associated with Cm-244. For each case, the difference between the flux (neutron or gamma) for all spent fuel pins and with a pin removed or replaced is calculated for each guide tube. Different detection criteria were established. The first was whether the relative error of the

Grid-to-rod flow-induced impact study for PWR fuel in reactor

DOE PAGES

Jiang, Hao; Qu, Jun; Lu, Roger Y.; ...

2016-06-10

The source for grid-to-rod fretting in a pressurized water nuclear reactor (PWR) is the dynamic contact impact from hydraulic flow-induced fuel assembly vibration. In order to support grid-to-rod fretting wear mitigation research, finite element analysis (FEA) was used to evaluate the hydraulic flow-induced impact intensity between the fuel rods and the spacer grids. Three-dimensional FEA models, with detailed geometries of the dimple and spring of the actual spacer grids along with fuel rods, were developed for flow impact simulation. The grid-to-rod dynamic impact simulation provided insights of the contact phenomena at grid-rod interface. Finally, it is an essential and effectivemore » way to evaluate contact forces and provide guidance for simulative bench fretting-impact tests.« less

Project-based learning applied to spacecraft power systems: a long-term engineering and educational program at UPM University

NASA Astrophysics Data System (ADS)

Pindado, Santiago; Cubas, Javier; Roibás-Millán, Elena; Sorribes-Palmer, Félix

2018-03-01

The IDR/UPM Institute is the research center responsible for the Master in Space Systems (MUSE) of Universidad Politécnica de Madrid (UPM). This is a 2-year (120 ECTS) master's degree focused on space technology. The UPMSat-2 satellite program has become an excellent educational framework in which the academic contents of the master are trained through project-based learning and following a multidisciplinary approach. In the present work, the educational projects developed and carried out in relation to spacecraft power systems at the IDR/UPM Institute are described. These projects are currently being developed in the framework represented by the aforementioned MUSE master's program and UPMSat-2.

Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

NASA Astrophysics Data System (ADS)

Rouf; Su'ud, Zaki

2016-08-01

Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

IMPACT OF FISSION PRODUCTS IMPURITY ON THE PLUTONIUM CONTENT IN PWR MOX FUELS

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

Gilles Youinou; Andrea Alfonsi

2012-03-01

This report presents the results of a neutronics analysis done in response to the charter IFCA-SAT-2 entitled 'Fuel impurity physics calculations'. This charter specifies that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies (UOX SNF) is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate PWR MOX fuel assemblies. Only non-gaseous FP have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1). This mixture of Pu and FP is called PuFP. Note that, in this preliminary analysis,more » the FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.« less

PREFACE: Festschrift to mark the sixtieth birthday of Professor Jens Lothe

NASA Astrophysics Data System (ADS)

Jøssang, Torstein; Barnett, David M.

1992-01-01

The collection of papers in this Festschrift represents the proceedings of a symposium held at the Norwegian Academy of Science and Letters on November 25-26, 1991, marking the occasion of the sixtieth birthday of Professor Jens Lothe. The symposium organizers attempted to invite contributors, either written, oral, or both, from a group of international scientists who have either collaborated with Jens in the past or whose work has had a significant impact in one of three areas in which Jens has focussed his own research interests, namely, statistical physics, elasticity and elastic waves, and the theory of dislocations in crystalline solids. The extent to which we have succeeded in obtaining a proper spectrum of contributors and contributions must be judged by the readers of this volume. It is rather rare in modern times to encounter a physicist such as Jens who has made seminal contributions in fields as diverse as the three included in this Festschrift. For this reason it is both historically interesting and instructive to follow the path that Jens Lothe's research career has taken him, since doing so clearly points out the international nature of the scientific endeavor and the fact that the search for scientific truth transcends national borders and governmental ideologies. Jens' postdoctoral studies at the University of Bristol in the late 1950s brought him in contact with an American postdoctoral student, John Hirth, who had worked on nucleation theory and condensation under the late Professor G M Pound at Carnegie-Mellon University. (Alex Maradudin, one of the contributors to the surface wave session of this symposium was also a postdoctoral fellow at Bristol at this time.) Both Lothe and Hirth had come to Bristol to acquaint themselves with dislocation theory; their first joint paper on double-kink nucleation theory was followed by numerous joint efforts, including their now-classic book Theory of Dislocations. Clearly, their interaction jelled. As legend

Conceptual design study of small long-life PWR based on thorium cycle fuel

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

Subkhi, M. Nurul; Su'ud, Zaki; Waris, Abdul

2014-09-30

A neutronic performance of small long-life Pressurized Water Reactor (PWR) using thorium cycle based fuel has been investigated. Thorium cycle which has higher conversion ratio in thermal region compared to uranium cycle produce some significant of {sup 233}U during burn up time. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.3, while the multi-energy-group diffusion calculations were optimized in whole core cylindrical two-dimension R-Z geometry by SRAC-CITATION. this study would be introduced thorium nitride fuel system which ZIRLO is the cladding material. The optimization of 350 MWt small long life PWRmore » result small excess reactivity and reduced power peaking during its operation.« less

Annual report, FY 1979 Spent fuel and fuel pool component integrity.

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

Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.

International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-..mu..m) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion.more » A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report.« less

Preparation, characterisation and viability of encapsulated Trichoderma harzianum UPM40 in alginate-montmorillonite clay.

PubMed

Adzmi, Fariz; Meon, Sariah; Musa, Mohamed Hanafi; Yusuf, Nor Azah

2012-01-01

Microencapsulation is a process by which tiny parcels of an active ingredient are packaged within a second material for the purpose of shielding the active ingredient from the surrounding environment. This study aims to determine the ability of the microencapsulation technique to improve the viability of Trichoderma harzianum UPM40 originally isolated from healthy groundnut roots as effective biological control agents (BCAs). Alginate was used as the carrier for controlled release, and montmorillonite clay (MMT) served as the filler. The encapsulated Ca-alginate-MMT beads were characterised using Fourier transform infrared spectroscopy (FTIR), thermogravimetric analysis (TGA) and scanning electron microscopy (SEM). The FTIR results showed the interaction between the functional groups of alginate and MMT in the Ca-alginate-MMT beads. Peaks at 1595, 1420 and 1020 cm(-1) characterised alginate, and peaks at 1028 and 453 cm(-1) characterised MMT; both sets of peaks appeared in the Ca-alginate-MMT FTIR spectrum. The TGA analysis showed an improvement in the thermal stability of the Ca-alginate-MMT beads compared with the alginate beads alone. SEM analysis revealed a homogeneous distribution of the MMT particles throughout the alginate matrix. T. harzianum UPM40 was successfully encapsulated in the Ca-alginate-MMT beads. Storage analysis of the encapsulated T. harzianum UPM40 showed that the low storage temperature of 5°C resulted in significantly (p < 0.05) better storage compared with room temperature (30°C).

AntiJen: a quantitative immunology database integrating functional, thermodynamic, kinetic, biophysical, and cellular data

PubMed Central

Toseland, Christopher P; Clayton, Debra J; McSparron, Helen; Hemsley, Shelley L; Blythe, Martin J; Paine, Kelly; Doytchinova, Irini A; Guan, Pingping; Hattotuwagama, Channa K; Flower, Darren R

2005-01-01

AntiJen is a database system focused on the integration of kinetic, thermodynamic, functional, and cellular data within the context of immunology and vaccinology. Compared to its progenitor JenPep, the interface has been completely rewritten and redesigned and now offers a wider variety of search methods, including a nucleotide and a peptide BLAST search. In terms of data archived, AntiJen has a richer and more complete breadth, depth, and scope, and this has seen the database increase to over 31,000 entries. AntiJen provides the most complete and up-to-date dataset of its kind. While AntiJen v2.0 retains a focus on both T cell and B cell epitopes, its greatest novelty is the archiving of continuous quantitative data on a variety of immunological molecular interactions. This includes thermodynamic and kinetic measures of peptide binding to TAP and the Major Histocompatibility Complex (MHC), peptide-MHC complexes binding to T cell receptors, antibodies binding to protein antigens and general immunological protein-protein interactions. The database also contains quantitative specificity data from position-specific peptide libraries and biophysical data, in the form of diffusion co-efficients and cell surface copy numbers, on MHCs and other immunological molecules. The uses of AntiJen include the design of vaccines and diagnostics, such as tetramers, and other laboratory reagents, as well as helping parameterize the bioinformatic or mathematical in silico modeling of the immune system. The database is accessible from the URL: . PMID:16305757

Design study of long-life PWR using thorium cycle

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

Subkhi, Moh. Nurul; Su'ud, Zaki; Waris, Abdul

2012-06-06

Design study of long-life Pressurized Water Reactor (PWR) using thorium cycle has been performed. Thorium cycle in general has higher conversion ratio in the thermal spectrum domain than uranium cycle. Cell calculation, Burn-up and multigroup diffusion calculation was performed by PIJ-CITATION-SRAC code using libraries based on JENDL 3.2. The neutronic analysis result of infinite cell calculation shows that {sup 231}Pa better than {sup 237}Np as burnable poisons in thorium fuel system. Thorium oxide system with 8%{sup 233}U enrichment and 7.6{approx} 8%{sup 231}Pa is the most suitable fuel for small-long life PWR core because it gives reactivity swing less than 1%{Delta}k/kmore » and longer burn up period (more than 20 year). By using this result, small long-life PWR core can be designed for long time operation with reduced excess reactivity as low as 0.53%{Delta}k/k and reduced power peaking during its operation.« less

Innovative teaching: Use of Screencast in the Agronomist Engineer High School of the UPM

NASA Astrophysics Data System (ADS)

Mediola, Maria Angeles; Aguado, Pedro Luis; Espejo, Rafael

2013-04-01

In the last academic courses, the Polytechnic University of Madrid (UPM) has supported the use of multimedia materials and methods in education and learning processes to improve the efficiency and impact obtained by faculty and students. With this aim during 2011-2012 course the multimedia method titled "Screencast" has been implemented in the subject "Plantas de Interés Agroalimentario" included in the curricula of the Agronomist Engineer High School. Next year will be apply in the subject "Soil Science" in the new degree. The Screencast tools allow record digital videos with sound directly into a computer so lecture and class can be recorded directly. The videos can be edited after including narrations, special effects as zoom, notes, images, etc. Screencast tools are simple use tools which are easy made tutorials, manuals, presentations and shows that help to students with different processes that are very hard to understand for students (1) (2). There are different Screencast tools in the market and after an evaluation process the most suitable for our need has been BB FlashBack Express (3) because is easy use, free and compatible with WEBCAM. This software allows export to Flash and AVI video formats. In our case the format chosen was the Flash format because the file sizes obtained were smaller than in AVI format. The use of BB FlashBack Express of the studied subject allowed make easy self-learning multimedia material and testing different methodologies and procedures for the use of this multimedia source in Internet. The BB FlashBack Express software was used during the course by teachers and students of this subject achieving a good improvement in the education and learning processes. The evaluation of the results obtained in the application of this method had shown that ability of students to use new technologies and spread their ideas has been increased as much into as outside classrooms. The materials made in this work had been joined to different

Reactivity loss validation of high burn-up PWR fuels with pile-oscillation experiments in MINERVE

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

Leconte, P.; Vaglio-Gaudard, C.; Eschbach, R.

2012-07-01

The ALIX experimental program relies on the experimental validation of the spent fuel inventory, by chemical analysis of samples irradiated in a PWR between 5 and 7 cycles, and also on the experimental validation of the spent fuel reactivity loss with bum-up, obtained by pile-oscillation measurements in the MINERVE reactor. These latter experiments provide an overall validation of both the fuel inventory and of the nuclear data responsible for the reactivity loss. This program offers also unique experimental data for fuels with a burn-up reaching 85 GWd/t, as spent fuels in French PWRs never exceeds 70 GWd/t up to now.more » The analysis of these experiments is done in two steps with the APOLLO2/SHEM-MOC/CEA2005v4 package. In the first one, the fuel inventory of each sample is obtained by assembly calculations. The calculation route consists in the self-shielding of cross sections on the 281 energy group SHEM mesh, followed by the flux calculation by the Method Of Characteristics in a 2D-exact heterogeneous geometry of the assembly, and finally a depletion calculation by an iterative resolution of the Bateman equations. In the second step, the fuel inventory is used in the analysis of pile-oscillation experiments in which the reactivity of the ALIX spent fuel samples is compared to the reactivity of fresh fuel samples. The comparison between Experiment and Calculation shows satisfactory results with the JEFF3.1.1 library which predicts the reactivity loss within 2% for burn-up of {approx}75 GWd/t and within 4% for burn-up of {approx}85 GWd/t. (authors)« less

Assessment for advanced fuel cycle options in CANDU

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

Morreale, A.C.; Luxat, J.C.; Friedlander, Y.

2013-07-01

The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a drivermore » fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.« less

JEN-1 Reactor Control System; SISTEMA DE CONTROL DEL REACTOR JEN-1

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

Cantillo, M.F.; Nuno, C.M.; Andreu, J.L.M.

1963-01-01

ABS>The JEN-1 3Mw power swimming pool reactor electrical control circuits are described. Start-up, power generation in the core, and shutdown are controlled by the reactor control system. This control system guarantees in each moment the safety conditions during reactor operation. Each circuit was represented by a scheme, complemented with a description of its function, components, and operation theory. Components described include: scram circuit; fission counter control circuit; servo control circuit; control circuit of safety sheets; control circuits of primary, secondary, and clean-up pump motors and tower fan motor; primary valve motor circuit; center cubicle alarm circuit; and process alarm circuit.more » (auth)« less

Effects of Newcastle Disease Virus Strains AF2240 and V4-UPM on Cytolysis and Apoptosis of Leukemia Cell Lines

PubMed Central

Alabsi, Aied M.; Bakar, Siti Aishah Abu; Ali, Rola; Omar, Abdul Rahman; Bejo, Mohd Hair; Ideris, Aini; Ali, Abdul Manaf

2011-01-01

Newcastle disease virus (NDV) is used as an antineoplastic agent in clinical tumor therapy. It has prompted much interest as an anticancer agent because it can replicate up to 10,000 times better in human cancer cells than in most normal cells. This study was carried out to determine the oncolytic potential of NDV strain AF2240 and V4-UPM on WEHI-3B leukemia cell line. Results from MTT cytotoxicity assay showed that the CD50 values for both strains were 2 and 8 HAU for AF2240 and V4-UPM, respectively. In addition, bromodeoxyuridine (BrdU) and trypan blue dye exclusion assays showed inhibition in cell proliferation after different periods. Increase in the cellular level of caspase-3 and detection of DNA laddering using agarose gel electrophoresis on treated cells with NDV confirmed that the mode of cell death was apoptosis. In addition, flow-cytometry analysis of cellular DNA content showed that the virus caused an increase in the sub-G1 region (apoptosis peaks). In conclusion, NDV strains AF2240 and V4-UPM caused cytolytic effects against WEHI-3B leukemic cell line. PMID:22272097

High-temperature compatibility between liquid metal as PWR fuel gap filler and stainless steel and high-density concrete

NASA Astrophysics Data System (ADS)

Wongsawaeng, Doonyapong; Jumpee, Chayanit; Jitpukdee, Manit

2014-08-01

In conventional nuclear fuel rods for light-water reactors, a helium-filled as-fabricated gap between the fuel and the cladding inner surface accommodates fuel swelling and cladding creep down. Because helium exhibits a very low thermal conductivity, it results in a large temperature rise in the gap. Liquid metal (LM; 1/3 weight portion each of lead, tin, and bismuth) has been proposed to be a gap filler because of its high thermal conductivity (∼100 times that of He), low melting point (∼100 °C), and lack of chemical reactivity with UO2 and water. With the presence of LM, the temperature drop across the gap is virtually eliminated and the fuel is operated at a lower temperature at the same power output, resulting in safer fuel, delayed fission gas release and prevention of massive secondary hydriding. During normal reactor operation, should an LM-bonded fuel rod failure occurs resulting in a discharge of liquid metal into the bottom of the reactor pressure vessel, it should not corrode stainless steel. An experiment was conducted to confirm that at 315 °C, LM in contact with 304 stainless steel in the PWR water chemistry environment for up to 30 days resulted in no observable corrosion. Moreover, during a hypothetical core-melt accident assuming that the liquid metal with elevated temperature between 1000 and 1600 °C is spread on a high-density concrete basement of the power plant, a small-scale experiment was performed to demonstrate that the LM-concrete interaction at 1000 °C for as long as 12 h resulted in no penetration. At 1200 °C for 5 h, the LM penetrated a distance of ∼1.3 cm, but the penetration appeared to stop. At 1400 °C the penetration rate was ∼0.7 cm/h. At 1600 °C, the penetration rate was ∼17 cm/h. No corrosion based on chemical reactions with high-density concrete occurred, and, hence, the only physical interaction between high-temperature LM and high-density concrete was from tiny cracks generated from thermal stress. Moreover

Fabrication of simulated DUPIC fuel

NASA Astrophysics Data System (ADS)

Kang, Kweon Ho; Song, Ki Chan; Park, Hee Sung; Moon, Je Sun; Yang, Myung Seung

2000-12-01

Simulated DUPIC fuel provides a convenient way to investigate the DUPIC fuel properties and behavior such as thermal conductivity, thermal expansion, fission gas release, leaching, and so on without the complications of handling radioactive materials. Several pellets simulating the composition and microstructure of DUPIC fuel are fabricated by resintering the powder, which was treated through OREOX process of simulated spent PWR fuel pellets, which had been prepared from a mixture of UO2 and stable forms of constituent nuclides. The key issues for producing simulated pellets that replicate the phases and microstructure of irradiated fuel are to achieve a submicrometre dispersion during mixing and diffusional homogeneity during sintering. This study describes the powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using the simulated spent PWR fuel. The homogeneity of additives in the powder was observed after attrition milling. The microstructure of the simulated spent PWR fuel agrees well with the other studies. The leading structural features observed are as follows: rare earth and other oxides dissolved in the UO2 matrix, small metallic precipitates distributed throughout the matrix, and a perovskite phase finely dispersed on grain boundaries.

Levey-Jennings Analysis Uncovers Unsuspected Causes of Immunohistochemistry Stain Variability.

PubMed

Vani, Kodela; Sompuram, Seshi R; Naber, Stephen P; Goldsmith, Jeffrey D; Fulton, Regan; Bogen, Steven A

Almost all clinical laboratory tests use objective, quantitative measures of quality control (QC), incorporating Levey-Jennings analysis and Westgard rules. Clinical immunohistochemistry (IHC) testing, in contrast, relies on subjective, qualitative QC review. The consequences of using Levey-Jennings analysis for QC assessment in clinical IHC testing are not known. To investigate this question, we conducted a 1- to 2-month pilot test wherein the QC for either human epidermal growth factor receptor 2 (HER-2) or progesterone receptor (PR) in 3 clinical IHC laboratories was quantified and analyzed with Levey-Jennings graphs. Moreover, conventional tissue controls were supplemented with a new QC comprised of HER-2 or PR peptide antigens coupled onto 8 μm glass beads. At institution 1, this more stringent analysis identified a decrease in the HER-2 tissue control that had escaped notice by subjective evaluation. The decrement was due to heterogeneity in the tissue control itself. At institution 2, we identified a 1-day sudden drop in the PR tissue control, also undetected by subjective evaluation, due to counterstain variability. At institution 3, a QC shift was identified, but only with 1 of 2 controls mounted on each slide. The QC shift was due to use of the instrument's selective reagent drop zones dispense feature. None of these events affected patient diagnoses. These case examples illustrate that subjective QC evaluation of tissue controls can detect gross assay failure but not subtle changes. The fact that QC issues arose from each site, and in only a pilot study, suggests that immunohistochemical stain variability may be an underappreciated problem.

On the condition of UO2 nuclear fuel irradiated in a PWR to a burn-up in excess of 110 MWd/kgHM

NASA Astrophysics Data System (ADS)

Restani, R.; Horvath, M.; Goll, W.; Bertsch, J.; Gavillet, D.; Hermann, A.; Martin, M.; Walker, C. T.

2016-12-01

Post-irradiation examination results are presented for UO2 fuel from a PWR fuel rod that had been irradiated to an average burn-up of 105 MWd/kgHM and showed high fission gas release of 42%. The radial distribution of xenon and the partitioning of fission gas between bubbles and the fuel matrix was investigated using laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS) and electron probe microanalysis. It is concluded that release from the fuel at intermediate radial positions was mainly responsible for the high fission gas release. In this region thermal release had occurred from the high burn-up structure (HBS) at some point after the sixth irradiation cycle. The LA-ICP-MS results indicate that gas release had also occurred from the HBS in the vicinity of the pellet periphery. It is shown that the gas pressure in the HBS pores is well below the pressure that the fuel can sustain.

Wirtgen Remixer surface recycling U.S. 90 Jennings, LA. : construction report.

DOT National Transportation Integrated Search

1991-02-01

This report documents the surface recycling of 7.4 miles of route U.S. 90 from the junction of La 99 to Jennings, Louisiana. The specialized recycling equipment was provided and operated by Remixer Contracting Co., Inc. of Austin, Texas. Roadway prod...

Lifestyle factors and dietary intake of Iranian postgraduate students in Universiti Putra Malaysia (UPM)

PubMed Central

Zarei, Maryam; Taib, Mohd Nasir Mohd; Zarei, Fatemeh

2013-01-01

Background: A student’s lifestyle can change notably in a foreign country. The objective of this study was to determine factors associated with the body weight status of Iranian postgraduate students in Universiti Putra Malaysia (UPM) 2009. Methods: A self-administered questionnaire was administered to 210 Iranian postgraduate students at UPM. Anthropometric factors also were measured using standard procedures. Body mass index (BMI) and waist-to-hip ratio (WHR) also were calculated. The chi-squared test, Spearman’s rho, and the Pearson product-moment correlation coefficient were used to determine the associations between the variables that were studied. Multiple linear regression analysis was used to measure the amount of influence a predictor variable had on a outcome variable. Results: There was no significant correlation between nutritional knowledge (P > 0.05), weight-management knowledge (P > 0.05), and smoking (P > 0.05) and BMI. There were statistically significant correlations between gender (P < 0.01), physical activity (P < 0.05), protein (P < 0.01), carbohydrate (P < 0.01), fat (P < 0.01), fiber (P < 0.01), vitamin C (P < 0.05), calcium (P < 0.01), and iron (P < 0.01) and BMI. There were also relationships between body fat (P < 0.01), waist circumference (P < 0.01), hip circumference (P < 0.01) and WHR (P < 0.01) and BMI. Conclusion: Our findings showed the need for a nutrition promotion program for the Iranian students to help them change their negative food habits and improve their lifestyles. PMID:26120404

Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

NASA Astrophysics Data System (ADS)

González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

2015-10-01

The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

Alleviation of Nitrogen and Sulfur Deficiency and Enhancement of Photosynthesis in Arabidopsis thaliana by Overexpression of Uroporphyrinogen III Methyltransferase (UPM1)

PubMed Central

Garai, Sampurna; Tripathy, Baishnab C.

2018-01-01

Siroheme, an iron-containing tetrapyrrole, is the prosthetic group of nitrite reductase (NiR) and sulfite reductase (SiR); it is synthesized from uroporphyrinogen III, an intermediate of chlorophyll biosynthesis, and is required for nitrogen (N) and sulfur (S) assimilation. Further, uroporphyrinogen III methyltransferase (UPM1), responsible for two methylation reactions to form dihydrosirohydrochlorin, diverts uroporphyrinogen III from the chlorophyll biosynthesis pathway toward siroheme synthesis. AtUPM1 [At5g40850] was used to produce both sense and antisense plants of Arabidopsis thaliana in order to modulate siroheme biosynthesis. In our experiments, overexpression of AtUPM1 signaled higher NiR (NII) and SiR gene and gene product expression. Increased NII expression was found to regulate and enhance the transcript and protein abundance of nitrate reductase (NR). We suggest that elevated NiR, NR, and SiR expression must have contributed to the increased synthesis of S containing amino acids in AtUPM1overexpressors, observed in our studies. We note that due to higher N and S assimilation in these plants, total protein content had increased in these plants. Consequently, chlorophyll biosynthesis increased in these sense plants. Higher chlorophyll and protein content of plants upregulated photosynthetic electron transport and carbon assimilation in the sense plants. Further, we have observed increased plant biomass in these plants, and this must have been due to increased N, S, and C assimilation. On the other hand, in the antisense plants, the transcript abundance, and protein content of NiR, and SiR was shown to decrease, resulting in reduced total protein and chlorophyll content. This led to a decrease in photosynthetic electron transport rate, carbon assimilation and plant biomass in these antisense plants. Under nitrogen or sulfur starvation conditions, the overexpressors had higher protein content and photosynthetic electron transport rate than the wild type

Development of modified MDA (M-MDA), PWR fuel cladding tube for high duty operation in future

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

Watanabe, Seiichi; Kido, Toshiya; Arakawa, Yasushi

2007-07-01

A new cladding material of M-MDA has been developed in order to prepare for a strong growing demand for advanced fuel which can maintain its integrity even under high duties due to more efficient operation such as higher burnup, higher LHR, and longer operation cycle which will contribute the suppression of environmental burdens like CO{sub 2} emission. The main aim of M-MDA is to have excellent corrosion resistance while the other properties are inherited from MDA, which has been adopted to the step 2 fuel, instead of Zry-4, of Japanese PWR plant whose upper limit of assembly discharged burnup ismore » 55 MWd/kgU. And we could confirm that the main aim of M-MDA was achieved by means of out-of-pile tests. In order to confirm improvement of corrosion resistance of M-MDA in the actual operation, irradiation test of M-MDA in the commercial reactor of Vandellos II is ongoing. The latest results of on-site examination after every end of cycle showed that oxide thickness of M-MDA-SR was much smaller than that of MDA at rod discharged burnup of approximately 60 MWd/kgU. The final irradiation cycle was completed on April 2007 and then we will obtain corrosion data of M-MDA over 70 MWd/kgU. M-MDA is a candidate alloy for advanced fuel under higher duty usage. (authors)« less

Optimization of burnable poison design for Pu incineration in fully fertile free PWR core

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

Fridman, E.; Shwageraus, E.; Galperin, A.

2006-07-01

The design challenges of the fertile-free based fuel (FFF) can be addressed by careful and elaborate use of burnable poisons (BP). Practical fully FFF core design for PWR reactor has been reported in the past [1]. However, the burnable poison option used in the design resulted in significant end of cycle reactivity penalty due to incomplete BP depletion. Consequently, excessive Pu loading were required to maintain the target fuel cycle length, which in turn decreased the Pu burning efficiency. A systematic evaluation of commercially available BP materials in all configurations currently used in PWRs is the main objective of thismore » work. The BP materials considered are Boron, Gd, Er, and Hf. The BP geometries were based on Wet Annular Burnable Absorber (WABA), Integral Fuel Burnable Absorber (IFBA), and Homogeneous poison/fuel mixtures. Several most promising combinations of BP designs were selected for the full core 3D simulation. All major core performance parameters for the analyzed cases are very close to those of a standard PWR with conventional UO{sub 2} fuel including possibility of reactivity control, power peaking factors, and cycle length. The MTC of all FFF cores was found at the full power conditions at all times and very close to that of the UO{sub 2} core. The Doppler coefficient of the FFF cores is also negative but somewhat lower in magnitude compared to UO{sub 2} core. The soluble boron worth of the FFF cores was calculated to be lower than that of the UO{sub 2} core by about a factor of two, which still allows the core reactivity control with acceptable soluble boron concentrations. The main conclusion of this work is that judicial application of burnable poisons for fertile free fuel has a potential to produce a core design with performance characteristics close to those of the reference PWR core with conventional UO{sub 2} fuel. (authors)« less

Suggestion on the safety classification of spent fuel dry storage in China’s pressurized water reactor nuclear power plant

NASA Astrophysics Data System (ADS)

Liu, Ting; Qu, Yunhuan; Meng, De; Zhang, Qiaoer; Lu, Xinhua

2018-01-01

China’s spent fuel storage in the pressurized water reactors(PWR) is stored with wet storage way. With the rapid development of nuclear power industry, China’s NPPs(NPPs) will not be able to meet the problem of the production of spent fuel. Currently the world’s major nuclear power countries use dry storage as a way of spent fuel storage, so in recent years, China study on additional spent fuel dry storage system mainly. Part of the PWR NPP is ready to apply for additional spent fuel dry storage system. It also need to safety classificate to spent fuel dry storage facilities in PWR, but there is no standard for safety classification of spent fuel dry storage facilities in China. Because the storage facilities of the spent fuel dry storage are not part of the NPP, the classification standard of China’s NPPs is not applicable. This paper proposes the safety classification suggestion of the spent fuel dry storage for China’s PWR NPP, through to the study on China’s safety classification principles of PWR NPP in “Classification for the items of pressurized water reactor nuclear power plants (GB/T 17569-2013)”, and safety classification about spent fuel dry storage system in NUREG/CR - 6407 in the United States.

Neutronics Studies of Uranium-bearing Fully Ceramic Micro-encapsulated Fuel for PWRs

DOE PAGES

George, Nathan M.; Maldonado, G. Ivan; Terrani, Kurt A.; ...

2014-12-01

Our study evaluated the neutronics and some of the fuel cycle characteristics of using uranium-based fully ceramic microencapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR lattice designs with FCM fuel have been developed that are expected to achieve higher specific burnup levels in the fuel while also increasing the tolerance to reactor accidents. The SCALE software system was the primary analysis tool used to model the lattice designs. A parametric study was performed by varying tristructural isotropic particle design features (e.g., kernel diameter, coating layer thicknesses, and packing fraction) to understand the impact on reactivity and resultingmore » operating cycle length. Moreover, to match the lifetime of an 18-month PWR cycle, the FCM particle fuel design required roughly 10% additional fissile material at beginning of life compared with that of a standard uranium dioxide (UO 2) rod. Uranium mononitride proved to be a favorable fuel for the fuel kernel due to its higher heavy metal loading density compared with UO 2. The FCM fuel designs evaluated maintain acceptable neutronics design features for fuel lifetime, lattice peaking factors, and nonproliferation figure of merit.« less

Hydrothermal synthesis of Ni 2FeBO 5 in near-supercritical PWR coolant and possible effects of neutron-induced 10B fission in fuel crud

NASA Astrophysics Data System (ADS)

Sawicki, Jerzy A.

2011-08-01

The hydrothermal synthesis of a nickel-iron oxyborate, Ni 2FeBO 5, known as bonaccordite, was investigated at pressures and temperatures that might occur at the surface of high-power fuel rods in PWR cores and in supercritical water reactors, especially during localized departures from nucleate boiling and dry-outs. The tests were performed using aqueous mixtures of nickel and iron oxides with boric acid or boron oxide, and as a function of lithium hydroxide addition, temperature and time of heating. At subcritical temperatures nickel ferrite NiFe 2O 4 was always the primary reaction product. High yield of Ni 2FeBO 5 synthesis started near critical water temperature and was strongly promoted by additions of LiOH up to Li/Fe and Li/B molar ratios in a range 0.1-1. The synthesis of bonaccordite was also promoted by other alkalis such as NaOH and KOH. The bonaccordite particles were likely formed by dissolution and re-crystallization by means of an intermediate nickel ferrite phase. It is postulated that the formation of Ni 2FeBO 5 in deposits of borated nickel and iron oxides on PWR fuel cladding can be accelerated by lithium produced in thermal neutron capture 10B(n,α) 7Li reactions. The process may also be aided in the reactor core by kinetic energy of α-particles and 7Li ions dissipated in the crud layer.

Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)

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

Shaver, Mark W.; Lanning, Donald D.

2010-02-01

The hypothesis of this paper is that the Zircaloy clad fuel source is minimal and that secondary startup neutron sources are the significant contributors of the tritium in the RCS that was previously assigned to release from fuel. Currently there are large uncertainties in the attribution of tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS). The measured amount of tritium in the coolant cannot be separated out empirically into its individual sources. Therefore, to quantify individual contributors, all sources of tritium in the RCS of a PWR must be understood theoretically and verified by the sum ofmore » the individual components equaling the measured values.« less

[Jennings Randolph Forum (3rd, Washington, District of Columbia, May 20-22, 1984).

ERIC Educational Resources Information Center

CAC Citizenship Education News, 1984

1984-01-01

This overview of the third annual Jennings Randolph Forum focuses on the role that election campaigns and the political process are capable of playing in citizenship education. Council for the Advancement of Citizenship (CAC) testimony on citizenship education follows a general overview of the conference proceedings. A set of mandates for…

Determination of uncertainties of PWR spent fuel radionuclide inventory based on real operational history data

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

Fast, Ivan; Bosbach, Dirk; Aksyutina, Yuliya

A requisite for the official approval of the safe final disposal of SNF is a comprehensive specification and declaration of the nuclear inventory in SNF by the waste supplier. In the verification process both the values of the radionuclide (RN) activities and their uncertainties are required. Burn-up (BU) calculations based on typical and generic reactor operational parameters do not encompass any possible uncertainties observed in real reactor operations. At the same time, the details of the irradiation history are often not well known, which complicates the assessment of declared RN inventories. Here, we have compiled a set of burnup calculationsmore » accounting for the operational history of 339 published or anonymized real PWR fuel assemblies (FA). These histories were used as a basis for a 'SRP analysis', to provide information about the range of the values of the associated secondary reactor parameters (SRP's). Hence, we can calculate the realistic variation or spectrum of RN inventories. SCALE 6.1 has been employed for the burn-up calculations. The results have been validated using experimental data from the online database - SFCOMPO-1 and -2. (authors)« less

Multirecycling of Plutonium from LMFBR Blanket in Standard PWRs Loaded with MOX Fuel

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

Sonat Sen; Gilles Youinou

2013-02-01

It is now well-known that, from a physics standpoint, Pu, or even TRU (i.e. Pu+M.A.), originating from LEU fuel irradiated in PWRs can be multirecycled also in PWRs using MOX fuel. However, the degradation of the isotopic composition during irradiation necessitates using enriched U in conjunction with the MOX fuel either homogeneously or heterogeneously to maintain the Pu (or TRU) content at a level allowing safe operation of the reactor, i.e. below about 10%. The study is related to another possible utilization of the excess Pu produced in the blanket of a LMFBR, namely in a PWR(MOX). In this casemore » the more Pu is bred in the LMFBR, the more PWR(MOX) it can sustain. The important difference between the Pu coming from the blanket of a LMFBR and that coming from a PWR(LEU) is its isotopic composition. The first one contains about 95% of fissile isotopes whereas the second one contains only about 65% of fissile isotopes. As it will be shown later, this difference allows the PWR fed by Pu from the LMFBR blanket to operate with natural U instead of enriched U when it is fed by Pu from PWR(LEU)« less

William Jennings Bryan: Boy Orator, Broken Man and the "Evolution" of America's Public Philosophy

ERIC Educational Resources Information Center

Murphy, Troy A.

2002-01-01

Perhaps more than any other figure in American history, William Jennings Bryan is remembered for specific and identifiable moments of rhetorical action: the much-revered 1896 "Cross of gold" speech and the much-maligned Scopes "monkey trial" of 1925. The dissonance between these two events, at least with respect to the ways in…

Toward a Pragmatist Epistemology: Arthur O. Lovejoy's and H. S. Jennings's Biophilosophical Responses to Neovitalism, 1909-1914.

PubMed

Russell, Doug

2015-01-01

The sustained interdisciplinary debate about neovitalism between two Johns Hopkins University colleagues, philosopher Arthur O. Lovejoy and experimental geneticist H. S. Jennings, in the period 1911-1914, was the basis for their theoretical reconceptualization of scientific knowledge as contingent and necessarily incomplete in its account of nature. Their response to Hans Driesch's neovitalist concept of entelechy, and his challenge to the continuity between biology and the inorganic sciences, resulted in a historically significant articulation of genetics and philosophy. This study traces the debate's shift of problem-focus away from neovitalism's threat to the unity of science - "organic autonomy," as Lovejoy put it - and toward the potential for development of a nonmechanististic, nonrationalist theory of scientific knowledge. The result was a new pragmatist epistemology, based on Lovejoy's and Jennings's critiques of the inadequacy of pragmatism's account of scientific knowledge. The first intellectual move, drawing on naturalism and pragmatism, was based on a reinterpretation of science as organized experience. The second, sparked by Henri Bergson's theory of creative evolution, and drawing together elements of Dewey's and James's pragmatisms, produced a new account of the contingency and necessary incompleteness of scientific knowledge. Prompted by the neovitalists' mix of a priori concepts and, in Driesch's case, and adherence to empiricism, Lovejoy's and Jennings's developing pragmatist epistemologies of science explored the interrelation between rationalism and empiricism.

Neutron Collar Evolution and Fresh PWR Assembly Measurements with a New Fast Neutron Passive Collar

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

Menlove, Howard Olsen; Geist, William H.; Root, Margaret A.

The passive neutron collar approach removes the effect of poison rods when using a 1mm Gd liner. This project sets out to solve the following challenges: BWR fuel assemblies have less mass and less neutron multiplication than PWR; and effective removal of cosmic ray spallation neutron bursts needed via QC tests.

Nuclear fuel performance: Trends, remedies and challenges

NASA Astrophysics Data System (ADS)

Rusch, C. A.

2008-12-01

It is unacceptable to have nuclear power plants unavailable or power restricted due to fuel reliability issues. 'Fuel reliability' has a much broader definition than just maintaining mechanical integrity and being leaker free - fuel must fully meet the specifications, impose no adverse impacts on plant operation and safety, and maintain quantifiable margins within design and operational envelopes. The fuel performance trends over the last decade are discussed and the significant contributors to reduced reliability experienced with commercial PWR and BWR designs are identified and discussed including grid-to-rod fretting and debris fretting in PWR designs and accelerated corrosion, debris fretting and pellet-cladding interaction in BWR designs. In many of these cases, the impacts have included not only fuel failures but also plant operating restrictions, forced shutdowns, and/or enhanced licensing authority oversight. Design and operational remedies are noted. The more demanding operating regimes and the constant quest to improve fuel performance require enhancements to current designs and/or new design features. Fuel users must continue to and enhance interaction with fuel suppliers in such areas as oversight of supplier design functions, lead test assembly irradiation programs and quality assurance oversight and surveillance. With the implementation of new designs and/or features, such fuel user initiatives can help to minimize the potential for performance problems.

Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations

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

Santamarina, A.; Bernard, D.; Blaise, P.

2013-07-01

This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO{sub 2}-Gd{sub 2}O{sub 3} poisoning worth, Efficiency ofmore » Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)« less

The Cup Anemometer, a Fundamental Meteorological Instrument for the Wind Energy Industry. Research at the IDR/UPM Institute

PubMed Central

Pindado, Santiago; Cubas, Javier; Sorribes-Palmer, Félix

2014-01-01

The results of several research campaigns investigating cup anemometer performance carried out since 2008 at the IDR/UPM Institute are included in the present paper. Several analysis of large series of calibrations were done by studying the effect of the rotor's geometry, climatic conditions during calibration, and anemometers' ageing. More specific testing campaigns were done regarding the cup anemometer rotor aerodynamics, and the anemometer signals. The effect of the rotor's geometry on the cup anemometer transfer function has been investigated experimentally and analytically. The analysis of the anemometer's output signal as a way of monitoring the anemometer status is revealed as a promising procedure for detecting anomalies. PMID:25397921

Experimental validation of the DARWIN2.3 package for fuel cycle applications

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

San-Felice, L.; Eschbach, R.; Bourdot, P.

2012-07-01

The DARWIN package, developed by the CEA and its French partners (AREVA and EDF) provides the required parameters for fuel cycle applications: fuel inventory, decay heat, activity, neutron, {gamma}, {alpha}, {beta} sources and spectrum, radiotoxicity. This paper presents the DARWIN2.3 experimental validation for fuel inventory and decay heat calculations on Pressurized Water Reactor (PWR). In order to validate this code system for spent fuel inventory a large program has been undertaken, based on spent fuel chemical assays. This paper deals with the experimental validation of DARWIN2.3 for the Pressurized Water Reactor (PWR) Uranium Oxide (UOX) and Mixed Oxide (MOX) fuelmore » inventory calculation, focused on the isotopes involved in Burn-Up Credit (BUC) applications and decay heat computations. The calculation - experiment (C/E-1) discrepancies are calculated with the latest European evaluation file JEFF-3.1.1 associated with the SHEM energy mesh. An overview of the tendencies is obtained on a complete range of burn-up from 10 to 85 GWd/t (10 to 60 GWcVt for MOX fuel). The experimental validation of the DARWIN2.3 package for decay heat calculation is performed using calorimetric measurements carried out at the Swedish Interim Spent Fuel Storage Facility for Pressurized Water Reactor (PWR) assemblies, covering a large burn-up (20 to 50 GWd/t) and cooling time range (10 to 30 years). (authors)« less

Multi level optimization of burnable poison utilization for advanced PWR fuel management

NASA Astrophysics Data System (ADS)

Yilmaz, Serkan

The objective of this study was to develop an unique methodology and a practical tool for designing burnable poison (BP) pattern for a given PWR core. Two techniques were studied in developing this tool. First, the deterministic technique called Modified Power Shape Forced Diffusion (MPSFD) method followed by a fine tuning algorithm, based on some heuristic rules, was developed to achieve this goal. Second, an efficient and a practical genetic algorithm (GA) tool was developed and applied successfully to Burnable Poisons (BPs) placement optimization problem for a reference Three Mile Island-1 (TMI-1) core. This thesis presents the step by step progress in developing such a tool. The developed deterministic method appeared to perform as expected. The GA technique produced excellent BP designs. It was discovered that the Beginning of Cycle (BOC) Kinf of a BP fuel assembly (FA) design is a good filter to eliminate invalid BP designs created during the optimization process. By eliminating all BP designs having BOC Kinf above a set limit, the computational time was greatly reduced since the evaluation process with reactor physics calculations for an invalid solution is canceled. Moreover, the GA was applied to develop the BP loading pattern to minimize the total Gadolinium (Gd) amount in the core together with the residual binding at End-of-Cycle (EOC) and to keep the maximum peak pin power during core depletion and Soluble boron concentration at BOC both less than their limit values. The number of UO2/Gd2O3 pins and Gd 2O3 concentrations for each fresh fuel location in the core are the decision variables and the total amount of the Gd in the core and maximum peak pin power during core depletion are in the fitness functions. The use of different fitness function definition and forcing the solution movement towards to desired region in the solution space accelerated the GA runs. Special emphasize is given to minimizing the residual binding to increase core lifetime as

Irradiation performance of (Th,Pu)O2 fuel under Pressurized Water Reactor conditions

NASA Astrophysics Data System (ADS)

Boer, B.; Lemehov, S.; Wéber, M.; Parthoens, Y.; Gysemans, M.; McGinley, J.; Somers, J.; Verwerft, M.

2016-04-01

This paper examines the in-pile safety performance of (Th,Pu)O2 fuel pins under simulated Pressurized Water Reactor (PWR) conditions. Both sol-gel and SOLMAS produced (Th,Pu)O2 fuels at enrichments of 7.9% and 12.8% in Pu/HM have been irradiated at SCK·CEN. The irradiation has been performed under PWR conditions (155 bar, 300 °C) in a dedicated loop of the BR-2 reactor. The loop is instrumented with flow and temperature monitors at inlet and outlet, which allow for an accurate measurement of the deposited enthalpy.

Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

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

McConnell, Paul E.; Koenig, Greg John; Uncapher, William Leonard

2016-05-01

This report describes the third set of tests (the “DCLa shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

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

McConnell, Paul E.; Koenig, Greg John; Uncapher, William Leonard

2016-05-12

This report describes the third set of tests (the “DCL a shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

Fuel thermal conductivity (FTHCON). Status report. [PWR; BWR

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

Hagrman, D. L.

1979-02-01

An improvement of the fuel thermal conductivity subcode is described which is part of the fuel rod behavior modeling task performed at EG and G Idaho, Inc. The original version was published in the Materials Properties (MATPRO) Handbook, Section A-2 (Fuel Thermal Conductivity). The improved version incorporates data which were not included in the previous work and omits some previously used data which are believed to come from cracked specimens. The models for the effect of porosity on thermal conductivity and for the electronic contribution to thermal coductivity have been completely revised in order to place these models on amore » more mechanistic basis. As a result of modeling improvements the standard error of the model with respect to its data base has been significantly reduced.« less

Advanced techniques for repair of irradiated PWR fuel assemblies

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

Knaab, H.; Westphal, M.

Kraftwerk union has recently designed and built a portable repair unit for use in nuclear power plants for repair of defective fuel assemblies where space limitations do not allow permanent installation of repair equipment. This new equipment is designed to be easily disassembled and decontaminated. The main component of the equipment is the fuel assembly reconstitution unit (FARU) which is placed on the floor of the spent fuel pool. The use of the FARU is described in the paper.

Progress in understanding fission-product behaviour in coated uranium-dioxide fuel particles

NASA Astrophysics Data System (ADS)

Barrachin, M.; Dubourg, R.; Kissane, M. P.; Ozrin, V.

2009-03-01

Supported by results of calculations performed with two analytical tools (MFPR, which takes account of physical and chemical mechanisms in calculating the chemical forms and physical locations of fission products in UO2, and MEPHISTA, a thermodynamic database), this paper presents an investigation of some important aspects of the fuel microstructure and chemical evolutions of irradiated TRISO particles. The following main conclusions can be identified with respect to irradiated TRISO fuel: first, the relatively low oxygen potential within the fuel particles with respect to PWR fuel leads to chemical speciation that is not typical of PWR fuels, e.g., the relatively volatile behaviour of barium; secondly, the safety-critical fission-product caesium is released from the urania kernel but the buffer and pyrolytic-carbon coatings could form an important chemical barrier to further migration (i.e., formation of carbides). Finally, significant releases of fission gases from the urania kernel are expected even in nominal conditions.

The fractalline properties of experimentally simulated PWR fuel crud

NASA Astrophysics Data System (ADS)

Dumnernchanvanit, I.; Mishra, V. K.; Zhang, N. Q.; Robertson, S.; Delmore, A.; Mota, G.; Hussey, D.; Wang, G.; Byers, W. A.; Short, M. P.

2018-02-01

The buildup of fouling deposits on nuclear fuel rods, known as crud, continues to challenge the worldwide fleet of light water reactors (LWRs). Crud may cause serious operational problems for LWRs, including axial power shifts, accelerated fuel clad corrosion, increased primary circuit radiation dose rates, and in some instances has led directly to fuel failure. Numerous studies continue to attempt to model and predict the effects of crud, but each makes critical assumptions regarding how to treat the complex, porous microstructure of crud and its resultant effects on temperature, pressure, and crud chemistry. In this study, we demonstrate that crud is indeed a fractalline porous medium using flowing loop experiments, validating the most recent models of its effects on LWR fuel cladding. This crud is shown to match that in other LWR-prototypical facilities through a porosity-fractal dimension scaling law. Implications of this result range from post-mortem analysis of the effects of crud on reactor fuel performance, to utilizing crud's fractalline dimensions to quantify the effectiveness of anti-fouling measures.

CHF considerations for highly moderated 100% MOX fuels PWRs

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

Saphier, D.; Raymond, P.

1995-09-01

A feasibility study on using 100% MOX fuel in a PWR with increased moderating ratio, RMA, was initiated. In the proposed design all the parameters were chosen identical to the French 1450MW PWR, except the fuel pin diameter which was reduced to achieve higher moderating ratios, V{sub M}/V{sub F}, where V{sub M} and V{sub F} are the moderator and fuel volume respectively. Moderating ratios from 2 to 4 were considered. In the present study the thermal-hydraulic feasibility of using fuel assemblies with smaller diameter fuel pins was investigated. The major design constrain in this study was the critical heat fluxmore » (CHF). In order to maintain the fuel pin integrity under nominal operating and transient conditions, the minimum DNBR, (Departure from Nucleate Boiling Ratio given by CHF/q{close_quotes}{sub local}, where q{close_quotes}{sub local} is the local heat flux), has to be above a given value. The limitations of the existing CHF correlations for the present study are outlined. Two designs based on the conventional 17x17 fuel assembly and on the advanced 19x19 assembly meeting the MDNBR criteria and satisfying the control margin requirements, are proposed.« less

Nuclear fuel element nut retainer cup. [PWR

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

Walton, L.A.

1977-07-19

A typical embodiment has an end fitting for a nuclear reactor fuel element that is joined to the control rod guide tubes by means of a nut plate assembly. The nut plate assembly has an array of nuts, each engaging the respective threaded end of the control rod guide tubes. The nuts, moreover, are retained on the plate during handling and before fuel element assembly by means of hollow cylindrical locking cups that are brazed to the plate and loosely circumscribe the individual enclosed nuts. After the nuts are threaded onto the respective guide tube ends, the locking cups aremore » partially deformed to prevent one or more of the nuts from working loose during reactor operation. The locking cups also prevent loose or broken end fitting parts from becoming entrained in the reactor coolant.« less

Development of burnup dependent fuel rod model in COBRA-TF

NASA Astrophysics Data System (ADS)

Yilmaz, Mine Ozdemir

The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN

Reflecting on Jens Rasmussen's legacy (2) behind and beyond, a 'constructivist turn'.

PubMed

Le Coze, Jean-Christophe

2017-03-01

This article is the second part of a study on the legacy of Jens Rasmussen. The first article, subtitled 'A Strong Program for a Hard Problem', looks back on his 30 years of scientific contribution, from 1969 to 2000. This second article explores and investigates some of the intellectual roots which influenced his thinking, using them as a basis to understand some limits and move forward. Indeed, historically oriented studies such as this one are not only tributes to researchers, but a way to differentiate and contrast our present situation with the past in order to integrate contemporary trends, be they theoretical or empirical, or oriented towards research and new models. In the first section of this article, I offer a synthesis of the background covered in the previous article, but I use a tree here as a graphical complement. Branches of the tree show the many fruitful directions opened by Jens Rasmussen, directions which inspired many researchers. In the second part, I address what I believe to be behind this wealth of engineering legacy: cybernetics. I contend that cybernetics has had a profound influence on his thinking and provided him key principles for his inspiring and successful models. To develop the tree image, one might say that cybernetics is the trunk of the tree. Finally, in the third part, I take the opportunity to explore the relevance of extending and sensitising his program to constructivist discourses. After an introduction to this discourse, identifying four types of constructivisms (cognitive, social, epistemological and anthropological), I characterise this move as a 'constructivist turn'. Copyright © 2015 Elsevier Ltd. All rights reserved.

Spent fuel data base: commercial light water reactors. [PWR; BWR

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

Hauf, M.J.; Kniazewycz, B.G.

1979-12-01

As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

Development of a new lattice physics code robin for PWR application

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

Zhang, S.; Chen, G.

2013-07-01

This paper presents a description of methodologies and preliminary verification results of a new lattice physics code ROBIN, being developed for PWR application at Shanghai NuStar Nuclear Power Technology Co., Ltd. The methods used in ROBIN to fulfill various tasks of lattice physics analysis are an integration of historical methods and new methods that came into being very recently. Not only these methods like equivalence theory for resonance treatment and method of characteristics for neutron transport calculation are adopted, as they are applied in many of today's production-level LWR lattice codes, but also very useful new methods like the enhancedmore » neutron current method for Dancoff correction in large and complicated geometry and the log linear rate constant power depletion method for Gd-bearing fuel are implemented in the code. A small sample of verification results are provided to illustrate the type of accuracy achievable using ROBIN. It is demonstrated that ROBIN is capable of satisfying most of the needs for PWR lattice analysis and has the potential to become a production quality code in the future. (authors)« less

Analysis of the return to power scenario following a LBLOCA in a PWR

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

Macian, R.; Tyler, T.N.; Mahaffy, J.H.

1995-09-01

The risk of reactivity accidents has been considered an important safety issue since the beginning of the nuclear power industry. In particular, several events leading to such scenarios for PWR`s have been recognized and studied to assess the potential risk of fuel damage. The present paper analyzes one such event: the possible return to power during the reflooding phase following a LBLOCA. TRAC-PF1/MOD2 coupled with a three-dimensional neutronic model of the core based on the Nodal Expansion Method (NEM) was used to perform the analysis. The system computer model contains a detailed representation of a complete typical 4-loop PWR. Thus,more » the simulation can follow complex system interactions during reflooding, which may influence the neutronics feedback in the core. Analyses were made with core models bases on cross sections generated by LEOPARD. A standard and a potentially more limiting case, with increased pressurizer and accumulator inventories, were run. In both simulations, the reactor reaches a stable state after the reflooding is completed. The lower core region, filled with cold water, generates enough power to boil part of the incoming liquid, thus preventing the core average liquid fraction from reaching a value high enough to cause a return to power. At the same time, the mass flow rate through the core is adequate to maintain the rod temperature well below the fuel damage limit.« 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

Probabilistic analysis on the failure of reactivity control for the PWR

NASA Astrophysics Data System (ADS)

Sony Tjahyani, D. T.; Deswandri; Sunaryo, G. R.

2018-02-01

The fundamental safety function of the power reactor is to control reactivity, to remove heat from the reactor, and to confine radioactive material. The safety analysis is used to ensure that each parameter is fulfilled during the design and is done by deterministic and probabilistic method. The analysis of reactivity control is important to be done because it will affect the other of fundamental safety functions. The purpose of this research is to determine the failure probability of the reactivity control and its failure contribution on a PWR design. The analysis is carried out by determining intermediate events, which cause the failure of reactivity control. Furthermore, the basic event is determined by deductive method using the fault tree analysis. The AP1000 is used as the object of research. The probability data of component failure or human error, which is used in the analysis, is collected from IAEA, Westinghouse, NRC and other published documents. The results show that there are six intermediate events, which can cause the failure of the reactivity control. These intermediate events are uncontrolled rod bank withdrawal at low power or full power, malfunction of boron dilution, misalignment of control rod withdrawal, malfunction of improper position of fuel assembly and ejection of control rod. The failure probability of reactivity control is 1.49E-03 per year. The causes of failures which are affected by human factor are boron dilution, misalignment of control rod withdrawal and malfunction of improper position for fuel assembly. Based on the assessment, it is concluded that the failure probability of reactivity control on the PWR is still within the IAEA criteria.

Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

DOE PAGES

Ham, Y.; Kerr, P.; Sitaraman, S.; ...

2016-05-05

Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reportedmore » the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.« less

Partial Defect Verification of Spent Fuel Assemblies by PDET: Principle and Field Testing in Interim Spent Fuel Storage Facility (CLAB) in Sweden

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

Ham, Y.S.; Kerr, P.; Sitaraman, S.

The need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called 'difficult-to-access' areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into 'difficult-to-access' areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported themore » successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17x17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly burnup levels. (authors)« less

Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

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

Ham, Y.; Kerr, P.; Sitaraman, S.

Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reportedmore » the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.« less

Multi-Pack Disposal Concepts for Spent Fuel (Revision 1)

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

Hardin, Ernest; Matteo, Edward N.; Hadgu, Teklu

2016-01-01

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media. Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement formore » extended ventilation) have been limited to the Yucca Mountain License Application Design. Thermal analysis showed that if “enclosed” concepts are constrained by peak package/buffer temperature, that waste package capacity is limited to 4 PWR assemblies (or 9 BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems. This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).« less

EMERALD REV.1. PWR Accident Activity Release

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

Brunot, W.K.; Fray, R.R.; Gillespie, S.G.

1975-10-01

The EMERALD program is designed for the calculation of radiation releases and exposures resulting from abnormal operation of a large pressurized water reactor (PWR). The approach used in EMERALD is similar to an analog simulation of a real system. Each component or volume in the plant which contains a radioactive material is represented by a subroutine which keeps track of the production, transfer, decay and absorption of radioactivity in that volume. During the course of the analysis of an accident, activity is transferred from subroutine to subroutine in the program as it would be transferred from place to place inmore » the plant. For example, in the calculation of the doses resulting from a loss-of-coolant accident the program first calculates the activity built up in the fuel before the accident, then releases some of this activity to the containment volume. Some of this activity is then released to the atmosphere. The rates of transfer, leakage, production, cleanup, decay, and release are read in as input to the program. Subroutines are also included which calculate the on-site and off-site radiation exposures at various distances for individual isotopes and sums of isotopes. The program contains a library of physical data for the twenty-five isotopes of most interest in licensing calculations, and other isotopes can be added or substituted. Because of the flexible nature of the simulation approach, the EMERALD program can be used for most calculations involving the production and release of radioactive materials during abnormal operation of a PWR. These include design, operational, and licensing studies.« less

Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

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

Billone, M. C.; Burtseva, T. A.

2016-08-30

The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

ACHILLES: Heat Transfer in PWR Core During LOCA Reflood Phase

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

2013-11-01

1. NAME AND TITLE OF DATA LIBRARY ACHILLES -Heat Transfer in PWR Core During LOCA Reflood Phase. 2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS N/A 3. CONTRIBUTOR AEA Technology, Winfrith Technology Centre, Dorchester DT2 8DH United Kingdom through the OECD Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France. 4. DESCRIPTION OF TEST FACILITY The most important features of the Achilles rig were the shroud vessel, which contained the test section, and the downcomer. These may be thought of as representing the core barrel and the annular downcomer in the reactor pressure vessel. The test section comprises a cluster of 69more » rods in a square array within a circular shroud vessel. The rod diameter and pitch (9.5 mm and 12.6 mm) were typical of PWR dimensions. The internal diameter of the shroud vessel was 128 mm. Each rod was electrically heated over a length of 3.66 m, which is typical of the nuclear heated length in a PWR fuel rod, and each contained 6 internal thermocouples. These were arranged in one of 8 groupings which concentrated the thermocouples in different axial zones. The spacer grids were at prototypic PWR locations. Each grid had two thermocouples attached to its trailing edge at radial locations. The axial power profile along the rods was an 11 step approximation to a "chopped cosine". The shroud vessel had 5 heating zones whose power could be independently controlled. 5. DESCRIPTION OF TESTS The Achilles experiments investigated the heat transfer in the core of a Pressurized Water Reactor during the re-flood phase of a postulated large break loss of coolant accident. The results provided data to validate codes and to improve modeling. Different types of experiments were carried out which included single phase cooling, re-flood under low flow conditions, level swell and re-flood under high flow conditions. Three series of experiments were performed. The first and the third used the same test section but the second used another test section

77 FR 60479 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-03

... Pressurized Water Reactor Spent Fuel in Transportation and Storage Casks AGENCY: Nuclear Regulatory Commission... 3, entitled, ``Burnup Credit in the Criticality Safety Analyses of PWR [Pressurized Water Reactor... water reactor spent nuclear fuel (SNF) in transportation packages and storage casks. SFST-ISG-8...

Probability of in-vessel steam explosion-induced containment failure for a KWU PWR

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

Esmaili, H.; Khatib-Rahbar, M.; Zuchuat, O.

During postulated core meltdown accidents in light water reactors, there is a likelihood for an in-vessel steam explosion when the melt contacts the coolant in the lower plenum. The objective of the work described in this paper is to determine the conditional probability of in-vessel steam explosion-induced containment failure for a Kraftwerk Union (KWU) pressurized water reactor (PWR). The energetics of the explosion depends on the mass of the molten fuel that mixes with the coolant and participates in the explosion and on the conversion of fuel thermal energy into mechanical work. The work can result in the generation ofmore » dynamic pressures that affect the lower head (and possibly lead to its failure), and it can cause acceleration of a slug (fuel and coolant material) upward that can affect the upper internal structures and vessel head and ultimately cause the failure of the upper head. If the upper head missile has sufficient energy, it can reach the containment shell and penetrate it. The analysis, must therefore, take into account all possible dissipation mechanisms.« less

Development of an extended-burnup Mark B design. First semi-annual progress report, July-December 1978. Report BAW-1532-1. [PWR

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

None

1979-10-01

The primary objective of this program is to develop and demonstrate an improved PWR fuel assembly design capable of batch average burnups of 45,000-50,000 MWd/mtU. To accomplish this, a number of technical areas must be investigated to verify acceptable extended-burnup fuel performance. This report is the first semi-annual progress report for the program, and it describes work performed during the July-December 1978 time period. Efforts during this period included the definition of a preliminary design for a high-burnup fuel rod, physics analyses of extended-burnup fuel cycles, studies of the physics characteristics of changes in fuel assembly metal-to-water ratios, and developmentmore » of a design concept for post-irradiation examination equipment to be utilized in examining high-burnup lead-test assemblies.« less

On-line detection of key radionuclides for fuel-rod failure in a pressurized water reactor.

PubMed

Qin, Guoxiu; Chen, Xilin; Guo, Xiaoqing; Ni, Ning

2016-08-01

For early on-line detection of fuel rod failure, the key radionuclides useful in monitoring must leak easily from failing rods. Yield, half-life, and mass share of fission products that enter the primary coolant also need to be considered in on-line analyses. From all the nuclides that enter the primary coolant during fuel-rod failure, (135)Xe and (88)Kr were ultimately chosen as crucial for on-line monitoring of fuel-rod failure. A monitoring system for fuel-rod failure detection for pressurized water reactor (PWR) based on the LaBr3(Ce) detector was assembled and tested. The samples of coolant from the PWR were measured using the system as well as a HPGe γ-ray spectrometer. A comparison showed the method was feasible. Finally, the γ-ray spectra of primary coolant were measured under normal operations and during fuel-rod failure. The two peaks of (135)Xe (249.8keV) and (88)Kr (2392.1keV) were visible, confirming that the method is capable of monitoring fuel-rod failure on-line. Copyright © 2016 Elsevier Ltd. All rights reserved.

Multi-pack Disposal Concepts for Spent Fuel (Rev. 0)

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

Hadgu, Teklu; Hardin, Ernest; Matteo, Edward N.

2015-12-01

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media (Hardin et al., 2012). Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are keptmore » open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design (CRWMS M&O, 1999). Thermal analysis showed that, if “enclosed” concepts are constrained by peak package/buffer temperature, waste package capacity is limited to 4 PWR assemblies (or 9-BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems (EnergySolution, 2015). This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).« less

Feasibility of recycling thorium in a fusion-fission hybrid/PWR symbiotic system

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

Josephs, J.M.

1980-12-31

A study was made of the economic impact of high levels of radioactivity in the thorium fuel cycle. The sources of this radioactivity and means of calculating the radioactive levels at various stages in the fuel cycle are discussed and estimates of expected levels are given. The feasibility of various methods of recycling thorium is discussed. These methods include direct recycle, recycle after storage for 14 years to allow radioactivity to decrease, shortening irradiation times to limit radioactivity build up, and the use of the window in time immediately after reprocessing where radioactivity levels are diminished. An economic comparison ismore » made for the first two methods together with the throwaway option where thorium is not recycled using a mass energy flow model developed for a CTHR (Commercial Tokamak Hybrid Reactor), a fusion fission hybrid reactor which serves as fuel producer for several PWR reactors. The storage option is found to be most favorable; however, even this option represents a significant economic impact due to radioactivity of 0.074 mills/kW-h which amounts to $4 x 10/sup 9/ over a 30 year period assuming a 200 gigawatt supply of electrical power.« less

PRESSURIZED WATER REACTOR (PWR) PROJECT TECHNICAL PROGRESS REPORT FOR THE PERIOD DECEMBER 24, 1959 TO FEBRUARY 23, 1960

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

None

-pressure-bonding appears to be a suitable means of incorporating void volume in fuel compartments of oxide plates. High density (99% T.D.) and improved microstructure of B/sub 4/C-SiC burnable poisons are achieved when small (2 micron) B/sub 4/C particle size powder is used ia hot pressing compacts. Measurements of the self-diffusion coefficients of uranium in UO/sub 2/ by the method of surface activity decrease were completed. Experiments on the diffusion of Xe/sup 133/ in Core 2--type UO/sup 2/ fuel platelets were completed. Diffusion anaeals carried out at 1000 deg C on samples from the X-3-1 and the 14-28 irradiation tests show that the apparent diffusion coefficient for Kr/sup 85/ incresses considerably with burnup. An average activation energy for thoron emanation in UO/sub 2/ was estimated to be 44 kcal/mole. An initial experiment on the release of helium from slightly irradiated B/sub 4/C at 900 deg C resulted in a diffusion coefficient for helium of 3.5 x 10/sup -8/ Physics: Calculatad values for seed-blanket power sharing as a function of PWR-1 Seed 1 life were compared with measured data obtained from thermal instrumentation at Shippingport. Two-dimensional depletion studies in the PWR-2 "composite cell" geometry were completed for seed assembly configurations having different radial fuel zoning. An eighth core representation is being employed for a two- dimensional depletion calculation of PWR-2. An analysis of the effect on the axial power distribution of the nonuniform temperature distribution in an 8 ft PWR-2 core loaded with 295 kg of U/sup 235/ indicated that local variations in power density of as much as 15% may occur, relative to the distribution that would exist if the axial temperature distribution were uniform. A technique was developed which makes possible an approximately correct description of the neutron capture rate within small rectangular boron wafers in diffusion theory calculations. Seed peaking factors measured in a five-cluster slab of PWR-2 mock- up

Multivariate analysis of gamma spectra to characterize used nuclear fuel

DOE PAGES

Coble, Jamie; Orton, Christopher; Schwantes, Jon

2017-01-17

The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

Multivariate analysis of gamma spectra to characterize used nuclear fuel

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

Coble, Jamie; Orton, Christopher; Schwantes, Jon

The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

Pool-site fuel inspection and examination techniques applied by the Kraftwerk Union Aktiengesellschaft Fuel Service. [PWR; BWR

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

Knaab, H.; Knecht, K.

The need for pool-site inspection and examination of fuel assemblies was recognized by Kraftwerk Union Aktiengesellschaft with the commissioning of the first nuclear power stations. A wet sipping method has demonstrated high reliability in detection of leaking fuel assemblies. The visual inspection system is a versatile tool. It can be supplemented by attaching devices for oxide thickness measurement or surface replication. Repair of leaking pressurized water reactor fuel assemblies has improved fuel utilization. Applied methods and typical results are described.

Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

NASA Astrophysics Data System (ADS)

Hartini, Entin; Andiwijayakusuma, Dinan

2014-09-01

This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.

The benefits of a fast reactor closed fuel cycle in the UK

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

Gregg, R.; Hesketh, K.

2013-07-01

The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size,more » so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since

Spent fuel burnup estimation by Cerenkov glow intensity measurement

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

Kuribara, Masayuki

1994-10-01

The Cerenkov glow images from irradiated fuel assemblies of boiling-water reactors (BWR) and pressurized-water reactors (PWR) are generally used for inspections. For this purpose, a new UV-I.I. CVD (ultra-violet light image intensifier Cerenkov viewing device), has been developed. This new device can measure the intensity of the Cerenkov glow from a spent fuel assembly, thus making it possible to estimate the burnup of the fuel assembly by comparing the Cerenkov glow intensity to the reference intensity. The experiment was carried out on BWR spent fuel assemblies and the results show that burnups are estimated within 20% accuracy compared to themore » declared burnups for the tested spent fuel assemblies for cooling times ranging from 900--2.000 d.« less

Performance of U3Si2 Fuel in a Reactivity Insertion Accident

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

Cheng, Lap Y.; Cuadra, Arantxa; Todosow, Michael

In this study we examined the performance of the U3Si2 fuel cladded with Zircaloy (Zr) in a reactivity insertion accident (RIA) in a PWR core. The power excursion as a result of a $1 reactivity insertion was calculated by a TRACE PWR plant model using point-kinetics, for alternative cores with UO2 and U3Si2 fuel assemblies. The point-kinetics parameters (feedback coefficients, prompt-neutron lifetime and group constants for six delayed-neutron groups) were obtained from beginning-of-cycle equilibrium full core calculations with PARCS. In the PARCS core calculations, the few-group parameters were developed utilizing the TRITON/NEWT tools in the SCALE package. In order tomore » assess the fuel response in finer detail (e.g. the maximum fuel temperature) the power shape and thermal boundary conditions from the TRACE/PARCS calculations were used to drive a BISON model of a fuel pin with U3Si2 and UO2 respectively. For a $1 reactivity transient both TRACE and BISON predicted a higher maximum fuel temperature for the UO2 fuel than the U3Si2 fuel. Furthermore, BISON is noted to calculate a narrower gap and a higher gap heat transfer coefficient than TRACE. This resulted in BISON predicting consistently lower fuel temperatures than TRACE. This study also provides a systematic comparison between TRACE and BISON using consistent transient boundary conditions. The TRACE analysis of the RIA only reflects the core-wide response in power. A refinement to the analysis would be to predict the local peaking in a three-dimensional core as a result of control rod ejection.« less

Fretting wear behaviors of a dual-cooled nuclear fuel rod under a simulated rod vibration

NASA Astrophysics Data System (ADS)

Lee, Young-Ho; Kim, Hyung-Kyu; Kang, Heung-Seok; Yoon, Kyung-Ho; Kim, Jae-Yong; Lee, Kang-Hee

2012-06-01

Recently, a dual-cooled fuel (i.e., annular fuel) that is compatible with current operating PWR plants has been proposed in order to realize both a considerable amount of power uprating and an increase of safety margins. As the design concept should be compatible with current operating PWR plants, however, it shows a narrow gap between the fuel rods when compared with current solid nuclear fuel arrays and needs to modify the spacer grid shapes and their positions. In this study, fretting wear tests have been performed to evaluate the wear resistance of a dual-cooled fuel by using a proposed spring and dimple of spacer grids that have a cantilever type and hemispherical shape, respectively. As a result, the wear volume of the spring specimen gradually increases as the contact condition is changed from a certain gap, just contact to positive force. However, in the dimple specimen, just contact condition shows a large wear volume. In addition, a circular rod motion at upper region of contact surface is gradually increased and its diametric size depends on the wear depth increase. Based on the test results, the fretting wear resistance of the proposed spring and dimple is analyzed by comparing the wear measurement results and rod motion in detail.

Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

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

Hartini, Entin, E-mail: entin@batan.go.id; Andiwijayakusuma, Dinan, E-mail: entin@batan.go.id

2014-09-30

This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuelmore » type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.« less

The effect of some organic substances on the mycelium of the fungus Ustilago nuda (Jens.) Rostr.

PubMed

Krátká, J

1976-01-01

Research was performed for studying the effect of some organic compounds, considered by many authors as the products ob barley seed metabolism generated after anaerobic seed treatment, on the mycelium of the fungus Ustilago nuda (Jens.) Rostr. The author examined the effectiveness of ethylacohol, acetaldehyde, acetic acid, succinic acid, lactic acid, and hydroquinone in concentrations from 1 M to 10(-6) M, and the effectiveness of extracts from disinfected seeds in doses from 10 g to 0.001 g/l. The effect of the mentioned solutions was examined as exerted on the growth of dicaryotic mycelium and on the growth of the haploid promycelium of the fungus. The dicaryotic mycelium of Ustilago nuda (Jens.) Rostr. was cultivated on potato agar with benzoic acid. The presence of the acid prevents mitosis, and the chlamydospores germinate on the nutritive medium with two fibres having binuclear cells. The haploid promycelium was cultivated on potato agar; chlamydospores germinated with one four-cell fibre, and individual cells are mononuclear and haploid. Only later, a dicarytic mycelium is created in a complex process. In all the substances used, the concentration of 1 M was found to stop further growth of mycelium. The concentration of 10(-1) M of acetic acid and hydroquinone also stopped growth, the same concentration of acetaldehyde, lactic acid, succinic acid, ethylacohol stimulated mycelium growth in comparison with the control. The concentration of 10(-6) M stimulated mycelium growth in a majority of cases. Extracts from disinfected seeds did not influence mycelium growth significantly in all cases in comparison with the control. The results were similar in the two types of mycelium.

PWR Facility Dose Modeling Using MCNP5 and the CADIS/ADVANTG Variance-Reduction Methodology

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

Blakeman, Edward D; Peplow, Douglas E.; Wagner, John C

2007-09-01

The feasibility of modeling a pressurized-water-reactor (PWR) facility and calculating dose rates at all locations within the containment and adjoining structures using MCNP5 with mesh tallies is presented. Calculations of dose rates resulting from neutron and photon sources from the reactor (operating and shut down for various periods) and the spent fuel pool, as well as for the photon source from the primary coolant loop, were all of interest. Identification of the PWR facility, development of the MCNP-based model and automation of the run process, calculation of the various sources, and development of methods for visually examining mesh tally filesmore » and extracting dose rates were all a significant part of the project. Advanced variance reduction, which was required because of the size of the model and the large amount of shielding, was performed via the CADIS/ADVANTG approach. This methodology uses an automatically generated three-dimensional discrete ordinates model to calculate adjoint fluxes from which MCNP weight windows and source bias parameters are generated. Investigative calculations were performed using a simple block model and a simplified full-scale model of the PWR containment, in which the adjoint source was placed in various regions. In general, it was shown that placement of the adjoint source on the periphery of the model provided adequate results for regions reasonably close to the source (e.g., within the containment structure for the reactor source). A modification to the CADIS/ADVANTG methodology was also studied in which a global adjoint source is weighted by the reciprocal of the dose response calculated by an earlier forward discrete ordinates calculation. This method showed improved results over those using the standard CADIS/ADVANTG approach, and its further investigation is recommended for future efforts.« less

Isotopic Details of the Spent Catawba-1 MOX Fuel Rods at ORNL

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

Ellis, Ronald James

The United States Department of Energy funded Shaw/AREVA MOX Services LLC to fabricate four MOX Lead Test Assemblies (LTA) from weapons-grade plutonium. A total of four MOX LTAs (including MX03) were irradiated in the Catawba Nuclear Station (Unit 1) Catawba-1 PWR which operated at a total thermal power of 3411 MWt and had a core with 193 total fuel assemblies. The MOX LTAs were irradiated along with Duke Energy s irradiation of eight Westinghouse Next Generation Fuel (NGF) LEU LTAs (ref.1) and the remaining 181 LEU fuel assemblies. The MX03 LTA was irradiated in the Catawba-1 PWR core (refs.2,3) duringmore » cycles C-16 and C-17. C-16 began on June 5, 2005, and ended on November 11, 2006, after 499 effective full power days (EFPDs). C-17 started on December 29, 2006, (after a shutdown of 48 days) and continued for 485 EFPDs. The MX03 and three other MOX LTAs (and other fuel assemblies) were discharged at the end of C-17 on May 3, 2008. The design of the MOX LTAs was based on the (Framatome ANP, Inc.) Mark-BW/MOX1 17 17 fuel assembly design (refs. 4,5,6) for use in Westinghouse PWRs, but with MOX fuel rods with three Pu loading ranges: the nominal Pu loadings are 4.94 wt%, 3.30 wt%, and 2.40 wt%, respectively, for high, medium, and low Pu content. The Mark-BW/MOX1 (MOX LTA) fuel assembly design is the same as the Advanced Mark-BW fuel assembly design but with the LEU fuel rods replaced by MOX fuel rods (ref. 5). The fabrication of the fuel pellets and fuel rods for the MOX LTAs was performed at the Cadarache facility in France, with the fabrication of the LTAs performed at the MELOX facility, also in France.« less

Finite Element Stress Analysis of Spent Nuclear Fuel Disposal Canister in a Deep Geological Repository

NASA Astrophysics Data System (ADS)

Kwon, Young Joo; Choi, Jong Won

This paper presents the finite element stress analysis of a spent nuclear fuel disposal canister to provide basic information for dimensioning the canister and configuration of canister components and consequently to suggest the structural analysis methodology for the disposal canister in a deep geological repository which is nowadays very important in the environmental waste treatment technology. Because of big differences in the pressurized water reactor (PWR) and the Canadian deuterium and uranium reactor (CANDU) fuel properties, two types of canisters are conceived. For manufacturing, operational reasons and standardization, however, both canisters have the same outer diameter and length. The construction type of canisters introduced here is a solid structure with a cast insert and a corrosion resistant overpack. The structural stress analysis is carried out using a finite element analysis code, NISA, and focused on the structural strength of the canister against the expected external pressures due to the swelling of the bentonite buffer and the hydrostatic head. The canister must withstand these large pressure loads. Consequently, canisters presented here contain 4 PWR fuel assemblies and 33×9 CANDU fuel bundles. The outside diameter of the canister for both fuels is 122cm and the cast insert diameter is 112cm. The total length of the canister is 483cm with the lid/bottom and the outer shell of 5cm.

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

DOE PAGES

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa; ...

2016-09-07

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

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

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

Comparison of fresh fuel experimental measurements to MCNPX calculations using self-interrogation neutron resonance densitometry

NASA Astrophysics Data System (ADS)

LaFleur, Adrienne M.; Charlton, William S.; Menlove, Howard O.; Swinhoe, Martyn T.

2012-07-01

A new non-destructive assay technique called Self-Interrogation Neutron Resonance Densitometry (SINRD) is currently being developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for Light Water Reactor (LWR) fuel assemblies. SINRD consists of four 235U fission chambers (FCs): bare FC, boron carbide shielded FC, Gd covered FC, and Cd covered FC. Ratios of different FCs are used to determine the amount of resonance absorption from 235U in the fuel assembly. The sensitivity of this technique is based on using the same fissile materials in the FCs as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. In this work, experimental measurements were performed in air with SINRD using a reference Pressurized Water Reactor (PWR) 15×15 low enriched uranium (LEU) fresh fuel assembly at LANL. The purpose of this experiment was to assess the following capabilities of SINRD: (1) ability to measure the effective 235U enrichment of the PWR fresh LEU fuel assembly and (2) sensitivity and penetrability to the removal of fuel pins from an assembly. These measurements were compared to Monte Carlo N-Particle eXtended transport code (MCNPX) simulations to verify the accuracy of the MCNPX model of SINRD. The reproducibility of experimental measurements via MCNPX simulations is essential to validating the results and conclusions obtained from the simulations of SINRD for LWR spent fuel assemblies.

The future in Agricultural Engineering: news degrees in the Universidad Politécnica de Madrid (UPM)

NASA Astrophysics Data System (ADS)

Cartagena, M. Carmen; Tarquis, A. M.; Vázquez, J.; Serrano, A.; Arce, A.

2010-05-01

The Bologna process is to improve the quality of education, mobility, diversity and the competitiveness and involves three fundamental changes: transform of the structure of titles, changing in methods of teaching and implementation of the systems of quality assurance. Engineer Agronomist at the Universidad Politécnica de Madrid (UPM) has been offered as a degree of five years with a total of 400 credits and seven optional orientations: Crop Production, Plant and Breeding Protection, Environment, Agricultural Economics, Animal Production, Rural Engineering and Food Technology. Actually, the Bologna plan creates three new degrees: Engineering and Science Agronomic, Food Engineering and Agro-Environmental Engineering, with 240 ECTS each one of them and with specific professional characteristics. The changes that involve the introduction of these new degrees is perhaps the largest occurred never at the Spanish university system, not only by the drastic transformation in the structure of titles, but also by the new changes that lie ahead in teaching methods. Among others we will comment the following ones: -A year decreased duration of studies and therefore incorporation into the market. - Elimination of the seven current guidelines to create three specific qualifications of degree. -Decrease of optional subjects and increase in credits for the basic subjects. - Inclusion of business practices. - Increase in the number of credits of final project. - Changes in methodologies and a higher involvement of teachers and students in the education.

PWR FLECHT SEASET 163-Rod Bundle Flow Blockage Task data report. NRC/EPRI/Westinghouse report No. 13, August-October 1982

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

Loftus, M J; Hochreiter, L E; McGuire, M F

This report presents data from the 163-Rod Bundle Blow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Systems Effects and Separate Effects Test Program (FLECHT SEASET). The task consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. These tests were designed to determine effects of flow blockage and flow bypass on reflooding behavior and to aid in the assessment of computational models in predicting the reflooding behavior of flow blockage in rod bundle arrays.

Thorium-based mixed oxide fuel in a pressurized water reactor: A feasibility analysis with MCNP

NASA Astrophysics Data System (ADS)

Tucker, Lucas Powelson

This dissertation investigates techniques for spent fuel monitoring, and assesses the feasibility of using a thorium-based mixed oxide fuel in a conventional pressurized water reactor for plutonium disposition. Both non-paralyzing and paralyzing dead-time calculations were performed for the Portable Spectroscopic Fast Neutron Probe (N-Probe), which can be used for spent fuel interrogation. Also, a Canberra 3He neutron detector's dead-time was estimated using a combination of subcritical assembly measurements and MCNP simulations. Next, a multitude of fission products were identified as candidates for burnup and spent fuel analysis of irradiated mixed oxide fuel. The best isotopes for these applications were identified by investigating half-life, photon energy, fission yield, branching ratios, production modes, thermal neutron absorption cross section and fuel matrix diffusivity. 132I and 97Nb were identified as good candidates for MOX fuel on-line burnup analysis. In the second, and most important, part of this work, the feasibility of utilizing ThMOX fuel in a pressurized water reactor (PWR) was first examined under steady-state, beginning of life conditions. Using a three-dimensional MCNP model of a Westinghouse-type 17x17 PWR, several fuel compositions and configurations of a one-third ThMOX core were compared to a 100% UO2 core. A blanket-type arrangement of 5.5 wt% PuO2 was determined to be the best candidate for further analysis. Next, the safety of the ThMOX configuration was evaluated through three cycles of burnup at several using the following metrics: axial and radial nuclear hot channel factors, moderator and fuel temperature coefficients, delayed neutron fraction, and shutdown margin. Additionally, the performance of the ThMOX configuration was assessed by tracking cycle length, plutonium destroyed, and fission product poison concentration.

Estimating irradiated nuclear fuel characteristics by nonlinear multivariate regression of simulated gamma-ray emissions

NASA Astrophysics Data System (ADS)

Åberg Lindell, M.; Andersson, P.; Grape, S.; Håkansson, A.; Thulin, M.

2018-07-01

In addition to verifying operator declared parameters of spent nuclear fuel, the ability to experimentally infer such parameters with a minimum of intrusiveness is of great interest and has been long-sought after in the nuclear safeguards community. It can also be anticipated that such ability would be of interest for quality assurance in e.g. recycling facilities in future Generation IV nuclear fuel cycles. One way to obtain information regarding spent nuclear fuel is to measure various gamma-ray intensities using high-resolution gamma-ray spectroscopy. While intensities from a few isotopes obtained from such measurements have traditionally been used pairwise, the approach in this work is to simultaneously analyze correlations between all available isotopes, using multivariate analysis techniques. Based on this approach, a methodology for inferring burnup, cooling time, and initial fissile content of PWR fuels using passive gamma-ray spectroscopy data has been investigated. PWR nuclear fuels, of UOX and MOX type, and their gamma-ray emissions, were simulated using the Monte Carlo code Serpent. Data comprising relative isotope activities was analyzed with decision trees and support vector machines, for predicting fuel parameters and their associated uncertainties. From this work it may be concluded that up to a cooling time of twenty years, the 95% prediction intervals of burnup, cooling time and initial fissile content could be inferred to within approximately 7 MWd/kgHM, 8 months, and 1.4 percentage points, respectively. An attempt aiming to estimate the plutonium content in spent UOX fuel, using the developed multivariate analysis model, is also presented. The results for Pu mass estimation are promising and call for further studies.

Nuclear Fuel Depletion Analysis Using Matlab Software

NASA Astrophysics Data System (ADS)

Faghihi, F.; Nematollahi, M. R.

Coupled first order IVPs are frequently used in many parts of engineering and sciences. In this article, we presented a code including three computer programs which are joint with the Matlab software to solve and plot the solutions of the first order coupled stiff or non-stiff IVPs. Some engineering and scientific problems related to IVPs are given and fuel depletion (production of the 239Pu isotope) in a Pressurized Water Nuclear Reactor (PWR) are computed by the present code.

Supporting the human life-raft in confronting the juggernaut of technology: Jens Rasmussen, 1961-1986.

PubMed

Kant, Vivek

2017-03-01

Jens Rasmussen's contribution to the field of human factors and ergonomics has had a lasting impact. Six prominent interrelated themes can be extracted from his research between 1961 and 1986. These themes form the basis of an engineering epistemology which is best manifested by his abstraction hierarchy. Further, Rasmussen reformulated technical reliability using systems language to enable a proper human-machine fit. To understand the concept of human-machine fit, he included the operator as a central component in the system to enhance system safety. This change resulted in the application of a qualitative and categorical approach for human-machine interaction design. Finally, Rasmussen's insistence on a working philosophy of systems design as being a joint responsibility of operators and designers provided the basis for averting errors and ensuring safe and correct system functioning. Copyright © 2016 Elsevier Ltd. All rights reserved.

Large-break LOCA, in-reactor fuel bundle Materials Test MT-6A

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

Wilson, C.L.; Hesson, G.M.; Pilger, J.P.

1993-09-01

This is a report on one of a series of experiments to simulates a loss-of-coolant accident (LOCA) using full-length fuel rods for pressurized water reactors (PWR). The experiments were conducted by Pacific Northwest Laboratory (PNL) under the LOCA simulation Program sponsored by the US Nuclear Regulatory Commission (NRC). The major objective of this program was causing the maximum possible expansion of the cladding on the fuel rods from a short-term adiabatic temperature transient to 1200 K (1700 F) leading to the rupture of the cladding; and second, by reflooding the fuel rods to determine the rate at which the fuelmore » bundle is cooled.« less

Application of Compton-suppressed self-induced XRF to spent nuclear fuel measurement

NASA Astrophysics Data System (ADS)

Park, Se-Hwan; Jo, Kwang Ho; Lee, Seung Kyu; Seo, Hee; Lee, Chaehun; Won, Byung-Hee; Ahn, Seong-Kyu; Ku, Jeong-Hoe

2017-11-01

Self-induced X-ray fluorescence (XRF) is a technique by which plutonium (Pu) content in spent nuclear fuel can be directly quantified. In the present work, this method successfully measured the plutonium/uranium (Pu/U) peak ratio of a pressurized water reactor (PWR)'s spent nuclear fuel at the Korea atomic energy research institute (KAERI)'s post irradiation examination facility (PIEF). In order to reduce the Compton background in the low-energy X-ray region, the Compton suppression system additionally was implemented. By use of this system, the spectrum's background level was reduced by a factor of approximately 2. This work shows that Compton-suppressed selfinduced XRF can be effectively applied to Pu accounting in spent nuclear fuel.

Differential Die-Away Instrument: Report on Fuel Assembly Mock-up Measurements with Neutron Generator

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

Goodsell, Alison Victoria; Swinhoe, Martyn Thomas; Henzl, Vladimir

2014-09-18

Fresh fuel experiments for the differential die-away (DDA) project were performed using a DT neutron generator, a 15x15 PWR fuel assembly, and nine 3He detectors in a water tank inside of a shielded cell at Los Alamos National Laboratory (LANL). Eight different fuel enrichments were created using low enriched (LEU) and depleted uranium (DU) dioxide fuel rods. A list-mode data acquisition system recorded the time-dependent signal and analysis of the DDA signal die-away time was performed. The die-away time depended on the amount of fissile material in the fuel assembly and the position of the detector. These experiments were performedmore » in support of the spent nuclear fuel Next Generation Safeguards Initiative DDA project. Lessons learned from the fresh fuel DDA instrument experiments and simulations will provide useful information to the spent fuel project.« less

Analysis and Implementation of Accident Tolerant Nuclear Fuels

NASA Astrophysics Data System (ADS)

Prewitt, Benjamin Joseph

To improve the reliability and robustness of LWR, accident tolerant nuclear fuels and cladding materials are being developed to possibly replace the current UO2/zirconium system. This research highlights UN and U3Si 2, two of the most favorable accident tolerant fuels being developed. To evaluate the commercial feasiblilty of these fuels, two areas of research were conducted. Chemical fabrication routes for both fuels were investigated in detail, considering UO2 and UF6 as potential starting materials. Potential pathways for industrial scale fabrication using these methods were discussed. Neutronic performance of 70%UN-30%U3Si2 composite was evaluated in MNCP using PWR assembly and core models. The results showed comparable performance to an identical UO2 fueled simulation with the same configuration. The parameters simulated for composite and oxide fuel include the following: fuel to moderator ratio curves; energy dependent flux spectra; temperature coefficients for fuel and moderator; delayed neutron fractions; power peaking factors; axial and radial flux profiles in 2D and 3D; burnup; critical boron concentration; and shutdown margin. Overall, the neutronic parameters suggest that the transition from UO2 to composite in existing nuclear systems will not require significant changes in operating procedures or modifications to standards and regulations.

Management of thermal peaking factors in CONFU-B PWR assemblies using neutron poisons and tailored enrichment

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

Visosky, M.; Hejzlar, P.; Kazimi, M.

2006-07-01

CONFU-B assemblies are PWR assemblies containing standard Uranium fuel rods and TRU bearing inert material fuel rods and are designed to achieve net TRU destruction over a 4.5-year irradiation. These highly heterogeneous assemblies tend to exhibit large intra-assembly power peaking factors (IAPPF). Neutronic strategies to reduce IAPPF are developed. The IAPPF are calculated at the assembly level using CASMO4, and these are used to calculate the most restrictive thermal margin (the Minimum Departure from Nucleate Boiling Ratio, MDNBR) using a whole-core VIPRE-01 model. This paper examines two strategies to manage the thermal margin of a CONFU-B assembly while retaining themore » TRU destruction performance: use of neutron poisons and tailored enrichment schemes. Burnable poisons can be used to suppress BOL reactivity of fresh CONFU-B assemblies with only minor impact on MDNBR and TRU destruction performance. Tailored enrichment, along with the use of soluble boron, can achieve significant improvements in MDNBR, but at some cost to TRU destruction performance. (authors)« less

The underwater coincidence counter (UWCC) for plutonium measurements in mixed oxide fuels

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

Eccleston, G.W.; Menlove, H.O.; Abhold, M.

1998-12-31

The use of fresh uranium-plutonium mixed oxide (MOX) fuel in light-water reactors (LWR) is increasing in Europe and Japan and it is necessary to verify the plutonium content in the fuel for international safeguards purposes. The UWCC is a new instrument that has been designed to operate underwater and nondestructively measure the plutonium in unirradiated MOX fuel assemblies. The UWCC can be quickly configured to measure either boiling-water reactor (BWR) or pressurized-water reactor (PWR) fuel assemblies. The plutonium loading per unit length is measured using the UWCC to precisions of less than 1% in a measurement time of 2 tomore » 3 minutes. Initial calibrations of the UWCC were completed on measurements of MOX fuel in Mol, Belgium. The MCNP-REN Monte Carlo simulation code is being benchmarked to the calibration measurements to allow accurate simulations for extended calibrations of the UWCC.« less

THE CALCULATION OF BURNABLE POISON CORRECTION FACTORS FOR PWR FRESH FUEL ACTIVE COLLAR MEASUREMENTS

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

Croft, Stephen; Favalli, Andrea; Swinhoe, Martyn T.

2012-06-19

Verification of commercial low enriched uranium light water reactor fuel takes place at the fuel fabrication facility as part of the overall international nuclear safeguards solution to the civilian use of nuclear technology. The fissile mass per unit length is determined nondestructively by active neutron coincidence counting using a neutron collar. A collar comprises four slabs of high density polyethylene that surround the assembly. Three of the slabs contain {sup 3}He filled proportional counters to detect time correlated fission neutrons induced by an AmLi source placed in the fourth slab. Historically, the response of a particular collar design to amore » particular fuel assembly type has been established by careful cross-calibration to experimental absolute calibrations. Traceability exists to sources and materials held at Los Alamos National Laboratory for over 35 years. This simple yet powerful approach has ensured consistency of application. Since the 1980's there has been a steady improvement in fuel performance. The trend has been to higher burn up. This requires the use of both higher initial enrichment and greater concentrations of burnable poisons. The original analytical relationships to correct for varying fuel composition are consequently being challenged because the experimental basis for them made use of fuels of lower enrichment and lower poison content than is in use today and is envisioned for use in the near term. Thus a reassessment of the correction factors is needed. Experimental reassessment is expensive and time consuming given the great variation between fuel assemblies in circulation. Fortunately current modeling methods enable relative response functions to be calculated with high accuracy. Hence modeling provides a more convenient and cost effective means to derive correction factors which are fit for purpose with confidence. In this work we use the Monte Carlo code MCNPX with neutron coincidence tallies to calculate the influence

The increase in fatigue crack growth rates observed for Zircaloy-4 in a PWR environment

NASA Astrophysics Data System (ADS)

Cockeram, B. V.; Kammenzind, B. F.

2018-02-01

Cyclic stresses produced during the operation of nuclear reactors can result in the extension of cracks by processes of fatigue. Although fatigue crack growth rate (FCGR) data for Zircaloy-4 in air are available, little testing has been performed in a PWR primary water environment. Test programs have been performed by Gee et al., in 1989 and Picker and Pickles in 1984 by the UK Atomic Energy Authority, and by Wisner et al., in 1994, that have shown an enhancement in FCGR for Zircaloy-2 and Zircaloy-4 in high-temperature water. In this work, FCGR testing is performed on Zircaloy-4 in a PWR environment in the hydrided and non-hydrided condition over a range of stress-intensity. Measurements of crack extension are performed using a direct current potential drop (DCPD) method. The cyclic rate in the PWR primary water environment is varied between 1 cycle per minute to 0.1 cycle per minute. Faster FCGR rates are observed in water in comparison to FCGR testing performed in air for the hydrided material. Hydrided and non-hydrided materials had similar FCGR values in air, but the non-hydrided material exhibited much lower rates of FCGR in a PWR primary water environment than for hydrided material. Hydrides are shown to exhibit an increased tendency for cracking or decohesion in a PWR primary water environment that results in an enhancement in FCGR values. The FCGR in the PWR primary water only increased slightly with decreasing cycle frequency in the range of 1 cycle per minute to 0.1 cycle per minute. Comparisons between the FCGR in water and air show the enhancement from the PWR environment is affected by the applied stress intensity.

Effects of materials and design on the criticality and shielding assessment of canister concepts for the disposal of spent nuclear fuel.

PubMed

Gutiérrez, Miguel Morales; Caruso, Stefano; Diomidis, Nikitas

2018-05-19

According to the Swiss disposal concept, the safety of a deep geological repository for spent nuclear fuel (SNF) is based on a multi-barrier system. The disposal canister is an important component of the engineered barrier system, aiming to provide containment of the SNF for thousands of years. This study evaluates the criticality safety and shielding of candidate disposal canister concepts, focusing on the fulfilment of the sub-criticality criterion and on limiting radiolysis processes at the outer surface of the canister which can enhance corrosion mechanisms. The effective neutron multiplication factor (k-eff) and the surface dose rates are calculated for three different canister designs and material combinations for boiling water reactor (BWR) canisters, containing 12 spent fuel assemblies (SFA), and pressurized water reactor (PWR) canisters, with 4 SFAs. For each configuration, individual criticality and shielding calculations were carried out. The results show that k-eff falls below the defined upper safety limit (USL) of 0.95 for all BWR configurations, while staying above USL for the PWR ones. Therefore, the application of a burnup credit methodology for the PWR case is required, being currently under development. Relevant is also the influence of canister material and internal geometry on criticality, enabling the identification of safer fuel arrangements. For a final burnup of 55MWd/kgHM and 30y cooling time, the combined photon-neutron surface dose rate is well below the threshold of 1 Gy/h defined to limit radiation-induced corrosion of the canister in all cases. Copyright © 2018 Elsevier Ltd. All rights reserved.

Detecting pin diversion from pressurized water reactors spent fuel assemblies

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

Ham, Young S.; Sitaraman, Shivakumar

Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and takingmore » the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.« less

Report on the PWR-radiation protection/ALARA Committee

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

Malone, D.J.

1995-03-01

In 1992, representatives from several utilities with operational Pressurized Water Reactors (PWR) formed the PWR-Radiation Protection/ALARA Committee. The mission of the Committee is to facilitate open communications between member utilities relative to radiation protection and ALARA issues such that cost effective dose reduction and radiation protection measures may be instituted. While industry deregulation appears inevitable and inter-utility competition is on the rise, Committee members are fully committed to sharing both positive and negative experiences for the benefit of the health and safety of the radiation worker. Committee meetings provide current operational experiences through members providing Plant status reports, and informationmore » relative to programmatic improvements through member presentations and topic specific workshops. The most recent Committee workshop was facilitated to provide members with defined experiences that provide cost effective ALARA performance.« less

Heuristic rules embedded genetic algorithm for in-core fuel management optimization

NASA Astrophysics Data System (ADS)

Alim, Fatih

The objective of this study was to develop a unique methodology and a practical tool for designing loading pattern (LP) and burnable poison (BP) pattern for a given Pressurized Water Reactor (PWR) core. Because of the large number of possible combinations for the fuel assembly (FA) loading in the core, the design of the core configuration is a complex optimization problem. It requires finding an optimal FA arrangement and BP placement in order to achieve maximum cycle length while satisfying the safety constraints. Genetic Algorithms (GA) have been already used to solve this problem for LP optimization for both PWR and Boiling Water Reactor (BWR). The GA, which is a stochastic method works with a group of solutions and uses random variables to make decisions. Based on the theories of evaluation, the GA involves natural selection and reproduction of the individuals in the population for the next generation. The GA works by creating an initial population, evaluating it, and then improving the population by using the evaluation operators. To solve this optimization problem, a LP optimization package, GARCO (Genetic Algorithm Reactor Code Optimization) code is developed in the framework of this thesis. This code is applicable for all types of PWR cores having different geometries and structures with an unlimited number of FA types in the inventory. To reach this goal, an innovative GA is developed by modifying the classical representation of the genotype. To obtain the best result in a shorter time, not only the representation is changed but also the algorithm is changed to use in-core fuel management heuristics rules. The improved GA code was tested to demonstrate and verify the advantages of the new enhancements. The developed methodology is explained in this thesis and preliminary results are shown for the VVER-1000 reactor hexagonal geometry core and the TMI-1 PWR. The improved GA code was tested to verify the advantages of new enhancements. The core physics code

EMERALD REVISION 1; PWR accident activity release. [IBM360,370; FORTRAN IV

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

Fowler, T.B.; Tobias, M.L.; Fox, J.N.

The EMERALD program is designed for the calculation of radiation releases and exposures resulting from abnormal operation of a large pressurized water reactor (PWR). The approach used in EMERALD is similar to an analog simulation of a real system. Each component or volume in the plant which contains a radioactive material is represented by a subroutine which keeps track of the production, transfer, decay and absorption of radioactivity in that volume. During the course of the analysis of an accident, activity is transferred from subroutine to subroutine in the program as it would be transferred from place to place inmore » the plant. For example, in the calculation of the doses resulting from a loss-of-coolant accident the program first calculates the activity built up in the fuel before the accident, then releases some of this activity to the containment volume. Some of this activity is then released to the atmosphere. The rates of transfer, leakage, production, cleanup, decay, and release are read in as input to the program. Subroutines are also included which calculate the on-site and off-site radiation exposures at various distances for individual isotopes and sums of isotopes. The program contains a library of physical data for the twenty-five isotopes of most interest in licensing calculations, and other isotopes can be added or substituted. Because of the flexible nature of the simulation approach, the EMERALD program can be used for most calculations involving the production and release of radioactive materials during abnormal operation of a PWR. These include design, operational, and licensing studies.IBM360,370; FORTRAN IV; OS/360,370 (IBM360,370); 520K bytes of memory are required..« less

Assessment of PWR Steam Generator modelling in RELAP5/MOD2. International Agreement Report

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

Putney, J.M.; Preece, R.J.

1993-06-01

An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed -- including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies willmore » still be present in the successor code RELAP5/MOD3.« less

Plasmid partition system of the P1par family from the pWR100 virulence plasmid of Shigella flexneri.

PubMed

Sergueev, Kirill; Dabrazhynetskaya, Alena; Austin, Stuart

2005-05-01

P1par family members promote the active segregation of a variety of plasmids and plasmid prophages in gram-negative bacteria. Each has genes for ParA and ParB proteins, followed by a parS partition site. The large virulence plasmid pWR100 of Shigella flexneri contains a new P1par family member: pWR100par. Although typical parA and parB genes are present, the putative pWR100parS site is atypical in sequence and organization. However, pWR100parS promoted accurate plasmid partition in Escherichia coli when the pWR100 Par proteins were supplied. Unique BoxB hexamer motifs within parS define species specificities among previously described family members. Although substantially different from P1parS from the P1 plasmid prophage of E. coli, pWR100parS has the same BoxB sequence. As predicted, the species specificity of the two types proved identical. They also shared partition-mediated incompatibility, consistent with the proposed mechanistic link between incompatibility and species specificity. Among several informative sequence differences between pWR100parS and P1parS is the presence of a 21-bp insert at the center of the pWR100parS site. Deletion of this insert left much of the parS activity intact. Tolerance of central inserts with integral numbers of helical DNA turns reflects the critical topology of these sites, which are bent by binding the host IHF protein.

Silicon carbide composite for light water reactor fuel assembly applications

NASA Astrophysics Data System (ADS)

Yueh, Ken; Terrani, Kurt A.

2014-05-01

The feasibility of using SiCf-SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

Testing and COBRA-SFS analysis of the VSC-17 ventilated concrete, spent fuel storage cask

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

McKinnon, M.A.; Dodge, R.E.; Schmitt, R.C.

1992-04-01

A performance test of a Pacific Sierra Nuclear VSC-17 ventilated concrete storage cask loaded with 17 canisters of consolidated PWR spent fuel generating approximately 15 kW was conducted. The performance test included measuring the cask surface, concrete, air channel surface, and fuel temperatures, as well as cask surface gamma and neutron dose rates. Testing was performed using vacuum, nitrogen, and helium backfill environments. Pretest predictions of cask thermal performance were made using the COBRA-SFS computer code. Analysis results were within 15{degrees}C of measured peak fuel temperature. Peak fuel temperature for normal operation was 321{degrees}C. In general, the surface dose ratesmore » were less than 30 mrem/h on the side of the cask and 40 mrem/h on the top of the cask.« less

Coupled Neutronics Thermal-Hydraulic Solution of a Full-Core PWR Using VERA-CS

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

Clarno, Kevin T; Palmtag, Scott; Davidson, Gregory G

2014-01-01

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a core simulator called VERA-CS to model operating PWR reactors with high resolution. This paper describes how the development of VERA-CS is being driven by a set of progression benchmark problems that specify the delivery of useful capability in discrete steps. As part of this development, this paper will describe the current capability of VERA-CS to perform a multiphysics simulation of an operating PWR at Hot Full Power (HFP) conditions using a set of existing computer codes coupled together in a novel method. Results for several single-assembly casesmore » are shown that demonstrate coupling for different boron concentrations and power levels. Finally, high-resolution results are shown for a full-core PWR reactor modeled in quarter-symmetry.« less

Characterization of LWRS Hybrid SiC-CMC-Zircaloy-4 Fuel Cladding after Gamma Irradiation

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

Isabella J van Rooyen

2012-09-01

The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory (INL) was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design (unfueled) in a simulated PWR water environment under irradiation conditions. The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability (LWRS) program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrixmore » composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing« less

Bias estimates used in lieu of validation of fission products and minor actinides in MCNP K eff calculations for PWR burnup credit casks

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

Mueller, Don E.; Marshall, William J.; Wagner, John C.

The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (k eff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the biasmore » due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of k eff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.« less

Integral Full Core Multi-Physics PWR Benchmark with Measured Data

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

Forget, Benoit; Smith, Kord; Kumar, Shikhar

In recent years, the importance of modeling and simulation has been highlighted extensively in the DOE research portfolio with concrete examples in nuclear engineering with the CASL and NEAMS programs. These research efforts and similar efforts worldwide aim at the development of high-fidelity multi-physics analysis tools for the simulation of current and next-generation nuclear power reactors. Like all analysis tools, verification and validation is essential to guarantee proper functioning of the software and methods employed. The current approach relies mainly on the validation of single physic phenomena (e.g. critical experiment, flow loops, etc.) and there is a lack of relevantmore » multiphysics benchmark measurements that are necessary to validate high-fidelity methods being developed today. This work introduces a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading and re-loading patterns. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from 58 instrumented assemblies. The benchmark description is now available online and has been used by many groups. However, much work remains to be done on the quantification of uncertainties and modeling sensitivities. This work aims to address these deficiencies and make this benchmark a true non-proprietary international benchmark for the validation of high-fidelity tools. This report details the BEAVRS uncertainty quantification for the first two cycle of operations and serves as the final report of the

EPRI/DOE High-Burnup Fuel Sister Rod Test Plan Simplification and Visualization

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

Saltzstein, Sylvia J.; Sorenson, Ken B.; Hanson, B. D.

The EPRI/DOE High-Burnup Confirmatory Data Project (herein called the “Demo”) is a multi-year, multi-entity test with the purpose of providing quantitative and qualitative data to show if high-burnup fuel mechanical properties change in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of common cladding alloys from the North Anna Nuclear Power Plant, loading them in an NRC-licensed TN-32B cask, drying them according to standard plant procedures, and then storing them on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and themore » mechanical properties of the rods will be tested and analyzed.« less

Advanced Pellet-Cladding Interaction Modeling using the US DOE CASL Fuel Performance Code: Peregrine

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

Montgomery, Robert O.; Capps, Nathan A.; Sunderland, Dion J.

The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermo-mechanical-chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale code thatmore » is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.« less

Advanced Pellet Cladding Interaction Modeling Using the US DOE CASL Fuel Performance Code: Peregrine

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

Jason Hales; Various

The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermomechanical- chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale codemore » that is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.« less

BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

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

Bevard, Bruce Balkcom; Mertyurek, Ugur; Belles, Randy

The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been storedmore » on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

DOEpatents

Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

1993-01-01

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

DOEpatents

Corletti, M.M.; Lau, L.K.; Schulz, T.L.

1993-12-14

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps. 1 figures.

CASMO5/TSUNAMI-3D spent nuclear fuel reactivity uncertainty analysis

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

Ferrer, R.; Rhodes, J.; Smith, K.

2012-07-01

The CASMO5 lattice physics code is used in conjunction with the TSUNAMI-3D sequence in ORNL's SCALE 6 code system to estimate the uncertainties in hot-to-cold reactivity changes due to cross-section uncertainty for PWR assemblies at various burnup points. The goal of the analysis is to establish the multiplication factor uncertainty similarity between various fuel assemblies at different conditions in a quantifiable manner and to obtain a bound on the hot-to-cold reactivity uncertainty over the various assembly types and burnup attributed to fundamental cross-section data uncertainty. (authors)

Development a computer codes to couple PWR-GALE output and PC-CREAM input

NASA Astrophysics Data System (ADS)

Kuntjoro, S.; Budi Setiawan, M.; Nursinta Adi, W.; Deswandri; Sunaryo, G. R.

2018-02-01

Radionuclide dispersion analysis is part of an important reactor safety analysis. From the analysis it can be obtained the amount of doses received by radiation workers and communities around nuclear reactor. The radionuclide dispersion analysis under normal operating conditions is carried out using the PC-CREAM code, and it requires input data such as source term and population distribution. Input data is derived from the output of another program that is PWR-GALE and written Population Distribution data in certain format. Compiling inputs for PC-CREAM programs manually requires high accuracy, as it involves large amounts of data in certain formats and often errors in compiling inputs manually. To minimize errors in input generation, than it is make coupling program for PWR-GALE and PC-CREAM programs and a program for writing population distribution according to the PC-CREAM input format. This work was conducted to create the coupling programming between PWR-GALE output and PC-CREAM input and programming to written population data in the required formats. Programming is done by using Python programming language which has advantages of multiplatform, object-oriented and interactive. The result of this work is software for coupling data of source term and written population distribution data. So that input to PC-CREAM program can be done easily and avoid formatting errors. Programming sourceterm coupling program PWR-GALE and PC-CREAM is completed, so that the creation of PC-CREAM inputs in souceterm and distribution data can be done easily and according to the desired format.

Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

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

R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag

2012-04-01

Pressurized Water Reactor (PWR) assemblies. In addition to consideration of this 'naive' use of TRISO fuel in LWRs, several refined options are briefly examined and others are identified for further consideration including the use of advanced, high density fuel forms and larger kernel diameters and TRISO packing fractions. The combination of 800 {micro}m diameter kernels of 20% enriched UN and 50% TRISO packing fraction yielded reactivity sufficient to achieve comparable burnup to present-day PWR fuel.« less

Irradiation behaviour of the large grained UO{sub 2} fuel pellet in the transient conditions

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

Kosaka, Yuji; Watanabe, Seiichi; Arakawa, Yasushi

2007-07-01

In order to achieve a high duty fuel rod design, it is the key issue to suppress the fission gas release from the view point of the fuel rod inner pressure design. The large grain UO{sub 2} pellet is one of the candidates to meet such a requirement by reducing the fission gas release especially at high power and/or high burnup. We have demonstrated the fuel performance of the large grain pellet in the PWR irradiation conditions, which was fabricated with no additive but with active UO{sub 2} powder through the conventional pelletizing process for the normal grain size pellet.more » According to the mechanism of the fission gas retention, there may be a concern about the larger gas bubble swelling of the large grain pellet at the power transient conditions which may increase the potential of the PCMI failure. In this paper, we focus on the differences of the dimensional change in comparison among the pellets with the different grain sizes at the power transient conditions. The power ramp tests were carried out on the high burnup fuel rods of normal and large grain pellet with no additive, which had been irradiated in the PWR conditions up to around 60 GWd/t at peak position. The detailed PIE results revealed that the volume increment due to the power ramp clearly showed the dependence on the grain size as well as the fission gas release and suggested that the larger grain with no additive may suppress the gas bubble swelling at the power transient conditions. According to the experimental results, it is concluded that the large grain pellet with no additive does not deteriorate the irradiation performance during the power transient conditions from the view point of the gas bubble swelling. (authors)« less

VERA Core Simulator Methodology for PWR Cycle Depletion

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

Kochunas, Brendan; Collins, Benjamin S; Jabaay, Daniel

2015-01-01

This paper describes the methodology developed and implemented in MPACT for performing high-fidelity pressurized water reactor (PWR) multi-cycle core physics calculations. MPACT is being developed primarily for application within the Consortium for the Advanced Simulation of Light Water Reactors (CASL) as one of the main components of the VERA Core Simulator, the others being COBRA-TF and ORIGEN. The methods summarized in this paper include a methodology for performing resonance self-shielding and computing macroscopic cross sections, 2-D/1-D transport, nuclide depletion, thermal-hydraulic feedback, and other supporting methods. These methods represent a minimal set needed to simulate high-fidelity models of a realistic nuclearmore » reactor. Results demonstrating this are presented from the simulation of a realistic model of the first cycle of Watts Bar Unit 1. The simulation, which approximates the cycle operation, is observed to be within 50 ppm boron (ppmB) reactivity for all simulated points in the cycle and approximately 15 ppmB for a consistent statepoint. The verification and validation of the PWR cycle depletion capability in MPACT is the focus of two companion papers.« less

Thorium Fuel Options for Sustained Transuranic Burning in Pressurized Water Reactors - 12381

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

Rahman, Fariz Abdul; Lee, John C.; Franceschini, Fausto

2012-07-01

As described in companion papers, Westinghouse is proposing the adoption of a thorium-based fuel cycle to burn the transuranics (TRU) contained in the current Used Nuclear Fuel (UNF) and transition towards a less radio-toxic high level waste. A combination of both light water reactors (LWR) and fast reactors (FR) is envisaged for the task, with the emphasis initially posed on their TRU burning capability and eventually to their self-sufficiency. Given the many technical challenges and development times related to the deployment of TRU burners fast reactors, an interim solution making best use of the current resources to initiate burning themore » legacy TRU inventory while developing and testing some technologies of later use is desirable. In this perspective, a portion of the LWR fleet can be used to start burning the legacy TRUs using Th-based fuels compatible with the current plants and operational features. This analysis focuses on a typical 4-loop PWR, with 17x17 fuel assembly design and TRUs (or Pu) admixed with Th (similar to U-MOX fuel, but with Th instead of U). Global calculations of the core were represented with unit assembly simulations using the Linear Reactivity Model (LRM). Several assembly configurations have been developed to offer two options that can be attractive during the TRU transmutation campaign: maximization of the TRU transmutation rate and capability for TRU multi-recycling, to extend the option of TRU recycling in LWR until the FR is available. Homogeneous as well as heterogeneous assembly configurations have been developed with various recycling schemes (Pu recycle, TRU recycle, TRU and in-bred U recycle etc.). Oxide as well as nitride fuels have been examined. This enabled an assessment of the potential for burning and multi-recycling TRU in a Th-based fuel PWR to compare against other more typical alternatives (U-MOX and variations thereof). Results will be shown indicating that Th-based PWR fuel is a promising option to multi

Thermal modeling of a vertical dry storage cask for used nuclear fuel

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

Li, Jie; Liu, Yung Y.

2016-05-01

Thermal modeling of temperature profiles of dry casks has been identified as a high-priority item in a U.S. Department of Energy gap analysis. In this work, a three-dimensional model of a vertical dry cask has been constructed for computer simulation by using the ANSYS/FLUENT code. The vertical storage cask contains a welded canister for 32 Pressurized Water Reactor (PWR) used-fuel assemblies with a total decay heat load of 34 kW. To simplify thermal calculations, an effective thermal conductivity model for a 17 x 17 PWR used (or spent)-fuel assembly was developed and used in the simulation of thermal performance. Themore » effects of canister fill gas (helium or nitrogen), internal pressure (1-6 atm), and basket material (stainless steel or aluminum alloy) were studied to determine the peak cladding temperature (PCT) and the canister surface temperatures (CSTs). The results showed that high thermal conductivity of the basket material greatly enhances heat transfer and reduces the PCT. The results also showed that natural convection affects both PCT and the CST profile, while the latter depends strongly on the type of fill gas and canister internal pressure. Of particular interest to condition and performance monitoring is the identification of canister locations where significant temperature change occurs after a canister is breached and the fill gas changes from high-pressure helium to ambient air. This study provided insight on the thermal performance of a vertical storage cask containing high-burnup fuel, and helped advance the concept of monitoring CSTs as a means to detect helium leakage from a welded canister. The effects of blockage of air inlet vents on the cask's thermal performance were studied. The simulation were validated by comparing the results against data obtained from the temperature measurements of a commercial cask.« less

Multidimensional effects in the thermal response of fuel rod simulators. [PWR

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

Dabbs, R.D.; Ott, L.J.

1980-01-01

One of the primary objectives of the Oak Ridge National Laboratory Pressurized-Water Reactor Blowdown Heat Transfer Separate-Effects Program is the determination of the transient surface temperature and surface heat flux of fuel pin simulators (FPSs) from internal thermocouple signals obtained during a loss-of-coolant experiment (LOCE) in the Thermal-Hydraulics Test Facility. This analysis requires the solution of the classical inverse heat conduction problem. The assumptions that allow the governing differential equation to be reduced to one dimension can introduce significant errors in the computed surface heat flux and surface temperature. The degree to which these computed variables are perturbed is addressedmore » and quantified.« less

MC21 analysis of the MIT PWR benchmark: Hot zero power results

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

Kelly Iii, D. J.; Aviles, B. N.; Herman, B. R.

2013-07-01

MC21 Monte Carlo results have been compared with hot zero power measurements from an operating pressurized water reactor (PWR), as specified in a new full core PWR performance benchmark from the MIT Computational Reactor Physics Group. Included in the comparisons are axially integrated full core detector measurements, axial detector profiles, control rod bank worths, and temperature coefficients. Power depressions from grid spacers are seen clearly in the MC21 results. Application of Coarse Mesh Finite Difference (CMFD) acceleration within MC21 has been accomplished, resulting in a significant reduction of inactive batches necessary to converge the fission source. CMFD acceleration has alsomore » been shown to work seamlessly with the Uniform Fission Site (UFS) variance reduction method. (authors)« less

Cyclic and SCC Behavior of Alloy 690 HAZ in a PWR Environment

NASA Astrophysics Data System (ADS)

Alexandreanu, Bogdan; Chen, Yiren; Natesan, Ken; Shack, Bill

The objective of this work is to determine the cyclic and stress corrosion cracking (SCC) crack growth rates (CGRs) in a simulated PWR water environment for Alloy 690 heat affected zone (HAZ). In order to meet the objective, an Alloy 152 J-weld was produced on a piece of Alloy 690 tubing, and the test specimens were aligned with the HAZ. The environmental enhancement of cyclic CGRs for Alloy 690 HAZ was comparable to that measured for the same alloy in the as-received condition. The two Alloy 690 HAZ samples tested exhibited maximum SCC CGR rates of 10-11 m/s in the simulated PWR environment at 320°C, however, on average, these rates are similar or only slightly higher than those for the as-received alloy.

Adaptation and Molecular Characterization of Two Malaysian Very Virulent Infectious Bursal Disease Virus Isolates Adapted in BGM-70 Cell Line

PubMed Central

Lawal, Nafi'u; Arshad, Siti Suri

2017-01-01

Two Malaysian very virulent infectious bursal disease virus (vvIBDV) strains UPM0081 and UPM190 (also known as UPMB00/81 and UPM04/190, respectively) isolated from local IBD outbreaks were serially passaged 12 times (EP12) in specific pathogen free (SPF) chicken embryonated eggs (CEE) by chorioallantoic membrane (CAM) route. The EP12 isolate was further adapted and serially propagated in BGM-70 cell line up to 20 passages (P20). Characteristic cytopathic effects (CPEs) were subtly observed at P1 in both isolates 72 hours postinoculation (pi). The CPE became prominent at P5 with cell rounding, cytoplasmic vacuoles, granulation, and detachment from flask starting from day 3 pi, up to 7 days pi with titers of 109.50 TCID50/mL and log109.80 TCID50/mL for UPM0081 and UPM190, respectively. The CPE became subtle at P17 and disappeared by P18 and P19 for UPM0081 and UPM190, respectively. However, the presence of IBDV was confirmed by immunoperoxidase, immunofluorescence, and RT-PCR techniques. Phylogenetic analysis showed that these two isolates were of the vvIBDV. It appears that a single mutation of UPM190 and UPM0081 IBDV isolates at D279N could facilitate vvIBDV strain adaptability in CEE and BGM-70 cultures. PMID:29230245

Simulation of differential die-away instrument’s response to asymmetrically burned spent nuclear fuel

DOE PAGES

Martinik, Tomas; Henzl, Vladimir; Grape, Sophie; ...

2015-03-04

Here, previous simulation studies of Differential Die–Away (DDA) instrument’s response to active interrogation of spent nuclear fuel from a pressurized water reactor (PWR) yielded promising results in terms of its capability to accurately measure or estimate basic spent fuel assembly (SFA) characteristics, such as multiplication, initial enrichment (IE) and burn-up (BU) as well as the total plutonium content. These studies were however performed only for a subset of idealized SFAs with a symmetric BU with respect to its longitudinal axis. Therefore, to complement the previous results, additional simulations have been performed of the DDA instrument’s response to interrogation of asymmetricallymore » burned spent nuclear fuel in order to determine whether detailed assay of SFAs from all 4 sides will be necessary in real life applications or whether a cost and time saving single sided assay could be used to achieve results of similar quality as previously reported in case of symmetrically burned SFAs.« less

Simulation of differential die-away instrument’s response to asymmetrically burned spent nuclear fuel

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

Martinik, Tomas; Henzl, Vladimir; Grape, Sophie

Here, previous simulation studies of Differential Die–Away (DDA) instrument’s response to active interrogation of spent nuclear fuel from a pressurized water reactor (PWR) yielded promising results in terms of its capability to accurately measure or estimate basic spent fuel assembly (SFA) characteristics, such as multiplication, initial enrichment (IE) and burn-up (BU) as well as the total plutonium content. These studies were however performed only for a subset of idealized SFAs with a symmetric BU with respect to its longitudinal axis. Therefore, to complement the previous results, additional simulations have been performed of the DDA instrument’s response to interrogation of asymmetricallymore » burned spent nuclear fuel in order to determine whether detailed assay of SFAs from all 4 sides will be necessary in real life applications or whether a cost and time saving single sided assay could be used to achieve results of similar quality as previously reported in case of symmetrically burned SFAs.« less

National Dam Inspection Program. Jennings Pond Dam (NDI I.D. PA-0891 DER I.D. 066-012) Susquehanna River Basin, Little Mehoopany Creek, Wyoming County, Pennsylvania. Phase I Inspection Report,

DTIC Science & Technology

1981-03-19

Area 7.9 square miles(1) b. Discharge at Dam Site ( cfs ) Maximum known flood at dam site Unknown Outlet conduit at maximum pool Unknown Gated spillway...700 cfs , based on the available 2.4-foot freeboard relative to the low spot on the left abutment. b. Experience Data. As previously stated, Jennings...in Appendix D. The inflow hydrograph for one-half PMF was found to have a peak flow of 6835 cfs . Computer input and summary of computer output are

Development of ECT/UT inspection system for bottom mounted instrumentation nozzle of PWR reactor vessels

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

Tanaka, H.; Fukui, S.; Iwahashi, Y.

1994-12-31

The development of inspection technique and tool for Bottom Mounted Instrument (BMI) nozzle of PWR plant was performed for countermeasure of leakage accident at incore instrument nozzle of Hamaoka-1 (BWR). MHI achieved the following development, of which object was PWR Plant R/V: (1) development of ECT/UT Multi-sensored Probe; (2) development of Inspection System (3) development of Data Processing System. The Inspection System had been functionally tested using full scale mock-up. As the result of the functional test, this system was confirmed to be very effective, and assumed to be hopeful for the actual application on site.

Development of a new bench for puncturing of irradiated fuel rods in STAR hot laboratory

NASA Astrophysics Data System (ADS)

Petitprez, B.; Silvestre, P.; Valenza, P.; Boulore, A.; David, T.

2018-01-01

A new device for puncturing of irradiated fuel rods in commercial power plants has been designed by Fuel Research Department of CEA Cadarache in order to provide experimental data of high precision on fuel pins with various designs. It will replace the current set-up that has been used since 1998 in hot cell 2 of STAR facility with more than 200 rod puncturing experiments. Based on this consistent experimental feedback, the heavy-duty technique of rod perforation by clad punching has been preserved for the new bench. The method of double expansion of rod gases is also retained since it allows upgrading the confidence interval of volumetric results obtained from rod puncturing. Furthermore, many evolutions have been introduced in the new design in order to improve its reliability, to make the maintenance easier by remote handling and to reduce experimental uncertainties. Tightness components have been studied with Sealing Laboratory Maestral at Pierrelatte so as to make them able to work under mixed pressure conditions (from vacuum at 10-5 mbar up to pressure at 50 bars) and to lengthen their lifetime under permanent gamma irradiation in hot cell. Bench ergonomics has been optimized to make its operating by remote handling easier and to secure the critical phases of a puncturing experiment. A high pressure gas line equipped with high precision pressure sensors out of cell can be connected to the bench in cell for calibration purposes. Uncertainty analyses using Monte Carlo calculations have been performed in order to optimize capacity of the different volumes of the apparatus according to volumetric characteristics of the rod to be punctured. At last this device is composed of independent modules which allow puncturing fuel pins out of different geometries (PWR, BWR, VVER). After leak tests of the device and remote handling simulation in a mock-up cell, several punctures of calibrated specimens have been performed in 2016. The bench will be implemented soon in hot

A preliminary evaluation of the ability of from-reactor casks to geometrically accommodate commercial LWR spent nuclear fuel

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

Andress, D.; Joy, D.S.; McLeod, N.B.

The Department of Energy has sponsored a number of cask design efforts to define several transportation casks to accommodate the various assemblies expected to be accepted by the Federal Waste Management System. At this time, three preliminary cask designs have been selected for the final design--the GA-4 and GA-9 truck casks and the BR-100 rail cask. In total, this assessment indicates that the current Initiative I cask designs can be expected to dimensionally accommodate 100% of the PWR fuel assemblies (other than the extra-long South Texas Fuel) with control elements removed, and >90% of the assemblies having the control elementsmore » as an integral part of the fuel assembly. For BWR assemblies, >99% of the assemblies can be accommodated with fuel channels removed. This paper summarizes preliminary results of one part of that evaluation related to the ability of the From-Reactor Initiative I casks to accommodate the physical and radiological characteristics of the Spent Nuclear Fuel projected to be accepted into the Federal Waste Management System. 3 refs., 5 tabs.« less

U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

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

Hu, Jianwei; Peterson, Joshua L.; Gauld, Ian C.

2016-09-01

Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decaymore » heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.« less

Vitamin D Counteracts an IL-23-Dependent IL-17A+IFN-γ+ Response Driven by Urban Particulate Matter.

PubMed

Mann, Elizabeth H; Ho, Tzer-Ren; Pfeffer, Paul E; Matthews, Nick C; Chevretton, Elfy; Mudway, Ian; Kelly, Frank J; Hawrylowicz, Catherine M

2017-09-01

Urban particulate matter (UPM) air pollution and vitamin D deficiency are detrimentally associated with respiratory health. This is hypothesized to be due in part to regulation of IL-17A, which UPM is reported to promote. Here, we used a myeloid dendritic cell (DC)-memory CD4 + T cell co-culture system to characterize UPM-driven IL-17A + cells, investigate the mechanism by which UPM-primed DCs promote this phenotype, and address evidence for cross-regulation by vitamin D. CD1c + myeloid DCs were cultured overnight with or without a reference source of UPM and/or active vitamin D (1,25[OH] 2 D 3 ) before they were co-cultured with autologous memory CD4 + T cells. Supernatants were harvested for cytokine analysis on Day 5 of co-culture, and intracellular cytokine staining was performed on Day 7. UPM-primed DCs increased the proportion of memory CD4 + T cells expressing the T helper 17 cell (Th17)-associated cytokines IL-17A, IL-17F, and IL-22, as well as IFN-γ, granulocyte-macrophage colony-stimulating factor, and granzyme B. Notably, a large proportion of the UPM-driven IL-17A + cells co-expressed these cytokines, but not IL-10, indicative of a proinflammatory Th17 profile. UPM-treated DCs expressed elevated levels of il23 mRNA and increased secretion of IL-23p40. Neutralization of IL-23 in culture reduced the frequency of IL-17A + IFN-γ + cells without affecting cell proliferation. 1,25(OH) 2 D 3 counteracted the UPM-driven DC maturation and inhibited the frequency of IL-17A + IFN-γ + cells, most prominently when DCs were co-treated with the corticosteroid dexamethasone, while maintaining antiinflammatory IL-10 synthesis. These data indicate that UPM might promote an inflammatory milieu in part by inducing an IL-23-driven proinflammatory Th17 response. Restoring vitamin D sufficiency may counteract these UPM-driven effects without obliterating important homeostatic immune functions.

Impact of nuclear data uncertainty on safety calculations for spent nuclear fuel geological disposal

NASA Astrophysics Data System (ADS)

Herrero, J. J.; Rochman, D.; Leray, O.; Vasiliev, A.; Pecchia, M.; Ferroukhi, H.; Caruso, S.

2017-09-01

In the design of a spent nuclear fuel disposal system, one necessary condition is to show that the configuration remains subcritical at time of emplacement but also during long periods covering up to 1,000,000 years. In the context of criticality safety applying burn-up credit, k-eff eigenvalue calculations are affected by nuclear data uncertainty mainly in the burnup calculations simulating reactor operation and in the criticality calculation for the disposal canister loaded with the spent fuel assemblies. The impact of nuclear data uncertainty should be included in the k-eff value estimation to enforce safety. Estimations of the uncertainty in the discharge compositions from the CASMO5 burn-up calculation phase are employed in the final MCNP6 criticality computations for the intact canister configuration; in between, SERPENT2 is employed to get the spent fuel composition along the decay periods. In this paper, nuclear data uncertainty was propagated by Monte Carlo sampling in the burn-up, decay and criticality calculation phases and representative values for fuel operated in a Swiss PWR plant will be presented as an estimation of its impact.

Alternate Fuel Cycle Technologies/Thorium Fuel Cycle Technology Programs. Quarterly report for period 1 April--30 June 1978

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

Vondra, B.L.

1978-08-01

Voloxidation and dissolution studies: rotary-kiln heat-transfer tests are under way using a small rotary kiln along with the development of a mathematical model to determine kiln-heat-flux profiles necessary to maintain a desired temperature gradient. The erosion/corrosion test for evaluating materials of construction is operational. Fuel from a BWR (Big Rock Point) yielded more fine solid residue on dissolution than in previous tests with PWR fuel. Two additional parametric voloxidation tests with H.B. Robinson fuel compared air vs pure oxygen atmospheres at 550{sup 0}C; overall tritium release and subsequent fuel dissolution were equivalent. Thorium dissolution studies: the dissolution rate of thoriamore » in fluoride-catalyzed 8 to 14 M HNO{sub 3} (100{sup 0}C) was max between 0.04 to 0.06 M HF; at higher fluoride concentrations, ThF{sub 4}.5H{sub 2}O precipitated. The rate of zircaloy dissolution continued to increase with increasing fluoride concentration. Stainless-steel-clad (Th,U)0{sub 2} fuel rods irradiated in the NRX reactor were sheared, voloxidized, and dissolved. {le}10% of the tritium was released during voloxidation in air at 600{sup 0}C. Carbon-14 removal from off-gas and fixation: carbon dioxide removal with Linde 13X molecular sieves to less than 100 ppB was experimentally verified using 300 ppM CO in air. Decontamination factors from 3000 to 7500 were obtained for CO{sub 2} removal in the gas-slurry stirred-tank reactor with CA(OH){sub 2}.or Ba(0H){sub 2}/sup .8H2O./. With Ba(OH){sub 2}.H{sub 2}0{sup 2} in a fixed-bed column, decontamination factors of about 30,000 were obtained.« less

Zebra: An advanced PWR lattice code

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

Cao, L.; Wu, H.; Zheng, Y.

2012-07-01

This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precisionmore » and a high efficiency. (authors)« less

Reactor physics behavior of transuranic-bearing TRISO-particle fuel in a pressurized water reactor

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

Pope, M. A.; Sen, R. S.; Ougouag, A. M.

2012-07-01

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) - only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space availablemore » for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is retained. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint. (authors)« less

Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

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

Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag

2012-04-01

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel,more » the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.« less

High-temperature Gas Reactor (HTGR)

NASA Astrophysics Data System (ADS)

Abedi, Sajad

2011-05-01

General Atomics (GA) has over 35 years experience in prismatic block High-temperature Gas Reactor (HTGR) technology design. During this period, the design has recently involved into a modular have been performed to demonstrate its versatility. This versatility is directly related to refractory TRISO coated - particle fuel that can contain any type of fuel. This paper summarized GA's fuel cycle studies individually and compares each based upon its cycle sustainability, proliferation-resistance capabilities, and other performance data against pressurized water reactor (PWR) fuel cycle data. Fuel cycle studies LEU-NV;commercial HEU-Th;commercial LEU-Th;weapons-grade plutonium consumption; and burning of LWR waste including plutonium and minor actinides in the MHR. results show that all commercial MHR options, with the exception of HEU-TH, are more sustainable than a PWR fuel cycle. With LEU-NV being the most sustainable commercial options. In addition, all commercial MHR options out perform the PWR with regards to its proliferation-resistance, with thorium fuel cycle having the best proliferation-resistance characteristics.

PREFACE: International Symposium on Ultrasound in the Control of Industrial Processes (UCIP 2012)

NASA Astrophysics Data System (ADS)

Segura, Luis Elvira; Resa López, Pablo; Salazar, Jordi; Benedito Fort, José Javier; Martínez Graullera, Óscar

2012-12-01

The following describes most of the presentations (both oral and poster) given at the International Symposium of Ultrasound in the Control of Industrial Processes (UCIP 2012) celebrated in Madrid between 18 and 20 April 2012. This event was intended to be a meeting point for scientists, engineers and professionals from all over the world in the field of ultrasonics applied to the characterization and control of materials and processes in the industry. More precisely, the topics included were: 1. Novel applications of ultrasound in the industry (including high-power ultrasound) Food science Biotechnology and microbiology Pharmaceutics and cosmetics Petrochemistry and civil engineering 2. New insights in the ultrasonic characterization of media: Fluids and emulsions Nano- and micro-particle dispersions Soft materials Porous bodies and inhomogeneous materials 3. New developments in ultrasonic measuring techniques: Acoustic microscopy Piezoelectric sensors Ultrasonic imaging Signal processing The symposium was organized by the Centro de Acústica Aplicada y Evaluación No Destructiva (CAEND, UPM-CSIC) in collaboration with the Universidad Politécnica de Cataluña, the Universidad Politécnica de Valencia and the University of Leeds. During the conference, 32 posters and 33 oral communications were presented. In addition, 4 invited lectures were imparted: 'Acoustic microscopy, spectroscopy and nanoparticle detection' by Dr Malcolm Povey; 'Acoustic and electroacoustic spectroscopy' by Dr Andrei Dukhin; 'High-Resolution Ultrasonic Spectroscopy and its application for material analysis by Dr Vitaly Buckin; 'Ultrasonic sensors for process applications - state of the art' by Dr Bern Henning; and three tutorials were given: 'PZFlex - Finite Element Analysis for Virtual Prototyping' by Weidlinger Associates; 'SITAU - A flexible architecture controlled by MATLAB for the development of ultrasonic applications' by DASEL; 'Ultra-SCATTERERTM (Acoustics Suite) - The R&D Tool for

Effects of Thermo-Mechanical Treatments on Deformation Behavior and IGSCC Susceptibility of Stainless Steels in Pwr Primary Water Chemistry

NASA Astrophysics Data System (ADS)

Nouraei, S.; Tice, D. R.; Mottershead, K. J.; Wright, D. M.

Field experience of 300 series stainless steels in the primary circuit of PWR plant has been good. Stress Corrosion Cracking of components has been infrequent and mainly associated with contamination by impurities/oxygen in occluded locations. However, some instances of failures have been observed which cannot necessarily be attributed to deviations in the water chemistry. These failures appear to be associated with the presence of cold-work produced by surface finishing and/or by welding-induced shrinkage. Recent data indicate that some heats of SS show an increased susceptibility to SCC; relatively high crack growth rates were observed even when the crack growth direction is orthogonal to the cold-work direction. SCC of cold-worked SS in PWR coolant is therefore determined by a complex interaction of material composition, microstructure, prior cold-work and heat treatment. This paper will focus on the interactions between these parameters on crack propagation in simulated PWR conditions.

Development of Neutron Energy Spectral Signatures for Passive Monitoring of Spent Nuclear Fuels in Dry Cask Storage

NASA Astrophysics Data System (ADS)

Harkness, Ira; Zhu, Ting; Liang, Yinong; Rauch, Eric; Enqvist, Andreas; Jordan, Kelly A.

2018-01-01

Demand for spent nuclear fuel dry casks as an interim storage solution has increased globally and the IAEA has expressed a need for robust safeguards and verification technologies for ensuring the continuity of knowledge and the integrity of radioactive materials inside spent fuel casks. Existing research has been focusing on "fingerprinting" casks based on count rate statistics to represent radiation emission signatures. The current research aims to expand to include neutron energy spectral information as part of the fuel characteristics. First, spent fuel composition data are taken from the Next Generation Safeguards Initiative Spent Fuel Libraries, representative for Westinghouse 17ˣ17 PWR assemblies. The ORIGEN-S code then calculates the spontaneous fission and (α,n) emissions for individual fuel rods, followed by detailed MCNP simulations of neutrons transported through the fuel assemblies. A comprehensive database of neutron energy spectral profiles is to be constructed, with different enrichment, burn-up, and cooling time conditions. The end goal is to utilize the computational spent fuel library, predictive algorithm, and a pressurized 4He scintillator to verify the spent fuel assemblies inside a cask. This work identifies neutron spectral signatures that correlate with the cooling time of spent fuel. Both the total and relative contributions from spontaneous fission and (α,n) change noticeably with respect to cooling time, due to the relatively short half-life (18 years) of the major neutron source 244Cm. Identification of this and other neutron spectral signatures allows the characterization of spent nuclear fuels in dry cask storage.

Regeneratively Cooled Liquid Oxygen/Methane Technology Development Between NASA MSFC and PWR

NASA Technical Reports Server (NTRS)

Robinson, Joel W.; Greene, Christopher B.; Stout, Jeffrey B.

2012-01-01

The National Aeronautics & Space Administration (NASA) has identified Liquid Oxygen (LOX)/Liquid Methane (LCH4) as a potential propellant combination for future space vehicles based upon exploration studies. The technology is estimated to have higher performance and lower overall systems mass compared to existing hypergolic propulsion systems. NASA-Marshall Space Flight Center (MSFC) in concert with industry partner Pratt & Whitney Rocketdyne (PWR) utilized a Space Act Agreement to test an oxygen/methane engine system in the Summer of 2010. PWR provided a 5,500 lbf (24,465 N) LOX/LCH4 regenerative cycle engine to demonstrate advanced thrust chamber assembly hardware and to evaluate the performance characteristics of the system. The chamber designs offered alternatives to traditional regenerative engine designs with improvements in cost and/or performance. MSFC provided the test stand, consumables and test personnel. The hot fire testing explored the effective cooling of one of the thrust chamber designs along with determining the combustion efficiency with variations of pressure and mixture ratio. The paper will summarize the status of these efforts.

Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

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

Bardet, Philippe; Ricciardi, Guillaume

2016-01-31

Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging tasks in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWRmore » fuel bundle behavior during seismic transients.« less

Sub-Scale Testing and Development of the J-2X Fuel Turbopump Inducer

NASA Technical Reports Server (NTRS)

Sargent, Scott R.; Becht, David G.

2011-01-01

In the early stages of the J-2X upper stage engine program, various inducer configurations proposed for use in the fuel turbopump (FTP) were tested in water. The primary objectives of this test effort were twofold. First, to obtain a more comprehensive data set than that which existed in the Pratt & Whitney Rocketdyne (PWR) historical archives from the original J-2S program, and second, to supplement that data set with information regarding the cavitation induced vibrations for both the historical J-2S configuration as well as those tested for the J-2X program. The J-2X FTP inducer, which actually consists of an inducer stage mechanically attached to a kicker stage, underwent 4 primary iterations utilizing sub-scaled test articles manufactured and tested in PWR's Engineering Development Laboratory (EDL). The kicker remained unchanged throughout the test series. The four inducer configurations tested retained many of the basic design features of the J-2S inducer, but also included variations on leading edge blade thickness and blade angle distribution, primarily aimed at improving suction performance at higher flow coefficients. From these data sets, the effects of the tested design variables on hydrodynamic performance and cavitation instabilities were discerned. A limited comparison of impact to the inducer efficiency was determined as well.

Preliminary Analysis of High-Flux RSG-GAS to Transmute Am-241 of PWR’s Spent Fuel in Asian Region

NASA Astrophysics Data System (ADS)

Budi Setiawan, M.; Kuntjoro, S.

2018-02-01

A preliminary study of minor actinides (MA) transmutation in the high flux profile RSG-GAS research reactor was performed, aiming at an optimal transmutation loading for present nuclear energy development. The MA selected in the analysis includes Am-241 discharged from pressurized water reactors (PWRs) in Asian region. Until recently, studies have been undertaken in various methods to reduce radiotoxicity from actinides in high-level waste. From the cell calculation using computer code SRAC2006, it is obtained that the target Am-241 which has a cross section of the thermal energy absorption in the region (group 8) is relatively large; it will be easily burned in the RSG-GAS reactor. Minor actinides of Am-241 which can be inserted in the fuel (B/T fuel) is 2.5 kg which is equivalent to Am-241 resulted from the partition of spent fuel from 2 units power reactors PWR with power 1000MW(th) operated for one year.

Graphical and statistical techniques for cardiac cycle time (phase) dependent changes in interbeat interval: problems with the Jennings et al. (1991) proposals.

PubMed

Barry, R J

1993-01-01

Two apparently new effects in human cardiac responding, "primary bradycardia" and "vagal inhibition", were first described by the Laceys. These effects have been considered by some researchers to reflect differential cardiac innervation, analogous to similar effects observed in animal preparations with direct vagal stimulation. However, it has been argued that such effects arise merely from the data-analytic techniques introduced by the Laceys, and hence are not genuine cardiac cycle effects. Jennings, van der Molen, Somsen and Ridderinkhoff (Psychophysiology, 28 (1991) 596-606) recently proposed a plotting technique and statistical procedure in an attempt to resolve this issue. The present paper demonstrates that the plotting technique fails to achieve their stated aim, since it identifies data from identical cardiac responses as showing cardiac-cycle effects. In addition, the statistical procedure is shown to be reducible to a trivial test of response occurrence. The implication of these demonstrations, in the context of other work, is that this area of investigation has reached a dead end.

Severe accident modeling of a PWR core with different cladding materials

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

Johnson, S. C.; Henry, R. E.; Paik, C. Y.

2012-07-01

The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCSmore » rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)« less

J-2X Fuel Turbopump Point of Departure: The Performance of the J-2s Fuel Turbopump Inducer

NASA Technical Reports Server (NTRS)

Sargent, S. R.; Becht, D. G.; Mulder, A. D.

2008-01-01

To aid the J-2X program design effort with detailed performance and environment information, the J-2S fuel turbopump (FTP) inducer has undergone a thorough test series in both water and hydrogen. The test series utilizes both inducer only and a complete pump configuration to assess the inducer interaction to the overall turbopump system. The test goals include verification of suction performance against heritage J-2S data, head production, effects of thermodynamic suppression head (TSH), and evaluation of the inducer dynamic pressure caused by cavitation instabilities. Test facilities at both Pratt & Whitney Rocketdyne (PWR) and NASA s Stennis Space Center (SSC) are employed for the testing. The inducer only water test effort conducted at PWR established performance curves for suction performance, head production, and efficiency over a wide operating range. Because the heritage J-2S suction performance data set is in hydrogen, it is desired to obtain current suction performance data in hydrogen as well, thus avoiding the reliance on a theoretical TSH correlation for direct comparison. This data then provides an empirically based TSH correlation allowing for the comparison of water test suction data to system suction requirements. The FTP testing performed at SSC provides these suction performance relationships as well as inlet duct dynamic pressures during liquid hydrogen operation. The test effort successfully confirms the heritage J-2S suction performance and establishes the amount of TSH between water and hydrogen operation at the design flow coefficient. Correlating data is also obtained for cavitating instability frequency content, illustrating the validity of using the wide flow range water test data to predict hydrogen performance.

Performance testing and analyses of the VSC-17 ventilated concrete cask. Final report

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

McKinnon, M.A.; Dodge, R.E.; Schmitt, R.C.

1992-05-01

This document details performance test which was conducted on a Pacific Sierra Nuclear VSC-17 ventilated concrete storage cask configured for pressurized-water reactor (PWR) spent fuel. The performance test consisted of loading the VSC-17 cask with 17 canisters of consolidated PWR spent fuel from Virginia Power`s Surry and Florida Power & Light Turkey Point reactors. Cask surface, concrete, air channel surfaces, and fuel canister guide tube temperatures were measured, as were cask surface gamma and neutron dose rates. Testing was performed with vacuum, nitrogen, and helium backfill environments in a vertical cask orientation. Data on spent fuel integrity were also obtained.

Rail-Cask Tests: Normal-Conditionsof- Transport Tests of Surrogate PWR Fuel Assemblies in an ENSA ENUN 32P Cask.

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

McConnell, Paul E.; Ross, Steven; Grey, Carissa Ann

This report describes tests conducted using a full-size rail cask, the ENSA ENUN 32P, involving handling of the cask and transport of the cask via truck, ships, and rail. The purpose of the tests was to measure strains and accelerations on surrogate pressurized water reactor fuel rods when the fuel assemblies were subjected to Normal Conditions of Transport within the rail cask. In addition, accelerations were measured on the transport platform, the cask cradle, the cask, and the basket within the cask holding the assemblies. These tests were an international collaboration that included Equipos Nucleares S.A., Sandia National Laboratories, Pacificmore » Northwest National Laboratory, Coordinadora Internacional de Cargas S.A., the Transportation Technology Center, Inc., the Korea Radioactive Waste Agency, and the Korea Atomic Energy Research Institute. All test results in this report are PRELIMINARY – complete analyses of test data will be completed and reported in FY18. However, preliminarily: The strains were exceedingly low on the surrogate fuel rods during the rail-cask tests for all the transport and handling modes. The test results provide a compelling technical basis for the safe transport of spent fuel.« less

In-field Calibration of a Fast Neutron Collar for the Measurement of Fresh PWR Fuel Assemblies

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

Swinhoe, Martyn Thomas; De Baere, Paul

2015-04-17

A new neutron collar has been designed for the measurement of fresh LEU fuel assemblies. This collar uses “fast mode” measurement to reduce the effect of burnable poison rods on the assay and thus reduce the dependence on the operator’s declaration. The new collar design reduces effect of poison rods considerably. Instead of 12 pins of 5.2% Gd causing a 20.4% effect, as in the standard thermal mode collar, they only cause a 3.2% effect in the new collar. However it has higher efficiency so that reasonably precise measurements can be made in 25 minutes, rather than the 1 hourmore » of previous collars. The new collar is fully compatible with the use of the standard data collection and analysis code INCC. This report describes the calibration that was made with a mock-up assembly at Los Alamos National Laboratory and with actual assemblies at the AREVA Fuel fabrication Plant in Lingen, Germany.« less

Fuel cladding behavior under rapid loading conditions

NASA Astrophysics Data System (ADS)

Yueh, K.; Karlsson, J.; Stjärnsäter, J.; Schrire, D.; Ledergerber, G.; Munoz-Reja, C.; Hallstadius, L.

2016-02-01

A modified burst test (MBT) was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident (RIA). A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile (BTD) transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor (PWR) cladding at both 296 and 553 K.

FUEL ASSEMBLY SHAKER TEST SIMULATION

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

Klymyshyn, Nicholas A.; Sanborn, Scott E.; Adkins, Harold E.

This report describes the modeling of a PWR fuel assembly under dynamic shock loading in support of the Sandia National Laboratories (SNL) shaker test campaign. The focus of the test campaign is on evaluating the response of used fuel to shock and vibration loads that a can occur during highway transport. Modeling began in 2012 using an LS-DYNA fuel assembly model that was first created for modeling impact scenarios. SNL’s proposed test scenario was simulated through analysis and the calculated results helped guide the instrumentation and other aspects of the testing. During FY 2013, the fuel assembly model was refinedmore » to better represent the test surrogate. Analysis of the proposed loads suggested the frequency band needed to be lowered to attempt to excite the lower natural frequencies of the fuel assembly. Despite SNL’s expansion of lower frequency components in their five shock realizations, pretest predictions suggested a very mild dynamic response to the test loading. After testing was completed, one specific shock case was modeled, using recorded accelerometer data to excite the model. Direct comparison of predicted strain in the cladding was made to the recorded strain gauge data. The magnitude of both sets of strain (calculated and recorded) are very low, compared to the expected yield strength of the Zircaloy-4 material. The model was accurate enough to predict that no yielding of the cladding was expected, but its precision at predicting micro strains is questionable. The SNL test data offers some opportunity for validation of the finite element model, but the specific loading conditions of the testing only excite the fuel assembly to respond in a limited manner. For example, the test accelerations were not strong enough to substantially drive the fuel assembly out of contact with the basket. Under this test scenario, the fuel assembly model does a reasonable job of approximating actual fuel assembly response, a claim that can be verified

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

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

Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

1978-02-01

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocitymore » and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.« less

Greenhouse gas (GHG) emission in organic farming. Approximate quantification of its generation at the organic garden of the School of Agricultural, Food and Biosystems Engineering (ETSIAAB) in the Technical University of Madrid (UPM)

NASA Astrophysics Data System (ADS)

Campos, Jorge; Barbado, Elena; Maldonado, Mariano; Andreu, Gemma; López de Fuentes, Pilar

2016-04-01

As it well-known, agricultural soil fertilization increases the rate of greenhouse gas (GHG) emission production such as CO2, CH4 and N2O. Participation share of this activity on the climate change is currently under study, as well as the mitigation possibilities. In this context, we considered that it would be interesting to know how this share is in the case of organic farming. In relation to this, a field experiment was carried out at the organic garden of the School of Agricultural, Food and Biosystems Engineering (ETSIAAB) in the Technical University of Madrid (UPM). The orchard included different management growing areas, corresponding to different schools of organic farming. Soil and gas samples were taken from these different sites. Gas samples were collected throughout the growing season from an accumulated atmosphere inside static chambers inserted into the soil. Then, these samples were carried to the laboratory and there analyzed. The results obtained allow knowing approximately how ecological fertilization contributes to air pollution due to greenhouse gases.

Primary water chemistry improvement for radiation exposure reduction at Japanese PWR Plants

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

Nishizawa, Eiichi

1995-03-01

Radiation exposure during the refueling outages at Japanese Pressurized Water Reactor (PWR) Plants has been gradually decreased through continuous efforts keeping the radiation dose rates at relatively low level. The improvement of primary water chemistry in respect to reduction of the radiation sources appears as one of the most important contributions to the achieved results and can be classified by the plant operation conditions as follows

Protective effects of chebulic acid on alveolar epithelial damage induced by urban particulate matter.

PubMed

Lee, Kyung-Won; Nam, Mi-Hyun; Lee, Hee-Ra; Hong, Chung-Oui; Lee, Kwang-Won

2017-07-19

Chebulic acid (CA) isolated from T. chebula, which has been reported for treating asthma, as a potent anti-oxidant resources. Exposure to ambient urban particulate matter (UPM) considered as a risk for cardiopulmonary vascular dysfunction. To investigate the protective effect of CA against UPM-mediated collapse of the pulmonary alveolar epithelial (PAE) cell (NCI-H441), barrier integrity parameters, and their elements were evaluated in PAE. CA was acquired from the laboratory previous reports. UPM was obtained from the National Institutes of Standards and Technology, and these were collected in St. Louis, MO, over a 24-month period and used as a standard reference. To confirm the protection of PAE barrier integrity, paracellular permeability and the junctional molecules were estimated with determination of transepithelial electrical resistance, Western Blotting, RT-PCR, and fluorescent staining. UPM aggravated the generation of reactive oxygen species (ROS) in PAE and also decreased mRNA and protein levels of junction molecules and barrier integrity in NCI-H441. However, CA repressed the ROS in PAE, also improved barrier integrity by protecting the junctional parameters in NCI-H411. These data showed that CA resulted in decreased UPM-induced ROS formation, and the protected the integrity of the tight junctions against UPM exposure to PAE barrier.

Parameter study on the influence of prepressurization on PWR fuel rod behavior during normal operation and hypothetical LOCAs

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

Brzoska, B.; Depisch, F.; Fuchs, H.P.

To analyze the influence of prepressurization on fuel rod behavior, a parametric study has been performed that considers the effects of as-fabricated fuel rod internal prepressure on the normal operation and postulated loss-of-coolant accident (LOCA) rod behavior of a 1300-MW(electric) Kraftwerk Union (KWU) standard pressurized water reactor nuclear power plant. A variation of the prepressure in the range from 15 to 35 bars has only a slight influence on normal operation behavior. Considering the LOCA behavior, only a small temperature increase results from prepressure reduction, while the core-wide straining behavior is improved significantly. The KWU prepressurization takes both conditions intomore » account.« less

Jens Esmark's Christiania (Oslo) meteorological observations 1816-1838: the first long-term continuous temperature record from the Norwegian capital homogenized and analysed

NASA Astrophysics Data System (ADS)

Hestmark, Geir; Nordli, Øyvind

2016-11-01

In 2010 we rediscovered the complete set of meteorological observation protocols made by Jens Esmark (1762-1839) during his years of residence in the Norwegian capital of Oslo (then Christiania). From 1 January 1816 to 25 January 1839, Esmark at his house in Øvre Voldgate in the morning, early afternoon and late evening recorded air temperature with state-of-the-art thermometers. He also noted air pressure, cloud cover, precipitation and wind directions, and experimented with rain gauges and hygrometers. From 1818 to the end of 1838 he twice a month provided weather tables to the official newspaper Den Norske Rigstidende, and thus acquired a semi-official status as the first Norwegian state meteorologist. This paper evaluates the quality of Esmark's temperature observations and presents new metadata, new homogenization and analysis of monthly means. Three significant shifts in the measurement series were detected, and suitable corrections are proposed. The air temperature in Oslo during this period is shown to exhibit a slow rise from 1816 towards 1825, followed by a slighter fall again towards 1838.

ADVANCED NUCLEAR FUEL CYCLE EFFECTS ON THE TREATMENT OF UNCERTAINTY IN THE LONG-TERM ASSESSMENT OF GEOLOGIC DISPOSAL SYSTEMS - EBS INPUT

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

Sutton, M; Blink, J A; Greenberg, H R

2012-04-25

The Used Fuel Disposition (UFD) Campaign within the Department of Energy's Office of Nuclear Energy (DOE-NE) Fuel Cycle Technology (FCT) program has been tasked with investigating the disposal of the nation's spent nuclear fuel (SNF) and high-level nuclear waste (HLW) for a range of potential waste forms and geologic environments. The planning, construction, and operation of a nuclear disposal facility is a long-term process that involves engineered barriers that are tailored to both the geologic environment and the waste forms being emplaced. The UFD Campaign is considering a range of fuel cycles that in turn produce a range of wastemore » forms. The UFD Campaign is also considering a range of geologic media. These ranges could be thought of as adding uncertainty to what the disposal facility design will ultimately be; however, it may be preferable to thinking about the ranges as adding flexibility to design of a disposal facility. For example, as the overall DOE-NE program and industrial actions result in the fuel cycles that will produce waste to be disposed, and the characteristics of those wastes become clear, the disposal program retains flexibility in both the choice of geologic environment and the specific repository design. Of course, other factors also play a major role, including local and State-level acceptance of the specific site that provides the geologic environment. In contrast, the Yucca Mountain Project (YMP) repository license application (LA) is based on waste forms from an open fuel cycle (PWR and BWR assemblies from an open fuel cycle). These waste forms were about 90% of the total waste, and they were the determining waste form in developing the engineered barrier system (EBS) design for the Yucca Mountain Repository design. About 10% of the repository capacity was reserved for waste from a full recycle fuel cycle in which some actinides were extracted for weapons use, and the remaining fission products and some minor actinides were

Preliminary design report: Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

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

None

1990-02-01

The purpose of this document is to provide information on burnup credit as applied to the preliminary design of the BR-100 shipping cask. There is a brief description of the preliminary basket design and the features used to maintain a critically safe system. Following the basket description is a discussion of various criticality analyses used to evaluate burnup credit. The results from these analyses are then reviewed in the perspective of fuel burnups expected to be shipped to either the final repository or a Monitored Retrievable Storage (MRS) facility. The hurdles to employing burnup credit in the certification of anymore » cask are then outlines and reviewed. the last section gives conclusions reached as to burnup credit for the BR-100 cask, based on our analyses and experience. All information in this study refers to the cask configured to transport PWR fuel. Boiling Water Reactor (BWR) fuel satisfies the criticality requirements so that burnup credit is not needed. All calculations generated in the preparation of this report were based upon the preliminary design which will be optimized during the final design. 8 refs., 19 figs., 16 tabs.« less

Effects of Ureaplasma parvum lipoprotein multiple-banded antigen on pregnancy outcome in mice.

PubMed

Uchida, Kaoru; Nakahira, Kumiko; Mimura, Kazuya; Shimizu, Takashi; De Seta, Francesco; Wakimoto, Tetsu; Kawai, Yasuhiro; Nomiyama, Makoto; Kuwano, Koichi; Guaschino, Secondo; Yanagihara, Itaru

2013-12-01

Ureaplasma spp. are members of the family Mycoplasmataceae and have been considered to be associated with chorioamnionitis and preterm delivery. However, it is unclear whether Ureaplasma spp. have virulence factors related to these manifestations. The purpose of the present study was to determine whether the immunogenic protein multiple-banded antigen (MBA) from Ureaplasma parvum is a virulence factor for preterm delivery. We partially purified MBA from a type strain and clinical isolates of U. parvum, and also synthesized a diacylated lipopeptide derived from U. parvum, UPM-1. Using luciferase assays, both MBA-rich fraction MRF and UPM-1 activated the NF-κB pathway via TLR2. UPM-1 upregulated IL-1β, IL-6, IL-12p35, TNF-α, MIP2, LIX, and iNOS in mouse peritoneal macrophage. MRF or UPM-1 was injected into uteri on day 15 of gestation on pregnant C3H/HeN mice. The intrauterine MRF injection group had a significantly higher incidence of intrauterine fetal death (IUFD; 38.5%) than the control group (14.0%). Interestingly, intrauterine injection of UPM-1 caused preterm deliveries at high concentration (80.0%). In contrast, a low concentration of UPM-1 induced a significantly higher rate of fetal deaths (55.2%) than the control group (14.0%). The placentas of the UPM-1 injection group showed neutrophil infiltration and increased iNOS protein expression. Our data indicate that MBA from the clinical isolate of U. parvum is a potential virulence factor for IUFD and preterm delivery in mice and that the N-terminal diacylated lipopeptide is essential for the initiation of inflammation. Copyright © 2013 Elsevier Ireland Ltd. All rights reserved.

Estimating probable flaw distributions in PWR steam generator tubes

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

Gorman, J.A.; Turner, A.P.L.

1997-02-01

This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regardingmore » uncertainties and assumptions in the data and analyses.« less

Carboxylic acids permeases in yeast: two genes in Kluyveromyces lactis.

PubMed

Lodi, Tiziana; Fontanesi, Flavia; Ferrero, Iliana; Donnini, Claudia

2004-09-15

Two new genes KlJEN1 and KlJEN2 were identified in Kluyveromyces lactis. The deduced structure of their products is typical of membrane-bound carriers and displays high similarity to Jen1p, the monocarboxylate permease of Saccharomyces cerevisiae. Both KlJEN1 and KlJEN2 are under the control of glucose repression mediated by FOG1 and FOG2, corresponding to S. cerevisiae GAL83 and SNF1 respectively, and KlCAT8, proteins involved in glucose signalling cascade in K. lactis. KlJEN1, but not KlJEN2, is induced by lactate. KlJEN2 in contrast is expressed at high level in ethanol and succinate. The physiological characterization of null mutants showed that KlJEN1 is the functional homologue of ScJEN1, whereas KlJEN2 encodes a dicarboxylic acids transporter. In fact, KlJen1p [transporter classification (TC) number: 2.A.1.12.2.] is required for lactate uptake and therefore for growth on lactate. KlJen2p is required for succinate transport, as demonstrated by succinate uptake experiments and by inability of Kljen2 mutant to grow on succinate. This carrier appears to transport also malate and fumarate because the Kljen2 mutant cannot grow on these substrates and the succinate uptake is competed by these carboxylic acids. We conclude that KlJEN2 is the first yeast gene shown to encode a dicarboxylic acids permease.

Consistent criticality and radiation studies of Swiss spent nuclear fuel: The CS2M approach.

PubMed

Rochman, D; Vasiliev, A; Ferroukhi, H; Pecchia, M

2018-06-15

In this paper, a new method is proposed to systematically calculate at the same time canister loading curves and radiation sources, based on the inventory information from an in-core fuel management system. As a demonstration, the isotopic contents of the assemblies come from a Swiss PWR, considering more than 6000 cases from 34 reactor cycles. The CS 2 M approach consists in combining four codes: CASMO and SIMULATE to extract the assembly characteristics (based on validated models), the SNF code for source emission and MCNP for criticality calculations for specific canister loadings. The considered cases cover enrichments from 1.9 to 5.0% for the UO 2 assemblies and 4.8% for the MOX, with assembly burnup values from 7 to 74 MWd/kgU. Because such a study is based on the individual fuel assembly history, it opens the possibility to optimize canister loadings from the point-of-view of criticality, decay heat and emission sources. Copyright © 2018 Elsevier B.V. All rights reserved.

Nuclear Data Uncertainties for Typical LWR Fuel Assemblies and a Simple Reactor Core

NASA Astrophysics Data System (ADS)

Rochman, D.; Leray, O.; Hursin, M.; Ferroukhi, H.; Vasiliev, A.; Aures, A.; Bostelmann, F.; Zwermann, W.; Cabellos, O.; Diez, C. J.; Dyrda, J.; Garcia-Herranz, N.; Castro, E.; van der Marck, S.; Sjöstrand, H.; Hernandez, A.; Fleming, M.; Sublet, J.-Ch.; Fiorito, L.

2017-01-01

The impact of the current nuclear data library covariances such as in ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, SCALE and TENDL, for relevant current reactors is presented in this work. The uncertainties due to nuclear data are calculated for existing PWR and BWR fuel assemblies (with burn-up up to 40 GWd/tHM, followed by 10 years of cooling time) and for a simplified PWR full core model (without burn-up) for quantities such as k∞, macroscopic cross sections, pin power or isotope inventory. In this work, the method of propagation of uncertainties is based on random sampling of nuclear data, either from covariance files or directly from basic parameters. Additionally, possible biases on calculated quantities are investigated such as the self-shielding treatment. Different calculation schemes are used, based on CASMO, SCALE, DRAGON, MCNP or FISPACT-II, thus simulating real-life assignments for technical-support organizations. The outcome of such a study is a comparison of uncertainties with two consequences. One: although this study is not expected to lead to similar results between the involved calculation schemes, it provides an insight on what can happen when calculating uncertainties and allows to give some perspectives on the range of validity on these uncertainties. Two: it allows to dress a picture of the state of the knowledge as of today, using existing nuclear data library covariances and current methods.

Annual progress report on the NSRR experiments, (21)

NASA Astrophysics Data System (ADS)

1992-05-01

Fuel behavior studies under simulated reactivity-initiated accident (RIA) conditions have been performed in the Nuclear Safety Research Reactor (NSRR) since 1975. This report gives the results of experiments performed from April, 1989 through March, 1990 and discussions of them. A total of 41 tests were carried out during this period. The tests are distinguished into pre-irradiated fuel tests and fresh fuel tests; the former includes 2 JMTR pre-irradiated fuel tests, 2 PWR pre-irradiated fuel tests, and 2 BWR pre-irradiated fuel tests, and the latter includes 6 standard fuel tests (6 SP(center dot)CP scoping tests), 7 power and cooling condition parameter tests (4 flow shrouded fuel tests, 1 bundle simulation test, 1 fully water-filled vessel test, 1 high pressure/high temperature loop test), 12 special fuel tests (3 stainless steel clad fuel tests, 3 improved PWR fuel tests, 6 improved BWR fuel tests), 3 severe fuel damage tests (1 high temperature flooding test, 1 flooding behavior observation test, 1 debris coolability test), 3 fast breeder reactor fuel tests (2 moderator material characteristic measurement tests, 1 fuel behavior observation test), and 2 miscellaneous tests (2 preliminary tests for pre-irradiated fuel tests).

Uncertainties for Swiss LWR spent nuclear fuels due to nuclear data

NASA Astrophysics Data System (ADS)

Rochman, Dimitri A.; Vasiliev, Alexander; Dokhane, Abdelhamid; Ferroukhi, Hakim

2018-05-01

This paper presents a study of the impact of the nuclear data (cross sections, neutron emission and spectra) on different quantities for spent nuclear fuels (SNF) from Swiss power plants: activities, decay heat, neutron and gamma sources and isotopic vectors. Realistic irradiation histories are considered using validated core follow-up models based on CASMO and SIMULATE. Two Pressurized and one Boiling Water Reactors (PWR and BWR) are considered over a large number of operated cycles. All the assemblies at the end of the cycles are studied, being reloaded or finally discharged, allowing spanning over a large range of exposure (from 4 to 60 MWd/kgU for ≃9200 assembly-cycles). Both UO2 and MOX fuels were used during the reactor cycles, with enrichments from 1.9 to 4.7% for the UO2 and 2.2 to 5.8% Pu for the MOX. The SNF characteristics presented in this paper are calculated with the SNF code. The calculated uncertainties, based on the ENDF/B-VII.1 library are obtained using a simple Monte Carlo sampling method. It is demonstrated that the impact of nuclear data is relatively important (e.g. up to 17% for the decay heat), showing the necessity to consider them for safety analysis of the SNF handling and disposal.

Design of a proteus lattice representative of a burnt and fresh fuel interface at power conditions in light water reactors

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

Hursin, M.; Perret, G.

The research program LIFE (Large-scale Irradiated Fuel Experiment) between PSI and Swissnuclear has been started in 2006 to study the interaction between large sets of burnt and fresh fuel pins in conditions representative of power light water reactors. Reactor physics parameters such as flux ratios and reaction rate distributions ({sup 235}U and {sup 238}U fissions and {sup 238}U capture) are calculated to estimate an appropriate arrangement of burnt and fresh fuel pins within the central element of the test zone of the zero-power research reactor PROTEUS. The arrangement should minimize the number of burnt fuel pins to ease fuel handlingmore » and reduce costs, whilst guaranteeing that the neutron spectrum in both burnt and fresh fuel regions and at their interface is representative of a large uniform array of burnt and fresh pins in the same moderation conditions. First results are encouraging, showing that the burnt/fresh fuel interface is well represented with a 6 x 6 bundle of burnt pins. The second part of the project involves the use of TSUNAMI, CASMO-4E and DAKOTA to perform parametric and optimization studies on the PROTEUS lattice by varying its pitch (P) and fraction of D{sub 2}O in moderator (F{sub D2O}) to be as representative as possible of a power light water reactor core at hot full power conditions at beginning of cycle (BOC). The parameters P and F{sub D2O} that best represent a PWR at BOC are 1.36 cm and 5% respectively. (authors)« less

Space and Time Distribution of Pu Isotopes inside The First Experimental Fuel Pin Designed for PWR and Manufactured in Indonesia

NASA Astrophysics Data System (ADS)

Suwardi; Setiawan, J.; Susilo, J.

2017-01-01

The first short fuel pin containing natural UO2 pellet in Zry4 cladding has been prepared and planned to be tested in power ramp irradiation. An irradiation test should be designed to allow an experiment can be performed safely and giving maximum results of many performance aspects of design and manufacturing. Performance analysis to the fuel specimen shows that the specimen is not match to be used for power ramp testing. Enlargement by 0.20 mm of pellet diameter has been proposed. The present work is evaluation of modified design for important aspect of isotopic Pu distribution during irradiation test, because generated Pu isotopes in natural UO2 fuel, contribute more power relative to the contribution by enriched UO2 fuel. The axial profile of neutrons flux have been chosen from both experimental measurement and model calculation. The parameters of ramp power has been obtained from statistical experiment data. A simplified and typical base-load commercial PHWR profile of LHR history has been chosen, to determine the minimum irradiation time before ramp test can be performed. The data design and Mat pro XI materials properties models have been chosen. The axial profile of neutrons flux has been accommodated by 5 slices of discrete pin. The Pu distribution of slice-4 with highest power rate has been chosen to be evaluated. The radial discretion of pellet and cladding and numerical parameter have been used the default best practice of TU. The results shows that Pu 239 increased rapidly. The maximum burn up of slice 4 at upper the median slice, it reached nearly 90% of maximum value at about 6000 h with peak of 0.8%a Pu/HM at 22000 h, which is higher than initial U 235. Each 240, 241 and 240 Pu grows slower and ends up to 0.4, 0.2 and 0.18 % respectively. This results can be used for verification of other aspect of fuel behavior in the modeling results and also can be used as guide and comparison to the future post irradiation examination for Pu isotopes distribution.

New developments and prospects on COSI, the simulation software for fuel cycle analysis

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

Eschbach, R.; Meyer, M.; Coquelet-Pascal, C.

2013-07-01

COSI, software developed by the Nuclear Energy Direction of the CEA, is a code simulating a pool of nuclear power plants with its associated fuel cycle facilities. This code has been designed to study various short, medium and long term options for the introduction of various types of nuclear reactors and for the use of associated nuclear materials. In the frame of the French Act for waste management, scenario studies are carried out with COSI, to compare different options of evolution of the French reactor fleet and options of partitioning and transmutation of plutonium and minor actinides. Those studies aimmore » in particular at evaluating the sustainability of Sodium cooled Fast Reactors (SFR) deployment and the possibility to transmute minor actinides. The COSI6 version is a completely renewed software released in 2006. COSI6 is now coupled with the last version of CESAR (CESAR5.3 based on JEFF3.1.1 nuclear data) allowing the calculations on irradiated fuel with 200 fission products and 100 heavy nuclides. A new release is planned in 2013, including in particular the coupling with a recommended database of reactors. An exercise of validation of COSI6, carried out on the French PWR historic nuclear fleet, has been performed. During this exercise quantities like cumulative natural uranium consumption, or cumulative depleted uranium, or UOX/MOX spent fuel storage, or stocks of reprocessed uranium, or plutonium content in fresh MOX fuel, or the annual production of high level waste, have been computed by COSI6 and compared to industrial data. The results have allowed us to validate the essential phases of the fuel cycle computation, and reinforces the credibility of the results provided by the code.« less

SCC Initiation Behavior of Alloy 182 in PWR Primary Water

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

Toloczko, Mychailo B.; Zhai, Ziqing; Bruemmer, Stephen M.

SCC initiation behavior of 15% cold forged specimens cut from four different alloy 182 weldments was investigated in 360°C simulated PWR primary water under constant load at the yield stress using direct current potential drop to perform in-situ monitoring of SCC initiation time. Within each weldment, one or more specimens underwent SCC initiation within 24 hours of reaching full load while some specimens had much longer initiation times, in a few cases exceeding 2500 hours. Detailed examinations were conducted on these specimens with a focus on different microstructural features such as preexisting defects, grain orientation and second phases, highlighting anmore » important role of microstructure in crack initiation of alloy 182.« less

Modeling local chemistry in PWR steam generator crevices

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

Millett, P.J.

1997-02-01

Over the past two decades steam generator corrosion damage has been a major cost impact to PWR owners. Crevices and occluded regions create thermal-hydraulic conditions where aggressive impurities can become highly concentrated, promoting localized corrosion of the tubing and support structure materials. The type of corrosion varies depending on the local conditions, with stress corrosion cracking being the phenomenon of most current concern. A major goal of the EPRI research in this area has been to develop models of the concentration process and resulting crevice chemistry conditions. These models may then be used to predict crevice chemistry based on knowledgemore » of bulk chemistry, thereby allowing the operator to control corrosion damage. Rigorous deterministic models have not yet been developed; however, empirical approaches have shown promise and are reflected in current versions of the industry-developed secondary water chemistry guidelines.« less

Short-term exposure to particulate matter induces arterial but not venous thrombosis in healthy mice.

PubMed

Emmerechts, J; Alfaro-Moreno, E; Vanaudenaerde, B M; Nemery, B; Hoylaerts, M F

2010-12-01

Epidemiological findings suggest an association between exposure to particulate matter (PM) and venous thrombo-embolism.   To investigate arterial vs. venous thrombosis, inflammation and coagulation in mice, (sub)acutely exposed to two types of PM. Various doses (25, 100 and 200 μg per animal) of urban particulate matter (UPM) or diesel exhaust particles (DEP) were intratracheally (i.t.) instilled in C57Bl6/n mice and several endpoints measured at 4, 10 and 24 h. Mice were also repeatedly exposed to 100 μg per animal on three consecutive days with endpoints measured 24 h after the last instillation. Exposure to 200 μg per mouse UPM enhanced arterial thrombosis, but neither UPM nor DEP significantly enhanced venous thrombosis. Both types of PM induced dose-dependent increases in broncho-alveolar lavage fluid (BALF) total cell numbers (mainly neutrophils) and cytokines (IL-6, KC, MCP-1, RANTES, MIP-1α), with peaks at 4 h and overall higher values for UPM than for DEP. Systemic inflammation was limited to increased serum IL-6 levels, 4 h after UPM. Both types of PM induced similar and dose-dependent but modest increases in factor (F)VII, FVIII and fibrinogen. Three repeated instillations did not or only modestly enhance the proinflammatory and procoagulant status. Compared with DEP, UPM induced more pronounced pulmonary inflammation, but both particle types triggered similar and mild short-term systemic effects. Hence, acute exposure to PM triggers activation of primary hemostasis in the mouse, but no substantial secondary hemostasis activation, resulting in arterial but not venous thrombogenicity. © 2010 International Society on Thrombosis and Haemostasis.

Design, Construction and Testing of an In-Pile Loop for PWR (Pressurized Water Reactor) Simulation.

DTIC Science & Technology

1987-06-01

computer modeling remains at best semiempirical (C-i), this large variation in scaling factor makes extrapolation of data impossible. The DIDO Water...in a full scale PWR are not practical. The reactor plant is not controlled to tolerances necessary for research, and utilities are reluctant to vary...MIT Reactor Safeguards Committee, in revision 1 to the PCCL Safety Evaluation Report (SER), for final approval to begin in-pile testing and

Uncertainty analysis on reactivity and discharged inventory for a pressurized water reactor fuel assembly due to {sup 235,238}U nuclear data uncertainties

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

Da Cruz, D. F.; Rochman, D.; Koning, A. J.

2012-07-01

This paper discusses the uncertainty analysis on reactivity and inventory for a typical PWR fuel element as a result of uncertainties in {sup 235,238}U nuclear data. A typical Westinghouse 3-loop fuel assembly fuelled with UO{sub 2} fuel with 4.8% enrichment has been selected. The Total Monte-Carlo method has been applied using the deterministic transport code DRAGON. This code allows the generation of the few-groups nuclear data libraries by directly using data contained in the nuclear data evaluation files. The nuclear data used in this study is from the JEFF3.1 evaluation, and the nuclear data files for {sup 238}U and {supmore » 235}U (randomized for the generation of the various DRAGON libraries) are taken from the nuclear data library TENDL. The total uncertainty (obtained by randomizing all {sup 238}U and {sup 235}U nuclear data in the ENDF files) on the reactor parameters has been split into different components (different nuclear reaction channels). Results show that the TMC method in combination with a deterministic transport code constitutes a powerful tool for performing uncertainty and sensitivity analysis of reactor physics parameters. (authors)« less

PWR steam generator chemical cleaning, Phase I. Final report

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

Rothstein, S.

1978-07-01

United Nuclear Industries (UNI) entered into a subcontract with Consolidated Edison Company of New York (Con Ed) on August 8, 1977, for the purpose of developing methods to chemically clean the secondary side tube to tube support crevices of the steam generators of Indian Point Nos. 1 and 2 PWR plants. This document represents the first reporting on activities performed for Phase I of this effort. Specifically, this report contains the results of a literature search performed by UNI for the purpose of determining state-of-the-art chemical solvents and methods for decontaminating nuclear reactor steam generators. The results of the searchmore » sought to accomplish two objectives: (1) identify solvents beyond those proposed at present by UNI and Con Ed for the test program, and (2) confirm the appropriateness of solvents and methods of decontamination currently in use by UNI.« less

Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

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

Field, Kevin G.; Howard, Richard H.

FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiationmore » tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO 2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO 2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys, hence promoting

The J-2X Fuel Turbopump - Design, Development, and Test

NASA Technical Reports Server (NTRS)

Tellier, James G.; Hawkins, Lakiesha V.; Shinguchi, Brian H.; Marsh, Matthew W.

2011-01-01

Pratt and Whitney Rocketdyne (PWR), a NASA subcontractor, is executing the design, development, test, and evaluation (DDT&E) of a liquid oxygen, liquid hydrogen two hundred ninety four thousand pound thrust rocket engine initially intended for the Upper Stage (US) and Earth Departure Stage (EDS) of the Constellation Program Ares-I Crew Launch Vehicle (CLV). A key element of the design approach was to base the new J-2X engine on the heritage J-2S engine with the intent of uprating the engine and incorporating SSME and RS-68 lessons learned. The J-2S engine was a design upgrade of the flight proven J-2 configuration used to put American astronauts on the moon. The J-2S Fuel Turbopump (FTP) was the first Rocketdyne-designed liquid hydrogen centrifugal pump and provided many of the early lessons learned for the Space Shuttle Main Engine High Pressure Fuel Turbopumps. This paper will discuss the design trades and analyses performed for the current J-2X FTP to increase turbine life; increase structural margins, facilitate component fabrication; expedite turbopump assembly; and increase rotordynamic stability margins. Risk mitigation tests including inducer water tests, whirligig turbine blade tests, turbine air rig tests, and workhorse gas generator tests characterized operating environments, drove design modifications, or identified performance impact. Engineering design, fabrication, analysis, and assembly activities support FTP readiness for the first J-2X engine test scheduled for July 2011.

Processing and microstructural characterisation of a UO2-based ceramic for disposal studies on spent AGR fuel

NASA Astrophysics Data System (ADS)

Hiezl, Z.; Hambley, D. I.; Padovani, C.; Lee, W. E.

2015-01-01

Preparation and characterisation of a Simulated Spent Nuclear Fuel (SIMFuel), which replicates the chemical state and microstructure of Spent Nuclear Fuel (SNF) discharged from a UK Advanced Gas-cooled Reactor (AGR) after a cooling time of 100 years is described. Given the relatively small differences in radionuclide inventory expected over longer time periods, the SIMFuel studied in this work is expected to be also representative of spent fuel after significantly longer periods (e.g. 1000 years). Thirteen stable elements were added to depleted UO2 and sintered to simulate the composition of fuel pellets after burn-ups of 25 and 43 GWd/tU and, as a reference, pure UO2 pellets were also investigated. The fission product distribution was calculated using the FISPIN code provided by the UK National Nuclear Laboratory. SIMFuel pellets were up to 92% dense and during the sintering process in H2 atmosphere Mo-Ru-Rh-Pd metallic precipitates and grey-phase ((Ba, Sr)(Zr, RE) O3 oxide precipitates) formed within the UO2 matrix. These secondary phases are present in real PWR and AGR SNF. Metallic precipitates are generally spherical and have submicron particle size (0.8 ± 0.7 μm). Spherical oxide precipitates in SIMFuel measured up to 30 μm in diameter, but no data were available in the public domain to compare this to AGR SNF. The grain size of actual AGR SNF (∼ 3-30 μm) is larger than that measured in AGR SIMFuel (∼ 2-5 μm).

Differences in allergic inflammatory responses between urban PM2.5 and fine particle derived from desert-dust in murine lungs.

PubMed

He, Miao; Ichinose, Takamichi; Kobayashi, Makoto; Arashidani, Keiichi; Yoshida, Seiichi; Nishikawa, Masataka; Takano, Hirohisa; Sun, Guifan; Shibamoto, Takayuki

2016-04-15

The biological and chemical natures of materials adsorbed onto fine particulate matter (PM2.5) vary by origin and passage routes. The exacerbating effects of the two samples-urban PM2.5 (U-PM2.5) collected during the hazy weather in a Chinese city and fine particles (ASD-PM2.5) collected during Asian sand dust (ASD) storm event days in Japan-on murine lung eosinophilia were compared to clarify the role of toxic materials in PM2.5. The amounts of β-glucan and mineral components were higher in ASD-PM2.5 than in U-PM2.5. On the other hand, organic chemicals, including polycyclic aromatic hydrocarbons (PAHs), were higher in U-PM2.5 than in ASD-PM2.5. When BALB/c mice were intratracheally instilled with U-PM2.5 and ASD-PM2.5 (total 0.4 mg/mouse) with or without ovalbumin (OVA), various biological effects were observed, including enhancement of eosinophil recruitment induced by OVA in the submucosa of the airway, goblet cell proliferation in the bronchial epithelium, synergic increase of OVA-induced eosinophil-relevant cytokines and a chemokine in bronchoalveolar lavage fluid, and increase of serum OVA-specific IgG1 and IgE. Data demonstrate that U-PM2.5 and ASD-PM2.5 induced allergic inflammatory changes and caused lung pathology. U-PM2.5 and ASD-PM2.5 increased F4/80(+) CD11b(+) cells, indicating that an influx of inflammatory and exudative macrophages in lung tissue had occurred. The ratio of CD206 positive F4/80(+) CD11b(+) cells (M2 macrophages) in lung tissue was higher in the OVA+ASD-PM2.5 treated mice than in the OVA+U-PM2.5 treated mice. These results suggest that the lung eosinophilia exacerbated by both PM2.5 is due to activation of a Th2-associated immune response along with induced M2 macrophages and the exacerbating effect is greater in microbial element (β-glucan)-rich ASD-PM2.5 than in organic chemical-rich U-PM2.5. Copyright © 2016 Elsevier Inc. All rights reserved.

Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

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

Zhang, Hongbin; Szilard, Ronaldo; Epiney, Aaron

Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance duringmore » LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.« less

Differences in allergic inflammatory responses between urban PM2.5 and fine particle derived from desert-dust in murine lungs

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

He, Miao, E-mail: hemiao@mail.cmu.edu.cn; Department of Health Sciences, Oita University of Nursing and Health Sciences, Oita 870-1201; Ichinose, Takamichi, E-mail: ichinose@oita-nhs.ac.jp

The biological and chemical natures of materials adsorbed onto fine particulate matter (PM2.5) vary by origin and passage routes. The exacerbating effects of the two samples—urban PM2.5 (U-PM2.5) collected during the hazy weather in a Chinese city and fine particles (ASD-PM2.5) collected during Asian sand dust (ASD) storm event days in Japan—on murine lung eosinophilia were compared to clarify the role of toxic materials in PM2.5. The amounts of β-glucan and mineral components were higher in ASD-PM2.5 than in U-PM2.5. On the other hand, organic chemicals, including polycyclic aromatic hydrocarbons (PAHs), were higher in U-PM2.5 than in ASD-PM2.5. When BALB/cmore » mice were intratracheally instilled with U-PM2.5 and ASD-PM2.5 (total 0.4 mg/mouse) with or without ovalbumin (OVA), various biological effects were observed, including enhancement of eosinophil recruitment induced by OVA in the submucosa of the airway, goblet cell proliferation in the bronchial epithelium, synergic increase of OVA-induced eosinophil-relevant cytokines and a chemokine in bronchoalveolar lavage fluid, and increase of serum OVA-specific IgG1 and IgE. Data demonstrate that U-PM2.5 and ASD-PM2.5 induced allergic inflammatory changes and caused lung pathology. U-PM2.5 and ASD-PM2.5 increased F4/80{sup +} CD11b{sup +} cells, indicating that an influx of inflammatory and exudative macrophages in lung tissue had occurred. The ratio of CD206 positive F4/80{sup +} CD11b{sup +} cells (M2 macrophages) in lung tissue was higher in the OVA + ASD-PM2.5 treated mice than in the OVA + U-PM2.5 treated mice. These results suggest that the lung eosinophilia exacerbated by both PM2.5 is due to activation of a Th2-associated immune response along with induced M2 macrophages and the exacerbating effect is greater in microbial element (β-glucan)-rich ASD-PM2.5 than in organic chemical-rich U-PM2.5. - Highlights: • The aggravating effects of urban-PM2.5 and desert-PM2.5 on lung eosinophilia were

SAS2H Generated Isotopic Concentrations For B&W 15X15 PWR Assembly (SCPB:N/A)

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

J.W. Davis

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.

Experimental evaluation of models for predicting Cherenkov light intensities from short-cooled nuclear fuel assemblies

NASA Astrophysics Data System (ADS)

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

2018-02-01

The Digital Cherenkov Viewing Device (DCVD) is a tool used by nuclear safeguards inspectors to verify irradiated nuclear fuel assemblies in wet storage based on the recording of Cherenkov light produced by the assemblies. One type of verification involves comparing the measured light intensity from an assembly with a predicted intensity, based on assembly declarations. Crucial for such analyses is the performance of the prediction model used, and recently new modelling methods have been introduced to allow for enhanced prediction capabilities by taking the irradiation history into account, and by including the cross-talk radiation from neighbouring assemblies in the predictions. In this work, the performance of three models for Cherenkov-light intensity prediction is evaluated by applying them to a set of short-cooled PWR 17x17 assemblies for which experimental DCVD measurements and operator-declared irradiation data was available; (1) a two-parameter model, based on total burnup and cooling time, previously used by the safeguards inspectors, (2) a newly introduced gamma-spectrum-based model, which incorporates cycle-wise burnup histories, and (3) the latter gamma-spectrum-based model with the addition to account for contributions from neighbouring assemblies. The results show that the two gamma-spectrum-based models provide significantly higher precision for the measured inventory compared to the two-parameter model, lowering the standard deviation between relative measured and predicted intensities from 15.2 % to 8.1 % respectively 7.8 %. The results show some systematic differences between assemblies of different designs (produced by different manufacturers) in spite of their similar PWR 17x17 geometries, and possible ways are discussed to address such differences, which may allow for even higher prediction capabilities. Still, it is concluded that the gamma-spectrum-based models enable confident verification of the fuel assembly inventory at the currently used

Radiotoxicity Characterization of Multi-Recycled Thorium Fuel - 12394

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

Franceschini, F.; Wenner, M.; Fiorina, C.

2012-07-01

As described in companion papers, Westinghouse is proposing the implementation of a thorium based fuel cycle to burn the transuranic (TRU) contained in the used nuclear fuel. The potential of thorium as a TRU burner is described in another paper presented at this conference. This paper analyzes the long-term impact of thorium on the front-end and backend of the fuel cycle. This is accomplished by an assessment of the isotopic make-up of Th in a closed cycle and its impact on representative metrics, such as radiotoxicity, decay heat and gamma heat. The behavior in both thermal and fast neutron energymore » ranges has been investigated. Irradiation in a Th fuel PWR has been assumed as representative of the thermal range, while a Th fuel fast reactor (FR) has been employed to characterize the behavior in the high-energy range. A comparison with a U-fuel closed-cycle FR has been undertaken in an attempt of a more comprehensive evaluation of each cycle's long-term potential. As the Th fuel undergoes multiple cycles of irradiation, the isotopic composition of the recycled fuel changes. Minor Th isotopes are produced; U-232 and Pa-231 build up; the U vector gradually shifts towards increasing amounts of U-234, U-235 etc., eventually leading to the production of non negligible amounts of TRU isotopes, especially Pu-238. The impact of the recycled fuel isotopic makeup on the in-core behavior is mild, and for some aspects beneficial, i.e. the reactivity swing during irradiation is reduced as the fertile characteristics of the fuel increase. On the other hand, the front and the back-end of the fuel cycle are negatively affected due to the presence of Th-228 and U-232 and the build-up of higher actinides (Pu-238 etc.). The presence of U-232 can also be seen as advantageous as it represents an obstacle to potential proliferators. Notwithstanding the increase in the short-term radiotoxicity and decay heat in the multi-recycled fuel, the Th closed cycle has some potentially

System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

NASA Astrophysics Data System (ADS)

Porter, Ian Edward

A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several

Fuel flexible fuel injector

DOEpatents

Tuthill, Richard S; Davis, Dustin W; Dai, Zhongtao

2015-02-03

A disclosed fuel injector provides mixing of fuel with airflow by surrounding a swirled fuel flow with first and second swirled airflows that ensures mixing prior to or upon entering the combustion chamber. Fuel tubes produce a central fuel flow along with a central airflow through a plurality of openings to generate the high velocity fuel/air mixture along the axis of the fuel injector in addition to the swirled fuel/air mixture.

The Impact of Operating Parameters and Correlated Parameters for Extended BWR Burnup Credit

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

Ade, Brian J.; Marshall, William B. J.; Ilas, Germina

Applicants for certificates of compliance for spent nuclear fuel (SNF) transportation and dry storage systems perform analyses to demonstrate that these systems are adequately subcritical per the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Parts 71 and 72. For pressurized water reactor (PWR) SNF, these analyses may credit the reduction in assembly reactivity caused by depletion of fissile nuclides and buildup of neutron-absorbing nuclides during power operation. This credit for reactivity reduction during depletion is commonly referred to as burnup credit (BUC). US Nuclear Regulatory Commission (NRC) staff review BUC analyses according to the guidancemore » in the Division of Spent Fuel Storage and Transportation Interim Staff Guidance (ISG) 8, Revision 3, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks.« less

PWR integral tie plate and locking mechanism

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

Flora, B.S.; Osborne, J.L.

1980-08-26

A locking mechanism for securing an upper tie plate to the tie rods of a nuclear fuel bundle is described. The mechanism includes an upper tie plate assembly and locking sleeves fixed to the ends of the tie rods. The tie plate is part of the upper tie plate assembly and is secured to the fuel bundle by securing the entire upper tie plate assembly to the locking sleeves fixed to the tie rods. The assembly includes, in addition to the tie plate, locking nuts for engaging the locking sleeves, retaining sleeves to operably connect the locking nuts to themore » assembly, a spring biased reaction plate to restrain the locking nuts in the locked position and a means to facilitate the removal of the entire assembly as a unit from the fuel bundle.« less

Fuel Exhaling Fuel Cell.

PubMed

Manzoor Bhat, Zahid; Thimmappa, Ravikumar; Devendrachari, Mruthyunjayachari Chattanahalli; Kottaichamy, Alagar Raja; Shafi, Shahid Pottachola; Varhade, Swapnil; Gautam, Manu; Thotiyl, Musthafa Ottakam

2018-01-18

State-of-the-art proton exchange membrane fuel cells (PEMFCs) anodically inhale H 2 fuel and cathodically expel water molecules. We show an unprecedented fuel cell concept exhibiting cathodic fuel exhalation capability of anodically inhaled fuel, driven by the neutralization energy on decoupling the direct acid-base chemistry. The fuel exhaling fuel cell delivered a peak power density of 70 mW/cm 2 at a peak current density of 160 mA/cm 2 with a cathodic H 2 output of ∼80 mL in 1 h. We illustrate that the energy benefits from the same fuel stream can at least be doubled by directing it through proposed neutralization electrochemical cell prior to PEMFC in a tandem configuration.

Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

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

Liu, Maolong; Ryals, Matthew; Ali, Amir

2016-08-01

A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentallymore » investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.« less

Fuel cell-fuel cell hybrid system

DOEpatents

Geisbrecht, Rodney A.; Williams, Mark C.

2003-09-23

A device for converting chemical energy to electricity is provided, the device comprising a high temperature fuel cell with the ability for partially oxidizing and completely reforming fuel, and a low temperature fuel cell juxtaposed to said high temperature fuel cell so as to utilize remaining reformed fuel from the high temperature fuel cell. Also provided is a method for producing electricity comprising directing fuel to a first fuel cell, completely oxidizing a first portion of the fuel and partially oxidizing a second portion of the fuel, directing the second fuel portion to a second fuel cell, allowing the first fuel cell to utilize the first portion of the fuel to produce electricity; and allowing the second fuel cell to utilize the second portion of the fuel to produce electricity.

Logistics Modeling of Emplacement Rate and Duration of Operations for Generic Geologic Repository Concepts

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

Kalinina, Elena Arkadievna; Hardin, Ernest

This study identified potential geologic repository concepts for disposal of spent nuclear fuel (SNF) and (2) evaluated the achievable repository waste emplacement rate and the time required to complete the disposal for these concepts. Total repository capacity is assumed to be approximately 140,000 MT of spent fuel. The results of this study provide an important input for the rough-order-of-magnitude (ROM) disposal cost analysis. The disposal concepts cover three major categories of host geologic media: crystalline or hard rock, salt, and argillaceous rock. Four waste package sizes are considered: 4PWR/9BWR; 12PWR/21BWR; 21PWR/44BWR, and dual purpose canisters (DPCs). The DPC concepts assumemore » that the existing canisters will be sealed into disposal overpacks for direct disposal. Each concept assumes one of the following emplacement power limits for either emplacement or repository closure: 1.7 kW; 2.2 kW; 5.5 kW; 10 kW; 11.5 kW, and 18 kW.« less

Advance Appropriations for Veteran’s Health Care: Issues and Options for Congress

DTIC Science & Technology

2009-04-28

care system. Utilization Projection Model (UPM) The UPM is based on the Milliman Health Cost Guidelines ( HCGs ), a proprietary set of utilization-rate...benchmarks derived from commercial data.31 The HCGs contain data on utilization for 37 of the 58 EHCPM health service categories. Milliman applies a...complex set of adjustments to the HCG data to reflect the health status of VA enrollees, their reliance on VA, and the relative efficiency of VA

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.

Posttest analysis of international standard problem 10 using RELAP4/MOD7. [PWR

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

Hsu, M.; Davis, C.B.; Peterson, A.C. Jr.

RELAP4/MOD7, a best estimate computer code for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This paper evaluates the capability of RELAP4/MOD7 to calculate refill/reflood phenomena. This evaluation uses the data of International Standard Problem 10, which is based on West Germany's KWU PKL refill/reflood experiment K9A. The PKL test facility represents a typical West German four-loop, 1300 MW pressurized water reactor (PWR) in reduced scale while maintaining prototypical volume-to-power ratio. The PKL facility was designed to specifically simulate the refill/reflood phase of amore » hypothetical loss-of-coolant accident (LOCA).« less

An Overview of Reactor Concepts, a Survey of Reactor Designs.

DTIC Science & Technology

1985-02-01

may be very different. HTGRs may use highly enriched uranium, thereby yielding better fuel economy and a reduc- tion of the actual core size for a...specific power level. The HTGR core may have fuel and control rods placed in graphite arrays similar to PWR core con- figuration, or they may have fuel ...rods are pulled out. A Peach Bottom core design is another HTGR design. This design is featured by the fuel pin’s ability to purge itself of fission

Alternative Fuels Data Center: Fuel Prices

Science.gov Websites

Report provides regional alternative and conventional fuel prices for biodiesel, compressed natural gas petroleum fuels (gasoline and diesel fuel) is the primary driver of liquid fuel prices. This is because the liquid fuels are used in non-dedicated vehicles and can be substituted out by petroleum fuels if their

Fuels research: Fuel thermal stability overview

NASA Technical Reports Server (NTRS)

Cohen, S. M.

1980-01-01

Alternative fuels or crude supplies are examined with respect to satisfying aviation fuel needs for the next 50 years. The thermal stability of potential future fuels is discussed and the effects of these characteristics on aircraft fuel systems are examined. Advanced fuel system technology and design guidelines for future fuels with lower thermal stability are reported.

Optimal arrangement of magnetic coils for functional magnetic stimulation of the inspiratory muscles in dogs.

PubMed

Lin, Vernon Weh-Hau; Zhu, Ercheng; Sasse, Scott A; Sassoon, Catherine; Hsiao, Ian N

2005-12-01

In an attempt to maximize inspiratory pressure and volume, the optimal position of a single or of dual magnetic coils during functional magnetic stimulation (FMS) of the inspiratory muscles was evaluated in twenty-three dogs. Unilateral phrenic magnetic stimulation (UPMS) or bilateral phrenic magnetic stimulation (BPMS), posterior cervical magnetic stimulation (PCMS), anterior cervical magnetic stimulation (ACMS) as well as a combination of PCMS and ACMS were performed. Trans-diaphragmatic pressure (Pdi), flow, and lung volume changes with an open airway were measured. Transdiaphragmatic pressure was also measured with an occluded airway. Changes in inspiratory parameters during FMS were compared with 1) electrical stimulation of surgically exposed bilateral phrenic nerves (BPES) and 2) ventral root electrical stimulation at C5-C7 (VRES C5-C7). Relative to the Pdi generated by BPES of 36.3 +/- 4.5 cm H2O (Mean +/- SEM), occluded Pdi(s) produced by UPMS, BPMS, PCMS, ACMS, and a combined PCMS + ACMS were 51.7%, 61.5%, 22.4%, 100.3%, and 104.5% of the maximal Pdi, respectively. Pdi(s) produced by UPMS, BPMS, PCMS, ACMS, and combined ACMS + PCMS were 38.0%, 45.2%, 16.5%, 73.8%, and 76.8%, respectively, of the Pdi induced by VRES (C5-C7) (48.0 +/- 3.9 cm H2O). The maximal Pdi(s) generated during ACMS and combined PCMS + ACMS were higher than the maximal Pdi(s) generated during UPMS, BPMS, or PCMS (p < 0.05). ACMS alone induced 129.8% of the inspiratory flow (73.0 +/- 9.4 L/ min) and 77.5% of the volume (626 +/- 556 ml) induced by BPES. ACMS and combined PCMS + ACMS produce a greater inspiratory pressure than UPMS, BPMS or PCMS. ACMS can be used to generate sufficient inspiratory pressure, flow, and volume for activation of the inspiratory muscles.

Corrosion fatigue characterization of reactor pressure vessel steels. [PWR; BWR

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

Van Der Sluys, W.A.

1982-12-01

During routine operation, light water reactor (LWR) pressure vessels are subjected to a variety of transients that result in time-varying stresses. Consequently, fatigue and environmentally-assisted fatigue are mechanisms of growth relevant to flaws in these pressure vessels. To provide a better understanding of the resistance of nuclear pressure vessel steels to these flaw growth processes, fracture mechanics data were generated on the rates of fatigue crack growth for SA508-2 and SA533B-1 steels in both room temperature air and 288/sup 0/C water. Areas investigated were: the relationship of crack growth rate to prior loading history; the effects of loading frequency andmore » R ratio (K/sub min//K/sub max/) on crack growth rate as a function of the stress intensity factor range (..delta..K); transient aspects of the fatigue crack growth behavior; the effect of material chemistry (sulphur content) on fatigue crack; and growth rate; water chemistry effects (high-purity water versus simulated pressurized water reactotr (PWR) primary coolant).« less

Reforming of fuel inside fuel cell generator

DOEpatents

Grimble, Ralph E.

1988-01-01

Disclosed is an improved method of reforming a gaseous reformable fuel within a solid oxide fuel cell generator, wherein the solid oxide fuel cell generator has a plurality of individual fuel cells in a refractory container, the fuel cells generating a partially spent fuel stream and a partially spent oxidant stream. The partially spent fuel stream is divided into two streams, spent fuel stream I and spent fuel stream II. Spent fuel stream I is burned with the partially spent oxidant stream inside the refractory container to produce an exhaust stream. The exhaust stream is divided into two streams, exhaust stream I and exhaust stream II, and exhaust stream I is vented. Exhaust stream II is mixed with spent fuel stream II to form a recycle stream. The recycle stream is mixed with the gaseous reformable fuel within the refractory container to form a fuel stream which is supplied to the fuel cells. Also disclosed is an improved apparatus which permits the reforming of a reformable gaseous fuel within such a solid oxide fuel cell generator. The apparatus comprises a mixing chamber within the refractory container, means for diverting a portion of the partially spent fuel stream to the mixing chamber, means for diverting a portion of exhaust gas to the mixing chamber where it is mixed with the portion of the partially spent fuel stream to form a recycle stream, means for injecting the reformable gaseous fuel into the recycle stream, and means for circulating the recycle stream back to the fuel cells.

Reforming of fuel inside fuel cell generator

DOEpatents

Grimble, R.E.

1988-03-08

Disclosed is an improved method of reforming a gaseous reformable fuel within a solid oxide fuel cell generator, wherein the solid oxide fuel cell generator has a plurality of individual fuel cells in a refractory container, the fuel cells generating a partially spent fuel stream and a partially spent oxidant stream. The partially spent fuel stream is divided into two streams, spent fuel stream 1 and spent fuel stream 2. Spent fuel stream 1 is burned with the partially spent oxidant stream inside the refractory container to produce an exhaust stream. The exhaust stream is divided into two streams, exhaust stream 1 and exhaust stream 2, and exhaust stream 1 is vented. Exhaust stream 2 is mixed with spent fuel stream 2 to form a recycle stream. The recycle stream is mixed with the gaseous reformable fuel within the refractory container to form a fuel stream which is supplied to the fuel cells. Also disclosed is an improved apparatus which permits the reforming of a reformable gaseous fuel within such a solid oxide fuel cell generator. The apparatus comprises a mixing chamber within the refractory container, means for diverting a portion of the partially spent fuel stream to the mixing chamber, means for diverting a portion of exhaust gas to the mixing chamber where it is mixed with the portion of the partially spent fuel stream to form a recycle stream, means for injecting the reformable gaseous fuel into the recycle stream, and means for circulating the recycle stream back to the fuel cells. 1 fig.

TRIGA MARK-II source term

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

Usang, M. D., E-mail: mark-dennis@nuclearmalaysia.gov.my; Hamzah, N. S., E-mail: mark-dennis@nuclearmalaysia.gov.my; Abi, M. J. B., E-mail: mark-dennis@nuclearmalaysia.gov.my

ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences ofmore » results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel.« less

Fuels processing for transportation fuel cell systems

NASA Astrophysics Data System (ADS)

Kumar, R.; Ahmed, S.

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

Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

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

Rempe, J. L.; Knudson, D. L.; Lutz, R. J.

The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure thatmore » critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

Development and Application of Laser Peening System for PWR Power Plants

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

Masaki Yoda; Itaru Chida; Satoshi Okada

2006-07-01

Laser peening is a process to improve residual stress from tensile to compressive in surface layer of materials by irradiating high-power laser pulses on the material in water. Toshiba has developed a laser peening system composed of Q-switched Nd:YAG laser oscillators, laser delivery equipment and underwater remote handling equipment. We have applied the system for Japanese operating BWR power plants as a preventive maintenance measure for stress corrosion cracking (SCC) on reactor internals like core shrouds or control rod drive (CRD) penetrations since 1999. As for PWRs, alloy 600 or 182 can be susceptible to primary water stress corrosion crackingmore » (PWSCC), and some cracks or leakages caused by the PWSCC have been discovered on penetrations of reactor vessel heads (RVHs), reactor bottom-mounted instrumentation (BMI) nozzles, and others. Taking measures to meet the unconformity of the RVH penetrations, RVHs themselves have been replaced in many PWRs. On the other hand, it's too time-consuming and expensive to replace BMI nozzles, therefore, any other convenient and less expensive measures are required instead of the replacement. In Toshiba, we carried out various tests for laser-peened nickel base alloys and confirmed the effectiveness of laser peening as a preventive maintenance measure for PWSCC. We have developed a laser peening system for PWRs as well after the one for BWRs, and applied it for BMI nozzles, core deluge line nozzles and primary water inlet nozzles of Ikata Unit 1 and 2 of Shikoku Electric Power Company since 2004, which are Japanese operating PWR power plants. In this system, laser oscillators and control devices were packed into two containers placed on the operating floor inside the reactor containment vessel. Laser pulses were delivered through twin optical fibers and irradiated on two portions in parallel to reduce operation time. For BMI nozzles, we developed a tiny irradiation head for small tubes and we peened the inner surface

PWR design for low doses in the United Kingdom: The present and the future

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

Zodiates, A.M.; Willcock, A.

1995-03-01

The Pressurizer Water Reactor (PWR) design chosen for adoption by Nuclear Electric plc was based on the Westinghouse Standard Nuclear Unit Power Plant System (SNUPPS). This design was developed to meet the United Kingdom (UK) requirements and those improvements are embodied in the Sizewell B plant. Nuclear Electric plc is now looking to the design of the future PWRs to be built in the UK. These PWRs will be based as replicas of the Sizewell B design, but attention will be given to reducing operator doses further. This paper details the approach in operator protection improvements incorporated at Sizewall B,more » presents the estimated annual collective dose, and identifies the approach being adopted to reduce further operator doses in future plants.« less

Electrochemical study of pre- and post-transition corrosion of Zr alloys in PWR coolant

NASA Astrophysics Data System (ADS)

Macák, Jan; Novotný, Radek; Sajdl, Petr; Renčiuková, Veronika; Vrtílková, Věra

Corrosion properties of Zr-Sn and Zr-Nb zirconium alloys were studied under simulated PWR conditions (or, more exactly, VVER conditions — boric acid, potassium hydroxide, lithium hydroxide) at temperatures up to 340°C and 15MPa using in-situ electrochemical impedance spectroscopy (EIS) and polarization measurements. EIS spectra were obtained in a wide range of frequencies (typically 100kHz — 100μHz). It enabled to gain information of both dielectric properties of oxide layers developing on the Zr-alloys surface and of the kinetics of the corrosion process and the associated charge and mass transfer phenomena. Experiments were run for more than 380 days; thus, the study of all the corrosion stages (pre-transition, transition, post-transition) was possible.

Alternative Fuels Data Center: CNG Vehicle Fueling Animation

Science.gov Websites

Vehicle Fueling Animation to someone by E-mail Share Alternative Fuels Data Center: CNG Vehicle Fueling Animation on Facebook Tweet about Alternative Fuels Data Center: CNG Vehicle Fueling Animation on Twitter Bookmark Alternative Fuels Data Center: CNG Vehicle Fueling Animation on Google Bookmark Alternative Fuels

The α-Arrestin Bul1p Mediates Lactate Transporter Endocytosis in Response to Alkalinization and Distinct Physiological Signals.

PubMed

Talaia, Gabriel; Gournas, Christos; Saliba, Elie; Barata-Antunes, Cláudia; Casal, Margarida; André, Bruno; Diallinas, George; Paiva, Sandra

2017-11-24

Eukaryotic α-arrestins connect environmental or stress signaling pathways to the endocytosis of plasma membrane transporters or receptors. The Saccharomyces cerevisiae lactate transporter Jen1p has been used as a model cargo for elucidating the mechanisms underlying endocytic turnover in response to carbon sources. Here, we discover a novel pathway of Jen1p endocytosis mediated by the α-arrestin Bul1p in response to the presence of cycloheximide or rapamycin, or prolonged growth in lactate. While cycloheximide or rapamycin modify cells pleiotropically, the major effect of prolonged growth in lactate was shown to be external pH alkalinization. Importantly, employment of specific inactive Jen1p versions showed that Bul1p-dependent endocytosis requires lactate transport, according to the signal imposed. Our results support a model where conformational changes of Jen1p, associated with substrate/H + symport, are critical for the efficiency of Bul1p-dependent Jen1p turnover. Copyright © 2017 Elsevier Ltd. All rights reserved.

Fuel processors for fuel cell APU applications

NASA Astrophysics Data System (ADS)

Aicher, T.; Lenz, B.; Gschnell, F.; Groos, U.; Federici, F.; Caprile, L.; Parodi, L.

The conversion of liquid hydrocarbons to a hydrogen rich product gas is a central process step in fuel processors for auxiliary power units (APUs) for vehicles of all kinds. The selection of the reforming process depends on the fuel and the type of the fuel cell. For vehicle power trains, liquid hydrocarbons like gasoline, kerosene, and diesel are utilized and, therefore, they will also be the fuel for the respective APU systems. The fuel cells commonly envisioned for mobile APU applications are molten carbonate fuel cells (MCFC), solid oxide fuel cells (SOFC), and proton exchange membrane fuel cells (PEMFC). Since high-temperature fuel cells, e.g. MCFCs or SOFCs, can be supplied with a feed gas that contains carbon monoxide (CO) their fuel processor does not require reactors for CO reduction and removal. For PEMFCs on the other hand, CO concentrations in the feed gas must not exceed 50 ppm, better 20 ppm, which requires additional reactors downstream of the reforming reactor. This paper gives an overview of the current state of the fuel processor development for APU applications and APU system developments. Furthermore, it will present the latest developments at Fraunhofer ISE regarding fuel processors for high-temperature fuel cell APU systems on board of ships and aircrafts.

KENNEDY SPACE CENTER, FLA. - The Honorable Toni Jennings (left), lieutenant governor of the state of Florida, and Frank T. Brogan, president of Florida Atlantic University, receive a briefing on the research that will be conducted in the Space Life Sciences Lab from Dr. Robert J. Ferl (right), director of Space Agriculture Biotechnology Research and Education (SABRE), University of Florida. Jennings and Brogan are speaking at a dedication and ribbon-cutting ceremony for the lab hosted by NASA-Kennedy Space Center and the state of Florida at the new lab. Completed in August, the facility encompasses more than 100,000 square feet and was formerly known as the Space Experiment Research and Processing Laboratory or SERPL. The state, through the Florida Space Authority, built the research lab which is host to NASA, NASA’s Life Sciences Services contractor Dynamac Corp., Bionetics Corp., and researchers from the University of Florida. Dynamac Corp. leases the facility. The Florida Space Research Institute is responsible for gaining additional tenants from outside the NASA community.

NASA Image and Video Library

2003-11-19

KENNEDY SPACE CENTER, FLA. - The Honorable Toni Jennings (left), lieutenant governor of the state of Florida, and Frank T. Brogan, president of Florida Atlantic University, receive a briefing on the research that will be conducted in the Space Life Sciences Lab from Dr. Robert J. Ferl (right), director of Space Agriculture Biotechnology Research and Education (SABRE), University of Florida. Jennings and Brogan are speaking at a dedication and ribbon-cutting ceremony for the lab hosted by NASA-Kennedy Space Center and the state of Florida at the new lab. Completed in August, the facility encompasses more than 100,000 square feet and was formerly known as the Space Experiment Research and Processing Laboratory or SERPL. The state, through the Florida Space Authority, built the research lab which is host to NASA, NASA’s Life Sciences Services contractor Dynamac Corp., Bionetics Corp., and researchers from the University of Florida. Dynamac Corp. leases the facility. The Florida Space Research Institute is responsible for gaining additional tenants from outside the NASA community.

Alternative Fuels Data Center: Ethanol Fueling Station Locations

Science.gov Websites

Fueling Station Locations to someone by E-mail Share Alternative Fuels Data Center: Ethanol Fueling Station Locations on Facebook Tweet about Alternative Fuels Data Center: Ethanol Fueling Station Locations on Twitter Bookmark Alternative Fuels Data Center: Ethanol Fueling Station Locations on Google

Electron Microscopy Characterizations and Atom Probe Tomography of Intergranular Attack in Alloy 600 Exposed to PWR Primary Water

NASA Astrophysics Data System (ADS)

Olszta, Matthew J.; Schreiber, Daniel K.; Thomas, Larry E.; Bruemmer, Stephen M.

Detailed examinations of intergranular attack (IGA) in alloy 600 were performed after exposure to simulated PWR primary water at 325°C for 500 h. High-resolution analyses of IGA characteristics were conducted on specimens with either a 1 µm diamond or 1200-grit SiC surface finish using scanning electron microscopy, transmission electron microscopy and atom probe tomography techniques. The diamond-polish finish with very little preexisting subsurface damage revealed attack of high-energy grain boundaries that intersected the exposed surface to depths approaching 2 µm. In all cases, IGA from the surface is localized oxidation consisting of porous, nanocrystalline MO-structure and spinel particles along with regions of faceted wall oxidation. Surprisingly, this continuous IG oxidation transitions to discontinuous, discrete Cr-rich sulfide particles up to 50 nm in diameter. In the vicinity of the sulfides, the grain boundaries were severely Cr depleted (to <1 at%) and enriched in S. The 1200 grit SiC finish surface exhibited a preexisting highly strained recrystallized layer of elongated nanocrystalline matrix grains. Similar IG oxidation and leading sulfide particles were found, but the IGA depth was typically confined to the near-surface ( 400 nm) recrystallized region. Difference in IGA for the two surface finishes indicates that the formation of grain boundary sulfides occurs during the exposure to PWR primary water. The source of S remains unclear, however it is not present as sulfides in the bulk alloy nor is it segregated to bulk grain boundaries.

Alternative Fuels Data Center: Natural Gas Fueling Stations

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Alternative Fuels Data Center: Fuel Cell Electric Vehicles

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Development of cement solidification process for sodium borate waste generated from PWR plants

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

Hirofumi Okabe; Tatsuaki Sato; Yuichi Shoji

2013-07-01

A cement solidification process for treating sodium borate waste produced in pressurized water reactor (PWR) plants was studied. To obtain high volume reduction and high mechanical strength of the waste, simulated concentrated borate liquid waste with a sodium / boron (Na/B) mole ratio of 0.27 was dehydrated and powdered by using a wiped film evaporator. To investigate the effect of the Na/B mole ratio on the solidification process, a sodium tetraborate decahydrate reagent with a Na/B mole ratio of 0.5 was also used. Ordinary portland cement (OPC) and some additives were used for the solidification. Solidified cement prepared from powderedmore » waste with a Na/B mole ratio 0.24 and having a high silica sand content (silica sand/cement>2) showed to improved uniaxial compressive strength. (authors)« less

Internal reforming fuel cell assembly with simplified fuel feed

DOEpatents

Farooque, Mohammad; Novacco, Lawrence J.; Allen, Jeffrey P.

2001-01-01

A fuel cell assembly in which fuel cells adapted to internally reform fuel and fuel reformers for reforming fuel are arranged in a fuel cell stack. The fuel inlet ports of the fuel cells and the fuel inlet ports and reformed fuel outlet ports of the fuel reformers are arranged on one face of the fuel cell stack. A manifold sealing encloses this face of the stack and a reformer fuel delivery system is arranged entirely within the region between the manifold and the one face of the stack. The fuel reformer has a foil wrapping and a cover member forming with the foil wrapping an enclosed structure.

Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks

DOE PAGES

Banerjee, Kaushik; Robb, Kevin R.; Radulescu, Georgeta; ...

2016-06-15

We completed a novel assessment to determine the unquantified and uncredited safety margins (i.e., the difference between the licensing basis and as-loaded calculations) available in as-loaded spent nuclear fuel (SNF) casks. This assessment was performed as part of a broader effort to assess issues and uncertainties related to the continued safety of casks during extended storage and transportability following extended storage periods. Detailed analyses crediting the actual as-loaded cask inventory were performed for each of the casks at three decommissioned pressurized water reactor (PWR) sites to determine their characteristics relative to regulatory safety criteria for criticality, thermal, and shielding performance.more » These detailed analyses were performed in an automated fashion by employing a comprehensive and integrated data and analysis tool—Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). Calculated uncredited criticality margins from 0.07 to almost 0.30 Δk eff were observed; calculated decay heat margins ranged from 4 to almost 22 kW (as of 2014); and significant uncredited transportation dose rate margins were also observed. The results demonstrate that, at least for the casks analyzed here, significant uncredited safety margins are available that could potentially be used to compensate for SNF assembly and canister structural performance related uncertainties associated with long-term storage and subsequent transportation. The results also suggest that these inherent margins associated with how casks are loaded could support future changes in cask licensing to directly or indirectly credit the margins. Work continues to quantify the uncredited safety margins in the SNF casks loaded at other nuclear reactor sites.« less

EPRI/DOE High Burnup Fuel Sister Pin Test Plan Simplification and Visualization

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

Saltzstein, Sylvia J.; Sorenson, Ken B.; Hanson, Brady

The EPRI/DOE High Burnup Confirmatory Data Project (herein called the "Demo") is a multi-year, multi-entity confirmation demonstration test with the purpose of providing quantitative and qualitative data to show how high-burnup fuel ages in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of four common cladding alloys from the North Anna Nuclear Power Plant, drying them according to standard plant procedures, and then storing them in an NRC-licensed TN-3 2B cask on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened andmore » the rods will be examined for signs of aging. Twenty-five rods from assemblies of similar claddings, in-reactor placement, and burnup histories (herein called "sister rods") have been shipped from the North Anna Nuclear Power Plant and are currently being nondestructively tested at Oak Ridge National Laboratory. After the non-destructive testing has been completed for each of the twenty-five rods, destructive analysis will be performed at ORNL, PNNL, and ANL to obtain mechanical data. Opinions gathered from the expert interviews, ORNL and PNNL Sister Rod Test Plans, and numerous meetings has resulted in the Simplified Test Plan described in this document. Some of the opinions and discussions leading to the simplified test plan are included here. Detailed descriptions and background are in the ORNL and PNNL plans in the appendices . After the testing described in this simplified test plan h as been completed , the community will review all the collected data and determine if additional testing is needed.« less

Technial Programme Committee

NASA Astrophysics Data System (ADS)

2014-06-01

Chairpersons Dr Dinesh Sathyamoorthy, Science & Technology Research Institute for Defence (STRIDE), Ministry of Defence, Malaysia Associate Professor Sr Dr Abdul Rashid Mohamed Shariff, Universiti Putra Malaysia (UPM), Malaysia Dr Ahmad Fikri Abdullah, Universiti Putra Malaysia (UPM), Malaysia Dr Farrah Melissa Muharram, Universiti Putra Malaysia (UPM), Malaysia Members Professor Dr Li Jing, Beijing Normal University, China Professor Dr Iyyanki Muralikrishna, Administrative Staff College of India (ASCI), India Professor Dr Alias Abdul Rahman, Universiti Teknologi Malaysia (UTM), Malaysia Professor Dr Ismat Mohamed El Hassan, King Saud University, Saudi Arabia Professor Dr George Miliaresis, Open University of Cyprus, Cyprus Professor Dr Christine Pohl, Universiti Teknologi Malaysia (UTM), Malaysia Professor Dr Mahender Kotha, Goa University, India Associate Professor Dr Paolo Gamba, University of Pavia, Italy Associate Professor Dr Behara Seshadri Daya Sagar, Indian Statistical Institute (ISI), India Associate Professor Sr Ranjit Singh, Infrastructure University Kuala Lumpur (IUKL), Malaysia Associate Professor Dr Abdul Nasir Matori, Universiti Teknologi Petronas (UTP), Malaysia Associate Dr Lucian Dragut, West University of Timişoara, Romania Associate Professor Dr Saied Pirasteh, Islamic Azad University, Iran Associate Professor Dr Peter Yuen, Cranfield University, United Kingdom Associate Professor Dr Lim Hwee San, Universiti Sains Malaysia (USM), Malaysia Associate Professor Dr Wayan Suparta, Universiti Kebangsaan Malaysia (UKM), Malaysia Associate Professor Dr Tuong Thuy Vu, The University of Nottingham Malaysia Campus, Malaysia Associate Professor Dr Maged Mahmoud Marghany, Universiti Teknologi Malaysia (UTM), Malaysia Associate Professor Dr Rami Al-Ruzouq, University of Sharjah, UAE Associate Professor Dr Biswajeet Pradhan, Universiti Putra Malaysia (UPM), Malaysia Associate Professor Dr Helmi Zulhaidi Mohd Shafri, Universiti Putra Malaysia (UPM), Malaysia

Multi-fuel reformers for fuel cells used in transportation. Phase 1: Multi-fuel reformers

NASA Astrophysics Data System (ADS)

1994-05-01

DOE has established the goal, through the Fuel Cells in Transportation Program, of fostering the rapid development and commercialization of fuel cells as economic competitors for the internal combustion engine. Central to this goal is a safe feasible means of supplying hydrogen of the required purity to the vehicular fuel cell system. Two basic strategies are being considered: (1) on-board fuel processing whereby alternative fuels such as methanol, ethanol or natural gas stored on the vehicle undergo reformation and subsequent processing to produce hydrogen, and (2) on-board storage of pure hydrogen provided by stationary fuel processing plants. This report analyzes fuel processor technologies, types of fuel and fuel cell options for on-board reformation. As the Phase 1 of a multi-phased program to develop a prototype multi-fuel reformer system for a fuel cell powered vehicle, the objective of this program was to evaluate the feasibility of a multi-fuel reformer concept and to select a reforming technology for further development in the Phase 2 program, with the ultimate goal of integration with a DOE-designated fuel cell and vehicle configuration. The basic reformer processes examined in this study included catalytic steam reforming (SR), non-catalytic partial oxidation (POX) and catalytic partial oxidation (also known as Autothermal Reforming, or ATR). Fuels under consideration in this study included methanol, ethanol, and natural gas. A systematic evaluation of reforming technologies, fuels, and transportation fuel cell applications was conducted for the purpose of selecting a suitable multi-fuel processor for further development and demonstration in a transportation application.

Fuel system for rotary distributor fuel injection pump

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

Klopfer, K.H.; Kelly, W.W.

1993-06-01

In a fuel injection pump having a drive shaft, a pump rotor driven by the drive shaft, reciprocating pumping means with periodic intake and pumping strokes to periodically receive an intake charge of fuel and deliver fuel at high pressure for fuel injection is described; a distributor head with a plurality of angularly spaced distributor outlets, the pump rotor providing a distributor rotor with a distributor port connected to the pumping means, the distributor rotor being rotatably mounted in the distributor head for sequential registration of the distributor port with the distributor outlets for distributing said high pressure delivery ofmore » fuel thereto; a fuel system for supplying fuel to the pumping means, having an end chamber at one end of the pump rotor and a fuel supply pump driven by the drive shaft and having an inlet and outlet, the supply pump outlet being connected to the end chamber for supplying fuel thereto, and a pressure regulator for regulating the fuel pressure in the end chamber; and a control valve connected between the pumping means and the end chamber and selectively opened during the intake strokes to supply fuel to the pumping means from the end chamber and during the pumping strokes to spill fuel from the pumping means into the end chamber to terminate said high pressure delivery of fuel; the improvement wherein the fuel system comprises a fuel return passage connected in series with the end chamber downstream thereof, wherein the pressure regulator is mounted in the return passage for regulating the upstream fuel pressure, including the upstream fuel pressure within the end chamber, and is connected for conducting excess fuel for return to the supply pump inlet, and wherein the supply pump is driven by the drive shaft to supply fuel at a rate exceeding the rate of said high pressure delivery of fuel for fuel injection and to provide excess fuel flow continuously through the end chamber and return passage to the pressure regulator.« less

Common cause evaluations in applied risk analysis of nuclear power plants. [PWR

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

Taniguchi, T.; Ligon, D.; Stamatelatos, M.

1983-04-01

Qualitative and quantitative approaches were developed for the evaluation of common cause failures (CCFs) in nuclear power plants and were applied to the analysis of the auxiliary feedwater systems of several pressurized water reactors (PWRs). Key CCF variables were identified through a survey of experts in the field and a review of failure experience in operating PWRs. These variables were classified into categories of high, medium, and low defense against a CCF. Based on the results, a checklist was developed for analyzing CCFs of systems. Several known techniques for quantifying CCFs were also reviewed. The information provided valuable insights inmore » the development of a new model for estimating CCF probabilities, which is an extension of and improvement over the Beta Factor method. As applied to the analysis of the PWR auxiliary feedwater systems, the method yielded much more realistic values than the original Beta Factor method for a one-out-of-three system.« less

146. FUEL LINE TO SKID 2 (FUEL LOADER) IN FUEL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

146. FUEL LINE TO SKID 2 (FUEL LOADER) IN FUEL CONTROL ROOM (215), LSB (BLDG. 751). LIQUID NITROGEN/HELIUM HEAT EXCHANGER ON RIGHT. - Vandenberg Air Force Base, Space Launch Complex 3, Launch Pad 3 East, Napa & Alden Roads, Lompoc, Santa Barbara County, CA

Alternative Fuels Data Center: Flexible Fuel Vehicle Conversions

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Conversions to someone by E-mail Share Alternative Fuels Data Center: Flexible Fuel Vehicle Conversions on Facebook Tweet about Alternative Fuels Data Center: Flexible Fuel Vehicle Conversions on Twitter Bookmark Alternative Fuels Data Center: Flexible Fuel Vehicle Conversions on Google Bookmark

Alternative Fuels Data Center: Biodiesel Fueling Station Locations

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Station Locations to someone by E-mail Share Alternative Fuels Data Center: Biodiesel Fueling Station Locations on Facebook Tweet about Alternative Fuels Data Center: Biodiesel Fueling Station Locations on Twitter Bookmark Alternative Fuels Data Center: Biodiesel Fueling Station Locations on Google

Benefits of barrier fuel on fuel cycle economics

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

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

1988-01-01

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

78 FR 62462 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard Program

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-22

... and Security Act of 2007 (EISA) to reduce the use of fossil fuels and encourage increased production... renewable fuel to replace or reduce the quantity of fossil fuel present in transportation fuel. Under EPA's... quantity of fossil fuel present in home heating oil or jet fuel.\\3\\ In essence, additional renewable fuel...

Fuel processor for fuel cell power system

DOEpatents

Vanderborgh, Nicholas E.; Springer, Thomas E.; Huff, James R.

1987-01-01

A catalytic organic fuel processing apparatus, which can be used in a fuel cell power system, contains within a housing a catalyst chamber, a variable speed fan, and a combustion chamber. Vaporized organic fuel is circulated by the fan past the combustion chamber with which it is in indirect heat exchange relationship. The heated vaporized organic fuel enters a catalyst bed where it is converted into a desired product such as hydrogen needed to power the fuel cell. During periods of high demand, air is injected upstream of the combustion chamber and organic fuel injection means to burn with some of the organic fuel on the outside of the combustion chamber, and thus be in direct heat exchange relation with the organic fuel going into the catalyst bed.

Alternative Fuels Data Center: Flexible Fuel Vehicles

Science.gov Websites

-ethanol blends, many vehicle owners don't realize their car is an FFV and that they have a choice of fuels Turbocharged GDI Vehicle and Fuel Economy and Emissions of a Vehicle Equipped with an Aftermarket Flexible-Fuel Fuel and Advanced Vehicle Inventory Clean Cities Alternative Fuel and Advanced Vehicle Inventory AFV

Fuel cells: principles, types, fuels, and applications.

PubMed

Carrette, L; Friedrich, K A; Stimming, U

2000-12-15

During the last decade, fuel cells have received enormous attention from research institutions and companies as novel electrical energy conversion systems. In the near future, they will see application in automotive propulsion, distributed power generation, and in low power portable devices (battery replacement). This review gives an introduction into the fundamentals and applications of fuel cells: Firstly, the environmental and social factors promoting fuel cell development are discussed, with an emphasis on the advantages of fuel cells compared to the conventional techniques. Then, the main reactions, which are responsible for the conversion of chemical into electrical energy in fuel cells, are given and the thermodynamic and kinetic fundamentals are stated. The theoretical and real efficiencies of fuel cells are also compared to that of internal combustion engines. Next, the different types of fuel cells and their main components are explained and the related material issues are presented. A section is devoted to fuel generation and storage, which is of paramount importance for the practical aspects of fuel cell use. Finally, attention is given to the integration of the fuel cells into complete systems. © 2000 WILEY-VCH Verlag GmbH, Weinheim, Fed. Rep. of Germany.

FuelCalc: A Method for Estimating Fuel Characteristics

Treesearch

Elizabeth Reinhardt; Duncan Lutes; Joe Scott

2006-01-01

This paper describes the FuelCalc computer program. FuelCalc is a tool to compute surface and canopy fuel loads and characteristics from inventory data, to support fuel treatment decisions by simulating effects of a wide range of silvicultural treatments on surface fuels and canopy fuels, and to provide linkages to stand visualization, fire behavior and fire effects...

Fuel Supply Defaults for Regional Fuels and Fuel Wizard Tool in MOVES201X

EPA Science Inventory

The fuel supply report documents the data and methodology used to derive the default gasoline, diesel and fuel-blend fuel properties, and their respective fuel market share in MOVES. The default market share of the individual fuels varies by calendar year, seasons, and several do...

Fuel system design concepts for broad property fuels

NASA Technical Reports Server (NTRS)

Versaw, E. F.

1984-01-01

The results of a study assessing the impact of using jet fuel with relaxed specification properties on an aircraft fuel system are given. The study objectives were to identify credible values for specific fuel properties which might be relaxed, to evolve advanced fuel system designs for airframe and engines which would permit use of the specified relaxed properties fuels, and to evaluate performance of the candidate advanced fuel systems and the relaxed property fuels in a typical transport aircraft. The study used, as a baseline, the fuel system incorporated in the Lockheed Tristar. This aircraft is powered by three RB.211-524 Rolls-Royce engines and incorporates a Pratt and Whitney ST6C-421 auxiliary power unit for engine starting and inflight emergency electrical power. The fuel property limits examined are compared with commercial Jet A kerosene and the NASA RFP fuel properties. A screening of these properties established that a higher freezing point and a lower thermal stability would impact fuel system design more significantly than any of the other property changes. Three candidate fuel systems which combine the ability to operate with fuels having both a high freeze point and a low thermal stability are described. All candidates employ bleed air to melt fuel freeze-out prior to starting the APU or an inoperable engine. The effects of incorporating these systems on aircraft weight and engine specific fuel consumption are given.

OptFuels: Fuel treatment optimization

Treesearch

Greg Jones

2011-01-01

Scientists at the USDA Forest Service, Rocky Mountain Research Station, in Missoula, MT, in collaboration with scientists at the University of Montana, are developing a tool to help forest managers prioritize forest fuel reduction treatments. Although several computer models analyze fuels and fire behavior, stand-level effects of fuel treatments, and priority planning...

Fuel economy of hybrid fuel-cell vehicles

NASA Astrophysics Data System (ADS)

Ahluwalia, Rajesh K.; Wang, X.; Rousseau, A.

The potential improvement in fuel economy of a mid-size fuel-cell vehicle by combining it with an energy storage system has been assessed. An energy management strategy is developed and used to operate the direct hydrogen, pressurized fuel-cell system in a load-following mode and the energy storage system in a charge-sustaining mode. The strategy places highest priority on maintaining the energy storage system in a state where it can supply unanticipated boost power when the fuel-cell system alone cannot meet the power demand. It is found that downsizing a fuel-cell system decreases its efficiency on a drive cycle which is compensated by partial regenerative capture of braking energy. On a highway cycle with limited braking energy the increase in fuel economy with hybridization is small but on the stop-and-go urban cycle the fuel economy can improve by 27%. On the combined highway and urban drive cycles the fuel economy of the fuel-cell vehicle is estimated to increase by up to 15% by hybridizing it with an energy storage system.

Alternative Fuels Data Center: Ethanol Flexible Fuel Vehicle Conversions

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An integrated approach for determining plutonium mass in spent fuel assemblies with nondestructive assay

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

Swinhoe, Martyn T; Tobin, Stephen J; Fensin, Mike L

2009-01-01

fuel can be part of a system that cost-effectively meets the burnup credit needs of a repository. Behind each of these reasons is a regulatory structure with MC&A requirements. In the case of the IAEA, the accountable quantity is elemental plutonium. The material in spent fuel (fissile isotopes, fission products, etc.) emits signatures that provide information about the content and history of the fuel. A variety of nondestructive assay (NDA) techniques are available to quantify these signatures. The effort presented in this paper is investigation of the capabilities of 12 NDA techniques. For these 12, none is conceptually capable of independently determining the Pu content in a spent fuel assembly while at the same time being able to detect the diversion of a significant quantity of rods. For this reason the authors are investigating the capability of 12 NDA techniques with the end goal of integrating a few techniques together into a system that is capable of measuring Pu mass in an assembly. The work described here is the beginning of what is anticipated to be a five year effort: (1) two years of modeling to select the best technologies, (2) one year fabricating instruments and (3) two years measuring spent fuel. This paper describes the first two years of this work. In order to cost effectively and robustly model the performance of the 12 NDA techniques, an 'assembly library' was created. The library contains the following: (a) A diverse range of PWR spent fuel assemblies (burnup, enrichment, cooling time) similar to that which exists in spent pools today and in the future. (b) Diversion scenarios that capture a range of possible rod removal options. (c) The spatial and isotopic detail needed to accurately quantify the capability of all the NDA techniques so as to enable integration. It is our intention to make this library available to other researchers in the field for inter-comparison purposes. The performance of each instrument will be quantified for the full

PRESSURIZED WATER REACTOR PROGRAM TECHNICAL PROGRESS REPORT FOR THE PERIOD MAY 5, 1955 TO JUNE 16, 1955

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

None

The current PWR plant and core parameters are listed. Resign requirements are briefly summarized for a radiation monitoring system, a fuel handling water system, a coolant purification system, an electrical power distribution system, and component shielding. Results of studies on thermal bowing and stressing of UO/sub 2/ are reported. A graph is presented of reactor power vs. reactor flow for various hot channel conditions. Development of U-- Mo and U-Nb alloys has been stopped because of the recent selection of UO/sub 2/ fuel material for the PWR core and blanket. The fabrication characteristics of UO/sub 2/ powders are being studied.more » Seamless Zircaloy-2 tubing has been tested to determine elastic limits, bursting pressures, and corrosion resistance. Fabrication techniques and tests for corrosion and defects in Zircaloy-clad U-Mo and UO/sub 2/ fuel rods are described. The preparation of UO/sub 2/ by various methods is being studied to determine which method produces a material most suitable for PWR fuel elements. The stability of UO/sub 2/ compacts in high temperature water and steam is being determined. Surface area and density measurements have been performed on samples of UO/sub 2/ powder prepared by various methods. Revelopment work on U-- Mo and U--Nb alloys has included studies of the effect on corrosion behavior of additions to the test water, additions to the alloys, homogenization of the alloys, annealing times, cladding, and fabrication techniques. Data are presented on relaxation in spring materials after exposure to a corrosive environment. Results are reported from loop and autoclave tests on fission product and crud deposition. Results of irradiation and corrosion testing of clad and unclad U--Mo and U-Nh alloys are described. The UO/sub 2/ irradiation program has included studies of dimensional changes, release of fission gases, and activity in the water surrounding the samples. A review of the methods of calculating reactor physics

Alternative Fuels Data Center: Natural Gas Fueling Infrastructure

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

UPM: unified policy-based network management

NASA Astrophysics Data System (ADS)

Law, Eddie; Saxena, Achint

2001-07-01

Besides providing network management to the Internet, it has become essential to offer different Quality of Service (QoS) to users. Policy-based management provides control on network routers to achieve this goal. The Internet Engineering Task Force (IETF) has proposed a two-tier architecture whose implementation is based on the Common Open Policy Service (COPS) protocol and Lightweight Directory Access Protocol (LDAP). However, there are several limitations to this design such as scalability and cross-vendor hardware compatibility. To address these issues, we present a functionally enhanced multi-tier policy management architecture design in this paper. Several extensions are introduced thereby adding flexibility and scalability. In particular, an intermediate entity between the policy server and policy rule database called the Policy Enforcement Agent (PEA) is introduced. By keeping internal data in a common format, using a standard protocol, and by interpreting and translating request and decision messages from multi-vendor hardware, this agent allows a dynamic Unified Information Model throughout the architecture. We have tailor-made this unique information system to save policy rules in the directory server and allow executions of policy rules with dynamic addition of new equipment during run-time.

Effect of hydrocarbon fuel type on fuel

NASA Technical Reports Server (NTRS)

Wong, E. L.; Bittker, D. A.

1982-01-01

A modified jet fuel thermal oxidation tester (JFTOT) procedure was used to evaluate deposit and sediment formation for four pure hydrocarbon fuels over the temperature range 150 to 450 C in 316-stainless-steel heater tubes. Fuel types were a normal alkane, an alkene, a naphthene, and an aromatic. Each fuel exhibited certain distinctive deposit and sediment formation characteristics. The effect of aluminum and 316-stainless-steel heater tube surfaces on deposit formation for the fuel n-decane over the same temperature range was investigated. Results showed that an aluminum surface had lower deposit formation rates at all temperatures investigated. By using a modified JFTOT procedure the thermal stability of four pure hydrocarbon fuels and two practical fuels (Jet A and home heating oil no. 2) was rated on the basis of their breakpoint temperatures. Results indicate that this method could be used to rate thermal stability for a series of fuels.

Conclusions and Recommendations Regarding the Deep Sea Hybrid Power Systems Initial Study

DTIC Science & Technology

2010-06-01

proton-exchange membrane fuel cells ( PEMFC ) powered with hydrogen and oxygen, similar to that used on proven subsurface vessels; (2) fuel-cells...AND STORAGE OPTIONS CONSIDERED FOR INITIAL STUDY NO. NOMENCLATURE DESCRIPTION 1 PWR Nuclear Reactor + Battery 2 FC1 PEMFC + Line for surface O2...Wellhead Gas + Reformer + Battery 3 FC2 PEMFC + Stored O2 + Wellhead Gas + Reformer + Battery 4 SV1 PEMFC + Submersible Vehicle for O2 Transport

Alternative Fuels Data Center: Test Your Alternative Fuel IQ

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Alternative Fuels Data Center: Alternative Fueling Station Locator

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Locate Stations Printable Version Share this resource Send a link to Alternative Fuels Data Center : Alternative Fueling Station Locator to someone by E-mail Share Alternative Fuels Data Center: Alternative Fuels Data Center: Alternative Fueling Station Locator on Digg Find More places to share Alternative

Alternative Fuels Data Center: Hydrogen Fueling Station Locations

Science.gov Websites

Hydrogen Printable Version Share this resource Send a link to Alternative Fuels Data Center : Hydrogen Fueling Station Locations to someone by E-mail Share Alternative Fuels Data Center: Hydrogen Fuels Data Center: Hydrogen Fueling Station Locations on Digg Find More places to share Alternative

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)

Neoclassical theory inside transport barriers in tokamaks

NASA Astrophysics Data System (ADS)

Shaing, K. C.; Hsu, C. T.

2012-02-01

Inside the transport barriers in tokamaks, ion energy losses sometimes are smaller than the value predicted by the standard neoclassical theory. This improvement can be understood in terms of the orbit squeezing theory in addition to the sonic poloidal E ×B Mach number Up,m that pushes the tips of the trapped particles to the higher energy. In general, Up,m also includes the poloidal component of the parallel mass flow speed. These physics mechanisms are the corner stones for the transition theory of the low confinement mode (L-mode) to the high confinement mode (H-mode) in tokamaks. Here, detailed transport fluxes in the banana regime are presented using the parallel viscous forces calculated earlier. It is found, as expected, that effects of orbit squeezing and the sonic Up,m reduce the ion heat conductivity. The former reduces it by a factor of |S|3/2 and the later by a factor of R(Up ,m2)exp(-Up ,m2) with R(Up ,m2), a rational function. Here, S is the orbit squeezing factor.

RELAP-7 Level 2 Milestone Report: Demonstration of a Steady State Single Phase PWR Simulation with RELAP-7

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

David Andrs; Ray Berry; Derek Gaston

The document contains the simulation results of a steady state model PWR problem with the RELAP-7 code. The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on INL's modern scientific software development framework - MOOSE (Multi-Physics Object-Oriented Simulation Environment). This report summarizes the initial results of simulating a model steady-state single phase PWR problem using the current version of the RELAP-7 code. The major purpose of this demonstration simulation is to show that RELAP-7 code can be rapidly developed to simulate single-phase reactor problems. RELAP-7more » is a new project started on October 1st, 2011. It will become the main reactor systems simulation toolkit for RISMC (Risk Informed Safety Margin Characterization) and the next generation tool in the RELAP reactor safety/systems analysis application series (the replacement for RELAP5). The key to the success of RELAP-7 is the simultaneous advancement of physical models, numerical methods, and software design while maintaining a solid user perspective. Physical models include both PDEs (Partial Differential Equations) and ODEs (Ordinary Differential Equations) and experimental based closure models. RELAP-7 will eventually utilize well posed governing equations for multiphase flow, which can be strictly verified. Closure models used in RELAP5 and newly developed models will be reviewed and selected to reflect the progress made during the past three decades. RELAP-7 uses modern numerical methods, which allow implicit time integration, higher order schemes in both time and space, and strongly coupled multi-physics simulations. RELAP-7 is written with object oriented programming language C++. Its development follows modern software design paradigms. The code is easy to read, develop, maintain, and couple with other codes. Most importantly, the modern software design allows the RELAP-7 code

Dictionary of Cancer Terms

MedlinePlus

... causes the cancer cells to grow. Also called AR. androgen receptor positive listen (AN-droh-jen reh- ... binding and actions of androgen hormones. Also called AR+. androgen suppression listen (AN-droh-jen suh-PREH- ...

A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

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

Herer, C.; Souyri, A.; Garnier, J.

1995-09-01

Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to themore » annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.« less

Pretest analysis of natural circulation on the PWR model PACTEL with horizontal steam generators

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

Kervinen, T.; Riikonen, V.; Ritonummi, T.

A new tests facility - parallel channel tests loop (PACTEL)- has been designed and built to simulate the major components and system behavior of pressurized water reactors (PWRs) during postulated small- and medium-break loss-of-coolant accidents. Pretest calculations have been performed for the first test series, and the results of these calculations are being used for planning experiments, for adjusting the data acquisition system, and for choosing the optimal position and type of instrumentation. PACTEL is a volumetrically scaled (1:305) model of the VVER-440 PWR. In all the calculated cases, the natural circulation was found to be effective in removing themore » heat from the core to the steam generator. The loop mass flow rate peaked at 60% mass inventory. The straightening of the loop seals increased the mass flow rate significantly.« less

Fuel cells

NASA Astrophysics Data System (ADS)

Hooie, D. T.; Harrington, B. C., III; Mayfield, M. J.; Parsons, E. L.

1992-07-01

The primary objective of DOE's Fossil Energy Fuel Cell program is to fund the development of key fuel cell technologies in a manner that maximizes private sector participation and in a way that will give contractors the opportunity for a competitive posture, early market entry, and long-term market growth. This summary includes an overview of the Fuel Cell program, an elementary explanation of how fuel cells operate, and a synopsis of the three major fuel cell technologies sponsored by the DOE/Fossil Energy Phosphoric Acid Fuel Cell program, the Molten Carbonate Fuel Cell program, and the Solid Oxide Fuel Cell program.

Method and device for feeding fuel in a fuel system

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

Williamson, E.

1988-07-26

This patent describes a device for feeding fuel in a fuel system for a liquid fuel engine, with the fuel system having a fuel tank, fuel lines, multiple microscreen fuel filters, a fuel pump, and engine fuel injectors, with the fuel tank having a fill opening having a perimeter, comprising, in combination: a ball having a size for overfitting and abutting with the perimeter of the fill opening of differing sizes, shapes, and constructions; and means for introducing air pressure greater than atmospheric through the ball and through the fill opening and into the fuel tank, with the ball havingmore » a solid cross section and being generally impermeable to air passage, with the ball being deformable to conform to the perimeter of the fill opening for sealingly engaging the perimeter of the fill opening and having a firmness for transmitting a force applied to the ball in the direction of the fill opening into a sealing force applied by the ball to the fill opening to balance opposing forces created by the introduction of air pressure into the fuel tank and for increasing the air pressure in the fuel tank acting on the fuel to increase the rate of fuel flow from the fuel tank into the fuel line for assisting the fuel pump in moving the fuel from the fuel tank through the fuel lines and through the microscreen filters to the engine fuel injectors while allowing an excessive air pressure to escape from the fill opening around the ball.« less

Alternate-Fueled Flight: Halophytes, Algae, Bio-, and Synthetic Fuels

NASA Technical Reports Server (NTRS)

Hendricks, R. C.

2012-01-01

Synthetic and biomass fueling are now considered to be near-term aviation alternate fueling. The major impediment is a secure sustainable supply of these fuels at reasonable cost. However, biomass fueling raises major concerns related to uses of common food crops and grasses (some also called "weeds") for processing into aviation fuels. These issues are addressed, and then halophytes and algae are shown to be better suited as sources of aerospace fuels and transportation fueling in general. Some of the history related to alternate fuels use is provided as a guideline for current and planned alternate fuels testing (ground and flight) with emphasis on biofuel blends. It is also noted that lessons learned from terrestrial fueling are applicable to space missions. These materials represent an update (to 2009) and additions to the Workshop on Alternate Fueling Sustainable Supply and Halophyte Summit at Twinsburg, Ohio, October 17 to 18, 2007.

Alternate-Fueled Flight: Halophytes, Algae, Bio-, and Synthetic Fuels

NASA Technical Reports Server (NTRS)

Hendricks, R. C.

2007-01-01

Synthetic and biomass fueling are now considered to be near-term aviation alternate fueling. The major impediment is a secure sustainable supply of these fuels at reasonable cost. However, biomass fueling raises major concerns related to uses of common food crops and grasses (some also called "weeds") for processing into aviation fuels. These issues are addressed, and then halophytes and algae are shown to be better suited as sources of aerospace fuels and transportation fueling in general. Some of the history related to alternate fuels use is provided as a guideline for current and planned alternate fuels testing (ground and flight) with emphasis on biofuel blends. It is also noted that lessons learned from terrestrial fueling are applicable to space missions. These materials represent an update and additions to the Workshop on Alternate Fueling Sustainable Supply and Halophyte Summit at Twinsburg, OH, Oct. 17 to 18, 2007 (ref. 1).

Alternative Fuels Data Center: E85 Flex Fuel Specification

Science.gov Websites

Flexible-Fuel Automotive Spark-Ignition Engines. Fuel retailers or fleets purchasing E85 should require , there is no concern with carrying over winter fuel into the summer months because flexible-fuel vehicles requirements. D5798-15 Standard Specification for Ethanol Fuel Blends for Flexible-Fuel Automotive Spark

Alternative Fuels Data Center: Fuel Cell Electric Vehicle Emissions

Science.gov Websites

Cell Electric Vehicle Emissions to someone by E-mail Share Alternative Fuels Data Center: Fuel Cell Electric Vehicle Emissions on Facebook Tweet about Alternative Fuels Data Center: Fuel Cell Electric Vehicle Emissions on Twitter Bookmark Alternative Fuels Data Center: Fuel Cell Electric Vehicle Emissions

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

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-30

... ENVIRONMENTAL PROTECTION AGENCY 40 CFR Part 80 [EPA-HQ-OAR-2005-0161; FRL-9169-9] RIN 2060-AQ31 Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard Program AGENCY...: June 24, 2010. Lisa P. Jackson, Administrator. PART 80--REGULATION OF FUELS AND FUEL ADDITIVES 0...

Checklist for transition to new highway fuel(s).

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

Risch, C.; Santini, D.J.

2011-12-15

Transportation is vital to the U.S. economy and society. As such, U.S. Presidents have repeatedly stated that the nation needs to reduce dependence on petroleum, especially for the highway transportation sector. Throughout history, highway transportation fuel transitions have been completed successfully both in United States and abroad. Other attempts have failed, as described in Appendix A: Historical Highway Fuel Transitions. Planning for a transition is critical because the changes can affect our nation's ability to compete in the world market. A transition will take many years to complete. While it is tempting to make quick decisions about the new fuel(s)more » of choice, it is preferable and necessary to analyze all the pertinent criteria to ensure that correct decisions are made. Doing so will reduce the number of changes in highway fuel(s). Obviously, changes may become necessary because of occurrences such as significant technology breakthroughs or major world events. With any and all of the possible transitions to new fuel(s), the total replacement of gasoline and diesel fuels is not expected. These conventional fuels are envisioned to coexist with the new fuel(s) for decades, while the revised fuel and vehicle infrastructures are implemented. The transition process must analyze the needs of the primary 'players,' which consist of the customers, the government, the fuel industry, and the automotive industry. To maximize the probability of future successes, the prime considerations of these groups must be addressed. Section 2 presents a succinct outline of the Checklist. Section 3 provides a brief discussion about the groupings on the Checklist.« less

Tensile and Fatigue Testing and Material Hardening Model Development for 508 LAS Base Metal and 316 SS Similar Metal Weld under In-air and PWR Primary Loop Water Conditions

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

Mohanty, Subhasish; Soppet, William; Majumdar, Saurin

This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in September 2015 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2015 report we presented a baseline mechanistic finite element model of a two-loop pressurized water reactor (PWR) for systemlevel heat transfer analysis and subsequent thermal-mechanical stress analysis and fatigue life estimation under reactor thermal-mechanical cycles. In the present report, we provide tensile and fatigue test data for 508 low-alloy steel (LAS) base metal,more » 508 LAS heat-affected zone metal in 508 LAS–316 stainless steel (SS) dissimilar metal welds, and 316 SS-316 SS similar metal welds. The test was conducted under different conditions such as in air at room temperature, in air at 300 oC, and under PWR primary loop water conditions. Data are provided on materials properties related to time-independent tensile tests and time-dependent cyclic tests, such as elastic modulus, elastic and offset strain yield limit stress, and linear and nonlinear kinematic hardening model parameters. The overall objective of this report is to provide guidance to estimate tensile/fatigue hardening parameters from test data. Also, the material models and parameters reported here can directly be used in commercially available finite element codes for fatigue and ratcheting evaluation of reactor components under in-air and PWR water conditions.« less

Microstructural Effects on SCC Initiation PWR Primary Water Cold-Worked Alloy 600

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

Zhai, Ziqing; Toloczko, Mychailo B.; Bruemmer, Stephen M.

SCC initiation behavior of one mill annealed alloy 600 plate heat was investigated in simulated PWR primary water under constant load at yield stress with in-situ direct current potential drop (DCPD) monitoring for crack initiation. Twelve specimens were tested at similar cold work levels among which three showed much shorter SCC initiation times (<400 hrs) than the others (>1200 hrs). Post-test examinations revealed that these three specimens all feature an inhomogeneous microstructure where the primary crack always nucleated along the boundary of large elongated grains protruding normally into the gauge. In contrast, such microstructure was either not observed or didmore » not extend deep enough into the gauge in the other specimens exhibiting ~3-6X longer initiation times. In order to better understand the role of this microstructural inhomogeneity in SCC initiation, high-resolution microscopy was performed to compare carbide morphology and strain distribution between the long grains and normal grains, and their potential effects on SCC initiation are discussed in this paper.« less

Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR)

NASA Astrophysics Data System (ADS)

Brovchenko, Mariya; Dechenaux, Benjamin; Burn, Kenneth W.; Console Camprini, Patrizio; Duhamel, Isabelle; Peron, Arthur

2017-09-01

The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR). The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

Low contaminant formic acid fuel for direct liquid fuel cell

DOEpatents

Masel, Richard I [Champaign, IL; Zhu, Yimin [Urbana, IL; Kahn, Zakia [Palatine, IL; Man, Malcolm [Vancouver, CA

2009-11-17

A low contaminant formic acid fuel is especially suited toward use in a direct organic liquid fuel cell. A fuel of the invention provides high power output that is maintained for a substantial time and the fuel is substantially non-flammable. Specific contaminants and contaminant levels have been identified as being deleterious to the performance of a formic acid fuel in a fuel cell, and embodiments of the invention provide low contaminant fuels that have improved performance compared to known commercial bulk grade and commercial purified grade formic acid fuels. Preferred embodiment fuels (and fuel cells containing such fuels) including low levels of a combination of key contaminants, including acetic acid, methyl formate, and methanol.

Organic fuel cells and fuel cell conducting sheets

DOEpatents

Masel, Richard I.; Ha, Su; Adams, Brian

2007-10-16

A passive direct organic fuel cell includes an organic fuel solution and is operative to produce at least 15 mW/cm.sup.2 when operating at room temperature. In additional aspects of the invention, fuel cells can include a gas remover configured to promote circulation of an organic fuel solution when gas passes through the solution, a modified carbon cloth, one or more sealants, and a replaceable fuel cartridge.

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.

Fuel characteristics pertinent to the design of aircraft fuel systems

NASA Technical Reports Server (NTRS)

Barnett, Henry C; Hibbard, R R

1953-01-01

Because of the importance of fuel properties in design of aircraft fuel systems the present report has been prepared to provide information on the characteristics of current jet fuels. In addition to information on fuel properties, discussions are presented on fuel specifications, the variations among fuels supplied under a given specification, fuel composition, and the pertinence of fuel composition and physical properties to fuel system design. In some instances the influence of variables such as pressure and temperature on physical properties is indicated. References are cited to provide fuel system designers with sources of information containing more detail than is practicable in the present report.

Alternative Fuels Data Center: Alternative Fuels Save Money in Indy

Science.gov Websites

Alternative Fuels Save Money in Indy to someone by E-mail Share Alternative Fuels Data Center : Alternative Fuels Save Money in Indy on Facebook Tweet about Alternative Fuels Data Center: Alternative Fuels Save Money in Indy on Twitter Bookmark Alternative Fuels Data Center: Alternative Fuels Save Money in

Fuel Cell Power Plants Renewable and Waste Fuels

DTIC Science & Technology

2011-01-13

of FuelCell Energy, Inc. Fuels Resources for DFC • Natural Gas and LNG • Propane • Biogas (by Anaerobicnaerobic Digestion) - Municipal Waste...FUEL RESOURCES z NATURAL GAS z PROPANE z DFC H2 (50-60%) z ETHANOL zWASTE METHANE z BIOGAS z COAL GAS Diversity of Fuels plus High Efficiency...trademarks (®) of FuelCell Energy, Inc. DFC Advantages for Biogas • More power for given amount of biogas : Higher efficiency than

Alternative Fuels Data Center: Hydrogen Fuel Cell Electric Vehicle

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Data Center: Hydrogen Fuel Cell Electric Vehicle Availability to someone by E-mail Share Alternative Fuels Data Center: Hydrogen Fuel Cell Electric Vehicle Availability on Facebook Tweet about Alternative Fuels Data Center: Hydrogen Fuel Cell Electric Vehicle Availability on Twitter Bookmark Alternative

Alternative Fuels Data Center: Indianapolis CNG Fueling Station Attracts

Science.gov Websites

Fueling Station Attracts Local Fleets, Turns into Profit Center on Facebook Tweet about Alternative Fuels Data Center: Indianapolis CNG Fueling Station Attracts Local Fleets, Turns into Profit Center on , Turns into Profit Center on Google Bookmark Alternative Fuels Data Center: Indianapolis CNG Fueling

Fuel dissipater for pressurized fuel cell generators

DOEpatents

Basel, Richard A.; King, John E.

2003-11-04

An apparatus and method are disclosed for eliminating the chemical energy of fuel remaining in a pressurized fuel cell generator (10) when the electrical power output of the fuel cell generator is terminated during transient operation, such as a shutdown; where, two electrically resistive elements (two of 28, 53, 54, 55) at least one of which is connected in parallel, in association with contactors (26, 57, 58, 59), a multi-point settable sensor relay (23) and a circuit breaker (24), are automatically connected across the fuel cell generator terminals (21, 22) at two or more contact points, in order to draw current, thereby depleting the fuel inventory in the generator.

Fuel supply device for supplying fuel to an engine combustor

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

Lindsay, M.H.; Kerr, W.B.

1990-05-29

This patent describes a variable flow rate fuel supply device for supplying fuel to an engine combustor. It comprises: fuel metering means having a fuel valve means for controlling the flow rate of fuel to the combustor; piston means for dividing a first cooling fluid chamber from a second cooling fluid chamber; coupling means for coupling the piston means to the fuel valve means; and cooling fluid supply means in communication with the first and second cooling fluid chamber for producing a first pressure differential across the piston means for actuating the fuel valve means in a first direction, andmore » for producing a second pressure differential across the piston means for actuating the valve means in a second direction opposite the first direction, to control the flow rate of the fuel through the fuel metering means and into the engine combustor; and means for positioning the fuel metering means within the second cooling air chamber enabling the cooling air supply means to both cool the fuel metering means and control the fuel supply rate of fuel supplied by the fuel metering means to the combustor.« less

75 FR 18907 - Sunshine Federal Register Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-13

...)-191, Assessment of Debris Accumulation on Pressurized Water Reactor (PWR) Sump Performance (Public... Fuel Cycle Oversight Process Revisions (Public Meeting). (Contact: Michael Raddatz, 301-492-3108.) This... Bolduc, Chief, Employee/Labor Relations and Work Life Branch, at 301-492-2230, TDD: 301-415-2100, or by e...

Apparatus and method for grounding compressed fuel fueling operator

DOEpatents

Cohen, Joseph Perry; Farese, David John; Xu, Jianguo

2002-06-11

A safety system for grounding an operator at a fueling station prior to removing a fuel fill nozzle from a fuel tank upon completion of a fuel filling operation is provided which includes a fuel tank port in communication with the fuel tank for receiving and retaining the nozzle during the fuel filling operation and a grounding device adjacent to the fuel tank port which includes a grounding switch having a contact member that receives physical contact by the operator and where physical contact of the contact member activates the grounding switch. A releasable interlock is included that provides a lock position wherein the nozzle is locked into the port upon insertion of the nozzle into the port and a release position wherein the nozzle is releasable from the port upon completion of the fuel filling operation and after physical contact of the contact member is accomplished.

Advanced fuel system technology for utilizing broadened property aircraft fuels

NASA Technical Reports Server (NTRS)

Reck, G. M.

1980-01-01

Factors which will determine the future supply and cost of aviation turbine fuels are discussed. The most significant fuel properties of volatility, fluidity, composition, and thermal stability are discussed along with the boiling ranges of gasoline, naphtha jet fuels, kerosene, and diesel oil. Tests were made to simulate the low temperature of an aircraft fuel tank to determine fuel tank temperatures for a 9100-km flight with and without fuel heating; the effect of N content in oil-shale derived fuels on the Jet Fuel Thermal Oxidation Tester breakpoint temperature was measured. Finally, compatibility of non-metallic gaskets, sealants, and coatings with increased aromatic content jet fuels was examined.

Carbon, fire, and fuels: The importance of fuels and fuel characterization and the status of wildland fire fuels data for the United States

NASA Astrophysics Data System (ADS)

French, N. H. F.; Prichard, S.; McKenzie, D.; Kennedy, M. C.; Billmire, M.; Ottmar, R. D.; Kasischke, E. S.

2016-12-01

Quantification of emissions of carbon during combustion relies on knowing three general variables: how much landscape is impacted by fire (burn area), how much carbon is in that landscape (fuel loading), and fuel properties that determine the fraction that is consumed (fuel condition). These variables also determine how much carbon remains at the site in the form of unburned organic material or char, and therefore drive post-fire carbon dynamics and pools. In this presentation we review the importance of understanding fuel type, fuel loading, and fuel condition for quantifying carbon dynamics properly during burning and for measuring and mapping fuels across landscapes, regions, and continents. Variability in fuels has been shown to be a major driver of uncertainty in fire emissions, but has had little attention until recently. We review the current state of fuel characterization for fire management and carbon accounting, and present a new approach to quantifying fuel loading for use in fire-emissions mapping and for improving fire-effects assessment. The latest results of a study funded by the Joint Fire Science Program (JFSP) are presented, where a fuel loading database is being built to quantify variation in fuel loadings, as represented in the Fuel Characteristic Classification System (FCCS), across the conterminous US and Alaska. Statistical assessments of these data at multiple spatial scales will improve tools used by fire managers and scientists to quantify fire's impact on the land, atmosphere, and carbon cycle.

78 FR 12005 - Regulation of Fuels and Fuel Additives: 2013 Renewable Fuel Standards; Public Hearing

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-21

... ENVIRONMENTAL PROTECTION AGENCY 40 CFR Part 80 [EPA-HQ-OAR-2012-0546; FRL-9784-4] RIN 2060-AR43 Regulation of Fuels and Fuel Additives: 2013 Renewable Fuel Standards; Public Hearing AGENCY: Environmental... Additives: 2013 Renewable Fuel Standards,'' which was published separately in the Federal Register on...

Fuel property effects on Navy aircraft fuel systems

NASA Technical Reports Server (NTRS)

Moses, C. A.

1984-01-01

Problems of ensuring compatibility of Navy aircraft with fuels that may be different than the fuels for which the equipment was designed and qualified are discussed. To avoid expensive requalification of all the engines and airframe fuel systems, methodologies to qualify future fuels by using bench-scale and component testing are being sought. Fuel blends with increasing JP5-type aromatic concentration were seen to produce less volume swell than an equivalent aromatic concentration in the reference fuel. Futhermore, blends with naphthenes, decalin, tetralin, and naphthalenes do not deviate significantly from the correlation line of aromatic blends, Similar results are found with tensile strenth and elongation. Other elastomers, sealants, and adhesives are also being tested.

40 CFR 80.8 - Sampling methods for gasoline, diesel fuel, fuel additives, and renewable fuels.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 40 Protection of Environment 17 2014-07-01 2014-07-01 false Sampling methods for gasoline, diesel... Provisions § 80.8 Sampling methods for gasoline, diesel fuel, fuel additives, and renewable fuels. The sampling methods specified in this section shall be used to collect samples of gasoline, diesel fuel...

77 FR 13009 - Regulation of Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-05

... ENVIRONMENTAL PROTECTION AGENCY 40 CFR Part 80 [EPA-HQ-OAR-2011-0542; FRL-9642-3] RIN 2060-AR07 Regulation of Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel Pathways Under... of Sec. 80.1426 to identify additional renewable fuel production pathways and pathway components that...

Alternative Fuels Data Center: About the Alternative Fuels Data Center

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About Printable Version Share this resource Send a link to Alternative Fuels Data Center: About the Alternative Fuels Data Center to someone by E-mail Share Alternative Fuels Data Center: About the Alternative Fuels Data Center on Facebook Tweet about Alternative Fuels Data Center: About the Alternative Fuels

Hydrogen and Fuel Cell Basics | Hydrogen and Fuel Cells | NREL

Science.gov Websites

Hydrogen and Fuel Cell Basics Hydrogen and Fuel Cell Basics NREL researchers are working to unlock the potential of hydrogen as a fuel and to advance fuel cell technologies for automobiles, equipment basics of NREL's hydrogen and fuel cell research and development. Fuel cell electric vehicles (FCEVs

Fuel weight and removal costs in fuel-break construction

Treesearch

James L. Murphy

1966-01-01

Three major fuel types were sampled during fuel-break construction on the west side of the Sierra Nevada, California. Fuel weight per acre ranged from 12.2. to 420.6 tons per acre on average. Fuel-break construction removed 27.9 to 40.5 percent of teh total fuel; costs ranged from $141.29 to $148.38 per acre and $1.16 to $4.43 per ton.

Inert matrix fuel in dispersion type fuel elements

NASA Astrophysics Data System (ADS)

Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

2006-06-01

The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

Molten Carbonate Fuel Cell Operation With Dual Fuel Flexibility

DTIC Science & Technology

2007-10-01

electrolyte membrane fuel cell ( PEMFC ). At the higher operating temperature, fuel reforming of natural gas can occur internally, eliminating the need...oxygen PAFC Phosphoric Acid Fuel Cell PEMFC Polymer Electrolyte Membrane Fuel Cell PDS Propane Desulfurization System ppm parts per million psig

40 CFR 600.206-93 - Calculation and use of fuel economy values for gasoline-fueled, diesel-fueled, electric, alcohol...

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Calculation and use of fuel economy values for gasoline-fueled, diesel-fueled, electric, alcohol-fueled, natural gas-fueled, alcohol dual fuel, and natural gas dual fuel vehicle configurations. 600.206-93 Section 600.206-93 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY ...

75 FR 15752 - Sunshine Act Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-30

...)--191, Assessment of Debris Accumulation on Pressurized Water Reactor (PWR) Sump Performance (Public... Fuel Cycle Oversight Process Revisions (Public Meeting). (Contact: Michael Raddatz, 301-492-3108.) This..., Employee/Labor Relations and Work Life Branch, at 301-492-2230, TDD: 301-415-2100, or by e-mail at angela...

Genomic Diversity in the Endosymbiotic Bacterium Rhizobium leguminosarum.

PubMed

Sánchez-Cañizares, Carmen; Jorrín, Beatriz; Durán, David; Nadendla, Suvarna; Albareda, Marta; Rubio-Sanz, Laura; Lanza, Mónica; González-Guerrero, Manuel; Prieto, Rosa Isabel; Brito, Belén; Giglio, Michelle G; Rey, Luis; Ruiz-Argüeso, Tomás; Palacios, José M; Imperial, Juan

2018-01-24

Rhizobium leguminosarum bv. viciae is a soil α-proteobacterium that establishes a diazotrophic symbiosis with different legumes of the Fabeae tribe. The number of genome sequences from rhizobial strains available in public databases is constantly increasing, although complete, fully annotated genome structures from rhizobial genomes are scarce. In this work, we report and analyse the complete genome of R. leguminosarum bv. viciae UPM791. Whole genome sequencing can provide new insights into the genetic features contributing to symbiotically relevant processes such as bacterial adaptation to the rhizosphere, mechanisms for efficient competition with other bacteria, and the ability to establish a complex signalling dialogue with legumes, to enter the root without triggering plant defenses, and, ultimately, to fix nitrogen within the host. Comparison of the complete genome sequences of two strains of R. leguminosarum bv. viciae , 3841 and UPM791, highlights the existence of different symbiotic plasmids and a common core chromosome. Specific genomic traits, such as plasmid content or a distinctive regulation, define differential physiological capabilities of these endosymbionts. Among them, strain UPM791 presents unique adaptations for recycling the hydrogen generated in the nitrogen fixation process.

Purchasing unconventional fuels

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

Doyal, J.

1995-09-01

The reasons to seek unconventional fuels are to either insure a fuel supply or to reduce existing fuel costs. The keys to successfully utilizing unconventional fuel are: (1) build as much flexibility as possible in the variety, quality, quantity and deliverability of the unconventional fuel sources that you can utilize; (2) provide maximum pricing flexibility; (3) fully commit to work with unconventional fuel suppliers; and to (4) consider unconventional fuel operations as a market opportunity. Unconventional fuels operations are well suited to marginal existing operations and can also supplement new startups with uncertain fuel supplies. Unconventional fuel operations can alsomore » help existing facilities generate above market profits for those willing to accommodate the wide market swings inherent in this fledgling industry.« less

Analytical methods in the high conversion reactor core design

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

Zeggel, W.; Oldekop, W.; Axmann, J.K.

High conversion reactor (HCR) design methods have been used at the Technical University of Braunschweig (TUBS) with the technological support of Kraftwerk Union (KWU). The present state and objectives of this cooperation between KWU and TUBS in the field of HCRs have been described using existing design models and current activities aimed at further development and validation of the codes. The hard physical and thermal-hydraulic boundary conditions of pressurized water reactor (PWR) cores with a high degree of fuel utilization result from the tight packing of the HCR fuel rods and the high fissionable plutonium content of the fuel. Inmore » terms of design, the problem will be solved with rod bundles whose fuel rods are adjusted by helical spacers to the proposed small rod pitches. These HCR properties require novel computational models for neutron physics, thermal hydraulics, and fuel rod design. By means of a survey of the codes, the analytical procedure for present-day HCR core design is presented. The design programs are currently under intensive development, as design tools with a solid, scientific foundation and with essential parameters that are widely valid and are required for a promising optimization of the HCR core. Design results and a survey of future HCR development are given. In this connection, the reoptimization of the PWR core in the direction of an HCR is considered a fascinating scientific task, with respect to both economic and safety aspects.« less

Alternative Fuels Data Center: Propane Fueling Infrastructure Development

Science.gov Websites

availability is the foundation for the acceptance of any fuel. Fleets depend on being able to locate fuel fuel systems and containers in vehicles, and the National Fire Prevention Association's NFPA 58

PWR-related integral safety experiments in the PKL 111 test facility SBLOCA under beyond-design-basis accident conditions

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

Weber, P.; Umminger, K.J.; Schoen, B.

1995-09-01

The thermal hydraulic behavior of a PWR during beyond-design-basis accident scenarios is of vital interest for the verification and optimization of accident management procedures. Within the scope of the German reactor safety research program experiments were performed in the volumetrically scaled PKL 111 test facility by Siemens/KWU. This highly instrumented test rig simulates a KWU-design PWR (1300 MWe). In particular, the latest tests performed related to a SBLOCA with additional system failures, e.g. nitrogen entering the primary system. In the case of a SBLOCA, it is the goal of the operator to put the plant in a condition where themore » decay heat can be removed first using the low pressure emergency core cooling system and then the residual heat removal system. The experimental investigation presented assumed the following beyond-design-basis accident conditions: 0.5% break in a cold leg, 2 of 4 steam generators (SGs) isolated on the secondary side (feedwater- and steam line-valves closed), filled with steam on the primary side, cooldown of the primary system using the remaining two steam generators, high pressure injection system only in the two loops with intact steam generators, if possible no operator actions to reach the conditions for residual heat removal system activation. Furthermore, it was postulated that 2 of the 4 hot leg accumulators had a reduced initial water inventory (increased nitrogen inventory), allowing nitrogen to enter the primary system at a pressure of 15 bar and nearly preventing the heat transfer in the SGs ({open_quotes}passivating{close_quotes} U-tubes). Due to this the heat transfer regime in the intact steam generators changed remarkably. The primary system showed self-regulating system effects and heat transfer improved again (reflux-condenser mode in the U-tube inlet region).« less

Fuel Cell and Hydrogen Technology Validation | Hydrogen and Fuel Cells |

Science.gov Websites

NREL Fuel Cell and Hydrogen Technology Validation Fuel Cell and Hydrogen Technology Validation The NREL technology validation team works on validating hydrogen fuel cell electric vehicles; hydrogen fueling infrastructure; hydrogen system components; and fuel cell use in early market applications such as

CERDEC Fuel Cell Team: Military Transitions for Soldier Fuel Cells

DTIC Science & Technology

2008-10-27

Fuel Cell (DMFC) (PEO Soldier) Samsung: 20W DMFC (CRADA) General Atomics & Jadoo: 50W Ammonia Borane Fueled PEMFC Current Fuel Cell Team Efforts...Continued Ardica: 20W Wearable PEMFC operating on Chemical Hydrides Spectrum Brands w/ Rayovac: Hydrogen Generators and Alkaline Fuel Cells for AA...100W Ammonia Borane fueled PEMFC Ultralife: 150W sodium borohydride fueled PEMFC Protonex: 250W RMFC and Power Manager (ARO) NanoDynamics: 250W SOFC

Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

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

Heidet, F.; Kim, T.; Grandy, C.

2012-07-30

Although thorium has long been considered as an alternative to uranium-based fuels, most of the reactors built to-date have been fueled with uranium-based fuel with the exception of a few reactors. The decision to use uranium-based fuels was initially made based on the technology maturity compared to thorium-based fuels. As a result of this experience, lot of knowledge and data have been accumulated for uranium-based fuels that made it the predominant nuclear fuel type for extant nuclear power. However, following the recent concerns about the extent and availability of uranium resources, thorium-based fuels have regained significant interest worldwide. Thorium ismore » more abundant than uranium and can be readily exploited in many countries and thus is now seen as a possible alternative. As thorium-based fuel technologies mature, fuel conversion from uranium to thorium is expected to become a major interest in both thermal and fast reactors. In this study the feasibility of fuel conversion in a fast reactor is assessed and several possible approaches are proposed. The analyses are performed using the Advanced Fast Reactor (AFR-100) design, a fast reactor core concept recently developed by ANL. The AFR-100 is a small 100 MW{sub e} reactor developed under the US-DOE program relying on innovative fast reactor technologies and advanced structural and cladding materials. It was designed to be inherently safe and offers sufficient margins with respect to the fuel melting temperature and the fuel-cladding eutectic temperature when using U-10Zr binary metal fuel. Thorium-based metal fuel was preferred to other thorium fuel forms because of its higher heavy metal density and it does not need to be alloyed with zirconium to reduce its radiation swelling. The various approaches explored cover the use of pure thorium fuel as well as the use of thorium mixed with transuranics (TRU). Sensitivity studies were performed for the different scenarios envisioned in order to

Development of Efficient Charge-Selective Materials for Bulk Heterojunction Polymer Solar Cells

DTIC Science & Technology

2015-01-15

Cho, D. S. Ginger and A. K. Y. Jen, Advanced Materials, 2014, 26, 6262-6267. 17. N. Cho, C.-Z. Li, H.-L. Yip and A. K. Y. Jen, Energy...Solar Cells”, Adv. Funct. Mater. 2015, ASAP. 4. J. H. Kim, P. W. Liang, S. T. Williams, N. C. Cho, C. C. Chueh, M. S. Glaz, D. S. Ginger , A. K.-Y...Kim, P. W. Liang, S. T. Williams, N. C. Cho, C. C. Chueh, M. S. Glaz, D. S. Ginger , A. K.-Y. Jen, 2015, ASAP. 7. P. W. Liang, C. C. Chueh, X. K

Fuel injector system

DOEpatents

Hsu, Bertrand D.; Leonard, Gary L.

1988-01-01

A fuel injection system particularly adapted for injecting coal slurry fuels at high pressures includes an accumulator-type fuel injector which utilizes high-pressure pilot fuel as a purging fluid to prevent hard particles in the fuel from impeding the opening and closing movement of a needle valve, and as a hydraulic medium to hold the needle valve in its closed position. A fluid passage in the injector delivers an appropriately small amount of the ignition-aiding pilot fuel to an appropriate region of a chamber in the injector's nozzle so that at the beginning of each injection interval the first stratum of fuel to be discharged consists essentially of pilot fuel and thereafter mostly slurry fuel is injected.

Alternative Fuels

DTIC Science & Technology

2009-06-11

equipment when supplying jet fuel not practicable or cost effective Unclassified 5 erna ve ue s ocus Petroleum Crude Oil (declining discovery / production...on Jet A/A-1 Approved fuels, DXXXX Unclassified 6 JP-8/5 (Commercial Jet Fuel, ASTM Spec) DARPA Alternative Jet Fuels • Agricultural crop oils ...canola, jatropha, soy, palm , etc.) Alternative fuels – University of North Dakota EERC – UOP – General Electric (GE) t i o n C o s t t i o n C o s t

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.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-23

... Fuels and Alternative Fueled Vehicles AGENCY: Federal Trade Commission (FTC or Commission). ACTION... Alternative Fuels and Alternative Fueled Vehicles'') to consolidate the FTC's alternative fueled vehicle (AFV...) established federal programs to encourage the development of alternative fuels and alternative fueled vehicles...

Fuel transfer system

DOEpatents

Townsend, Harold E.; Barbanti, Giancarlo

1994-01-01

A nuclear fuel bundle fuel transfer system includes a transfer pool containing water at a level above a reactor core. A fuel transfer machine therein includes a carriage disposed in the transfer pool and under the water for transporting fuel bundles. The carriage is selectively movable through the water in the transfer pool and individual fuel bundles are carried vertically in the carriage. In a preferred embodiment, a first movable bridge is disposed over an upper pool containing the reactor core, and a second movable bridge is disposed over a fuel storage pool, with the transfer pool being disposed therebetween. A fuel bundle may be moved by the first bridge from the reactor core and loaded into the carriage which transports the fuel bundle to the second bridge which picks up the fuel bundle and carries it to the fuel storage pool.

Alternative Fuels Data Center: Emerging Fuels

Science.gov Websites

quantities. More research is needed to characterize the impacts of these fuels, such as necessary vehicle modifications, required fueling infrastructure, human health impacts, greenhouse gas emissions, and tailpipe

Alternative Fuels Data Center: Vehicle Parts and Equipment to Conserve Fuel

Science.gov Websites

Vehicle Parts and Equipment to Conserve Fuel to someone by E-mail Share Alternative Fuels Data Center: Vehicle Parts and Equipment to Conserve Fuel on Facebook Tweet about Alternative Fuels Data Center: Vehicle Parts and Equipment to Conserve Fuel on Twitter Bookmark Alternative Fuels Data Center

Impact of hydrogen SAE J2601 fueling methods on fueling time of light-duty fuel cell electric vehicles

DOE PAGES

Reddi, Krishna; Elgowainy, Amgad; Rustagi, Neha; ...

2017-05-16

Hydrogen fuel cell electric vehicles (HFCEVs) are zero-emission vehicles (ZEVs) that can provide drivers a similar experience to conventional internal combustion engine vehicles (ICEVs), in terms of fueling time and performance (i.e. power and driving range). The Society of Automotive Engineers (SAE) developed fueling protocol J2601 for light-duty HFCEVs to ensure safe vehicle fills while maximizing fueling performance. This study employs a physical model that simulates and compares the fueling performance of two fueling methods, known as the “lookup table” method and the “MC formula” method, within the SAE J2601 protocol. Both the fueling methods provide fast fueling of HFCEVsmore » within minutes, but the MC formula method takes advantage of active measurement of precooling temperature to dynamically control the fueling process, and thereby provides faster vehicle fills. Here, the MC formula method greatly reduces fueling time compared to the lookup table method at higher ambient temperatures, as well as when the precooling temperature falls on the colder side of the expected temperature window for all station types. Although the SAE J2601 lookup table method is the currently implemented standard for refueling hydrogen fuel cell vehicles, the MC formula method provides significant fueling time advantages in certain conditions; these warrant its implementation in future hydrogen refueling stations for better customer satisfaction with fueling experience of HFCEVs.« less

Impact of hydrogen SAE J2601 fueling methods on fueling time of light-duty fuel cell electric vehicles

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

Reddi, Krishna; Elgowainy, Amgad; Rustagi, Neha

Hydrogen fuel cell electric vehicles (HFCEVs) are zero-emission vehicles (ZEVs) that can provide drivers a similar experience to conventional internal combustion engine vehicles (ICEVs), in terms of fueling time and performance (i.e. power and driving range). The Society of Automotive Engineers (SAE) developed fueling protocol J2601 for light-duty HFCEVs to ensure safe vehicle fills while maximizing fueling performance. This study employs a physical model that simulates and compares the fueling performance of two fueling methods, known as the “lookup table” method and the “MC formula” method, within the SAE J2601 protocol. Both the fueling methods provide fast fueling of HFCEVsmore » within minutes, but the MC formula method takes advantage of active measurement of precooling temperature to dynamically control the fueling process, and thereby provides faster vehicle fills. Here, the MC formula method greatly reduces fueling time compared to the lookup table method at higher ambient temperatures, as well as when the precooling temperature falls on the colder side of the expected temperature window for all station types. Although the SAE J2601 lookup table method is the currently implemented standard for refueling hydrogen fuel cell vehicles, the MC formula method provides significant fueling time advantages in certain conditions; these warrant its implementation in future hydrogen refueling stations for better customer satisfaction with fueling experience of HFCEVs.« less

Impact of Reprocessed Uranium Management on the Homogeneous Recycling of Transuranics in PWRs

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

Youinou, Gilles J.

This article presents the results of a neutronics analysis related to the homogeneous recycling of transuranics (TRU) in PWRs with a MOX fuel using enriched uranium instead of depleted uranium. It also addresses an often, if not always, overlooked aspect related to the recycling of TRU in PWRs, namely the use of reprocessed uranium. From a neutronics point of view, it is possible to multi-recycle the entirety of the plutonium with or without neptunium and americium in a PWR fleet using MOX-EU fuel in between one third and two thirds of the fleet. Recycling neptunium and americium with plutonium significantlymore » decreases the decay heat of the waste stream between 100 to 1,000 years compared to those of an open fuel cycle or when only plutonium is recycled. The uranium present in MOX-EU used fuel still contains a significant amount of 235uranium and recycling it makes a major difference on the natural uranium needs. For example, a PWR fleet recycling its plutonium, neptunium and americium in MOXEU needs 28 percent more natural uranium than a reference UO 2 open cycle fleet generating the same energy if the reprocessed uranium is not recycled and 19 percent less if the reprocessed uranium is recycled back in the reactors, i.e. a 47 percent difference.« less

Impact of Reprocessed Uranium Management on the Homogeneous Recycling of Transuranics in PWRs

DOE PAGES

Youinou, Gilles J.

2017-05-04

This article presents the results of a neutronics analysis related to the homogeneous recycling of transuranics (TRU) in PWRs with a MOX fuel using enriched uranium instead of depleted uranium. It also addresses an often, if not always, overlooked aspect related to the recycling of TRU in PWRs, namely the use of reprocessed uranium. From a neutronics point of view, it is possible to multi-recycle the entirety of the plutonium with or without neptunium and americium in a PWR fleet using MOX-EU fuel in between one third and two thirds of the fleet. Recycling neptunium and americium with plutonium significantlymore » decreases the decay heat of the waste stream between 100 to 1,000 years compared to those of an open fuel cycle or when only plutonium is recycled. The uranium present in MOX-EU used fuel still contains a significant amount of 235uranium and recycling it makes a major difference on the natural uranium needs. For example, a PWR fleet recycling its plutonium, neptunium and americium in MOXEU needs 28 percent more natural uranium than a reference UO 2 open cycle fleet generating the same energy if the reprocessed uranium is not recycled and 19 percent less if the reprocessed uranium is recycled back in the reactors, i.e. a 47 percent difference.« less

Alternative fuels

NASA Technical Reports Server (NTRS)

Grobman, J. S.; Butze, H. F.; Friedman, R.; Antoine, A. C.; Reynolds, T. W.

1977-01-01

Potential problems related to the use of alternative aviation turbine fuels are discussed and both ongoing and required research into these fuels is described. This discussion is limited to aviation turbine fuels composed of liquid hydrocarbons. The advantages and disadvantages of the various solutions to the problems are summarized. The first solution is to continue to develop the necessary technology at the refinery to produce specification jet fuels regardless of the crude source. The second solution is to minimize energy consumption at the refinery and keep fuel costs down by relaxing specifications.

Alternative Fuels Data Center: How Do Flexible Fuel Cars Work Using

Science.gov Websites

Ethanol? Flexible Fuel Cars Work Using Ethanol? to someone by E-mail Share Alternative Fuels Data Center: How Do Flexible Fuel Cars Work Using Ethanol? on Facebook Tweet about Alternative Fuels Data Center: How Do Flexible Fuel Cars Work Using Ethanol? on Twitter Bookmark Alternative Fuels Data

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.

Fuel transfer system

DOEpatents

Townsend, H.E.; Barbanti, G.

1994-03-01

A nuclear fuel bundle fuel transfer system includes a transfer pool containing water at a level above a reactor core. A fuel transfer machine therein includes a carriage disposed in the transfer pool and under the water for transporting fuel bundles. The carriage is selectively movable through the water in the transfer pool and individual fuel bundles are carried vertically in the carriage. In a preferred embodiment, a first movable bridge is disposed over an upper pool containing the reactor core, and a second movable bridge is disposed over a fuel storage pool, with the transfer pool being disposed therebetween. A fuel bundle may be moved by the first bridge from the reactor core and loaded into the carriage which transports the fuel bundle to the second bridge which picks up the fuel bundle and carries it to the fuel storage pool. 6 figures.

Development of PEM fuel cell technology at international fuel cells

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

Wheeler, D.J.

1996-04-01

The PEM technology has not developed to the level of phosphoric acid fuel cells. Several factors have held the technology development back such as high membrane cost, sensitivity of PEM fuel cells to low level of carbon monoxide impurities, the requirement to maintain full humidification of the cell, and the need to pressurize the fuel cell in order to achieve the performance targets. International Fuel Cells has identified a hydrogen fueled PEM fuel cell concept that leverages recent research advances to overcome major economic and technical obstacles.

Microbial fuel cell treatment of fuel process wastewater

DOEpatents

Borole, Abhijeet P; Tsouris, Constantino

2013-12-03

The present invention is directed to a method for cleansing fuel processing effluent containing carbonaceous compounds and inorganic salts, the method comprising contacting the fuel processing effluent with an anode of a microbial fuel ell, the anode containing microbes thereon which oxidatively degrade one or more of the carbonaceous compounds while producing electrical energy from the oxidative degradation, and directing the produced electrical energy to drive an electrosorption mechanism that operates to reduce the concentration of one or more inorganic salts in the fuel processing effluent, wherein the anode is in electrical communication with a cathode of the microbial fuel cell. The invention is also directed to an apparatus for practicing the method.

Solid oxide fuel cell generator with removable modular fuel cell stack configurations

DOEpatents

Gillett, James E.; Dederer, Jeffrey T.; Zafred, Paolo R.; Collie, Jeffrey C.

1998-01-01

A high temperature solid oxide fuel cell generator produces electrical power from oxidation of hydrocarbon fuel gases such as natural gas, or conditioned fuel gases, such as carbon monoxide or hydrogen, with oxidant gases, such as air or oxygen. This electrochemical reaction occurs in a plurality of electrically connected solid oxide fuel cells bundled and arrayed in a unitary modular fuel cell stack disposed in a compartment in the generator container. The use of a unitary modular fuel cell stack in a generator is similar in concept to that of a removable battery. The fuel cell stack is provided in a pre-assembled self-supporting configuration where the fuel cells are mounted to a common structural base having surrounding side walls defining a chamber. Associated generator equipment may also be mounted to the fuel cell stack configuration to be integral therewith, such as a fuel and oxidant supply and distribution systems, fuel reformation systems, fuel cell support systems, combustion, exhaust and spent fuel recirculation systems, and the like. The pre-assembled self-supporting fuel cell stack arrangement allows for easier assembly, installation, maintenance, better structural support and longer life of the fuel cells contained in the fuel cell stack.

Alternative Fuels

EPA Pesticide Factsheets

Alternative fuels include gaseous fuels such as hydrogen, natural gas, and propane; alcohols such as ethanol, methanol, and butanol; vegetable and waste-derived oils; and electricity. Overview of alternative fuels is here.

Prototype Demonstration of Gamma- Blind Tensioned Metastable Fluid Neutron/Multiplicity/Alpha Detector – Real Time Methods for Advanced Fuel Cycle Applications

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

McDeavitt, Sean M.

represent the more significant hazards from a radiation protection standpoint. However, alpha and neutron emitting uranium and TRU elements represent the more significant safeguards and security concerns. Table 1.1 presents a representative PWR inventory of the uranium and actinide isotopes present in a used fuel assembly. The uranium and actinide isotopes (chiefly the Pu, Am and Cm elements) are all emitters of alpha particles and some of them release significant quantities of neutrons through spontaneous fissions« less

ENVIRONMENTAL TECHNOLOGY VERIFICATION REPORT: ENVIROFUELS DIESEL FUEL CATALYZER FUEL ADDITIVE

EPA Science Inventory

EPA's Environmental Technology Verification Program has tested EnviroFuels diesel fuel additive, called the Diesel Fuel Catalyzer. EnviroFuels has stated that heavy-duty on and off road diesel engines are the intended market for the catalyzer. Preliminary tests conducted indicate...

Assessment of bio-fuel options for solid oxide fuel cell applications

NASA Astrophysics Data System (ADS)

Lin, Jiefeng

Rising concerns of inadequate petroleum supply, volatile crude oil price, and adverse environmental impacts from using fossil fuels have spurred the United States to promote bio-fuel domestic production and develop advanced energy systems such as fuel cells. The present dissertation analyzed the bio-fuel applications in a solid oxide fuel cell-based auxiliary power unit from environmental, economic, and technological perspectives. Life cycle assessment integrated with thermodynamics was applied to evaluate the environmental impacts (e.g., greenhouse gas emission, fossil energy consumption) of producing bio-fuels from waste biomass. Landfill gas from municipal solid wastes and biodiesel from waste cooking oil are both suggested as the promising bio-fuel options. A nonlinear optimization model was developed with a multi-objective optimization technique to analyze the economic aspect of biodiesel-ethanol-diesel ternary blends used in transportation sectors and capture the dynamic variables affecting bio-fuel productions and applications (e.g., market disturbances, bio-fuel tax credit, policy changes, fuel specification, and technological innovation). A single-tube catalytic reformer with rhodium/ceria-zirconia catalyst was used for autothermal reformation of various heavy hydrocarbon fuels (e.g., diesel, biodiesel, biodiesel-diesel, and biodiesel-ethanol-diesel) to produce a hydrogen-rich stream reformates suitable for use in solid oxide fuel cell systems. A customized mixing chamber was designed and integrated with the reformer to overcome the technical challenges of heavy hydrocarbon reformation. A thermodynamic analysis, based on total Gibbs free energy minimization, was implemented to optimize the operating environment for the reformations of various fuels. This was complimented by experimental investigations of fuel autothermal reformation. 25% biodiesel blended with 10% ethanol and 65% diesel was determined to be viable fuel for use on a truck travelling with

Elimination of fuel pressure fluctuation and multi-injection fuel mass deviation of high pressure common-rail fuel injection system

NASA Astrophysics Data System (ADS)

Li, Pimao; Zhang, Youtong; Li, Tieshuan; Xie, Lizhe

2015-03-01

The influence of fuel pressure fluctuation on multi-injection fuel mass deviation has been studied a lot, but the fuel pressure fluctuation at injector inlet is still not eliminated efficiently. In this paper, a new type of hydraulic filter consisting of a damping hole and a chamber is developed for elimination of fuel pressure fluctuation and multi-injection fuel mass deviation. Linear model of the improved high pressure common-rail system(HPCRS) including injector, the pipe connecting common-rail with injector and the hydraulic filter is built. Fuel pressure fluctuation at injector inlet, on which frequency domain analysis is conducted through fast Fourier transformation, is acquired at different target pressure and different damping hole diameter experimentally. The linear model is validated and can predict the natural frequencies of the system. Influence of damping hole diameter on fuel pressure fluctuation is analyzed qualitatively based on the linear model, and it can be inferred that an optimal diameter of the damping hole for elimination of fuel pressure fluctuation exists. Fuel pressure fluctuation and fuel mass deviation under different damping hole diameters are measured experimentally, and it is testified that the amplitude of both fuel pressure fluctuation and fuel mass deviation decreases first and then increases with the increasing of damping hole diameter. The amplitude of main injection fuel mass deviation can be reduced by 73% at most under pilot-main injection mode, and the amplitude of post injection fuel mass deviation can be reduced by 92% at most under main-post injection mode. Fuel mass of a single injection increases with the increasing of the damping hole diameter. The hydraulic filter proposed by this research can be potentially used to eliminate fuel pressure fluctuation at injector inlet and improve the stability of HPCRS fuel injection.

Turbine combustor with fuel nozzles having inner and outer fuel circuits

DOEpatents

Uhm, Jong Ho; Johnson, Thomas Edward; Kim, Kwanwoo

2013-12-24

A combustor cap assembly for a turbine engine includes a combustor cap and a plurality of fuel nozzles mounted on the combustor cap. One or more of the fuel nozzles would include two separate fuel circuits which are individually controllable. The combustor cap assembly would be controlled so that individual fuel circuits of the fuel nozzles are operated or deliberately shut off to provide for physical separation between the flow of fuel delivered by adjacent fuel nozzles and/or so that adjacent fuel nozzles operate at different pressure differentials. Operating a combustor cap assembly in this fashion helps to reduce or eliminate the generation of undesirable and potentially harmful noise.

Reforming options for hydrogen production from fossil fuels for PEM fuel cells

NASA Astrophysics Data System (ADS)

Ersoz, Atilla; Olgun, Hayati; Ozdogan, Sibel

PEM fuel cell systems are considered as a sustainable option for the future transport sector in the future. There is great interest in converting current hydrocarbon based transportation fuels into hydrogen rich gases acceptable by PEM fuel cells on-board of vehicles. In this paper, we compare the results of our simulation studies for 100 kW PEM fuel cell systems utilizing three different major reforming technologies, namely steam reforming (SREF), partial oxidation (POX) and autothermal reforming (ATR). Natural gas, gasoline and diesel are the selected hydrocarbon fuels. It is desired to investigate the effect of the selected fuel reforming options on the overall fuel cell system efficiency, which depends on the fuel processing, PEM fuel cell and auxiliary system efficiencies. The Aspen-HYSYS 3.1 code has been used for simulation purposes. Process parameters of fuel preparation steps have been determined considering the limitations set by the catalysts and hydrocarbons involved. Results indicate that fuel properties, fuel processing system and its operation parameters, and PEM fuel cell characteristics all affect the overall system efficiencies. Steam reforming appears as the most efficient fuel preparation option for all investigated fuels. Natural gas with steam reforming shows the highest fuel cell system efficiency. Good heat integration within the fuel cell system is absolutely necessary to achieve acceptable overall system efficiencies.

Solid oxide fuel cell generator with removable modular fuel cell stack configurations

DOEpatents

Gillett, J.E.; Dederer, J.T.; Zafred, P.R.; Collie, J.C.

1998-04-21

A high temperature solid oxide fuel cell generator produces electrical power from oxidation of hydrocarbon fuel gases such as natural gas, or conditioned fuel gases, such as carbon monoxide or hydrogen, with oxidant gases, such as air or oxygen. This electrochemical reaction occurs in a plurality of electrically connected solid oxide fuel cells bundled and arrayed in a unitary modular fuel cell stack disposed in a compartment in the generator container. The use of a unitary modular fuel cell stack in a generator is similar in concept to that of a removable battery. The fuel cell stack is provided in a pre-assembled self-supporting configuration where the fuel cells are mounted to a common structural base having surrounding side walls defining a chamber. Associated generator equipment may also be mounted to the fuel cell stack configuration to be integral therewith, such as a fuel and oxidant supply and distribution systems, fuel reformation systems, fuel cell support systems, combustion, exhaust and spent fuel recirculation systems, and the like. The pre-assembled self-supporting fuel cell stack arrangement allows for easier assembly, installation, maintenance, better structural support and longer life of the fuel cells contained in the fuel cell stack. 8 figs.

Alternative aircraft fuels

NASA Technical Reports Server (NTRS)

Longwell, J. P.; Grobman, J.

1978-01-01

In connection with the anticipated impossibility to provide on a long-term basis liquid fuels derived from petroleum, an investigation has been conducted with the objective to assess the suitability of jet fuels made from oil shale and coal and to develop a data base which will allow optimization of future fuel characteristics, taking energy efficiency of manufacture and the tradeoffs in aircraft and engine design into account. The properties of future aviation fuels are examined and proposed solutions to problems of alternative fuels are discussed. Attention is given to the refining of jet fuel to current specifications, the control of fuel thermal stability, and combustor technology for use of broad specification fuels. The first solution is to continue to develop the necessary technology at the refinery to produce specification jet fuels regardless of the crude source.

Cognitive Work Analysis: Foundations, Extensions, and Challenges

DTIC Science & Technology

2011-11-01

1 . Introduction A quarter of a century has passed since Jens Rasmussen published his first...achievements is the development of the cognitive work analysis framework itself, which has its origins in the work of Jens Rasmussen and his colleagues...intentionally blank UNCLASSIFIED DSTO-GD-0680 Contents 1 . INTRODUCTION

Simulation of German PKL refill/reflood experiment K9A using RELAP4/MOD7. [PWR

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

Hsu, M.T.; Davis, C.B.; Behling, S.R.

This paper describes a RELAP4/MOD7 simulation of West Germany's Kraftwerk Union (KWU) Primary Coolant Loop (PKL) refill/reflood experiment K9A. RELAP4/MOD7, a best-estimate computer program for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This study was the first major simulation using RELAP4/MOD7 since its release by the Idaho National Engineering Laboratory (INEL). The PKL facility is a reduced scale (1:134) representation of a typical West German four-loop 1300 MW pressurized water reactor (PWR). A prototypical scale of the total volume to power ratio wasmore » maintained. The test facility was designed specifically for an experiment simulating the refill/reflood phase of a Loss-of-Coolant Accident (LOCA).« less

Fuel loads and fuel type mapping

USGS Publications Warehouse

Chuvieco, Emilio; Riaño, David; Van Wagtendonk, Jan W.; Morsdof, Felix; Chuvieco, Emilio

2003-01-01

Correct description of fuel properties is critical to improve fire danger assessment and fire behaviour modeling, since they guide both fire ignition and fire propagation. This chapter deals with properties of fuel that can be considered static in short periods of time: biomass loads, plant geometry, compactness, etc. Mapping these properties require a detail knowledge of vegetation vertical and horizontal structure. Several systems to classify the great diversity of vegetation characteristics in few fuel types are described, as well as methods for mapping them with special emphasis on those based on remote sensing images.

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)

Tanker Fuel Consolidation: Impact of Fuel Efficiency on ATO Resiliency

DTIC Science & Technology

2011-06-01

states, “Unnecessary risk comes without a commensurate return in terms of real benefits or available opportunities…The most logical choices for... bingo fuel‟ and is not made available to receiver aircraft. Almost all tankers in the problem set had „extra‟ fuel remaining above bingo fuel after...transfer fuel above its bingo fuel, ensuring its ability to return to its departure airfield. The receiving tanker will only accept fuel up to a level of

Fuel Cell and Hydrogen Technologies Program | Hydrogen and Fuel Cells |

Science.gov Websites

NREL Fuel Cell and Hydrogen Technologies Program Fuel Cell and Hydrogen Technologies Program Through its Fuel Cell and Hydrogen Technologies Program, NREL researches, develops, analyzes, and validates fuel cell and hydrogen production, delivery, and storage technologies for transportation

75 FR 13800 - Sunshine Federal Register Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-23

..., Assessment of Debris Accumulation on Pressurized Water Reactor (PWR) Sump Performance (Public Meeting..., 2010. Week of April 26, 2010--Tentative Thursday, April 29, 2010 9:30 a.m. Briefing on the Fuel Cycle... and Work Life Branch, at 301-492-2230, TDD: 301-415-2100, or by e-mail at [email protected

Sipping fuel and saving lives: increasing fuel economy withoutsacrificing safety

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

Gordon, Deborah; Greene, David L.; Ross, Marc H.

The public, automakers, and policymakers have long worried about trade-offs between increased fuel economy in motor vehicles and reduced safety. The conclusion of a broad group of experts on safety and fuel economy in the auto sector is that no trade-off is required. There are a wide variety of technologies and approaches available to advance vehicle fuel economy that have no effect on vehicle safety. Conversely, there are many technologies and approaches available to advance vehicle safety that are not detrimental to vehicle fuel economy. Congress is considering new policies to increase the fuel economy of new automobiles in ordermore » to reduce oil dependence and reduce greenhouse gas emissions. The findings reported here offer reassurance on an important dimension of that work: It is possible to significantly increase the fuel economy of motor vehicles without compromising their safety. Automobiles on the road today demonstrate that higher fuel economy and greater safety can co-exist. Some of the safest vehicles have higher fuel economy, while some of the least safe vehicles driven today--heavy, large trucks and SUVs--have the lowest fuel economy. At an October 3, 2006 workshop, leading researchers from national laboratories, academia, auto manufacturers, insurance research industry, consumer and environmental groups, material supply industries, and the federal government agreed that vehicles could be designed to simultaneously improve safety and fuel economy. The real question is not whether we can realize this goal, but the best path to get there. The experts' studies reveal important new conclusions about fuel economy and safety, including: (1) Vehicle fuel economy can be increased without affecting safety, and vice versa; (2) Reducing the weight and height of the heaviest SUVs and pickup trucks will simultaneously increase both their fuel economy and overall safety; and (3) Advanced materials can decouple size from mass, creating important new

Edward W. Jennings | NREL

Science.gov Websites

practices promoting a safe and supportive working environment. His research interests include environmental, safety, and health (ESH) and facility issues, he is responsible for implementing safe work plan; working directly with awardees by performing on-site validations to collect and analyze enzyme

Impact of future fuel properties on aircraft engines and fuel systems

NASA Technical Reports Server (NTRS)

Rudey, R. A.; Grobman, J. S.

1978-01-01

The effect of modifications in hydrocarbon jet fuels specifications on engine performance, component durability and maintenance, and aircraft fuel system performance is discussed. Specific topics covered include: specific fuel consumption; ignition at relight limits; exhaust emissions; combustor liner temperatures; carbon deposition; gum formation in fuel nozzles, erosion and corrosion of turbine blades and vanes; deposits in fuel system heat exchangers; and pumpability and flowability of the fuel. Data that evaluate the ability of current technology aircraft to accept fuel specification changes are presented, and selected technological advances that can reduce the severity of the problems are described and discussed.

Alternative Fuels Data Center: Hydrogen Fueling Infrastructure Development

Science.gov Websites

market. As the market expands, fueling infrastructure and vehicle rollout will need to grow together Locations by State More Hydrogen Data | All Maps & Data Publications 2016 Vehicle Technologies Market Report State of the States: Fuel Cells in America 2016, 7th Edition 2014 Fuel Cell Technologies Market

DoD use of Domestically-Produced Alternative Fuels and Alternative Fuel Vehicles

DTIC Science & Technology

2014-04-10

based fuels and biodiesel , in DoD vehicles; (2) current and projected actions by the DoD to increase the use of alternative fuels in vehicles; (3) a...fuels and vehicles. 15. SUBJECT TERMS alternative fuel infrastructure, electric vehicles, biofuels, ethanol, biodiesel , drop-in, synthetic fuel...of: (1) use and potential use of domestically-produced alternative fuels including but not limited to, natural gas based fuels and biodiesel , in DoD

Late Bilinguals Share Syntax Unsparingly between L1 and L2: Evidence from Crosslinguistically Similar and Different Constructions

ERIC Educational Resources Information Center

Hwang, Heeju; Shin, Jeong-Ah; Hartsuiker, Robert J.

2018-01-01

Languages often use different constructions to convey the same meaning. For example, the meaning of a causative construction in English ("Jen had her computer fixed") is conveyed using an active structure in Korean ("Jen-NOM her computer-ACC fixed"), and yet little is known about how bilinguals represent and process such…

TESTING AND ACCEPTANCE OF FUEL PLATES FOR RERTR FUEL DEVELOPMENT EXPERIMENTS

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

J.M. Wight; G.A. Moore; S.C. Taylor

2008-10-01

This paper discusses how candidate fuel plates for RERTR Fuel Development experiments are examined and tested for acceptance prior to reactor insertion. These tests include destructive and nondestructive examinations (DE and NDE). The DE includes blister annealing for dispersion fuel plates, bend testing of adjacent cladding, and microscopic examination of archive fuel plates. The NDE includes Ultrasonic (UT) scanning and radiography. UT tests include an ultrasonic scan for areas of “debonds” and a high frequency ultrasonic scan to determine the "minimum cladding" over the fuel. Radiography inspections include identifying fuel outside of the maximum fuel zone and measurements and calculationsmore » for fuel density. Details of each test are provided and acceptance criteria are defined. These tests help to provide a high level of confidence the fuel plate will perform in the reactor without a breach in the cladding.« less

Heating subsurface formations by oxidizing fuel on a fuel carrier

DOEpatents

Costello, Michael; Vinegar, Harold J.

2012-10-02

A method of heating a portion of a subsurface formation includes drawing fuel on a fuel carrier through an opening formed in the formation. Oxidant is supplied to the fuel at one or more locations in the opening. The fuel is combusted with the oxidant to provide heat to the formation.

Aviation fuels outlook

NASA Technical Reports Server (NTRS)

Momenthy, A. M.

1980-01-01

Options for satisfying the future demand for commercial jet fuels are analyzed. It is concluded that the most effective means to this end are to attract more refiners to the jet fuel market and encourage development of processes to convert oil shale and coal to transportation fuels. Furthermore, changing the U.S. refineries fuel specification would not significantly alter jet fuel availability.

Alternative Fuels Data Center: How Do Bi-fuel Propane Vehicles Work?

Science.gov Websites

Vehicles Work? Bi-fuel propane vehicles typically use a spark-ignited internal combustion engine. A bi-fuel stored on board and the driver can switch between the fuels. The vehicle is equipped with fuel tanks Propane vehicle image Key Components of a Bi-fuel Propane Vehicle Battery: The battery provides

40 CFR 80.602 - What records must be kept by entities in the NRLM diesel fuel, ECA marine fuel, and diesel fuel...

Code of Federal Regulations, 2013 CFR

2013-07-01

... in the NRLM diesel fuel, ECA marine fuel, and diesel fuel additive production, importation, and... (CONTINUED) AIR PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel... fuel additive production, importation, and distribution systems? (a) Records that must be kept by...

40 CFR 80.602 - What records must be kept by entities in the NRLM diesel fuel, ECA marine fuel, and diesel fuel...

Code of Federal Regulations, 2014 CFR

2014-07-01

... in the NRLM diesel fuel, ECA marine fuel, and diesel fuel additive production, importation, and... (CONTINUED) AIR PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel... fuel additive production, importation, and distribution systems? (a) Records that must be kept by...

40 CFR 80.602 - What records must be kept by entities in the NRLM diesel fuel, ECA marine fuel, and diesel fuel...

Code of Federal Regulations, 2011 CFR

2011-07-01

... in the NRLM diesel fuel, ECA marine fuel, and diesel fuel additive production, importation, and... (CONTINUED) AIR PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel... fuel additive production, importation, and distribution systems? (a) Records that must be kept by...

40 CFR 80.602 - What records must be kept by entities in the NRLM diesel fuel, ECA marine fuel, and diesel fuel...

Code of Federal Regulations, 2012 CFR

2012-07-01

... in the NRLM diesel fuel, ECA marine fuel, and diesel fuel additive production, importation, and... (CONTINUED) AIR PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel... fuel additive production, importation, and distribution systems? (a) Records that must be kept by...

40 CFR 80.602 - What records must be kept by entities in the NRLM diesel fuel, ECA marine fuel, and diesel fuel...

Code of Federal Regulations, 2010 CFR

2010-07-01

... in the NRLM diesel fuel, ECA marine fuel, and diesel fuel additive production, importation, and... (CONTINUED) AIR PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel... fuel additive production, importation, and distribution systems? (a) Records that must be kept by...

National Fuel Cell Technology Evaluation Center | Hydrogen and Fuel Cells |

Science.gov Websites

NREL National Fuel Cell Technology Evaluation Center National Fuel Cell Technology Evaluation Center The National Fuel Cell Technology Evaluation Center (NFCTEC) at NREL's Energy Systems Integration Cell Technology Evaluation Center to process and analyze data for a variety of hydrogen and fuel cell

SiC Composite for Fuel Structure Applications

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

Yueh, Ken

Extensive evaluation was performed to determine the suitability of using SiC composite as a boiling water reactor (BWR) fuel channel material. A thin walled SiC composite box, 10 cm in dimension by approximately 1.5 mm wall thickness was fabricated using chemical vapor deposition (CVD) for testing. Mechanical test results and performance evaluations indicate the material could meet BWR channel mechanical design requirement. However, large mass loss of up to 21% was measured in in-pile corrosion test under BWR-like conditions in under 3 months of irradiation. A fresh sister sample irradiated in a follow-up cycle under PWR conditions showed no measureablemore » weight loss and thus supports the hypothesis that the oxidizing condition of the BWR-like coolant chemistry was responsible for the high corrosion rate. A thermodynamic evaluation showed SiC is not stable and the material may oxidize to form SiO 2 and CO 2. Silica has demonstrated stability in high temperature steam environment and form a protective oxide layer under severe accident conditions. However, it does not form a protective layer in water under normal BWR operational conditions due to its high solubility. Corrosion product stabilization by modifying the SiC CVD surface is an approach evaluated in this study to mitigate the high corrosion rate. Titanium and zirconium have been selected as stabilizing elements since both TiSiO 4 and ZrSiO 4 are insoluble in water. Corrosion test results in oxygenated water autoclave indicate TiSiO4 does not form a protective layer. However, zirconium doped test samples appear to form a stable continuous layer of ZrSiO 4 during the corrosion process. Additional process development is needed to produce a good ZrSiC coating to verify functionality of the mitigation concept.« less

Alternative Fuels Data Center: State Alternative Fuel and Advanced Vehicle

Science.gov Websites

Laws and Incentives: 2014 Year in Review State Alternative Fuel and Advanced Vehicle Laws and Fuel and Advanced Vehicle Laws and Incentives: 2014 Year in Review on Facebook Tweet about Alternative Fuels Data Center: State Alternative Fuel and Advanced Vehicle Laws and Incentives: 2014 Year in Review

FUEL3-D: A Spatially Explicit Fractal Fuel Distribution Model

Treesearch

Russell A. Parsons

2006-01-01

Efforts to quantitatively evaluate the effectiveness of fuels treatments are hampered by inconsistencies between the spatial scale at which fuel treatments are implemented and the spatial scale, and detail, with which we model fire and fuel interactions. Central to this scale inconsistency is the resolution at which variability within the fuel bed is considered. Crown...

Fuel Cell Technology Status Analysis | Hydrogen and Fuel Cells | NREL

Science.gov Websites

Technology Status Analysis Fuel Cell Technology Status Analysis Get Involved Fuel cell developers interested in collaborating with NREL on fuel cell technology status analysis should send an email to NREL's Technology Validation Team at techval@nrel.gov. NREL's analysis of fuel cell technology provides objective

Fuels characterization studies. [jet fuels

NASA Technical Reports Server (NTRS)

Seng, G. T.; Antoine, A. C.; Flores, F. J.

1980-01-01

Current analytical techniques used in the characterization of broadened properties fuels are briefly described. Included are liquid chromatography, gas chromatography, and nuclear magnetic resonance spectroscopy. High performance liquid chromatographic ground-type methods development is being approached from several directions, including aromatic fraction standards development and the elimination of standards through removal or partial removal of the alkene and aromatic fractions or through the use of whole fuel refractive index values. More sensitive methods for alkene determinations using an ultraviolet-visible detector are also being pursued. Some of the more successful gas chromatographic physical property determinations for petroleum derived fuels are the distillation curve (simulated distillation), heat of combustion, hydrogen content, API gravity, viscosity, flash point, and (to a lesser extent) freezing point.

Development of new UV-I. I. Cerenkov Viewing Device

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

Kuribara, Masayuki; Nemoto, Koshichi

1994-02-01

The Cerenkov glow images from boiling-water reactors (BWR) and pressurized-water reactors (PWR) irradiated fuel assemblies are generally used for inspections. However, sometimes it is difficult or impossible to identify the image by the conventional Cerenkov Viewing Device (CVD), because of the long cooling time and/or low burnup. Now a new UV-I.I. (Ultra-Violet light Image Intensifier) CVD has been developed, which can detect the very weak Cerenkov glow from spent fuel assemblies. As this new device uses the newly developed proximity focused type UV-I.I., Cerenkov photons are used efficiently, producing better quality Cerenkov glow images. Moreover, since the image is convertedmore » to a video signal, it is easy to improve the signal to noise ratio (S/N) by an image processor. The new CVD was tested at BWR and PWR power plants in Japan, with fuel burnups ranging from 6,200--33,000 MWD/MTU (megawatt days per metric ton of uranium) and cooling times ranging from 370 to 6,200 d. The tests showed that the new CVD is superior to the conventional STA/CRIEPI CVD, and could detect very feeble Cerenkov glow images using an image processor.« less

ADVANCEMENTS IN TIME-SPECTRA ANALYSIS METHODS FOR LEAD SLOWING-DOWN SPECTROSCOPY

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

Smith, Leon E.; Anderson, Kevin K.; Gesh, Christopher J.

2010-08-11

Direct measurement of Pu in spent nuclear fuel remains a key challenge for safeguarding nuclear fuel cycles of today and tomorrow. Lead slowing-down spectroscopy (LSDS) is an active nondestructive assay method that has the potential to provide independent, direct measurement of Pu and U isotopic mass with an uncertainty lower than the approximately 10 percent typical of today’s confirmatory assay methods. Pacific Northwest National Laboratory’s (PNNL) previous work to assess the viability of LSDS for the assay of pressurized water reactor (PWR) assemblies indicated that the method could provide direct assay of Pu-239 and U-235 (and possibly Pu-240 and Pu-241)more » with uncertainties less than a few percent, assuming suitably efficient instrumentation, an intense pulsed neutron source, and improvements in the time-spectra analysis methods used to extract isotopic information from a complex LSDS signal. This previous simulation-based evaluation used relatively simple PWR fuel assembly definitions (e.g. constant burnup across the assembly) and a constant initial enrichment and cooling time. The time-spectra analysis method was founded on a preliminary analytical model of self-shielding intended to correct for assay-signal nonlinearities introduced by attenuation of the interrogating neutron flux within the assembly.« less

PRELIMINARY COUPLING OF THE MONTE CARLO CODE OPENMC AND THE MULTIPHYSICS OBJECT-ORIENTED SIMULATION ENVIRONMENT (MOOSE) FOR ANALYZING DOPPLER FEEDBACK IN MONTE CARLO SIMULATIONS

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

Matthew Ellis; Derek Gaston; Benoit Forget

In recent years the use of Monte Carlo methods for modeling reactors has become feasible due to the increasing availability of massively parallel computer systems. One of the primary challenges yet to be fully resolved, however, is the efficient and accurate inclusion of multiphysics feedback in Monte Carlo simulations. The research in this paper presents a preliminary coupling of the open source Monte Carlo code OpenMC with the open source Multiphysics Object-Oriented Simulation Environment (MOOSE). The coupling of OpenMC and MOOSE will be used to investigate efficient and accurate numerical methods needed to include multiphysics feedback in Monte Carlo codes.more » An investigation into the sensitivity of Doppler feedback to fuel temperature approximations using a two dimensional 17x17 PWR fuel assembly is presented in this paper. The results show a functioning multiphysics coupling between OpenMC and MOOSE. The coupling utilizes Functional Expansion Tallies to accurately and efficiently transfer pin power distributions tallied in OpenMC to unstructured finite element meshes used in MOOSE. The two dimensional PWR fuel assembly case also demonstrates that for a simplified model the pin-by-pin doppler feedback can be adequately replicated by scaling a representative pin based on pin relative powers.« less

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

Vidal, Jean-Marc; Eschbach, Romain; Launay, Agnes

CEA and AREVA-NC have developed and used a depletion code named CESAR for 30 years. This user-friendly industrial tool provides fast characterizations for all types of nuclear fuel (PWR / UOX or MOX or reprocess Uranium, BWR / UOX or MOX, MTR and SFR) and the wastes associated. CESAR can evaluate 100 heavy nuclides, 200 fission products and 150 activation products (with Helium and Tritium formation). It can also characterize the structural material of the fuel (Zircalloy, stainless steel, M5 alloy). CESAR provides depletion calculations for any reactor irradiation history and from 3 months to 1 million years of coolingmore » time. CESAR5.3 is based on the latest calculation schemes recommended by the CEA and on an international nuclear data base (JEFF-3.1.1). It is constantly checked against the CEA referenced and qualified depletion code DARWIN. CESAR incorporates the CEA qualification based on the dissolution analyses of fuel rod samples and the 'La Hague' reprocessing plant feedback experience. AREVA-NC uses CESAR intensively at 'La Hague' plant, not only for prospective studies but also for characterizations at different industrial facilities all along the reprocessing process and waste conditioning (near 150 000 calculations per year). CESAR is the reference code for AREVA-NC. CESAR is used directly or indirectly with other software, data bank or special equipment in many parts of the La Hague plants. The great flexibility of CESAR has rapidly interested other projects. CESAR became a 'tool' directly integrated in some other softwares. Finally, coupled with a Graphical User Interface, it can be easily used independently, responding to many needs for prospective studies as a support for nuclear facilities or transport. An English version is available. For the principal isotopes of U and Pu, CESAR5 benefits from the CEA experimental validation for the PWR UOX fuels, up to a burnup of 60 GWd/t and for PWR MOX fuels, up to 45 GWd/t. CESAR version 5.3 uses the

Carbon fuel particles used in direct carbon conversion fuel cells

DOEpatents

Cooper, John F.; Cherepy, Nerine

2012-10-09

A system for preparing particulate carbon fuel and using the particulate carbon fuel in a fuel cell. Carbon particles are finely divided. The finely dividing carbon particles are introduced into the fuel cell. A gas containing oxygen is introduced into the fuel cell. The finely divided carbon particles are exposed to carbonate salts, or to molten NaOH or KOH or LiOH or mixtures of NaOH or KOH or LiOH, or to mixed hydroxides, or to alkali and alkaline earth nitrates.

Carbon Fuel Particles Used in Direct Carbon Conversion Fuel Cells

DOEpatents

Cooper, John F.; Cherepy, Nerine

2008-10-21

A system for preparing particulate carbon fuel and using the particulate carbon fuel in a fuel cell. Carbon particles are finely divided. The finely dividing carbon particles are introduced into the fuel cell. A gas containing oxygen is introduced into the fuel cell. The finely divided carbon particles are exposed to carbonate salts, or to molten NaOH or KOH or LiOH or mixtures of NaOH or KOH or LiOH, or to mixed hydroxides, or to alkali and alkaline earth nitrates.

Carbon fuel particles used in direct carbon conversion fuel cells

DOEpatents

Cooper, John F [Oakland, CA; Cherepy, Nerine [Oakland, CA

2011-08-16

A system for preparing particulate carbon fuel and using the particulate carbon fuel in a fuel cell. Carbon particles are finely divided. The finely dividing carbon particles are introduced into the fuel cell. A gas containing oxygen is introduced into the fuel cell. The finely divided carbon particles are exposed to carbonate salts, or to molten NaOH or KOH or LiOH or mixtures of NaOH or KOH or LiOH, or to mixed hydroxides, or to alkali and alkaline earth nitrates.

Carbon fuel particles used in direct carbon conversion fuel cells

DOEpatents

Cooper, John F [Oakland, CA; Cherepy, Nerine [Oakland, CA

2012-01-24

A system for preparing particulate carbon fuel and using the particulate carbon fuel in a fuel cell. Carbon particles are finely divided. The finely dividing carbon particles are introduced into the fuel cell. A gas containing oxygen is introduced into the fuel cell. The finely divided carbon particles are exposed to carbonate salts, or to molten NaOH or KOH or LiOH or mixtures of NaOH or KOH or LiOH, or to mixed hydroxides, or to alkali and alkaline earth nitrates.

Fuel characteristics pertinent to the design of aircraft fuel systems, Supplement I : additional information on MIL-F-7914(AER) grade JP-5 fuel and several fuel oils

NASA Technical Reports Server (NTRS)

Barnett, Henry C; Hibbard, Robert R

1953-01-01

Since the release of the first NACA publication on fuel characteristics pertinent to the design of aircraft fuel systems (NACA-RM-E53A21), additional information has become available on MIL-F7914(AER) grade JP-5 fuel and several of the current grades of fuel oils. In order to make this information available to fuel-system designers as quickly as possible, the present report has been prepared as a supplement to NACA-RM-E53A21. Although JP-5 fuel is of greater interest in current fuel-system problems than the fuel oils, the available data are not as extensive. It is believed, however, that the limited data on JP-5 are sufficient to indicate the variations in stocks that the designer must consider under a given fuel specification. The methods used in the preparation and extrapolation of data presented in the tables and figures of this supplement are the same as those used in NACA-RM-E53A21.

Analysis of metal tolerance in Rhizobium leguminosarum strains isolated from an ultramafic soil.

PubMed

Rubio-Sanz, Laura; Brito, Belén; Palacios, Jose

2018-02-01

Natural habitats containing high amounts of heavy metals provide a valuable source of bacteria adapted to deal with metal toxicity. A functional analysis of the population of legume endosymbiotic bacteria in an ultramafic soil was undertaken by studying a collection of Rhizobium leguminosarum bv viciae (Rlv) isolates obtained using pea as trap plant. One of the isolates, Rlv UPM1137, was selected on the basis of its higher tolerance to nickel and cobalt and presence of inducible mechanisms for such tolerance. A random transposon mutagenesis of Rlv UPM1137 allowed the generation of 14 transposant derivatives with increased nickel sensitivity; five of these transposants were also more sensitive to cobalt. Sequencing of the insertion sites revealed that one of the transposants (D2250) was affected in a gene homologous to the cation diffusion facilitator gene dmeF first identified in the metal-resistant bacterium Cupriavidus metallidurans CH34. The symbiotic performance of D2250 and two other transposants bearing single transposon insertions was unaffected under high-metal conditions, suggesting that, in contrast to previous observations in other Rlv strain, metal tolerance in UPM1137 under symbiotic conditions might be supported by functional redundancy between several mechanisms. © FEMS 2018. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

Fuel Production from Seawater and Fuel Cells Using Seawater.

PubMed

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

2017-11-23

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

Evaluation of CASL boiling model for DNB performance in full scale 5x5 fuel bundle with spacer grids

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

Kim, Seung Jun

As one of main tasks for FY17 CASL-THM activity, Evaluation study on applicability of the CASL baseline boiling model for 5x5 DNB application is conducted and the predictive capability of the DNB analysis is reported here. While the baseline CASL-boiling model (GEN- 1A) approach has been successfully implemented and validated with a single pipe application in the previous year’s task, the extended DNB validation for realistic sub-channels with detailed spacer grid configurations are tasked in FY17. The focus area of the current study is to demonstrate the robustness and feasibility of the CASL baseline boiling model for DNB performance inmore » a full 5x5 fuel bundle application. A quantitative evaluation of the DNB predictive capability is performed by comparing with corresponding experimental measurements (i.e. reference for the model validation). The reference data are provided from the Westinghouse Electricity Company (WEC). Two different grid configurations tested here include Non-Mixing Vane Grid (NMVG), and Mixing Vane Grid (MVG). Thorough validation studies with two sub-channel configurations are performed at a wide range of realistic PWR operational conditions.« less

76 FR 18066 - Regulation of Fuels and Fuel Additives: Changes to Renewable Fuel Standard Program

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-01

... ENVIRONMENTAL PROTECTION AGENCY 40 CFR Part 80 Regulation of Fuels and Fuel Additives: Changes to Renewable Fuel Standard Program CFR Correction In Title 40 of the Code of Federal Regulations, Parts 72 to 80, revised as of July 1, 2010, on page 1160, in Sec. 80.1466, in paragraph (h)(1), the equation is...

Methanol Fuel Cell

NASA Technical Reports Server (NTRS)

Voecks, G. E.

1985-01-01

In proposed fuel-cell system, methanol converted to hydrogen in two places. External fuel processor converts only part of methanol. Remaining methanol converted in fuel cell itself, in reaction at anode. As result, size of fuel processor reduced, system efficiency increased, and cost lowered.

Fuel economy and life-cycle cost analysis of a fuel cell hybrid vehicle

NASA Astrophysics Data System (ADS)

Jeong, Kwi Seong; Oh, Byeong Soo

The most promising vehicle engine that can overcome the problem of present internal combustion is the hydrogen fuel cell. Fuel cells are devices that change chemical energy directly into electrical energy without combustion. Pure fuel cell vehicles and fuel cell hybrid vehicles (i.e. a combination of fuel cell and battery) as energy sources are studied. Considerations of efficiency, fuel economy, and the characteristics of power output in hybridization of fuel cell vehicle are necessary. In the case of Federal Urban Driving Schedule (FUDS) cycle simulation, hybridization is more efficient than a pure fuel cell vehicle. The reason is that it is possible to capture regenerative braking energy and to operate the fuel cell system within a more efficient range by using battery. Life-cycle cost is largely affected by the fuel cell size, fuel cell cost, and hydrogen cost. When the cost of fuel cell is high, hybridization is profitable, but when the cost of fuel cell is less than 400 US$/kW, a pure fuel cell vehicle is more profitable.

FUEL ELEMENT SUPPORT

DOEpatents

Wyman, W.L.

1961-06-27

The described cylindrical fuel element has longitudinally spaced sets of short longitudinal ribs circumferentially spaced from one another. The ribs support the fuel element in a coolant tube so that there is an annular space for coolant flow between the fuel element and the interior of the coolant tube. If the fuel element grows as a result of reactor operation, the circumferential distribution of the ribs maintains the uniformity of the annular space between the coolant tube and the fuel element, and the collapsibility of the ribs prevents the fuel element from becoming jammed in the coolant tube.

Risk-Informed External Hazards Analysis for Seismic and Flooding Phenomena for a Generic PWR

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

Parisi, Carlo; Prescott, Steve; Ma, Zhegang

This report describes the activities performed during the FY2017 for the US-DOE Light Water Reactor Sustainability Risk-Informed Safety Margin Characterization (LWRS-RISMC), Industry Application #2. The scope of Industry Application #2 is to deliver a risk-informed external hazards safety analysis for a representative nuclear power plant. Following the advancements occurred during the previous FYs (toolkits identification, models development), FY2017 focused on: increasing the level of realism of the analysis; improving the tools and the coupling methodologies. In particular the following objectives were achieved: calculation of buildings pounding and their effects on components seismic fragility; development of a SAPHIRE code PRA modelsmore » for 3-loops Westinghouse PWR; set-up of a methodology for performing static-dynamic PRA coupling between SAPHIRE and EMRALD codes; coupling RELAP5-3D/RAVEN for performing Best-Estimate Plus Uncertainty analysis and automatic limit surface search; and execute sample calculations for demonstrating the capabilities of the toolkit in performing a risk-informed external hazards safety analyses.« less

Fuel system technology overview

NASA Technical Reports Server (NTRS)

Friedman, R.

1980-01-01

Fuel system research and technology studies are being conducted to investigate the correlations and interactions of aircraft fuel system design and environment with applicable characteristics of the fuel. Topics include: (1) analysis of in-flight fuel temperatures; (2) fuel systems for high freezing point fuels; (3) experimental study of low temperature pumpability; (4) full scale fuel tank simulation; and (5) rapid freezing point measurement.

Aviation-fuel lubricity evaluation

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

Not Available

1988-07-01

Fuel-system components have experienced problems with the slipperiness or lubricity of the fuel back to the early 1960's. As a consequence of the level of refinement necessary for the PWA 523 fuel (now designated MIL-T-38219 grade JP-7) to obtain its high-temperature stability, many of the polar compounds contributing to lubricity had been removed, resulting in abnormal hydraulic fuel-pump wear. A lubricity-enhancing compound was developed (PWA 536) to eliminate the wear problem. High-pressure piston-type fuel pumps were one of the first parts of the engine fuel system to exhibit problems related to fuel properties. One early problem manifested itself as corrosionmore » of silver-plated slipper pads and was related to carryover of residual-chlorides fuel. Fuel controls were another part of the engine fuel system susceptible to fuel properties. Lack of lubricity agents caused fuel control sliding servo valves to stick.« less

14 CFR 29.979 - Pressure refueling and fueling provisions below fuel level.

Code of Federal Regulations, 2012 CFR

2012-01-01

..., DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Fuel System § 29.979 Pressure refueling and fueling provisions below fuel level. (a) Each fueling connection... from that tank in case of malfunction of the fuel entry valve. (b) For systems intended for pressure...

14 CFR 29.979 - Pressure refueling and fueling provisions below fuel level.

Code of Federal Regulations, 2010 CFR

2010-01-01

..., DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Fuel System § 29.979 Pressure refueling and fueling provisions below fuel level. (a) Each fueling connection... from that tank in case of malfunction of the fuel entry valve. (b) For systems intended for pressure...

14 CFR 29.979 - Pressure refueling and fueling provisions below fuel level.

Code of Federal Regulations, 2014 CFR

2014-01-01

..., DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Fuel System § 29.979 Pressure refueling and fueling provisions below fuel level. (a) Each fueling connection... from that tank in case of malfunction of the fuel entry valve. (b) For systems intended for pressure...

14 CFR 29.979 - Pressure refueling and fueling provisions below fuel level.

Code of Federal Regulations, 2013 CFR

2013-01-01

..., DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Fuel System § 29.979 Pressure refueling and fueling provisions below fuel level. (a) Each fueling connection... from that tank in case of malfunction of the fuel entry valve. (b) For systems intended for pressure...

14 CFR 29.979 - Pressure refueling and fueling provisions below fuel level.

Code of Federal Regulations, 2011 CFR

2011-01-01

..., DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Fuel System § 29.979 Pressure refueling and fueling provisions below fuel level. (a) Each fueling connection... from that tank in case of malfunction of the fuel entry valve. (b) For systems intended for pressure...

Automotive Fleet Fuel Consumption Model : Fuel For

DOT National Transportation Integrated Search

1978-01-01

The computer model described in this report is a tool for determining the fuel conservation benefits arising from various hypothetical schedules of new car fuel economy standards. (Portions of this document are not fully legible)

Nuclear fuel element

DOEpatents

Zocher, Roy W.

1991-01-01

A nuclear fuel element and a method of manufacturing the element. The fuel element is comprised of a metal primary container and a fuel pellet which is located inside it and which is often fragmented. The primary container is subjected to elevated pressure and temperature to deform the container such that the container conforms to the fuel pellet, that is, such that the container is in substantial contact with the surface of the pellet. This conformance eliminates clearances which permit rubbing together of fuel pellet fragments and rubbing of fuel pellet fragments against the container, thus reducing the amount of dust inside the fuel container and the amount of dust which may escape in the event of container breach. Also, as a result of the inventive method, fuel pellet fragments tend to adhere to one another to form a coherent non-fragmented mass; this reduces the tendency of a fragment to pierce the container in the event of impact.

Fuel nozzle assembly

DOEpatents

Johnson, Thomas Edward [Greer, SC; Ziminsky, Willy Steve [Simpsonville, SC; Lacey, Benjamin Paul [Greer, SC; York, William David [Greer, SC; Stevenson, Christian Xavier [Inman, SC

2011-08-30

A fuel nozzle assembly is provided. The assembly includes an outer nozzle body having a first end and a second end and at least one inner nozzle tube having a first end and a second end. One of the nozzle body or nozzle tube includes a fuel plenum and a fuel passage extending therefrom, while the other of the nozzle body or nozzle tube includes a fuel injection hole slidably aligned with the fuel passage to form a fuel flow path therebetween at an interface between the body and the tube. The nozzle body and the nozzle tube are fixed against relative movement at the first ends of the nozzle body and nozzle tube, enabling the fuel flow path to close at the interface due to thermal growth after a flame enters the nozzle tube.

Fossil Fuels.

ERIC Educational Resources Information Center

Crank, Ron

This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

Solid Oxide Fuel Cells Operating on Alternative and Renewable Fuels

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

Wang, Xiaoxing; Quan, Wenying; Xiao, Jing

2014-09-30

This DOE project at the Pennsylvania State University (Penn State) initially involved Siemens Energy, Inc. to (1) develop new fuel processing approaches for using selected alternative and renewable fuels – anaerobic digester gas (ADG) and commercial diesel fuel (with 15 ppm sulfur) – in solid oxide fuel cell (SOFC) power generation systems; and (2) conduct integrated fuel processor – SOFC system tests to evaluate the performance of the fuel processors and overall systems. Siemens Energy Inc. was to provide SOFC system to Penn State for testing. The Siemens work was carried out at Siemens Energy Inc. in Pittsburgh, PA. Themore » unexpected restructuring in Siemens organization, however, led to the elimination of the Siemens Stationary Fuel Cell Division within the company. Unfortunately, this led to the Siemens subcontract with Penn State ending on September 23rd, 2010. SOFC system was never delivered to Penn State. With the assistance of NETL project manager, the Penn State team has since developed a collaborative research with Delphi as the new subcontractor and this work involved the testing of a stack of planar solid oxide fuel cells from Delphi.« less

Effect of broadened-specification fuels on aircraft engines and fuel systems

NASA Technical Reports Server (NTRS)

Rudey, R. A.

1979-01-01

A wide variety of studies on the potential effects of broadened-specification fuels on future aircraft engines and fuel systems are summarized. The compositions and characteristics of aircraft fuels that may be derived from current and future crude-oil sources are described, and the most critical properties that may effect aircraft engines and fuel systems are identified and discussed. The problems that are most likely to be encountered because of changes in selected fuel properties are explored; and the related effects on engine performance, component durability and maintenance, and aircraft fuel-system performance are examined. The ability of current technology to accept possible future fuel specification changes is assessed and selected technological advances that can reduce the severity of the potential problems are illustrated.

Effect of broadened-specification fuels on aircraft engines and fuel systems

NASA Technical Reports Server (NTRS)

Rudey, R. A.

1979-01-01

A wide variety of studies on the potential effects of broadened-specification fuels on future aircraft engines and fuel systems are summarized. The compositions and characteristics of aircraft fuels that may be derived from current and future crude-oil sources are described, and the most critical properties that may affect aircraft engines and fuel systems are identified and discussed. The problems that are most likely to be encountered because of changes in selected fuel properties are described; and the related effects on engine performance, component durability and maintenance, and aircraft fuel-system performance are discussed. The ability of current technology to accept possible future fuel-specification changes is discussed, and selected technological advances that can reduce the severity of the potential problems are illustrated.

Highly durable, coking and sulfur tolerant, fuel-flexible protonic ceramic fuel cells.

PubMed

Duan, Chuancheng; Kee, Robert J; Zhu, Huayang; Karakaya, Canan; Chen, Yachao; Ricote, Sandrine; Jarry, Angelique; Crumlin, Ethan J; Hook, David; Braun, Robert; Sullivan, Neal P; O'Hayre, Ryan

2018-05-01

Protonic ceramic fuel cells, like their higher-temperature solid-oxide fuel cell counterparts, can directly use both hydrogen and hydrocarbon fuels to produce electricity at potentially more than 50 per cent efficiency 1,2 . Most previous direct-hydrocarbon fuel cell research has focused on solid-oxide fuel cells based on oxygen-ion-conducting electrolytes, but carbon deposition (coking) and sulfur poisoning typically occur when such fuel cells are directly operated on hydrocarbon- and/or sulfur-containing fuels, resulting in severe performance degradation over time 3-6 . Despite studies suggesting good performance and anti-coking resistance in hydrocarbon-fuelled protonic ceramic fuel cells 2,7,8 , there have been no systematic studies of long-term durability. Here we present results from long-term testing of protonic ceramic fuel cells using a total of 11 different fuels (hydrogen, methane, domestic natural gas (with and without hydrogen sulfide), propane, n-butane, i-butane, iso-octane, methanol, ethanol and ammonia) at temperatures between 500 and 600 degrees Celsius. Several cells have been tested for over 6,000 hours, and we demonstrate excellent performance and exceptional durability (less than 1.5 per cent degradation per 1,000 hours in most cases) across all fuels without any modifications in the cell composition or architecture. Large fluctuations in temperature are tolerated, and coking is not observed even after thousands of hours of continuous operation. Finally, sulfur, a notorious poison for both low-temperature and high-temperature fuel cells, does not seem to affect the performance of protonic ceramic fuel cells when supplied at levels consistent with commercial fuels. The fuel flexibility and long-term durability demonstrated by the protonic ceramic fuel cell devices highlight the promise of this technology and its potential for commercial application.

Fuel Vapor Pressures and the Relation of Vapor Pressure to the Preparation of Fuel for Combustion in Fuel Injection Engines

NASA Technical Reports Server (NTRS)

Joachim, William F; Rothrock, A M

1930-01-01

This investigation on the vapor pressure of fuels was conducted in connection with the general research on combustion in fuel injection engines. The purpose of the investigation was to study the effects of high temperatures such as exist during the first stages of injection on the vapor pressures of several fuels and certain fuel mixtures, and the relation of these vapor pressures to the preparation of the fuel for combustion in high-speed fuel injection engines.

Fuel tank integrity research : fuel tank analyses and test plans

DOT National Transportation Integrated Search

2013-04-15

The Federal Railroad Administrations Office of Research : and Development is conducting research into fuel tank : crashworthiness. Fuel tank research is being performed to : determine strategies for increasing the fuel tank impact : resistance to ...

Thermodynamic analysis of biofuels as fuels for high temperature fuel cells

NASA Astrophysics Data System (ADS)

Milewski, Jarosław; Bujalski, Wojciech; Lewandowski, Janusz

2011-11-01

Based on mathematical modeling and numerical simulations, applicativity of various biofuels on high temperature fuel cell performance are presented. Governing equations of high temperature fuel cell modeling are given. Adequate simulators of both solid oxide fuel cell (SOFC) and molten carbonate fuel cell (MCFC) have been done and described. Performance of these fuel cells with different biofuels is shown. Some characteristics are given and described. Advantages and disadvantages of various biofuels from the system performance point of view are pointed out. An analysis of various biofuels as potential fuels for SOFC and MCFC is presented. The results are compared with both methane and hydrogen as the reference fuels. The biofuels are characterized by both lower efficiency and lower fuel utilization factors compared with methane. The presented results are based on a 0D mathematical model in the design point calculation. The governing equations of the model are also presented. Technical and financial analysis of high temperature fuel cells (SOFC and MCFC) are shown. High temperature fuel cells can be fed by biofuels like: biogas, bioethanol, and biomethanol. Operational costs and possible incomes of those installation types were estimated and analyzed. A comparison against classic power generation units is shown. A basic indicator net present value (NPV) for projects was estimated and commented.

Thermodynamic analysis of biofuels as fuels for high temperature fuel cells

NASA Astrophysics Data System (ADS)

Milewski, Jarosław; Bujalski, Wojciech; Lewandowski, Janusz

2013-02-01

Based on mathematical modeling and numerical simulations, applicativity of various biofuels on high temperature fuel cell performance are presented. Governing equations of high temperature fuel cell modeling are given. Adequate simulators of both solid oxide fuel cell (SOFC) and molten carbonate fuel cell (MCFC) have been done and described. Performance of these fuel cells with different biofuels is shown. Some characteristics are given and described. Advantages and disadvantages of various biofuels from the system performance point of view are pointed out. An analysis of various biofuels as potential fuels for SOFC and MCFC is presented. The results are compared with both methane and hydrogen as the reference fuels. The biofuels are characterized by both lower efficiency and lower fuel utilization factors compared with methane. The presented results are based on a 0D mathematical model in the design point calculation. The governing equations of the model are also presented. Technical and financial analysis of high temperature fuel cells (SOFC and MCFC) are shown. High temperature fuel cells can be fed by biofuels like: biogas, bioethanol, and biomethanol. Operational costs and possible incomes of those installation types were estimated and analyzed. A comparison against classic power generation units is shown. A basic indicator net present value (NPV) for projects was estimated and commented.

Alternative Fuels Data Center

Science.gov Websites

on AddThis.com... Fuel Properties Search Fuel Properties Comparison Create a custom chart comparing Custom Chart Fuel Chart Icon Download the complete fuel comparison chart. More fuel information

NONDESTRUCTIVE EXAMINATION OF FUEL PLATES FOR THE RERTR FUEL DEVELOPMENT EXPERIMENTS

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

N.E. Woolstenhulme; S.C. Taylor; G.A. Moore

2012-09-01

Nuclear fuel is the core component of reactors that is used to produce the neutron flux required for irradiation research purposes as well as commercial power generation. The development of nuclear fuels with low enrichments of uranium is a major endeavor of the RERTR program. In the development of these fuels, the RERTR program uses nondestructive examination (NDE) techniques for the purpose of determining the properties of nuclear fuel plate experiments without imparting damage or altering the fuel specimens before they are irradiated in a reactor. The vast range of properties and information about the fuel plates that can bemore » characterized using NDE makes them highly useful for quality assurance and for analyses used in modeling the behavior of the fuel while undergoing irradiation. NDE is also particularly useful for creating a control group for post-irradiation examination comparison. The two major categories of NDE discussed in this paper are X-ray radiography and ultrasonic testing (UT) inspection/evaluation. The radiographic scans are used for the characterization of fuel meat density and homogeneity as well as the determination of fuel location within the cladding. The UT scans are able to characterize indications such as voids, delaminations, inclusions, and other abnormalities in the fuel plates which are generally referred to as debonds as well as to determine the thickness of the cladding using ultrasonic acoustic microscopy methods. Additionally, the UT techniques are now also being applied to in-canal interim examination of fuel experiments undergoing irradiation and the mapping of the fuel plate surface profile to determine fuel swelling. The methods used to carry out these NDE techniques, as well as how they operate and function, are described along with a description of which properties are characterized.« less

77 FR 61281 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard and Diesel...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-09

... fuel produced by transmix processors. These amendments will allow locomotive and marine diesel fuel produced by transmix processors to meet a maximum 500 parts per million (ppm) sulfur standard provided that... processors while having a neutral or net positive environmental impact. EPA is also amending the fuel marker...

75 FR 14669 - Regulation of Fuels and Fuel Additives: Changes to Renewable Fuel Standard Program

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-26

...Under the Clean Air Act Section 211(o), as amended by the Energy Independence and Security Act of 2007 (EISA), the Environmental Protection Agency is required to promulgate regulations implementing changes to the Renewable Fuel Standard program. The revised statutory requirements specify the volumes of cellulosic biofuel, biomass-based diesel, advanced biofuel, and total renewable fuel that must be used in transportation fuel. This action finalizes the regulations that implement the requirements of EISA, including the cellulosic, biomass- based diesel, advanced biofuel, and renewable fuel standards that will apply to all gasoline and diesel produced or imported in 2010. The final regulations make a number of changes to the current Renewable Fuel Standard program while retaining many elements of the compliance and trading system already in place. This final rule also implements the revised statutory definitions and criteria, most notably the new greenhouse gas emission thresholds for renewable fuels and new limits on renewable biomass feedstocks. This rulemaking marks the first time that greenhouse gas emission performance is being applied in a regulatory context for a nationwide program. As mandated by the statute, our greenhouse gas emission assessments consider the full lifecycle emission impacts of fuel production from both direct and indirect emissions, including significant emissions from land use changes. In carrying out our lifecycle analysis we have taken steps to ensure that the lifecycle estimates are based on the latest and most up-to-date science. The lifecycle greenhouse gas assessments reflected in this rulemaking represent significant improvements in analysis based on information and data received since the proposal. However, we also recognize that lifecycle GHG assessment of biofuels is an evolving discipline and will continue to revisit our lifecycle analyses in the future as new information becomes available. EPA plans to ask the National

Integrated control of transporter endocytosis and recycling by the arrestin-related protein Rod1 and the ubiquitin ligase Rsp5.

PubMed

Becuwe, Michel; Léon, Sébastien

2014-11-07

After endocytosis, membrane proteins can recycle to the cell membrane or be degraded in lysosomes. Cargo ubiquitylation favors their lysosomal targeting and can be regulated by external signals, but the mechanism is ill-defined. Here, we studied the post-endocytic trafficking of Jen1, a yeast monocarboxylate transporter, using microfluidics-assisted live-cell imaging. We show that the ubiquitin ligase Rsp5 and the glucose-regulated arrestin-related trafficking adaptors (ART) protein Rod1, involved in the glucose-induced internalization of Jen1, are also required for the post-endocytic sorting of Jen1 to the yeast lysosome. This new step takes place at the trans-Golgi network (TGN), where Rod1 localizes dynamically upon triggering endocytosis. Indeed, transporter trafficking to the TGN after internalization is required for their degradation. Glucose removal promotes Rod1 relocalization to the cytosol and Jen1 deubiquitylation, allowing transporter recycling when the signal is only transient. Therefore, nutrient availability regulates transporter fate through the localization of the ART/Rsp5 ubiquitylation complex at the TGN.

46 CFR 154.706 - Cargo boil-off as fuel: Fuel lines.

Code of Federal Regulations, 2013 CFR

2013-10-01

... SAFETY STANDARDS FOR SELF-PROPELLED VESSELS CARRYING BULK LIQUEFIED GASES Design, Construction and Equipment Cargo Pressure and Temperature Control § 154.706 Cargo boil-off as fuel: Fuel lines. (a) Gas fuel lines must not pass through accommodation, service, or control spaces. Each gas fuel line passing...

46 CFR 154.706 - Cargo boil-off as fuel: Fuel lines.